Text

Emergency License Amendment Request 09-04 Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs) for Unit 2 Cycle 15
	ML092580627
Person / Time
Site:	Diablo Canyon
Issue date:	09/03/2009
From:	Becker J; Pacific Gas & Electric Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	DCL-09-062
	Download: ML092580627 (32)
	v • d • e

Pacific Gas and Electric Company' James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601 P. 0. Box 56 Avila Beach, CA 93424 805.545.3462 September 3, 2009 Internal: 691.3462 Fax: 805.545.6445 PG&E Letter DCL-09-062 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Diablo Canyon Unit 2 Docket No. Docket No. 50-323, OL-DPR-82

Subject:

Emerqency License Amendment Request 09-04 Revision to Technical Specification 3.7.1, "Main Steam Safety Valves (MSSVs) for Unit 2 Cycle 15"

Dear Commissioners and Staff:

Pursuant to 10 CFR 50.90, Pacific Gas and Electric (PG&E) hereby requests a License Amendment to Facility Operating License No. DPR-82 for Unit 2 of the Diablo.Canyon Power Plant (DCPP). The enclosed License Amendment Request (LAR) proposes a one-time change to Technical Specification (TS) 3.7.1, "Main Steam Safety Valves (MSSVs)," Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs," to allow an increase in the Power Range Neutron Flux High setpoint from 87 percent rated thermal power (RTP) to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 main steam relief valve (RV) RV-224 inoperable.

The enclosure contains a description of the proposed changes, the supporting technical analyses, and the no significant hazards consideration determination.

Attachments 1 and 2 contain marked-up and retyped (clean) TS pages, respectively. provides the marked-up TS Bases changes. TS Bases changes will be implemented pursuant to TS 5.5.14, "Technical Specifications Bases Control Program," at the time this amendment is implemented.

PG&E has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

This LAR is submitted on an emergency basis to allow operation of DCPP Unit 2 at the licensed core power level while in TS 3.7.1, Required Action A.1, until corrective action can be taken to exit the required action during the upcoming Unit 2 fifteenth refueling outage. Unit 2 has one MSSV inoperable, and is currently operating at A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway -

Comanche Peak

Diabto Canyon

Palo Verde

San Onofre -

South Texas Project

Wolf Creek Pacific Gas and Electric Company September 3, 2009 PG&E Letter DCL-09-062 U:S. Nuclear Regulatory Comm!ssion ATTN: Document Control Desk Washington, D.C. 20555-0001 Diablo Canyon Unit 2 Docket No. Docket No. 50-323, OL-DPR-82 James R. Becker Site Vice President

Subject:

Emergency License Amendment Request09-04 Diablo Canyon Power Plant Mail Code 104/5/601

p. O. Box 56 Avila Beach, CA 93424 805.545.3462 Internal: 691.3462 Fax: 805.545.6445 10 CFR 50.90 Revision to Technical Specification 3.7.1, "Main Steam Safety Valves (MSSVs) for Unit 2 Cycle 15"

Dear Commissioners and Staff:

Pursuant to 10 CFR 50.90, Pacific Gas and Electric (PG&E) hereby requests a License Amendment to Facility Operating License No. DPR-82 for Unit 2 of the Diablo Canyon Power Plant (DCPP). The enclosed License Amendment Request

, (LAR) proposes a one-time change to Technical Specification (TS) 3.7.1, "Main Steam Safety Valves (MSSVs)," Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs," to allow an increase in the Power Range Neutron Flux High setpoint from 87 percent rated thermal power (RTP) to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 main steam relief valve (RV) RV-224 inoperable.

The enclosure contains a description of the proposed changes, the supporting technical analyses, and the no significant hazards consideration determination.

Attachments 1 and 2 contain marked-up and retyped (clean) TS pages, respectively. provides the marked-up TS Bases changes. TS Bases changes will be implemented pursuant to TS 5.5.14, "Technical Specifications Bases Control Program," at the time this amendment is implemented.

PG&E has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

This LAR is submitted on an emergency basis to allow operation of DCPP Unit 2 at the licensed core power level while in TS 3.7.1, Required Action A.1, until corrective action can be taken to exit the required action during the upcoming Unit 2 fifteenth refueling outage. Unit 2 has one MSSV inoperable, and is currently operating at A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

Comanche Peak

Diablo Canyon

Palo Verde

San Onofre

South Texas Project

Wolf Creek

Document Control Desk September 3, 2009 Page 2 PG&E Letter DCL-09-062 approximately 80 percent RTP. Approval of this request would allow Unit 2 to operate at 100 percent RTP for the remainder of Cycle 15.

Therefore, PG&E requests approval of this LAR no later than September 10, 2009.

Also, PG&E requests the license amendment(s) be made effective upon NRC issuance, to be implemented within 5 days from the date of issuance.

In accordance with administrative procedures and the Quality Assurance Program Manual, the proposed amendment has been reviewed by the Plant Staff Review Committee and approved by the Station Director.

Pursuant to 10 CFR 50.91, a copy of this proposed amendment is being sent to the California Department of Public Health.

PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.

This letter includes no revisions to existing regulatory commitments.

If you have any questions or require additional information, please contact Larry Parker at (805) 545-3386.

I state under penalty of perjury that the foregoing is true and correct.

Executed on September 3, 2009.

James R. B~ee Site Vice President kjse/4328/N50265286 Enclosure cc:

Gary W. Butner, California Department of Public Health Elmo E. Collins, NRC Region IV Diablo Distribution cc/enc:

Michael S. Peck, NRC, Senior Resident Inspector Alan B. Wang, NRC Project Manager, Office of NRR A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway Comanche Peak

Diablo Canyon Palo Verde

San Onofre

South Texas Project

Wolf Creek Document Control Desk September 3, 2009 Page 2 PG&E Letter DCL-09-062 approximately 80 percent RTP. Approval of this request would allow Unit 2 to operate at 100 percent RTP for the remainder of Cycle 15.

Therefore, PG&E requests approval of this LAR no later than September 10, 2009.

Also, PG&E requests the license amendment(s) be made effective upon NRC issuance, to be implemented within 5 days from the date of issuance.

In accordance with administrative procedures and the Quality Assurance Program Manual, the proposed amendment has been reviewed by the Plant Staff Review Committee and approved by the Station Director.

Pursuant to 10 CFR 50.91, a copy of this proposed amendment is being sent to the California Department of Public Health.

PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.

This letter includes no revisions to existing regulatory commitments.

If you have any questions or require additional information, please contact Larry Parker at (805) 545-3386.

I state under penalty of perjury that the foregoing is true and correct.

Executed on September 3, 2009.

er Site Vice President kjse/4328/N50265286 Enclosure cc:

Gary W. Butner, California Department of Public Health Elmo E. Collins, NRC Region IV Diablo Distribution cc/enc:

Michael S. Peck, NRC, Senior Resident Inspector Alan B. Wang, NRC Project Manager, Office of NRR A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway. Comanche Peak

Diablo Canyon. Palo Verde. San Onofre. South Texas Project. Wolf Creek

Enclosure PG&E Letter DCL-09-062 Evaluation of the Proposed Change

Subject:

Emerqency License Amendment Request 09-04 Revision to Technical Specification 3.7.1, "Main Steam Safety Valves (MSSVs) for Unit 2 Cycle 15"

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATIONS

6. REFERENCES ATTACHMENTS:

1. Technical Specification Page Markups

2. Retyped Technical Specification Pages

3. Technical Specification Bases Page Markups Enclosure PG&E Letter DCL-09-062 Evaluation of the Proposed Change

Subject:

Emergency License Amendment Request 09-04 Revision to Technical Specification 3.7.1! "Main Steam Safety Valves (MSSVs) for Unit 2 Cycle 15"

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION.

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4* Conclusions

5. ENVIRONMENTAL CONSIDERATIONS

6. REFERENCES ATTACHMENTS:

1. Technical Specification Page Markups

2. Retyped Technical Specification Pages

3. Technical Specification Bases Page Markups

Enclosure PG&E Letter DCL-09-062

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating License DPR-82 for Unit 2 of the Diablo Canyon Power Plant (DCPP).

This License Amendment Request (LAR) proposes a one-time change to Technical Specification (TS) 3.7.1, "Main Steam Safety Valves (MSSVs)," Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs," to allow an increase in the Power Range Neutron Flux High setpoint from 87 percent rated thermal power (RTP) to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 main steam (MS) relief valve (RV) RV-224 inoperable.

This LAR is submitted on an emergency basis to allow operation of DCPP Unit 2 at the licensed core power level while in TS 3.7.1, Required Action A.1, until corrective action can be taken to exit the required action during the upcoming Unit 2 fifteenth refueling outage. Unit 2 has one MSSV inoperable, and is currently operating at approximately 80 percent RTP. Approval of this request would allow Unit 2 to operate at 100 percent RTP for the remainder of current Cycle 15.

The TS 3.3.1 power range neutron flux high nominal trip setpoint per TS Table 3.3.1-1, "Reactor Trip System Instrumentation," is 109 percent RTP. No change to this setpoint is requested by this LAR. PG&E has previously made a commitment to the NRC that the Technical Specification Task Force (TSTF) -493 "Clarify Application of Setpoint Methodology for LSSS Functions," changes will be made to the applicable Reactor Trip System and Engineered Safety Feature Actuation System functions in a separate LAR that will be submitted after TSTF-493 is approved by the NRC (reference PG&E Letter DCL-07-002, "License Amendment Request 07-01, Revision to Technical Specifications to Support Steam Generator Replacement," dated January 11, 2007).

2.

DETAILED DESCRIPTION Proposed Amendment TS 3.7.1, "Main Steam Safety Valves (MSSVs)," Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs,"

is revised to allow an increase in the Power Range Neutron Flux High'setpoint from 87 percent RTP to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 MS-2-RV-224 inoperable.

The proposed TS changes are noted on the marked-up TS page provided in. The proposed retyped TS is provided in Attachment 2. TS Bases changes are included for information only in Attachment 3.

1 Enclosure PG&E Letter DCL-09-062

1.

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating License DPR-82 for Unit 2 of the Diablo Canyon Power Plant (DCPP).

This License Amendment Request (LAR) proposes a one-time change to Technical Specification (TS) 3.7.1, "Main Steam Safety Valves (MSSVs)," Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High SetpointWith Inoperable MSSVs," to allow an increase in the Power Range Neutron Flux High setpoint from 87 percent rated thermal power (RTP) to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 main steam (MS) relief valve (RV) RV-224 inoperable.

\\...

This LAR is submitted on an emergency basis to allow operation of DCPP Unit 2 at the licensed core power level while in TS 3.7.1, Required Action A.1, until corrective action can be taken to exit the required action during the upcoming Unit 2 fifteenth refueling outage. Unit 2 has one MSSV inoperable, and is currently operating at approximately 80 percent RTP. Approval of this request would allow Unit 2 to operate at 100 percent RTP for the remainder of current Cycle 15.

The TS 3.3.1 power range neutron flux high nominal trip setpoint per TS Table 3.3.1-1, "Reactor Trip System Instrumentation," is 109 percent RTP. No change to this setpoint is requested by this LAR. PG&E has previously made a commitment to the NRC that the Technical Specification Task Force (TSTF) -493 "Clarify Application of Setpoint Methodology for LSSS Functions," changes will be made to the applicable Reactor Trip System and Engineered Safety Feature Actuation System functions in a separate LAR that will be submitted after TSTF-493 is approved by the NRC (reference PG&E Letter DCL:'07 -002, "License Amendment Request 07-01, Revision to Technical Specifications to Support Steam Generator Replacement," dated January 11,2007).

2.

DETAILED DESCRIPTION Proposed Amendment TS 3.7.1, "Main Steam Safety Valves (MSSVs)," Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs,"

is revised to allow an increase in the Power Range Neutron Flux High'setpoint from,87 percent RTP to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 MS-2-RV-224 inoperable.

The proposed TS changes are noted on the marked-up TS page provided in. The proposed retyped TS is provided in Attachment 2.. TS Bases changes are included for information only in Attachment 3.

)

1

Enclosure PG&E Letter DCL-09-062

System Description

The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurizing the reactor coolant pressure boundary (RCPB) by providing a heat sink for the removal of energy from the reactor coolant system (RCS) if the preferred heat sink, provided by the condenser and circulating water system, is not available.

Five MSSVs are located on each main steam header, outside containment, upstream of the main steam isolation valves (MSIVs), as described in the Final Safety Analysis Report (FSAR), Section 10.3.1. The MSSVs must have sufficient capacity to limit the secondary system pressure to less than or equal to 110 percent of the steam generator (SG) design pressure. The MSSV design includes staggered setpoints, according to TS 3.7.1, Table 3.7.1-2, so that only the needed valves will actuate. Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves during an overpressure event.

The accident analysis requires that five MSSVs per SG be operable to provide overpressure protection for design basis transients occurring at 102 percent RTP.

