View source for DCL-12-035, Report for 10 CFR 50.59, Changes, Test, and Experiments, for the Period January 1, 2010, Through December 31, 2011

{{Adams
| number = ML12110A110
| issue date = 04/18/2012
| title = Report for 10 CFR 50.59, Changes, Test, and Experiments, for the Period January 1, 2010, Through December 31, 2011
| author name = Becker J
| author affiliation = Pacific Gas &amp; Electric Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000275, 05000323
| license number = DPR-080, DPR-082
| contact person = 
| case reference number = PG&amp;E Letter DCL-12-035
| package number = ML121100121
| document type = Letter type:DCL
| page count = 67
}}

=Text=
{{#Wiki_filter:Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Pacific Gas and Electric Company James R. Becker      Diablo Canyon Power Plant Site Vice President  Mail Code 104/6 P. O. Box 56 Avila Beach, CA 93424 805 .545. 3462 April 18, 2012                                                                                     Internal: 691.3462 Fax: 805 .545.6445 PG&E Letter DCL-12-035 U.S. Nuclear Regulatory Commission                                                      10 CFR*50.59 Attention: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Report of 10 CFR 50.59, "Changes, Tests, and Experiments," for the Period January 1,2010, through December 31,2011

==Dear Commissioners and Staff:==

Pursuant to 10 CFR 50.59, "Changes, tests, and experiments," Pacific Gas and Electric Company (PG&E) is enclosing the. 10 CFR 50.59 Report for Diablo Canyon Power Plant, Units 1 and 2, for the period January 1, 2010, through December 31, 2011. In accordance with 10 CFR 50.59(d)(2), Enclosures 1 and 2 provide a summary of all changes, tests, and experiments performed in accordance with 10 CFR 50.59 during this period. contains a summary of one change performed in accordance with 10 CFR 50.59 that contains security-related information. PG&E requests the enclosure be withheld from public disclosure as security-related information under 10 CFR 2.390.
Evaluations performed in accordance with 10 CFR 50.59 are performed as part of PG&E's licensing basis impact evaluation (LBIE) process. Since the LBIE process is used to perform reviews for compliance with regulations in addition to 10 CFR 50.59, some LBIEs do not include a 10 CFR 50.59 evaluation and, therefore are not included in this report.
The Plant Staff Review Committee has reviewed the referenced LBIEs and has concurred with the determination regarding NRC approval as specified in the summaries.
PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.
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 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
m I
Document Control Desk April 18, 2012
  & Page 2 PG&E Letter DCL-12-035 If you have any questions or require additional information, please contact Mr. Tom Baldwin at (805) 545-4720.
Sincer~~                                                   _ _ _,,"
James R. Becker Site Vice President
                             ~~
d ngd/4955/50045548 Enclosures cc:               Diablo Distribution cc/enc:           Elmo E. Collins, NRC Region IV Michael S. Peck, NRC Senior Resident Inspector Joseph M. Sebrowsky, Project Manager NRR Alan B. Wang, Project Manager 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 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 REPORT OF 10 CFR 50.59, "CHANGES, TESTS, AND EXPERIMENTS,"
for the Period January 1, 2010, through December 31, 2011 Pacific Gas and Electric Company Diablo Canyon Power Plant, Units 1 and 2 Docket Nos. 50-275 and 50-323  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 Index of LBIEs LBIE    Title                                                                                                            Page 10-007  Abandon Boric Acid (BA) Evaporator and Perform Auxiliary Control Board Modifications .............................................................................. 1 10-008  Unit 1 Safety Injection (51) Test Header Project, Revision 1 ........................... 4 10-009  Upgrade Safety Parameter Display System (SPDS) ......................................... 7 10-011  Allow Bypass of P-12 Interlock .......................................................................... 9 10-013  Maximum Flow from ECCS .............................................................................. 11 10-015  Unit 2 Containment Fan Cooler Unit (CFCU) Anti-Reverse Rotation Device (ARRD) ............................... .................................................................... 12 10-016  Unit 1 CFCU ARRD ....................... ..................................................................... 14 10-017  Backup Spent Fuel Pool (SFP) Cooling System ............................................. 15 10-018  Units 1 and 2 Temporary Vital Area Fencing .................................................. 18 10-019  Pressurizer Level Increase in Modes 3, 4, and 5 UFSAR ............................... 18 10-020  Discontinue Use of Diablo Creek Water .......................................................... 20 10-021  Integrated Head Assembly (IHA) ...................................................................... 23 10-022  Prompt Operability Assessment (POA) Update per SAP Notification (SAPN) 50301167-7 for MMD M000085 Unit 1 Cycle 17 Reload Evaluation ........................ ..................................................................... 24 10-023  UFSAR 15.4.3 Steam Generator Tube Rupture (SGTR) Revised Analysis for Margin to Overfill (MTO) .............................................................. 26 10-024  Unit 1 Replace Reactor Coolant Drain Tank (RCDT) Concrete Hatch Covers with Grating ....................................... ........................................ 30 10-025  Auxiliary Building Control Board Replacement - Phase 4B ......................... 32 10-026  Unit 2 Safety Injection (51) Test Header Project and Unit 2 51 Travel Stop Project ..................... ...................................................................... 34 11-003  Blast and Bullet Resistant Enclosure (BBRE) Guard Towers ....................... 37 11-005  Unit 2 Replace RCDT Concrete Hatch Covers with Grating .......................... 39 11-006  230 kV System Dual Unit Trip Licensing and Design Bases Change (document only change) ..................................................................... 41 11-007  Unit 1 and Unit 2 First Level Undervoltage Load Shed Relay Setpoint Change ....................... .... .. ................................................................... 43 11-008   POA Update per SAPN 50301167-7 for MMD M000084 Unit 2 Cycle Reload Evaluation ........................ ..................................................................... 47 11-009   Replacement of LBIE Review - Gross Failed Fuel Detector (GFFD)
Removal .............................................................................................................49 11-010   Replace Unit 1 7100 Process Controls System (PCS) ................................... 50 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 LBIE     Title                                                                                            Page 11-013  Component Cooling Water (CCW) Inventory Control Procedures .... .... .... .... 52 11-014  Residual Heat Removal (RHR) System Pressurization Due to Small Break Loss of Coolant Accident (SBLOCA) .. .... .... ........................... .... 53 11-015  Unit 1 Polar Crane Modification ....................................................................... 56 11-017  Units 1 and 2 Auxiliary and Fuel Handling Building Ventilation System (AFHBVS) Control System .................................................................. 58 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 10-007      Abandon Boric Acid (BA) Evaporator and Perform Auxiliary Control Board Modifications Reference Document No.:           DDP 1000000320, Revision 0 Reference Document