Purpose for Proposed Amendment This LAR is submitted on an emergency basis to allow Unit 2 operation at the licensed core power level while in TS 3.7.1, Required Action A.1, until corrective action can be taken to exit the required action during the upcoming Unit 2 fifteenth refueling outage. Unit 2 has one MSSV inoperable (MSSV MS-2-RV-224), and is currently operating at approximately 80 percent RTP. Approval of this request would allow Unit 2 to operate at 100 percent RTP for the remainder of Cycle 15 with only MSSV MS-2-RV-224 inoperable. Unit 2 Cycle 15 began on April 12, 2008.

Continued operation with less than all five MSSVs operable for each SG is permissible, if thermal power is limited such that the relief capacity of the remaining MSSVs is sufficient to limit the secondary system pressure to less than or equal to 110 percent of the SG design pressure. This is accomplished by restricting thermal power if necessary and reducing the Power Range Neutron Flux trip setpoint so that the energy transfer to the most limiting SG is not greater than the available relief capacity in that SG.

In the unlikely event an additional MSSV, besides currently inoperable MSSV MS-2-RV-224, becomes inoperable during Unit 2 Cycle 15, the requested

2 System Description

Enclosure PG&E Letter DCL-09-062 The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurizing the reactor coolant pressure boundary (RCPS) by providing a heat sink for the removal of energy from the reactor coolant system (RCS) if the preferred heat sink, provided by the condenser and circulating water system, is not available.

Five MSSVs are located on each main steam header, outside containment, upstream of the main steam isolation valves (MSIVs), as described in the Final Safety Analysis Report (FSAR), Section 10.3.1. The MSSVs must have sufficient capacity to limit the secondary system pressure to less than or equal to 110 percent of the steam generator (SG) design pressure. The MSSV design includes staggered setpoints, according to TS 3.7.1, Table 3.7.1-2, so that only the needed valves will actuate. Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves during an overpressure event.

The accident analysis requires that five MSSVs per SG be operable to provide overpressure protection for design basis transients occurring at 102 percent RTP..

Purpose for Proposed Amendment This LAR is submitted on an emergency basis to allow Unit 2 operation at the licensed core power level while in TS 3.7.1, Required Action A.1,until corrective action can be taken to exit the required action during the upcoming Unit 2 fifteenth refueling outage. Unit 2 has one MSSV inoperable (MSSV MS-2-RV-224), and is currently operating at approximately 80 percent RTP. Approval of this request would allow Unit 2 to operate at 100 percent RTP for the remainder of Cycle 15 with only MSSV MS-2-RV-224 inoperable. Unit 2 Cycle 15 began on April 12, 2008.

Continued operation with less than all five MSSVs operable for each SG is permissible, if thermal power is limited such that the relief capacity of the remaining MSSVs is sufficient to limit the secondary system pressure to less than or equal to 110 percent of the SG design pressure. This is accomplished by restricting thermal power if necessary and reducing the Power Range Neutron Flux trip setpoint so that the energy transfer to the most limiting SG is not greater than the available relief capacity in that SG.

In the unlikely event an additional MSSV, besides currentlyinoperable MSSV MS-2-RV-224, becomes inoperable during Unit 2 Cycle 15, the requested 2

Enclosure PG&E Letter DCL-09-062 TS change would no longer be applicable and the Power Range Neutron Flux High setpoint will be required to be reduced per the requirements of TS Table 3.7.1-1.

Justification and Basis for the Emerqency Circumstances DCPP entered TS 3.7.1, Required Action A.1, at 1245 on August 26, 2009, when the spring on DCPP Unit 2 MSSV MS-2-RV-224 was found broken and the valve was declared inoperable. Unit 2 power was reduced to approximately 80 percent RTP and the Power Range Neutron Flux setpoint was reduced to 87 percent RTP as required by TS Table 3.7.1-1. The unit is operated at a power level below the Power Range Neutron Flux setpoint to ensure that nuclear instrumentation instrument and channel uncertainties do not result in an inadvertent Power Range Neutron Flux High reactor trip signal.

The broken spring was found during valve testing on August 26, 2009, and Notification 50264402 was initiated to enter the condition into the DCPP Corrective Action Program. MS-2-RV-224 was last replaced with a refurbished relief valve in March 1998. Testing is performed every other cycle, and is normally performed as preoutage work. MS-2-RV-224 was last tested per TS Surveillance Requirement 3.7.1.1 in 2006 and the as-found and as-left setpoints were within specification with no adjustments required.

MS-2-RV-224 has the highest specified lift setting of the five RVs per SG.

TS Table 3.7.1-2 specifies 1115 psig + 3 percent for the lift setting. During the August 26, 2009, testing, the valve initially lifted 7 percent low and was adjusted.

In the next test it lifted 4.5 percent low and was adjusted. In the third test it lifted 2.5 percent low and was within the TS Table 3.7.1-2 limit, but lower than the as-left TS Surveillance Requirement 3.7.1 lift setting tolerance of 1 percent, so it was further adjusted. The spring crack was found prior to performance of the as-left test. The valve is currently gagged closed as a conservative measure.

The nature of the degraded condition is such that PG&E could not have predicted the failure.

This LAR is submitted on an emergency basis to allow an increase in power of DCPP Unit 2 from approximately 80 percent RTP to 100 percent RTP for the remainder of Cycle 15 with only MS-2-RV-224 inoperable.

Risk-Informed Licensing Change The requested change. in the LAR is not a risk-informed licensing change.

3 Enclosure PG&E Letter DCL-09-062 TS change would no longer be applicable and the Power Range Neutron Flux High setpoint will be required to be reduced per the requirements of TS Tabl'e 3.7.1-1.

Justification and Basis for the Emergency Circumstances DCPP entered TS 3.7.1, Required Action A.1, at 1245 on August 26, 2009, when the spring on DCPP Unit 2 MSSV MS-2-RV-224 was found broken and the valve was declared inoperable. Unit 2 power was reduced to approximately 80 percent RTP and the Power Range Neutron Flux setpoint was reduced to 87 percent RTP as required by TS Table 3.7.1-1. The unit is operated at a power level below the Power Range Neutron Flux setpoint to ensure that nuclear instrumentation instrument and channel uncertainties do not result in an inadvertent Power Range Neutron Flux High reactor trip signal.

The broken spring was found during valve testing on August 26, 2009, and Notification 50264402 was initiated to enter the condition into the DCPP Corrective Action Program. MS-2-RV-224 was last replaced with a refurbished relief valve in March 1998: Testing is performed every other cycle, and is normally performed as preoutage work. MS-2-RV-224 was last tested per TS Surveillance Requirement 3.7.1.1 in 2006 and the as-found and as-left setpoints were within specification with no adjustments required.

MS-2-RV-224 has the highest specified lift setting of the five RVs per SG.

TS Table 3.7.1-2 specifies 1115 psig.:!:. 3 percent for the lift setting. During the August 26, 2009, testing, the valve initially lifted 7 percent low and was adjusted.

In the next test it lifted 4.5 percent low and was adjusted. In the third test it lifted 2.5 percent low and was within the TS Table 3.7.1-2 limit, but lower than the

. as-left TS Surveillance Requirement 3.7.1 lift setting tolerance of 1 percent, so it was further adjusted. The spring crack was found prior to performance of the as-left test. The valve is currently gagged closed as a conservative measure.

The nature of the degraded condition is such that PG&E could not have predicted

" the failure.

This LAR is submitted on an emergency basis to allow an increase in power of

, DCPP Unit ifrom approximately 80 percent RTP to 100 percent RTP for the remainder of Cycle 15 with only MS-2-RV.:224 inoperable.

Risk-Informed Licensing Change The requested change, in the LAR is not a risk-informed licensing change.

3

Enclosure PG&E Letter DCL-09-062 3

TECHNICAL EVALUATION System Design Basis The design basis for the MSSVs is to limit the secondary system pressure to less than or equal to 110 percent of design pressure for any anticipated operational occurrence (AOO) or accident considered in the Design Basis Accident (DBA) and transient analysis.

The events that challenge the relieving capacity of the MSSVs, and thus RCS pressure, are those characterized as decreased heat removal events, which are presented in the FSAR, Sections 15.2 and 15.3. Of these, the full power turbine trip without steam dump is the limiting AOO with respect to secondary system pressure. The analysis of this event also assumes termination of normal feedwater flow to the SGs.

The safety analysis demonstrates that the transient response for turbine trip occurring from full power without a direct reactor trip presents no hazard to the integrity of the RCS or the main steam supply system (MSSS).

One turbine trip analysis is performed assuming primary system pressure control via operation of the pressurizer relief valves and sprays. Two cases are performed, one to evaluate the limiting affect on the fuel and one to evaluate the limiting effect on secondary pressure. The analysis results demonstrate that the departure from nucleate boiling and the secondary pressure design bases are met. Another analysis is performed assuming no primary system pressure control, but crediting reactor trip on high pressurizer pressure and operation of the pressurizer safety valves. This analysis demonstrates that the maximum RCS pressure does not exceed 110 percent of the design pressure. All cases analyzed demonstrate that the MSSVs maintain MSSS integrity by limiting the maximum steam pressure to less than 110 percent of the SG design pressure.

Safety Analyses With One MSSV Inoperable The analysis that confirmed the acceptability of the 87 percent RTP maximum allowable Power Range Neutron Flux High setpoint with a minimum of four MSSVs operable per loop contained in current TS 3.7.1, Table 3.7.1-1, is discussed in this section.

The limiting FSAR Condition II accident for overpressure concerns is a loss of external load/turbine trip (LOL/TT). The event was analyzed with the RETRAN02/ModOO4 computer program to demonstrate the adequacy of the MSSVs to maintain the MSSS lower than 1210 psia, or 110 percent of the 1085 psig SG design pressure.

4 Enclosure PG&E Letter DCL-09-062 3

TECHNICAL EVALUATION System Design Basis The design basis for the MSSVs is to limit the secondary system pressure to less than or equal to 110 percent of design pressure for any anticipated operational occurrence (AOO) or accident considered in the Design Basis Accident (DBA) and transient analysis.

The events that challenge the relieving capacity of the MSSVs, and thus RCS pressure, are' those characterized as decreased heat removal events, which are presented in the FSAR, Sections 15.2 and 15.3. Of these, the full power turbine trip without steam dump is the limiting AOO with respect to secondary system pressure. The analysis of this event also assumes termination of normal feedwater flow to the SGs.

The safety analysis demonstrates that the transient response for turbine trip occurring from full power without a direct reactor trip presents no hazard to the integrity of the RCS or the main steam supply system (MSSS).

One turbine trip analysis is performed assuming primary system 'pressure control via operation of the pressurizer relief valves and sprays. Two cases are performed, one to evaluate the limiting affect on the fuel and one to evaluate the limiting effect on secondary pressure. The analysis results demonstrate that the departure from nucleate boiling and the secondary pressure design bases are met. Another analysis is performed assuming no primary system pressure control, but crediting reactor trip on high pressurizer pressure and operation of the pressurizer safety valves. This analysis demonstrates that the maximum RCS pressure does not exceed 110 percent of the design pressure. All cases analyzed demonstrate that the MSSVs maintain MSSS integrity by limiting the maximum steam pressure to less than 110 percent of the SG design pressure.

Safety Analyses With One MSSV Inoperable The analysis that confirmed the acceptability of the 87 percent RTP maximum allowable Power Range Neutron Flux High setpoint with a minimum of four MSSVs operable per loop contained in current TS 3.7.1, Table 3.7.1-1, is discussed in this section.

The limiting FSAR Condition II accident for overpressure concerns is a loss of external load/turbine trip (LOLITT). The event was analyzed with the RETRAN02/Mod004 computer program to demonstrate the adequacy of the MSSVs to maintain the MSSS lower than 1210 psia, or 110 percent of the 1085 psig SG design pressure.

4

Enclosure PG&E Letter DCL-09-062 To determine the effect of only four MSSVs per SG being available, PG&E calculation N-1 14, "Over-Pressure Study for One MSSV Per Loop Unavailable,"

dated March 10, 1994, was performed to document the analysis. The analysis assumed a 3 percent tolerance for all the available MSSVs. The MSSV on each SG with the lowest nominal setpoint was assumed unavailable, and the Unit 2 model was used because of the higher Unit 2 thermal rating at the time the calculation was performed.

The PG&E Calculation N-1 14 analysis considered the algorithm contained in Westinghouse Nuclear Safety Advisory Letter (NSAL)94-001, "Operation at Reduced Power Levels with Inoperable MSSVs," dated January 20, 1994.

NSAL 94-001 indicated that the original methodology used to determine the maximum allowable number of inoperable MSSVs at Various power levels was deficient and provided a revised algorithm. NRC Information Notice 94-60, "Potential Overpressurization of Main Steam System," August 22, 1994, provided the information contained in NSAL 94-001 to licensees. The NSAL 94-001 algorithm assumes that the reactor trip following main turbine trip does not occur, and that the entire heat generated by the reactor is removed by the steam flow through the MSSVs. No credit is taken for main feedwater (MFW) flow or the subcooled'water in the SG secondary side, which would provide a cooling effect on the SGs; thereby, contributing to a reduction in SG pressure. Similarly, no credit is taken for the cooling and pressure reducing effects of auxiliary feedwater (AFW) flow because SG pressure would reach a maximum before AFW flow was introduced into the SGs.

The MSSVs are assumed to have two active and one passive failure modes.

The active failure modes are spurious opening and failure to close once opened.

The passive failure mode is failure to open on demand. At DCPP, a passive failure is 'not postulated to occur for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following an accident as previously discussed in PG&E Letter DCL-97-105, "License Amendment Request 97-06, Revision of Technical Specification 3.7.1.1, Table 3.7-1 and Associated Bases - Reduced Power Operation Levels for Inoperable Main Steam Safety Valves (MSSVs)" dated December 23, 1997. Therefore, when developing the algorithm and determining the required reduction in power level, it is not assumed that an MSSV fails to open on demand after the Power Range Neutron Flux High setpoint is reduced.

The analysis assumed the plant operated at 100 percent RTP including consideration of a 2 percent RTP uncertainty. The analysis value used for the Power Range Neutron Flux High setpoints was 118 percent. The 100 percent RTP was conservatively used as the basis for the Power Range Neutron Flux High setpoint and was lowered an additional 6 percent RTP to account for instrument and channel uncertainties. This adjustment resulted in a required Power Range Neutron Flux High setpoint of 94 percent RTP with one MSSV inoperable. Since the required setpoint of 94 percent RTP was higher than the 5

I Enclosure PG&E Letter DCL-09-062 To determine the effect of only four MSSVs per SG being available, PG&E calculation N-114, "Over-Pressure Study for One MSSV Per Loop Unavailable,"

dated March 10, 1994, was performed to document the analysis. The analysis assumed a 3 percent tolerance for all the available MSSVs. The MSSV on each SG with the lowest nominal setpoint was assumed unavailable, and the Unit 2 model was used because of the higher Unit 2 thermal rating at the time the calculation was performed.

The PG&E Calculation N-114 analysis considered the algorithm contained in Westinghouse Nuclear Safety Advisory Letter (NSAL)94-001, "Operation at Reduced Power Levels with Inoperable MSSVs," dated January 20,1994.

NSAL 94-001 indicated that the original methodology used to determine the maximum allowable number of inoperable MSSVs at various power levels was deficient and provided a revised algorithm. NRC Information Notice 94-60, "Potential Overpressurization of Main Steam System," August 22, 1994, provided the information contained in NSAL 94-001 to licensees. The NSAL 94-001 algorithm assumes that the reactor trip following main turbine trip does not occur, and that the entire heat generated by the reactor is removed by the steam flow through the MSSVs. No credit is taken for main feedwater (MFW) flow or the subcooled'water in the SG secondary side, which would provide a cooling effect on the SGs; thereby, contributing to a reduction in SG pressure. Similarly, no credit is taken for the cooling and pressure reducing effects of auxiliary feedwater (AFW) flow because SG pressure would reach a maximum before AFW flow was introduced into the SGs.

The MSSVs are assumed to have two active and one passive failure modes.

The active failure modes are spurious opening and failure to close once opened.

The passive failure mode is failure to open on demand. At DCPP, a passive failure is hot postulated to occur for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following an accident as previously discussed in PG&E Letter DCL-97-1 05, "License Amendment Request 97-06, Revision of Technical Specification 3.7.1.1, Table 3.7-1 and Associated Bases - Reduced Power Operation Levels for Inoperable Main Steam Safety Valves (MSSVs)" dated December 23,1997. Therefore, when developing the algorithm and determining the required reduction in power level, it is not assumed that an MSSV fails to open on demand after the Power Range Neutron Flux High setpoint is reduced.

i The analysis assumed the plant operated at 100 percent RTP including consideration of a 2 percent RTP uncertainty. The analysis value used for the Power Range Neutron Flux High setpoints was 118 percent. The 100 percent RTP was conservatively used as the basis for the Power Range Neutron Flux High setpoint and was lowered an additional 6 percent RTP to account for instrument and channel uncertainties. This adjustment resulted in a required Power Range Neutron Flux High setpoint of 94 percent RTP with one MSSV inoperable. Since the required setpoint of 94 percent RTP was higher than the 5

Enclosure PG&E Letter DCL-09-062 current TS required setpoint of 87 percent RTP, the TS required setpoint of 87 percent RTP was determined to be conservative and was not revised to provide additional TS conservatism.

An evaluation using the NSAL 94-001 algorithm for 2 and 3 MSSVs inoperable per SG, was also performed to determine the required Power Range Neutron Flux High setpoints. After subtracting an instrument and channel inaccuracy of 6 percent RTP based on a channel statistical analysis, the maximum allowable Power Range Neutron Flux High setpoints were determined to be 47 percent RTP and 29 percent RTP for 2 or 3 inoperable MSSVs per SG, respectively.

The results of the analyses were submitted to the NRC in PGE Letter DCL-97-105 and were approved by the NRC in License Amendment No. 125 to Facility Operating License No. DPR-80 and License Amendment No. 123 to Facility Operating License No. DPR-82, "Issuance of Amendments for Diablo Canyon Power Plant, Unit No. 1 (TAC No. MA0397) and Unit No. 2 (TAC No.

MA0398), dated May 28, 1998.

Revised Safety Analyses With One MSSV Inoperable A revised RETRAN/02Mod005.2 analysis has been performed to determine the required Power Range Neutron Flux High setpoint for the one MSSV currently inoperable for Unit 2 Cycle 15.

This revised analysis uses a more recent version of the RETRAN/02 computer code, Mod 5.2, than the RETRAN/02 Mod 4 code used in the original PG&E Calculation N-1 14 that provides the basis for current TS Table 3.7.1-1. There have been no substantial changes in the RETRAN/02 code versions or modeling options used for the reanalysis and the two RETRAN/02 versions have been verified to generate essentially identical results. The PG&E RETRAN/02Mod005.2 model used for the revised analysis was previously reviewed and approved for use for the spurious Safety Injection analysis that was submitted in PG&E Letter DCL-02-115, "License Amendment Request (LAR) 01-08, Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves; Pressurizer Safety Valve Loop Seal Temperature," dated September 24, 2002, and approved by the NRC in License Amendment No. 171 to Facility Operating License No. DPR-80 and License Amendment No. 172 to Facility Operating License No. DPR-82, "Diablo Canyon Power Plant, Unit Nos. 1 and 2

- Issuance of Amendment RE: Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves (TAC Nos. MB6758 MB6759)," dated July 2, 2004.

The revised analysis uses the same RETRAN/02ModOO5.2 model approved for use for the spurious Safety Injection analysis in License Amendments 171 and 172 except the model has been updated to incorporate the new thermal design 6

Enclosure PG&E Letter DCL-09-062 current TS required setpoint of 87 percent RTP, the TS required setpoint of 87 percent RTP was determined to be conservative and was not revised to provide additional TS conservatism.

An evaluation using the NSAL 94-001 algorithm for 2 and 3 MSSVs inoperable per SG, was also performed to determine the required Power Range Neutron Flux High setpoints. After subtracting an instrument and channel inaccuracy of 6 percent RTP based on a channel statistical analysis, the maximum allowable Power Range Neutron Flux High setpoints were determined to be 47 percent RTP and 29 percent RTP for 2 or 3 inoperable MSSVs per SG, respectively.

The results of the analyses were submitted to the NRC in PGE Letter DCL-97 -105 and were approved by the NRC in License Amendment No. 125 to Facility Operating License No. DPR-80 and License Amendment No. 123 to Facility Operating License No. DPR-82, I'ssuance of Amendments for Diablo Canyon Power Plant, Unit No.1 (TAC No. MA0397) and Unit No.2 (TAC No.

MA0398), dated May 28, 1998.

Revised Safety Analyses With One MSSV Inoperable A revised RETRAN/02Mod005.2 analysis has been performed to determine the required Power Range Neutron Flux High setpoint for the one MSSV currently inoperable for Unit 2 Cycle 15.,

This revised analysis uses a more recent version of the RETRAN/02 computer code, Mod 5.2, than the RETRAN/02 Mod 4 code used in the original PG&E Calculation N-114that provides the basis for current TS Table 3.7.1-1. There have been no substantial changes in the RETRAN/02 code versions or modeling options used for the reanalysis and the two RETRAN/02 versions have been verified to generate essentially identical results. The PG&E RETRAN/02Mod005.2 model used for the revised analysis was previously reviewed and approved for use for the spurious Safety Injection analysis that was submitted in PG&E Letter DCL-02-115, "License AmendmentRequest (LAR) 01-08, Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves; Pressurizer Safety Valve' Loop Seal Temperature," dated September 24, 2002, and approved by the NRC in License Amendment No. 171 to Facility Operating License No. DPR-80 and LicenseAmendment No. 172 to Facility Operating License No. DPR-82, "Diablo Canyon Power Plant, Unit Nos. 1 and 2

- Issuance of Amendment RE: Credit for Automatic Actuation of Pressurizer Power qperated Relief Valves (TAC Nos. MB6758 MB6759)," dated July 2, 2004.

The revised analysis uses the same RETRAN/02Mod005.2 model approved for use for the spurious Safety Injection analysis in License Amendments 171 and 172 except the model has been updated to incorporate the new thermal design 6

\\

Enclosure PG&E Letter DCL-09-062 characteristics of the Replacement SGs (RSGs) that have been installed on both DCPP Units. The LOL/TT analysis uses the same conservative LOL/TT methodology as submitted in PG&E Letter DCL-95-141, "License Amendment Request 95-04, Revision of Technical Specification 3/4.4.2.2 - Technical Specifications Administrative Changes," dated June 29, 1995, and approved by the NRC in License Amendment No. 107 to Facility Operating License No. DPR-80 and License Amendment No. 106 to Facility Operating License No. DPR-82, "Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No. 1 (TAC No. M92767) and No. 2 (TAC No. M92768)," dated August 23, 1995, and as currently established in the DCPP FSAR Section 15.2.7 LOL/TT analysis.

The LOL/TT analysis in PG&E calculation N-1 14 assumed a simple conservative safety valve model based on a 3 percent drift in the lift setpoint and a 3 percent additional increase in the accumulation pressure until the safety valve was assumed to be at full flow conditions. The revised RETRAN/02Mod005.2 LOL/TT analysis uses an enhanced MSSV model to more accurately model valve behavior related to full open accumulation conditions. The enhanced MSSV model is consistent with the latest Westinghouse methodology that still assumes a 3 percent setpoint drift but assumes a 5 pound per square inch (psi) accumulation to full open instead of an additional 3 percent, which was considered overly conservative. Westinghouse has established that the 5 psi accumulation more accurately models the actual valve behavior as demonstrated through extensive valve testing. The enhanced MSSV model was used for all of the non-Loss-of-Coolant Accident RETRAN safety analyses performed by Westinghouse for the SG Replacement Project (SGRP) for both Units. This model change included in the Westinghouse SGRP analyses was incorporated into the DCPP design basis via 10 CFR 50.59.

The DCPP RETRAN/02ModOO5.2 MSSV model conservatively accounts for the pressure drop effects of the MSSV inlet header on all MSSVs even though only two of the four main steam lines have these MSSV inlet headers installed.

The LOL/TT evaluation for the inoperable MSSV is performed using the same conservative assumptions and uncertainties applied to the FSAR Section 15.2.7 analysis, as summarized in Table 1. The slightly revised MSSV lift and full open setpoints that include the 5 psi accumulation and conservatively account for the MSSV inlet header pressure drop effects are summarized in Table 2. The latest FSAR 15.2.7 LOL/TT analysis evaluates two Departure form Nucleate Boiling (DNB) scenarios and two overpressure scenarios. The DNB scenarios are not limiting with respect to the SG peak pressure limits and do not require evaluation for the inoperable MSSV. The FSAR LOL/TT overpressure analysis evaluates two separate scenarios. Scenario 1 assumes the automatic pressurizer pressure control system does not function since this maximizes the RCS peak pressure during the event. Scenario 2 assumes that the automatic pressurizer pressure 7

Enclosure PG&E Letter DCL-09-062 characteristics of the Replacement SGs (RSGs) that have been installed on both DCPP Units. The LOLfTT analysis uses the same conservative LOLfTT metho~ology as submitted in PG&E Letter DCL-95-141, "License Amendment.

Request 95-04, Revision of Technical. Specification 3/4.4.2.2 - Technical Specifications Administrative Changes," dated June 29, 1995, and approved by the NRC in License Amendment No.1 07 to Facility Operating License No. DPR-80 and License Amendment No.1 06 to Facility Operating License No. DPR-82, "Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No.1 (TAC No. M92767) and No.2 (TAC No. M92768)," dated August 23, 1995, and as currently established in the DCPP FSAR Section. 15.2.7 LOLfTT analysis.

The LOLfTT analysis in PG&E calculation N-114 assumed a simple conservative safety valve model based on a 3 percent drift in the lift setpoint and a 3 percent additional increase in the accumulation pressure until the safety valve was assumed to be at full flow conditions. The revised RETRAN/02Mod005.2 LOLITT analysis uses an enhanced MSSV model to more accurately model valve behavior related to full open accumulation conditions. The enhanced MSSV model is consistent with the latest Westinghouse methodology that still assumes a 3 percent setpoint drift but assumes a 5 pound per square inch (psi) accumulation to full open instead of an additional 3 percent, which was considered overly conservative. Westinghouse has established that the 5 psi accumulation more accurately models the actual valve behavior as demonstrated through extensive valve testing. The enhanced MSSV model waS used for all of the non-Loss-of-Coolant Accident RETRAN safety analyses performed by Westinghouse for the SG Replacement Project (SGRP) for both Units.. This model change included in the Westinghouse SGRP analyses was incorporated into the DCPP design basis via 10 CFR 50.59.

The DCPP RETRAN/02Mod005.2 MSSV model conservatively accounts for the pressure drop effects of the MSSV inlet header on all MSSVs even though only two of the four main steam lines have these MSSV inlet headers installed.

The LOLfTT evaluation for the inoperable MSSV is performed using the same conservative assumptions and uncertainties applied to the FSAR Section 15.2.7 analysis, as summarized in Table 1. The slightly revised MSSV lift and full open setpoints that include the 5 psi accumulation and conservatively account for the MSSV inlet header pressure drop effects are summ~rized in Table 2. The latest FSAR 15.2.7 LOLfTT analysis evaluates two Departure form Nucleate Boiling (DNB) scenarios and two overpressure scenarios. The DNB scenarios are not limiting with respect to the SG peak pressure limits and do not require evaluation for the inoperable MSSV. The FSAR LOLfTT overpressure analysis evaluates two separate scenarios. Scenario 1 assumes the automatic pressurizer pressure control system does not function since this maximizes the RCS peak pressure during the event.. Scenario 2 assumes that the automatic pressurizer pressure 7

Enclosure PG&E Letter DCL-09-062 control system functions since this delays the reactor trip and results in the maximum primary to secondary heat transfer to the SG resulting in the peak secondary side pressure. Since the RCS peak pressure occurs before the MSSVs open in the Scenario 1 LOL/TT, this case is not impacted and only Scenario 2 requires evaluation for the effects of an inoperable MSSV.

The FSAR LOL/TT overpressure analysis and the Westinghouse Letter NSAL 94-001 methodology are conservatively bounding based on minimizing the subcooled liquid heat removal capability of the SGs. Therefore, the LOL/TT methodology has established the conservatism of the immediate isolation of MFW and no credit forAFW flow during the event. These conservative assumptions have been maintained for this evaluation.

The latest FSAR Section 15.2.7 analysis evaluated LOL/TT cases for both Unit 1 and Unit 2 with 0 percent SG tube plugging (SGTP) and 10 percent SGTP, respectively. The LOL/TT results showed that Unit 2 was more limiting than Unit 1 since it has a slightly higher design RCS vessel average temperature value and that 0 percent SGTP is more limiting for peak SG pressure since it maximizes the primary to secondary heat transfer during the event. Therefore, only the Unit 2 Case 2 with 0 percent SGTP requires evaluation to determine the bounding effects of an inoperable MSSV for the TS 3.7.1, Table 3.7.1-1 maximum allowable Power Range Neutron Flux High setpoint.

Three separate cases are performed to ensure the evaluation bounds the range of potential conditions and events related to the inoperable MSSV on Unit 2 Cycle 15 operation. The first case (Case 1) evaluates a LOL/TT event with the inoperable MSSV and assuming the most positive moderator temperature coefficient (MTC) of +5pcm/F to determine the reduction in the Power Range Neutron Flux High setpoint that is required to bound any potential over power transient. The second case (Case 2) evaluates a LOL/TT event with the inoperable MSSV and the near end of life (EOL) cycle conditions and the associated negative MTC that currently exist on Unit 2. This evaluation demonstrates that Unit 2 can operate at 100 percent RTP for the remainder of Cycle 15 and maintain the peak SG pressure within the applicable ASME limit for any LOL event that does not result in a high flux reactor trip. The third case (Case 3) evaluates the effect of an inadvertent MSIV closure on the single SG with the inoperable MSSV to ensure this asymmetric flow condition does not cause the SG pressure to exceed 110 percent of the design value. Each case was evaluated to ensure the peak SG pressure remains within 110 percent of the ASME code design value, a limit of 1210 psia.

The sequence of events for the three cases is listed in Table 3 and the peak pressure results are shown in Table 4. These cases demonstrate that Unit 2 can operate at a core power level of 102 percent with a Power Range Neutron Flux 8

Enclosure PG&E Letter DCL-09-062 control system functions since this delays the reactor trip and results in the maximum primary to secondary heat transfer to the SG resulting in the peak secondary side pressure. Since the RCS peak pressure occurs before the MSSVs open in the Scenario 1 LOLITT, this case is not impacted and only Scenario 2 requires evaluation for the effects of an inoperable MSSV.

The FSAR LOLITT overpressure analysis and the Westinghouse Letter NsAL 94-001 methodology are conservatively bounding based on minimizing the subcooled liquid heat removal capability of the SGs. Therefore, the LOLITT methodology has established the conservatism of the immediate isolation of MFW and no credit for AFW flow during the event. These conservative assumptions have been maintained for this evaluation.

The latest FSAR Section 15.2.7 analysis evaluated LOLITT cases for both Unit 1 and Unit 2 with 0 percent SG tube plugging (SGTP) and 10 percent SGTP, respectively. The LOLITT results showed that Unit 2 was more limiting than Unit 1 since it has a slightly higher design RCS vessel average temperature value and that 0 percent SGTP is more limiting for peak SG pressure since it maximizes the primary to secondary heat transfer during the event. Therefore, only the Unit 2 Case 2 with 0 percent SGTP requires evaluation to determine the bounding effects of an inoperable MSSV for the TS 3.7.1, Table 3.7.1-1 maximum allowable Power Range Neutron Flux High setpoint.

Three separate cases are performed to ensure the evaluation bounds the range of potential conditions and events related to the inoperable MSSV on Unit 2 Cycle 15 operation. The first case (Case 1) evaluates a LOLITT event with the inoperable MSSV and assuming the most positive moderator temperature coefficient (MTC) of +5pcm/F to determine the reduction in the Power Range Neutron Flux High setpoint that is required to bound any potential over power transient. The second case (Case 2) evaluates a LOLITT event with the inoperable MSSV and the near end of life (EOL) cycle conditions and the associated negative MTC that currently exist on Unit 2. This evaluation demonstrates that Unit 2 can operate at 100 percent RTP for the remainder of Cycle 15 and maintain the peak SG pressure within the applicable ASME limit for any LOL event that does not result in a high flux reactor trip. The third case (Case 3) evaluates the effect of an inadvertent MSIV closure on the single SG with the inoperable MSSV to ensure this asymmetric flo~ condition does not cause the SG pressure to exceed 110 percent of the design value. Each case was evaluated to ensure the peak SG pressure remains within 110 percent of the ASME code design value, a limit of 1210 psia.

The sequence of events for the three cases is listed in Table 3 and the peak pressure results are shown in Table 4. These cases demonstrate that Unit 2 can operate at a core power level of 102 percent with a Power Range Neutron. Flux 8

Enclosure PG&E Letter DCL-09-062 High setpoint analysis value of 112 percent and ensure that the peak SG pressure does not exceed the limit of 1210 psia.

The LOL/TT with an inoperable MSSV results are essentially the same as those presented in the overpressure case in FSAR Section 15.2.7, except for a slight change in the sequence of events due to crediting the reduced Power Range Neutron Flux High trip setpoint. This event has an insignificant impact and the overall plant response remains characteristic to the results shown in FSAR Figures 15.2.7-10 through 15.2.7-12. The inadvertent closure of an MSIV is not presented in the FSAR since it is bounded by the full power LOL FSAR analysis.

The inadvertent closure of an MSIV event has been evaluated to ensure this asymmetric condition on a single SG with an inoperable MSSV also remains bounded. The RETRAN evaluation conservatively precludes any control or protection system actions when the MSIV is closed.and the model evaluates the worst case condition associated with the reactor remaining at full power with the affected SG reaching a new steady state condition with steam relief available only through the four highest setpoint MSSVs.

Case 1 determined that a Power Range Neutron Flux High setpoint analysis value of 112 percent RTP was required to ensure the SG peak pressure remains within the 1210 psia limit for any potential over power transients. The Case 1 analysis assumes 102 percent RTP, which bounds a 2 percent uncertainty in core power level (bounds plant operation at 100 percent RTP) and a Power Range Neutron Flux High setpoint of 112 percent. The analysis value for the Power Range Neutron Flux High setpoint needs to be reduced an additional 6 percent RTP to account for instrument and channel uncertainties. This adjustment results in a required Power Range Neutron Flux High setpoint of 106 percent RTP with the inoperable MSSV MS-2-RV-224 to ensure the remaining MSSVs are capable of providing sufficient pressure relief capacity.

This will allow Unit 2 operation at the licensed core power level (i.e., 100 percent RTP) while in TS 3.7.1, Required Action A.1, for the remainder of Cycle 15 with only MS-2-RV-224 inoperable.

The determination of the 6 percent RTP uncertainty for the Power Range Neutron Flux High setpoints is contained in the current PG&E setpoint methodology document WCAP-1 1082. The setpoint methodology for DCPP is contained in WCAP-1 1082, Revision 6, "Westinghouse Setpoint Methodology for Protection Systems, Diablo Canyon Units 1 & 2, 24 Month Fuel Cycle Evaluation," (Proprietary) dated February 2003. The WCAP-1 1082, Revision 6, setpoint methodology was submitted to the NRC in PG&E Letter DCL-03-1 11, "License Amendment Request 03-12, Revision to Technical Specifications 3.3.1,

'RTS Instrumentation,' and 3.3.2, 'ESFAS Instrumentation,"' dated September 12, 2003. WCAP-1 1082, Revision 6, was approved by the NRC for DCPP by Amendment No. 178 to Facility Operating License No. DPR-80 and 9

Enclosure PG&E Letter DCL-09-062 High setpoint analysis value of 112 percent and ensure that the peak SG pressure does not exceed the limit of 1210 psia.

The LOLITT with an inoperable MSSV results are essentially the same as those presented in the overpressure case in FSAR Section 15.2.7, except for a slight change in the sequence of events due to crediting the reduced Power Range Neutron Flux High trip setpoint. This event has an insignificant impact and the overall plant response remains characteristic to the results shown in FSAR Figures 15.2.7-10 through 15.2.7-12. The inadvertent closure of an MSIV is not presented in the FSAR since it is bounded by the full power LOL FSAR analysis.

The inadvertent closure of an MSIV event has been evaluated to ensure this asymmetric condition on a single SG with an inoperable MSSV also remains bounded. The RETRAN evaluation conservatively precludes any control or protection system actions when the MSIV is closed.and the model evaluates the worst case condition associated with the reactor remaining at full power with the affected SG reaching a new steady state condition with steam relief available only through the four highest setpoint MSSVs.

Case 1 determined that a Power Range Neutron Flux High setpoint analysis value of 112 percent RTP was required to ensure the SG peak pressure remains*

within the 1210 psia limit for any potential over power transients. The Case 1 analysis assumes 102 percent RTP, which bounds a 2 percent uncertainty in core power level (bounds plant operation at 100 percent *RTP) and a Power Range Neutron Flux High setpoint of 112 percent. The analysis value for the Power Range Neutron Flux High setpoint needs to be reduced an additional 6 percent RTP to account for instrument and channel uncertainties. This adjustment results in a required Power Range Neutron Flux High setpoint of 106 percent RTP with the inoperable MSSV MS-2-RV-224 to ensure the remaining MSSVs are capable of providing sufficient pressure relief capacity.

This will allow Unit 2 operation at the licensed core power level (i.e., 100 percent RTP) while in TS 3.7.1, Required Action A.1, for the remainder of Cycle 15 with only MS-2-RV-224 inoperable.

The determination of the 6 percent RTP uncertainty for the Power Range Neutron Flux High setpoints is contained in the current PG&E setpoint methodology document WCAP-11082. The setpoint methodology for DCPP is contained in WCAP-11082, Revision 6, "Westinghouse Setpoint Methodology for Protection Systems, Diablo Canyon Units 1 & 2, 24 Month Fuel Cycle.

Evaluation," (Proprietary) dated February 2003. The WCAP-11082, Revision 6, setpoint methodology was submitted to the NRC in PG&E Letter DCL';03-111, "License Amendment Request 03-12, Revision to Technical Specifications 3.3.1,

'RTS Instrumentation,' and 3.3.2, 'ESFAS Instrumentation,'" dated

. September 12, 2003. WCAP-11082, Revision 6, was approved by the NRC for DCPP by Amendment No. 178 to Facility Operating License No. DPR.. 80 and 9

Enclosure PG&E Letter DCL-09-062 Amendment No. 180 to Facility Operating License No. DPR 82 in letter "Issuance of Amendment Re: Revised Technical Specifications 3.3.1 'Reactor Trip System (RTS) Instrumentation' and 3.3.2, 'Engineered Safety Features Actuation System (ESFAS) Instrumentation' (TAC Nos. MC0893 and M0894)," dated December 2, 2004.

Since the required Power Range Neutron Flux High setpoint with the MSSV inoperable of 106 percent RTP is higher than the current TS Table 3.7.1-1 required setpoint of 87 percent RTP, the analysis supports revision of the TS required setpoint to 106 percent for the remainder of Unit 2 Cycle 15 with only MSSV MS-2-RV-224 inoperable. The TS Bases changes reflect the revised LOL/TT analysis assumptions and results.

Assessment of Margin for Revised Safety Analyses With One MSSV Inoperable The revised RETRAN/02ModOO5.2 LOL/TT analysis for the Unit 2 inoperable MSSV establishes there is significant margin for the specific time in core life for the current Unit 2 operating conditions. Since Unit 2 Cycle 15 is near the end of cycle life, the core MTC is very negative and the LOL/TT event would result in a significant power reduction during the transient. Therefore, an evaluation of the current Unit 2 cycle 15 conditions establishes that Unit 2 maintains analysis margin to the peak SG pressure limits.

In addition, the DCPP Units 1 and 2 have not requested an increase to the core thermal power to a value above the original 3411 megawatts thermal (MWt) design rating of the nuclear steam supply system (NSSS). Millstone Unit 3, a similar 4-loop designed plant, has obtained an NRC approved increase to the licensed core power of 7 percent. Therefore, with the requested TS change, DCPP Unit 2 still maintains reasonable thermal margin to the secondary side relief capacity and associated pressure limits.

10 Enclosure PG&E Letter DCL-09-062 Amendment No. 180 to Facility Operating License No. DPR 82 in letter "Issuance of Amendment Re: Revised Technical Specifications 3.3.1 'Reactor Trip System (RTS) Instrumentation' and 3.3.2, 'Engineered Safety Features Actuation System (ESFAS) Instrumentation' (TAC Nos. MC0893 and M0894)," dated December2, 2004.

Since the required Power Range Neutron Flux High setpoint with the MSSV inoperable of 106 percent RTP is higher than the current TS Table 3.7.1-1 required setpoint of 87 percent RTP, the analysis supports revision of the TS required setpoint to 106 percent for the remainder of Unit 2 Cycle 15 with only MSSV MS-2-RV-224 inoperable. The TS Bases changes reflect the revised LOLITT analysis assumptions and results.

Assessment of Margin for Revised Safety Analyses With One MSSV Inoperable The revised RETRAN/02Mod005.2 LOLITT analysis for the Unit 2 inoperable MSSV establishes there is significant margin for the specific time in core life for the current Unit 2 operating conditions. Since Unit 2 Cycle 15 is near the end of cycle life, the core MTC is very negative and the LOLITT event would result in a significant power reduction during the transient. Therefore, an evaluation of the current Unit 2 cycle 15 conditions establishes that Unit 2 maintains analysis margin to the peak SG pressure limits.

In addition, the DCPP Units 1 and 2 have not requested an increase to the core thermal power to a value above the original 3411 megawatts thermal (MWt) design rating of the nuclear steam supply system (NSSS). Millstone Unit 3, a similar 4-loop designed plant, has obtained an NRC approved increase to the licensed core power of 7 percent. Therefore, with the requested TS change, DCPP Unit 2 still maintains reasonable thermal margin to the secondary side relief capacity and associated pressure limits.

(

10

Enclosure PG&E Letter DCL-09-062 Table 1 LOL/TT Reanalysis Assumptions RETRAN Model Parameter Assumed Value Core Power(%)

102 Bounds 2% uncertainty Core Power (MWt) 3479.2 Bounds 2% uncertainty Initial Pressurizer Pressure (psia) 2249.7 - 2189.7 Bounds 60 psi uncertainty Initial Pressurizer Level (%)

67.4 Bounds 6.3% uncertainty Initial RCS Tavg (OF) 582.6 Bounds 5 OF uncertainty Thermal Design RCS Flow (gpm) 354,000 High Neutron Flux Reactor Trip (%)

112 Bounds 6% uncertainty High Neutron Flux Reactor Trip Delay (sec) 0.5 High Pressurizer Pressure Reactor Trip (psia) 2460 Low Pressurizer Pressure Reactor Trip (psia) 1845 High/Low Pressurizer Pressure Reactor Trip Delay 2.0 (sec)

Delayed Neutron Beta Effective

.007337 Normalized Reactor Trip Reactivity (% Ak/k)

-4.0 Most Positive MTC (pcm/°F) Case 1 and Case 3

+5.0 Near EOL Negative MTC (pcm/°F) Case 2

-10.0 to -18.0 Doppler Temperature Coefficient (pcm/°F)

-0.91 Core Decay Heat ANS 1979 + 2 PSV Lift Setpoint (psia) @ 1% drift 2525.

PSV delay with loop seal purge time (sec) 1.272 Steam Dump Control System OFF AFW Sytem OFF 11 Enclosure PG&E letter DCl-09-062 Table 1 LOLITT Reanalysis Assumptions RETRAN Mod~'. Parameter Assumed Value Core Power (%)

102 Bounds 2% uncertainty Core Power (MWt) 3479.2 Bounds 2% uncertainty Initial Pressurizer Pressure (psia) 2249.7 - 2189.7 Bounds 60 psi uncertainty Initial Pressurizer level (%)

67.4 Bounds 6.3% uncertainty Initial RCS Tav9 (OF) 582.6 Bounds 5 of uncertainty Thermal Design RCS Flow (gpm) 354,000 High Neutron Flux Reactor Trip (%)

112 Bounds 6% uncertainty High Neutron Flux Reactor Trip Delay (sec) 0.5 High Pressurizer Pressure Reactor Trip (psia) 2460 low Pressurizer Pressure Reactor Trip (psia) 1845 High/low Pressurizer Pressure Reactor Trip Delay 2.0 (sec)

Delayed Neutron Beta Effective

.007337 Normalized Reactor Trip Reactivity (% ~k/k)

-4.0

\\

Most Positive MTC (pcm/oF) Case 1 and Case 3

+5.0 Near EOl Negative MTC (pcm/oF) Case 2

-10.0 to -18.0 Doppler Temperature Coefficient (pcm/oF)

-0.91

~

Core Decay Heat ANS 1979 + 2 cr PSV Lift Setpoint (psia) @ 1 % drift 2525.

PSV delay with loop seal purge time (sec) 1.272 Steam Dump Control System OFF AFWSytem OFF 11

Enclosure PG&E Letter DCL-09-062 Table 2 MSSV Reanalvsis SetDoints MSSV Tech Spec Lift Full Open with Rated flow at Lift Setpoint 5 psi full Setpoint with 3%

accumulation accumulation (psig)

Drift (psia) * (psia)*

(lb/hr) 1 1065 1111.7 1120.4 803790 2

1078 1125.2 1139.2 813471 3

1090 1138.9 1159.2 822408 4

1103 1156.2 1180.7 832090 5

1115 1174.4 1197.5 841027 Includes pressure drop effects of MSSV header as a function of MSSV flow Table 3 RETRAN Unit 2 Seauence of Event Results Case Event Sequence Case 1 Case 2 Case 3 MSSV5 MSSV5 MSIV Overpower EOL MTC Closure LOL / MSIV Closure Occurs

0.0 0.0 0.0 Pressurizer Power Operated 3.5 NA NA Relief Valves Open (sec)

MSSVs With Lowest Setpoint 9.1 9.1 8.9 Opens (sec)

Reactor Trip (sec) 15.2 20.9 NA High Flux Low Pressure Peak Power Occurs (sec) 16.4 NA 39.0 MSSVs With Highest Setpoint 18.8 15.6 18.3 Opens (sec)

Peak SG Pressure Occurs (sec) 20.6 24.6 33.0

All event times are relative to LOL/MSIV initiation, which was simulated to occur at ten seconds in order to provide an appropriate steady state evaluation period for the model.

12 Enclosure PG&E Letter DCL-09-062 Table 2 MSSV Reanalysis Setpoints MSSV Tech Spec Lift Full Open with Rated flow at Lift Setpoint 5 psi full Setpoint with 3%

accumulation accumulation (psig)

Drift (psia) *

(psia)*

(Ib/hr) 1 1065 1111.7 1120.4 803790 2

1078 1125.2 1139.2 813471 3

1090 1138.9 1159.2 822408 4

1103 1156.2 1180.7 832090 5

1115 1174.4 1197.5 841027

Includes pressure drop effects of MSSV header as a function of MSSV flow Table 3 RETRAN Unit 2 Sequence of Event Results Case Event Sequence Case 1 Case 2 Case 3 MSSV5 MSSV5 MSIV Overpower EOl MTC Closure LOL / MSIV Closure Occurs

0.0 0.0 0.0 Pressurizer Power Operated 3.5 NA NA Relief Valves Open (sec)

MSSVs With Lowest Setpoint

.9.1 9.1 8.9 Opens (sec)

Reactor Trip (sec) 15.2 20.9 NA High Flux Low Pressure Peak Power Occurs (sec) 16.4 NA 39.0 MSSVs With Highest Setpoint 18.8 15.6 18.3 Opens (sec)

Peak SG Pressure Occurs (sec) 20.6 24.6 33.0

All event times are relative to LOLIMSIV initiation, which was simulated to occur at ten seconds in order to provide an appropriate steady state evaluation period for the model.

12

Enclosure PG&E Letter DCL-09-062 Table 4 RETRAN Peak SG Pressure Results Case Case 1 Case 2 Case 3 MSSV5 MSSV5 MSIV Overpower EOL MTC Closure Peak Core Power at 115.6 76.4 101.4 Reactor Trip (%)

Peak SG Pressure (psia) 1204.2 1186.5 1198.3 Evaluation of Other Analyses In addition to the LOL/TT analysis, the impact of 100 percent RTP operation with one MSSV inoperable for the remainder of the current Unit 2 Cycle 15 has been evaluated for the following analyses:

Large break loss-of-coolant accident (LOCA)

Small break LOCA Post LOCA Non-LOCA Containment integrity Mass and energy releases SG tube rupture Margin to trip analyses Instrument uncertainties The evaluation determined that the conclusions of these current licensing basis analyses remain valid for operation of DCPP Unit 2 for the remainder of Cycle 15 with MSSV MS-2-RV-224 inoperable at 100 percent RTP while maintaining the RCS vessel average temperature between 565'F and 577.6°F.

Impact on AFW System The inoperable MSSV MS-2-RV-224 is the highest lift setpoint valve in the loop and there is no impact on AFW system design flow requirements since they are based on the MSSV with the lowest lift setpoint and are not impacted by the inoperable MSSV MS-2-RV-224.

Conclusion A revised RETRAN/02ModOO5.2 analysis has been performed and determined the required Power Range Neutron Flux High setpoint for remainder of Unit 2 13 Enclosure PG&E Letter DCL-09-062 Table 4 RETRAN Peak SG Pressure Results Case Case 1 Case 2 Case 3 MSSV5 MSSV5 MSIV Overpower EOl MTC Closure Peak Core Power at 115.6 76.4 101.4 Reactor Trip (%)

Peak SG Pressure (psia) 1204.2 1186.5 1198.3 Evaluation of Other Analyses In addition t9 the LOLITT analysis, the impact of 100 percent RTP operation with one MSSV inoperable for the remainder of the current Unit 2 Cycle 15 has been evaluated for the following analyses:

Large break loss-of-coolant accident (LOCA)

Small break LOCA Post LOCA Non-LOCA Containment integrity Mass and energy releases SG tube rupture Margin to trip analyses Instrument uncertainties The evaluation determined that the conclusions of these current licensing basis analyses remain valid for operation of DCPP Unit 2 for the remainder of Cycle 15 with MSSV MS-2-RV-224 inoperable at 100 percent RTP* while maintaining the RCS vessel average temperature between 565°F and 577.6°F.

Impact on AFW System The inoperable MSSV MS-2-RV-224 is the highest lift setpoint valve in the loop and there is no impact on AFW system design flow requirements since they are based on the MSSV with the lowest lift setpoint and are not impacted by the inoperable MSSV MS-2-RV-224.

Conclusion A revised RETRAN/02Mod005.2 analysis has been performed and determined the required Power Range Neutron Flux High setpoint for remainder of Unit 2

/

13

Enclosure PG&E Letter DCL-09-062 Cycle 15 with the MSSV MS-2-RV-224 inoperable. The analysis bounds plant operation at 100 percent RTP, considering a 2 percent RTP uncertainty, with a reduced Power Range Neutron Flux High setpoint analysis value of 112 percent RTP. With a reduction of 6 percent RTP to account for instrument and channel uncertainties, it has been determined a Power Range Neutron Flux High setpoint of 106 percent is acceptable for Unit 2 Cycle 15 operation at the licensed core power level while in TS 3.7.1, Required Action A.1, with only MSSV MS-2-RV-224 inoperable. The results of the analysis showed that the peak pressures in the SGs are lower than 110 percent of the 1085 psig SG design pressure with the one MSSV inoperable and the remaining MSSVs are capable of providing sufficient pressure relief capacity. The analysis supports a TS Table 3.7.1-1 maximum allowable Power Range Neutron Flux High setpoint of 106 percent RTP for only MSSV MS-2-RV-224 inoperable for the remainder of Unit 2 Cycle 15. With the proposed TS change, the remaining Unit 2 MSSVs are capable of providing sufficient pressure relief capacity.

In summary, the health and safety of the public will not be adversely affected.

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria

1.

ASME Boiler and Pressure Vessel Code,Section III, 1968.

2.

Title 10 to Code of Federal Regulations Part 50.36.

The ASME Boiler and Pressure Vessel Code,Section III, provides the design basis for the MSSVs and limits the secondary system pressure to less than or equal to 110 percent of design pressure for any anticipated operational occurrence or accident considered in the Design Basis Accident and transient analysis. The MSSVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

With the TS changes requested in this LAR, the Unit 2 MSSVs will continue to limit the secondary system pressure to less than or equal to 110 percent of design pressure for any AOO or accident considered in the Design Basis Accident and Transient analysis and the MSSVs continue to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

14 Enclosure PG&E Letter DCL-09-062

)

Cycle 15 with the MSSV MS-2-RV-224 inoperable. The analysis bounds plant operation at 100 percent RTP, considering a 2 perqent RTP uncertainty, with a reduced Power Range Neutron Flux High setpoint analysis value of 112 percent RTP. With a reduction of 6 percent RTP to account for instrument and channel uncertainties, it has been determined a Power Range Neutron Flux High setpoint of 106 percent is acceptable for Unit 2 Cycle 15 operation at the licensed core power level while in TS 3.7.1, Required Action A.1, with only MSSV MS-2-RV-224 inoperable. The results of the analysis showed that the peak pressures in the SGs are lower than 110 percent of the 1085 psig SG design pressure with the one MSSV inoperable and the remaining MSSVs are capable of providing sufficient pressure relief capacity. The analysis supports a TS Table 3.7.1-1 maximum allowable Power Range Neutron Flux High setpoint of 106 percent RTP for only MSSV MS-2-RV-224 inoperable for the remainder of Unit 2 Cycle 15. With the proposed TS change, the remaining Unit 2 MSSVs are capable of providing sufficient pressure relief capacity.

, In summary, the health and safety of the public will not be adversely affected.

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria

1.

ASME Boiler and Pressure Vessel Code,Section III, 1968.

2.

Title 10 to Code of Federal Regulations Part 50.36.

The ASME Boiler and Pressure Vessel Code,Section III, provides the design basis for the MSSVs and limits the secondary system pressure to less than or equal to 110 percent of design pressure for any anticipated operational occurrence or accident considered in the Design Basis Accident and transient analysis. The MSSVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

With the TS changes requested in this LAR, the Unit 2 MSSVs will continue to limit the secondary system pressure to less than or equal to 110 percent of design pressure for any AOO or accident considered in the Design Basis Accident and Transient analysis and the MSSVs continue to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

14

Enclosure PG&E Letter DCL-09-062 4.2 Precedent Similar changes to the TS 3.7.1 maximum allowable Power Range Neutron Flux High setpoint versus minimum number of MSSVs per SG Operable have been previously approved for DCPP in Reference 3.

4.3 No Significant Hazards Consideration PG&E 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

This License Amendment Request (LAR) proposes a one-time change to Technical Specification (TS) 3.7.1, "Main Steam Safety Valves (MSSVs),"

Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs" to allow an increase in the Power Range Neutron Flux High setpoint from 87 percent rated thermal power (RTP) to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 main steam relief valve 224 (MS-2-RV-224) inoperable.

The increase in the Power Range Neutron Flux High setpoint TS value does not initiate an accident. Technician adjustments to lower the Power Range Neutron Flux High setpoint could cause a reactor trip, however, this action is already a TS requirement. Thus, increasing the TS setpoint value from the current value will not change the requirement for a technician to adjust the setpoints downward when MSSVs become inoperable, and therefore, will not increase the probability of a reactor trip.

With the increase in the Power Range Neutron Flux High setpoint with only MS-2-RV-224 inoperable during Unit 2 Cycle 15 the remaining MSSVs will continue to prevent overpressure of the main steam leads and Steam Generators (SGs), and remove adequate heat from the reactor coolant system (RCS).

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

15 4.2 Precedent Enclosure PG&E Letter DCL-09-062 Similar changes to the TS 3.7.1 maximum allowable Power Range.

Neutron Flux High setpoint versus minimum number of MSSVs perSG Operable have been previously approved for DCPP in Reference 3.

4.3 No Significant Hazards Consideration PG&E 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

This License Amendment Request (LAR) proposes a one-time change to Technical Specification (TS) 3.i.1, "Main Steam SafetyValves (MSSVs),"

Table 3.7.1-1, "Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs" to allow an increase in the Power Range Neutron Flux High setpoint from 87 percent rated thermal power (RTP) to 106 percent RTP for Unit 2 Cycle 15 with only Unit 2 main steam relief valve 224 (MS-2-RV-224) inoperable.

The increase in the Power Range Neutron Flux High setpoint TS value does not initiate an accident. Technician adjustments to lower the Power Range Neutron Flux High setpoint could cause a reactor trip, however; this action is already a TS requirement. Thus, increasing the TS setpoint value from the current value will not change the requirement for a technician to adjust the setpoints downward when MSSVs become inoperable, and therefore, will not increase the probability of a reactor trip.

With the increase in the Power Range Neutron Flux High setpoint with only MS-2-RV-224 inoperable during Unit 2 Cycle 15 the remaining MSSVs will continue to prevent overpressure of the main steam leads and Steam Generators (SGs), and remove adequate heat from the reactor coolant system (RCS).

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

15

Enclosure PG&E Letter DCL-09-062

2.

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

The increase in the Power Range Neutron Flux High setpoint TS value with only MS-2-RV-224 inoperable during Unit 2 Cycle 15 does not initiate an accident and does not change the method by which any safety-related system performs the function.

The proposed change does not result in plant operation outside the limits previously considered, nor allow the progression of transients or accidents in a manner different than previously considered.

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

3.

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

The RCS pressure boundary is applicable to the proposed change. With the proposed change all relevant event acceptance criteria were found to be satisfied. Therefore, the proposed change does not involve a reduction in a margin of safety.

With the proposed change, the MSSVs will still prevent SG pressure from exceeding 110 percent of SG design pressure in accordance with the ASME code. The conclusions for the Final Safety Analysis Report accident analyses are unaffected by the change, remain valid, and provide margin.

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

Based on the above safety evaluation, PG&E concludes that the change proposed by this LAR satisfies the no significant hazards consideration standards of 10 CFR 50.92(c), and accordingly a no significant hazards finding is justified.

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 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.

16 Enclosure PG&E Letter DCL-09-062

2.

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

The increase in the Power Range Neutron Flux High setpoint TS value with only MS-2-RV-224 inoperable during Unit 2 Cycle 15 does not initiate an accident and does not change the method by which any safety-related system performs the function.

The proposed change does not result in plant operation outside the limits previously considered, nor allow the progression of transients or accidents in a manner different than previously considered.

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

3.

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

The RCS pressure boundary is applicable to the proposed change. With the proposed change all relevant event acceptance criteria were found to be satisfied. Therefore, the proposed change does not involve a reduction in a margin of safety.

With the proposed change, the MSSVs will still prevent SG pressure from exceeding 110 percent of SG design pressure in accordance with the ASME code. The conclusions for the Final Safety Analysis Report accident analyses are unaffected by the change, remain valid, and provide margin.

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

Based on the above safety evaluation, PG&E concludes that the change proposed by this LAR satisfies the no significant hazards consideration standards of 10 CFR 50.92(c), and accordingly a no significant hazards finding is justified.

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 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.

16

Enclosure PG&E Letter DCL-09-062

5.

ENVIRONMENTAL CONSIDERATION PG&E has evaluated the proposed amendment and has determined that the proposed amendment does not involve: (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in individual or cumulative 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 REFERENCES

1.

PG&E Letter DCL-07-002, "License Amendment Request 07-01, Revision to Technical Specifications to Support Steam Generator Replacement," dated January 11,2007.

2.

Westinghouse Nuclear Safety Advisory Letter (NSAL)94-001, Operation at Reduced Power Levels with Inoperable MSSVs, January 20, 1994.

3.

NRC Information Notice 94-60, Potential Overpressurization of Main Steam System, August 22, 1994.

4.

PGE Letter DCL-97-105, "License Amendment Request 97-06, Revision of Technical Specification 3.7.1.1, Table 3.7-1 and Associated Bases-Reduced Power Operation Levels for Inoperable Main Steam Safety Valves (MSSVs)," dated December 23, 1997.

5.

License Amendment No. 125 to Facility Operating License No. DPR-80 and License Amendment No. 123 to Facility Operating License No. DPR-82, "Issuance of Amendments for Diablo Canyon Power Plant, Unit No. 1 (TAC No. MA0397) and Unit No. 2 (TAC No. MA0398)," dated May 28, 1998.

6.

PG&E 'Letter DCL-02-115, "License Amendment Request (LAR) 01-08, Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves; Pressurizer Safety Valve Loop Seal Temperature," dated September 24, 2002.

7.

License Amendment No. 171 to Facility Operating License No. DPR-80 and License Amendment No. 172 to Facility Operating License No. DPR-82, "Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Issuance of Amendment RE: Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves (TAC Nos. MB6758 MB6759)," dated July 2, 2004.

8.

PG&E Letter DCL-95-141, "License Amendment Request 95-04, Revision of Technical Specification 3/4.4.2.2 - Technical Specifications Administrative Changes," dated June 29, 1995.

9.

License Amendment No. 107 to Facility Operating License No. DPR-80 and License Amendment No. 106 to Facility Operating License No. DPR-82, 17 Enclosure PG&E Letter DCL-09-062

5.

ENVIRONMENTAL CONSIDERATION J

PG&E has evaluated the proposed amendment and has determined that the proposed amendment does not involve: (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in individual or cumulative 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 REFERENCES

1.

PG&E Letter DCL-07-002, "License Amendment Request 07-01, Revision to Technical Specifications to Support Steam Generator Replacement," dated January 11, 2007.

2.

Westinghouse Nuclear Saf~ty Advisory Letter (NSAL)94-001, Operation at Reduced Power Levels with Inoperable MSSVs, January20, 1994.

3.

NRC Information Notice 94-60, Potential Overpressurization of Main Steam System, August 22, 1994.

4.

PGE Letter DCL-97-105, "License Amendment Request 97-06, Revision of Technical Specification 3.7.1.1, Table 3.7-1 and Associated Bases-Reduced Power Operation Levels for Inoperable Main Steam Safety Valves (MSSVs)," dated December 23, 1997.

5.

License Amendment No. 125 to Facility Operating License No. DPR-80 and License Amendment No. 123 to Facility Operating License No. DPR-82, "Issuance of Amendments for Diablo Canyon Power Plant, Unit No.1 (TAC No. MA0397) and Unit No.2 (TAC No. MA0398)," dated May 28,1998.

6.

PG&E Letter DCL-02-115, "License Amendment Request (LAR) 01-08, Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves;,

Pressurizer Sa:fety Valve Loop Seal Temperature," dated September 24, 2002.

7.

License Amendment No. 171 to Facility Operating License No. DPR-80 and License Amendment No. 172 to Facility Operating License No. DPR-82, "Diablo Canyon Power Plant, Unit Nos. 1 and 2 -Issuance of Amendment RE: Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves (TAC Nos. MB6758 MB6759)," dated July 2, 2004.

8.

PG&E Letter DCL-95-141, "License Amendment Request 95-04, Revision of Technical Specification 3/4.4.2.2 - Technical* Specifications Administrative Changes," dated June 29, 1995.

9.

License Amendment No.1 07 to Facility Operating License No. DPR-80 and License Amendment No.1 06 to Facility Operating License No. DPR-82, 17

Enclosure PG&E Letter DCL-09-062 "Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No.

1 (TAC No. M92767) and No. 2 (TAC No. M92768)," dated August 23, 1995.

10. WCAP-1 1082, Revision 6, "Westinghouse Setpoint Methodology for Protection Systems, Diablo Canyon Units 1 & 2, 24 Month Fuel Cycle Evaluation," (Proprietary) dated February 2003.

11, PG&E Letter DCL-03-1 11, "License Amendment Request 03-12, Revision to Technical Specifications 3.3.1, 'RTS Instrumentation,' and 3.3.2, 'ESFAS Instrumentation,"' dated September 12, 2003.

12, License Amendment No. 178 to Facility Operating License No. DPR-80 and Amendment No. 180 to Facility Operating License No. DPR 82, "Issuance of Amendment Re: Revised Technical Specifications 3.3.1 'Reactor Trip System (RTS) Instrumentation' and 3.3.2, 'Engineered Safety Features Actuation System (ESFAS) Instrumentation' (TAC Nos. MC0893 and MC0894)," dated December 2, 2004.

18 Enclosure PG&E Letter DCL-09-062 "Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No.

1 (TAC No. M92767) and No.2 (TAC No. M92768)," dated August 23, 1995.

10. WCAP-11082, Revision 6, "Westinghouse Setpoint Methodology for Protection Systems, Diablo Canyon Units 1 & 2,24 Month Fuel Cycle Evaluation," (Proprietary) dated February 2003.

11. PG&E Letter DCL-03-111, "License Amendment Request 03-12, Revision to Technical Specifications 3.3.1, 'RTS Instrumentation,' and 3.3.2, 'ESFAS Instrumentation,'" dated September 12, 2003.

12.

License Amendment No. 178 to Facility Operating License No. DPR-80 and Amendment No. 180 to Facility Operating License No. DPR 82, "Issuance of Amendment Re: Revised Technical Specifications 3.3.1 'Reactor Trip System (RTS) Instrumentation' and 3.3.2, 'Engineered Safety Features Actuation System (ESFAS) Instrumentation' (TAC Nos. MC0893 and MC0894)," dated December 2,2004.

18

Enclosure PG&E Letter DCL-09-062 Proposed Technical Specification Changes (marked-up)

Page: 3.7-2 Enclosure PG&E Letter DCL-09-062 Proposed Technical Specification Changes (marked-up)

Page: 3.7-2

'/

MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs MINIMUM NUMBER OF MSSVs PER STEAM MAXIMUM ALLOWABLE POWER RANGE GENERATOR REQUIRED OPERABLE NEUTRON FLUX HIGH SETPOINT

%RTP 4

87*

3 47*

2 29*

Unless the reactor trip system breakers are in the open position.

- A Maximum Allowable Power Range Neutron Flux High Setpoint of 106% RTP may be used for Unit 2 Cycle 15 with only MS-2-RV-224 inoperable.

I A-4-I DIABLO CANYON - UNITS 1 & 2 3.7-2 Unit 1 - Amendment No. 435 142 Unit 2 - Amendment No. 4-5 +4 Table 3.7.1-1 (page 1 of 1)

MSSVs 3.7.1 Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs MINIMUM NUMBER OF MSSVs PER STEAM MAXIMUM ALLOWABLE POWER RANGE GENERATOR REQUIRED OPERABLE NEUTRON FLUX HIGH SETPOINT 4

3 2

%RTP 87*

47*

29*

Unless the reactor trip system breakers are in the open position.

- A Maximum Allowable Power Range Neutron Flux High Setpoint of 106% RTP may be used for Unit 2 Cycle 15 with only MS-2-RV-224 inoperable.

DIABLO CANYON - UNITS 1 & 2 3.7-2

/

Unit 1 - Amendment No. +Je 142 Unit 2 - Amendment No. +Je t4Z-I t

+

4-I

Enclosure PG&E Letter DCL-09-062 Proposed Technical Specification Changes (retyped)

Remove Page Insert Page 3.7-2 3.7-2 Enclosure PG&E Letter DCL-09-062

. Proposed Technical Specification Changes (re~yped)

Remove Page Insert Page 3.7-2 3.7-2

/

MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs MINIMUM NUMBER OF MSSVs PER STEAM MAXIMUM ALLOWABLE POWER RANGE GENERATOR REQUIRED OPERABLE NEUTRON FLUX HIGH SETPOINT

%RTP 4

87* **

3 47*

2 29*

Unless the reactor trip system breakers are in the open position.

- A Maximum Allowable Power Range Neutron Flux High Setpoint of 106% RTP may be used for Unit 2 Cycle 15 with only MS-2-RV-224 inoperable.

DIABLO CANYON - UNITS 1 & 2 3.7-2 Unit 1 - Amendment No. 4-3, 142 Unit 2 - Amendment No. 4-35,-1-42, Table 3.7.1-1 (page 1 of 1)

MSSVs*

3.7.1 Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs MINIMUM NUMBER OF MSSVs PER STEAM MAXIMUM ALLOWABLE POWER RANGE GENERATOR REQUIRED OPERABLE NEUTRON FLUX HIGH SETPOINT 4

3 2

%RTP 87* **

47*

29*

Unless the reactor trip system breakers are in the open position.

- A Maximum Allowable Power Range Neutron Flux High Setpoint of 106%. RTP may be used for Unit 2 Cycle 15 with only MS-2-RV-224 inoperable.

DIABLO CANYON - UNITS 1 & 2 I

3.7-2 Unit 1 - Amendment No. ~, 142 Unit 2 - Amendment No. ~, -'/-4Z,

Enclosure PG&E Letter DCL-09-062 Changes to Technical Specification Bases Pages (For information only)

MSSVs B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs)

BASES BACKGROUND The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurizing the reactor coolant pressure boundary (RCPB) by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the preferred heat sink, provided by the Condenser and Circulating Water System, is not available.

Five MSSVs are located on each main steam header, outside containment, upstream of the main steam isolation valves, as described in the FSAR, Section 10.3.1 (Ref. 1). The MSSVs must have sufficient capacity to limit the secondary system pressure to

< 110% of the steam generator design pressure. The MSSV design includes staggered setpoints, according to Table 3.7.1-2 in the accompanying LCO, so that only the needed valves will actuate.

Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves during an overpressure event.

APPLICABLE SAFETY ANALYSES The design basis for the MSSVs comes from Reference 2 and its purpose is to limit the secondary system pressure to < 110% of design pressure for any anticipated operational occurrence (AOO) or accident considered in the Design Basis Accident (DBA) and transient analysis.

The events that challenge the relieving capacity of the MSSVs, and thus RCS pressure, are those characterized as decreased heat removal events, which are presented in the FSAR, Section 15.2 and 15.3 (Ref. 3). Of these, the full power turbine trip without steam dump is the limiting AOO with respect to secondary system pressure. This event also terminates normal feedwater flow to the steam generators.

The safety analysis demonstrates that the transient response for turbine trip occurring from full power without a direct reactor trip presents no hazard to. the integrity of the RCS or the Main Steam System.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 5 1

MSSVs B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs)

BASES BACKGROUND APPLICABLE SAFETY ANALYSES The primary purpose of the MSSVsis to provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurizing the reactor coolant pressure boundary (RCPB) by providing a heat sink for the removal of energy from the Reactor Coolant System (RCS) if the preferred heat sink, provided by the Condenser and Circulating Water System, is not available.

Five MSSVs are located on each main steam header, outside containment, upstream of the main steam isolation valves, as described in the FSAR, Section 10.3.1 (Ref. 1). The MSSVs must have sufficient capacity to limit the secondary system pressure to

110% of the steam generator design pressure. The MSSV design includes staggered setpoints, according to Table 3.7.1-2 in the accompanying LCO, so that only the needed valves will actuate.

Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves during an overpressure event.

. The design basis for the MSSVs comes from Reference 2 and its purpose is to limit the secondary system pressure to :::; 110% of design pressure for any anticipated operational occurrence (AOO) or accident considered in the Design Basis Accident (DBA) and transient analysis.

The events that challenge the relieving capacity of the MSSVs, and thus RCS pressure, are those characterized as decreased heat removal events, which are presented in the FSAR, Section 15.2 and 15.3 (Ref. 3). Of these, the full power turbine trip withbut steam dump is the limiting AOOwith respect to secondary system pressure. This event also terminates normal feedwater flow to the steam generators.

The safety analysis demonstrates that the transient response for turbine trip occurring from full power without a direct reactor trip presents no hazard to the integrity of the RCS or the Main Steam System.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 5 1

MSSVs B 3.7.1 BASES APPLICABLE SAFETY ANALYSES (continued)

One turbine trip analysis is performed assuming primary system pressure control via operation of the pressurizer relief valves and sprays. The analysis demonstrates that the DNB design basis is met.

Another analysis is performed assuming no primary system pressure control, but crediting reactor trip on high pressurizer pressure and operation of the pressurizer safety valves. This analysis demonstrates that the maximum RCS pressure does not exceed 110% of the design pressure. All cases analyzed demonstrate that the MSSVs maintain Main Steam System integrity by limiting the maximum steam pressure to less than 110% of the steam generator design pressure.

The MSSVs are assumed to have two active and one passive failure modes. The active failure modes are spurious opening, and failure to reclose once opened. The passive failure mode is failure to open upon demand.

The MSSVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The accident analysis requires that five MSSVs per steam generator be OPERABLE to provide overpressure protection for design basis transients occurring at 102% RTP. The LCO requires that five MSSVs per steam generator be OPERABLE in compliance with Reference 2.

The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances, to relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is verified by periodic surveillance testing in accordance with the Inservice Testing Program.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB, or Main Steam System integrity.

APPLICABILITY In MODES 1, 2, and 3, five MSSVs per steam generator are required to be OPERABLE to limit secondary pressure.

In MODES 4 and 5, there are no credible transients requiring the MSSVs. The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

.(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 5

.2 BASES APPLICABLE SAFETY ANALYSES

( continued)

LCO APPLICABILITY r

MSSVs B 3.7.1 One turbine trip analysis is performed assuming primary system pressure control via operation of the pressurizer relief valves and sprays. The analysis demonstrates that the DNB design basis is met.

Another analysis is performed assuming no primary system pressure control, but crediting reactor trip on high pressurizer pressure and operation of the pressurizer safety valves. This analysis demonstrates that the maximum RCS pressure does not exceed 110% of the design pressure. All cases analyzed demonstrate that the MSSVs maintain Main Steam System integrity by limiting the maximum steam pressure to less than 110% of the steam generator design pressure.

The MSSVs are assumed to have two active and one passive failure modes. The active failure modes are spurious opening, and failure to reciose once opened. The passive failure mode is failure to open upon demand.

The MSSVs satisfy Criterion 3 of 1 0 CFR 50.36(c)(2)(ii).

The accident analysis requires that five MSSVs per steam generator be' OPERABLE to provide overpressure protection for design basis transients occurring at 102% RTP. The LCO requires that five MSSVs per steam generator be OPERABLE in compliance with Reference 2.

The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances, to relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is verified by periodic surveillance testing in accordance with the Inservice Testing Program.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB, or Main Steam System integrity.

In MODES 1, 2, and 3, five MSSVs per steam generator are required to be OPERABLE to limit secondary pressure..

In MODES 4 and 5, there are no credible transients requiring the MSSVs. The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

.( continued)

DIABLO CANYON - UNITS 1 & 2 Revision 5

.2

MSSVs B 3.7.1 BASES (continued)

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

A.1 With one or more MSSVs inoperable, action must be taken so that the available MSSV relieving capacity meets Reference 2 requirements.

Continued operation with less than all five MSSVs OPERABLE for each steam generator is permissible, if THERMAL POWER is limited to the relief capacity of the remaining MSSVs. This is accomplished by restricting THERMAL POWER and the Power Range Neutron Flux trip setpoint so that the energy transfer to the most limiting steam generator is not greater than the available relief capacity in that steam generator.

The Reactor Trip Setpoint reductions applied in TS Table 3.7.1-1 are derived on the following bases:

One MSSV Inoperable The limiting FSAR Condition II accident for overpressure concerns is a loss of external load/turbine trip. The event is analyzed with the RETRAN-02 computer program to demonstrate the adequacy of the MSSVs to maintain the main steam system lower than 1210 psia, or 110% of the 1085 psig SG design pressure.

In a PG&E calculation, the transient is reanalyzed to determine the effect of only four MSSVs per SG being available. The analysis assumes a 3% tolerance for all the available MSSVs. The MSSV on each SG with the lowest nominal setpoint was assumed unavailable, and the Unit 2 model is used because of its higher design RCS average temperaturethemral rat~ig. The results of the calculation show that the peak pressures in the SGs are lower than 1210 psia, or 110% of the 1085 psig SG design pressure (Ref. 8).

Thus, with one MSSV inoperable per SG, the remaining MSSVs are capable of providing sufficient pressure relief capacity for the plant to' operate at 100% RATED THERMAL POWER (RTP). However, the value applied to the high neutron flux trip setpoints must be lowered an additional 6% RTP to account for instrument and channel uncertainties (Ref. 7). This adjustment results in a setpoint of 94% RTP; however, the setpoint will remain at 87% RTP for additional conservatism.

For Unit 2 Cycle 15 with only MS-2-RV-224 inoperable, the required high neutron flux trip setpoint is 106% RTP to ensure the remaining MSSVs are capable of providing sufficient pressure relief capacity for plant operation at 100% RTP (Ref. 10).

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 5 3

BASES (continued)

ACTIONS MSSVs B 3.7.1 The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

A.1 With one or more MSSVs inoperable, action must be taken so that the available MSSV relieving capacity meets Reference 2 requirements.

Continued operation with less than all five MSSVs OPERABLE for each steam generator is permissible, if THERMAL POWER is limited to the relief capacity of the remaining MSSVs. This is accomplished by restricting THERMAL POWER and the Power Range Neutron Flux trip setpoint so that the energy transfer to the most limiting steam generator is not greater than the available relief capacity in that steam generator.

The Reactor Trip Setpoint reductions applied in TS Table 3.7.1-1 are derived on the following.bases:

One MSSV Inoperable The limiting FSAR Condition II accident for overpressure concerns is a loss of external load/turbine trip. The event is analyzed with the RETRAN-02 computer program to demonstrate the adequacy of the MSSVs to maintain the main steam system lower than 1210 psia, or 110% of the 1085 psig SG design pressure.

In a PG&E calculation, the transient is reanalyzed to determine the effect of only four MSSVs per SG being available. The analysis assumes a 3% tolerance for all the available MSSVs. The MSSV on each SG with the lowest nominal setpoint was assumed unavailable, and the Unit 2 model is used because of its higher design RCS average temperaturethermal rating. The results of the calculation show that the peak pressures in the SGs are lower than 1210 psia, or 110% of the 1085 psig SG design pressure (Ref. 8).

Thus, with one MSSV inoperable per SG, the remaining MSSVs are capable of providing sufficient pressure relief capacity for the plant to I operate at 1 00% RATED THERMAL POWER (RTP). However, the valu~applied to the high neutron flux trip setpoints must be lowered an additional 6% RTP to account for instrument and channel uncertainties (Ref. 7). This adjustment results in a setpoint of 94% RTP; however, the setpoint will remain at 87% RTP for additional conservatism..

For Unit 2 Cycle 15 with only MS*2*RV*224 inoperable, the required high neutron flux trip setpoint is 106% RTP to ensure the remaining MSSVs are capable of providing sufficient pressure relief capacity for plant operation at 100% RTP (Ref. 10).

(continued)

DIABLO CANYON - UNITS 1 & 2 11 Revision 5 3

MSSVs B 3.7.1 BASES ACTIONS A.1 (continued)

More than One MSSV Inoperable For more than one MSSV on each loop inoperable, the following Westinghouse algorithm contained in NSAL 94-001 (Ref. 4) is used:

(wghfgN)

Hi 4

=

(100/Q)--------

K where:

Hi 4

=

Safety Analysis PR high neutron flux setpoint, percent Q,

=

Nominal NSSS power rating of the plant (including reactor coolant pump heat), MWt K

=

Conversion factor, 947.82 (Btu/sec)/MWt W=

Minimum total steam flow rate capability of the operable MSSVs on any one SG at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in lb/sec. For example, if the maximum number of inoperable MSSVs per SG is three, then ws should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the three highest capacity MSSVs.

hfg

=

heat of vaporization for steam at the highest MSSV opening pressure including tolerance and" accumulation, as appropriate, Btu/Ibm N

=

Number of loops in plant For the case of two and three inoperable MSSVs per SG, the setpoints derived are 53% and 35% RTP, respectively. However, the values applied to the high neutron flux trip setpoints must be lowered an additional 6% RTP to account for instrument and channel uncertainties (Ref. 7), which results in setpoints of 47% and 29% RTP, respectively (Ref. 9).

When a MSSV(s) is inoperable, the power must be reduced in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to a value less than or equal to the value specified in table 3.7.1-1, corresponding to the number of OPERABLE MSSVs.

The Power Range Neutron Flux-high trip setpoint must also be reduced in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , to less than or equal to the value specified in Table 3.7.1-1, corresponding to the number of OPERABLE MSSVs.

(continued)

DIABLO CANYON -'UNITS 1 & 2 Revision 5 4

BASES ACTIONS

./

A.1 (continued)

MSSVs B 3.7.1 More than One MSSV Inoperable For more than one MSSV on each loop inoperable, the following Westinghouse algorithm contained in NSAL 94-001 (Ref. 4) is used:

(wghfgN)

Hi ~

=

(100IQ)----------------

where:

Hi~

=

Q

=

K

=

Ws

=

K Safety Analysis PR high neutron flux setpoint, percent Nominal NSSS power rating of the plant (including reactor coolant pump heat), MWt Conversion factor, 947.82 (Btu/sec)/MWt Minimum total steam flow rate capability of the operable MSSVs on anyone SG at the highest MSSV opening pressure including tolerance and accumulation, as app'fopriate, in Ib/sec. For example, if the maximum number of inoperable MSSVs per SG is three, then Ws should be a summation of the capacity of the operable MSSVs at the highest operable MSSVoperating pressure, excluding the three highest capacity MSSVs.

heat of vaporization for steam at the highest MSSVopening pressure including tolerance and accumulation, as appropriate, Btu/lbm N

=

Number of loops in plant For the case of two and three inoperable MSSVs per SG, the setpoints derived are 53% and 35% RTP, respectively_ However, the values applied to the high neutron flux trip setpoints must be lowered an additional 6% RTP to account for instrument and channel uncertainties (Ref. 7), which results in setpoints of 47% and 29% RTP, respectively (Ref. 9). -

When a MSSV(s) is inoperable, the power must be reduced in 4' hours to a value less than or equal to the value specified in table 3.7.1-1, corresponding to the number of OPERABLE MSSVs.

The Power Range Neutron Flux-high trip setpoint must also be reduced in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , to less than or equal to the value specified in Table 3.7.1-1, corresponding to the number of OPERABLE MSSVs;

( continued)

DIABLO CANYON -'UNITS 1 & 2 Revision 5 4

MSSVs B 3.7.1 BASES ACTIONS A.1 (continued)

The allowed Completion Time is reasonable base on operating experience to complete the Required Actions in an orderly manner without challenging unit systems.

B.1 and B.2 If THERMAL POWER and Power Range Neutron Flux Trip are not reduced as required by Table 3.7.1-1 within the associated Completion Time, or if one or more steam generators have less than two MSSVs OPERABLE, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must beý placed in at.

least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with the Inservice Testing Program. The ASME Code,Section XI (Ref. 5), requires that safety and relief valve tests be performed in accordance with ASME OM Code Appendix I (Ref. 6). According to Reference 6, the following tests are required:

a.

Visual examination;

b.

Seat tightness determination;

c.

Setpoint pressure determination (lift setting);

d.

Compliance with owner's seat tightness criteria; and

e.

Verification of the balancing device integrity on balanced valves.

The ASME OM Code requires that all valves be tested every 5 years, and a minimum of 20% of the valves be tested every 24 months. The ASME Code specifies the activities and frequencies necessary to satisfy the requirements.-Table 3.7.1-2 allows a +/- 3% setpoint' (as-found lift point) tolerance on the valves for OPERABILITY (with the exception of the lowest set MSSV setpoint, which is (+3%/-2%);

however, the valves are reset to + 1 %during the Surveillance to allow for drift. The lift settings, according to Table. 3.7.1-2 in the accompanying LCO, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 5 5

BASES ACTIONS SURVEILLANCE REQUIREMENTS J

A.1 (continued)

MSSVs B 3.7.1 The allowed Completion Time is reasonable base on operating experience to complete the Required Actions in an orderly manner without challenging unit systems.

B.1 and'B.2 If THERMAL POWER and Power Range Neutron Flux Trip are not' reduced as required by Table 3.7.1-1 within the associated Completion

,, Time, or if one or more st~am generators have less than two MSSVs

, OPERABLE, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be' placed in at, least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and, in MODE 4.within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are re'asonable, based on operating experience, to reach the-required unit conditions from full power conditions inan orderly manner and without challenging unit systems.

SR 3.7.1.1 This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with the Inservice Testing Program. The ASME Code,Section XI (Ref. 5), requires that safety and relief valve tests be performed in accordance 'with ASME OM Code Appendix I (Ref. 6). According to Reference 6, the following tests are required:

a.

Visual examination;

b.

Seat tightness determination;

c.

' Setpoint pressure determination (lift setting);

d.

Compliance with owner's seat tightnf3ss criteria; and

e.

Verification of the balancing device integrity on balanced valves.

The ASME OM Code requires that all valv.es be tested every 5 years, '

and a minimum of 20% of the \\lalves be tested every 24 months. The

,ASME Code specifies the activities and frequencies necessary to satisfy the requirements. ' Table 3.7.1-2 allows a +/- 3% setpoinf '

(as-found lift point) tolerance on the valves for OPERABILITY (with the exception of the lowest set MSSV setpoint, which is (+3%/-2%);

however, the valves are reset to +/- 1% during the Surveillance to ai/ow for drift. The lift settings, according to Table 3. 7.1-2 in the aexompanying LCO, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

( continued)

DIABLO CANYON - UNITS 1 & 2

'Revision 5 5

MSSVs B 3.7.1 BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure. If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

REFERENCES

1.

FSAR, Section 10.3.1.

2.

ASME Boiler and Pressure Vessel Code,Section III, 1968..

3.

FSAR, Section 15.2 and 15.3.

4.

Westinghouse Nuclear Safety Advisory Letter NSAL-94-001, "Operation at Reduced Power Levels with Inoperable MSSVs," January 20, 1994 (included in NRC Information Notice IN-94-60, "Potential Overpressurization of the Main Steam System," August 22, 19941.

5.

ASME, Boiler and Pressure Vessel Code,Section XI.

6.

ASME Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including 2002 and 2003 Addenda.

7.

Westinghouse Report WCAP-1 1082, Revision 65, "Westinghouse Setpoint Methodology for Protection Systems Diablo Canyon Units 1 and 2, 24 Month Fuel Cycle Program."

8.

PG&E Design Calculation N-1 14, "Over-Pressure Study for One MSSV Per Loop Unavailable", dated 3/10/94.

9.

PG&E Design Calculation N-1 15, "Reduced Power Levels for A Number of MSSVs Inoperable", dated 3/14/94.

10.

PG&E Design Calculation STA-279, Revision 0, "RETRAN Loss of Load With an Inoperable MSSV".

DIABLO CANYON - UNITS 1 & 2 Revision 5 6

BASES SURVEILLANCE REQUIREMENTS REFERENCES SR 3. 7.1.1 (continued)

MSSVs B 3.7.1 This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure. If the MSSVs are not tested i at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

1.

FSAR, Section 10.3.1.

2.

ASME Boiler and Pressure Vessel Code, Section 111,1968..

3.

FSAR, Section 15.2 and 15.3.

4.

Westinghouse Nuclear Safety Advisory Letter NSAL-94-001,

"Operation at Reduced Power Levels with Inoperable MSSVs," January 20, 1994 (included in NRC Information Notice IN-94-60, "Potential Overpressurization of the Main Steam System," August 22, 1994}.

5.

ASME, Boiler and Pressure Vessel Code,Section XI.

6.

ASME Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including 2002 and 2003 Addenda.

7.

Westinghouse Report WCAP-11 082, Revision §e, "Westinghouse Setpoint Methodology for Protection Systems Diablo Canyon Units 1 and 2, 24 Month Fuel Cycle Program."

8.

PG&E Design Calculation N-114, "Over-Pressure Study for One MSSV Per Loop Unavailable", dated 3/10/94. '

9.

PG&E Design Calculation N-115, "Reduced Power Levels for A Number of MSSVs Inoperable", dated 3/14/94.

10.PG&E Design Calculation STA-279, Revision 0, "RETRAN Loss of Load With an Inoperable MSSV".

(.

I DIABLO CANYON - UNITS 1 & 2 Revision 5 6

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Initiation Request Acceptance... Administration Questions 8 September 2009 Answers 8 September 2009 Results Approval Other: ML092470412, ML092470416, ML092600769, ML092600799
MONTHYEAR2009-09-03 [Table View]

DCL-09-062, Emergency License Amendment Request 09-04 Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs) for Unit 2 Cycle 15

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DCL-09-062, Emergency License Amendment Request 09-04 Revision to Technical Specification 3.7.1, Main Steam Safety Valves (Mssvs) for Unit 2 Cycle 15

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