==Title:==
Auxiliary Control Board Phase 4A-Boric Acid Evaporator Activity

== Description:==

This design change implements Phase 4A of the Auxiliary Control Board Obsolescence Management upgrade.
This design change abandons in place the BA evaporator which is part of the chemical volume control system (CVCS) including the BA condensate demineralizers, boric acid concentrates holding tank (BACHT) and associated pumps. Abandonment in place of the BA evaporator would involve minimal costs compared to the complete removal of the equipment and would eliminate the future cost of modification to the Auxiliary Control Board, which includes controls and indication for the BA evaporator. The abandoned equipment will be drained and isolated from active systems. The design change would define abandonment boundary isolation valves and/or cutting and capping piping (or use of blind flange) and draining at the system interface. This includes CVCS system interface isolation with the component cooling water (CCW), auxiliary steam, liquid radwaste, primary make-up water, and nitrogen system. Providing equipment abandonment eliminates the BA evaporator controls and indication on the Auxiliary Control Board. This design change also abandons the pump controls and heat trace and controls to keep BACHT water warm. It also reduces the maintenance cost of maintaining the BA evaporator equipment for future phases of the Auxiliary Control Board design. The system is no longer being used, and the abandonment will be in accordance with plant procedures.
Except for the CCW piping, all piping and pipe supports associated with this design change are Class II. The existing CCW piping that feeds the abandoned BA evaporator and auxiliary steam drain receiver vent condenser heat exchangers is Class I piping and will be isolated via the supply and return valves on the 10 inch line and mechanically blocked by the addition of in line flanges with a blank plate. The portions of CCW lines and valves that are abandoned in place will no longer be in service. The existing BA evaporator and auxiliary steam drain receiver vent condenser heat exchangers are Class II (QA Class S) and will no longer be in service. Coordination of the piping seismic update has been reviewed and accepted.
Other miscellaneous modifications include Auxiliary Control Board control 1
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Enclosure 1 PG&E Letter DCL-12-035 modifications to associated panels and to provide panel preparation for the next Auxiliary Control Board Design Phase 4B:
* Remove auxiliary building horns and beacons that are no longer being used.
o Remove Horn/Alarm relays - POB 1.
o Remove Local horns/beacons (EAAX).
o Remove Call Box POB2.
* Remove BA Evaporator instrument and electrical indication and controllers that have been abandoned.
o Remove valve, instrument controls, relays and indication.
o Remove instrument tubing to controllers and extraneous valves.
o Remove sloping console associated BA Evaporator.
o Disconnect annunciation from POB 1 and POB 2 related to the BA evaporator. Associated Annunciators PK61 and PK66 will be removed during Phase 4B o Provide foreign material exclusion to cover holes left by removal of instrument and electrical panel components until patch work can be completed during Phase 4B.
o Abandon in place Heat Trace Circuits 10191, 20102, 20104 (225 F alarms), and other heat trace circuits associated with the BA evaporator and concentrates holding tank.
o Abandon in place BACHT 0-1 and 0-2 heat trace and warming circuits.
o Delete concentrates holding tank controls on existing digital control system.
* Extend power distribution (set up availability to provide connectivity during future Phase 4B) o POCV1 - Breakers/terminals o POCV2 - Breakers/terminals o POB1 - Breakers/terminals o POB2 - Breakers/terminals
* Extend conduit and wire pulls (set up availability to provide connectivity during future Phase 4B) o Pressure transmitters to Panel PM - 129 (Unit 1 and 2) o Conduit and wire pulls for containment isolation valve indication
* POLRI1 to POB 1
* POLRI3 to POB 1
* Add new chassis - determine supports and mounting (set up availability to provide connectivity during future Phase 4B) 2  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 o  POCV1-IO (10 slot) o  POCV2-IO (10 slot) o  POB1-I0 1 (1 3 slot) o  POB1-102 (10 slot) o  POB2-IO (10 slot)
* Clean up POCV1 (set up availability to provide connectivity during future Phase 4B) o Remove empty Panduit
* Remove 16,17, 18, 19, and 20
* Replace Kepco Power Supply Chassis RA-60 (parallel wire redundant, hot swap connectors) in PAXBNET with a Kepco RA-63 chassis (independent, hot swap connectors). This is to prevent taking out both power supplies in the event of a short.
* Replace keyboard-video-mouse (KVM) in PAXBNET with a new rack KVM. The old KVM is obsolete.
Summary of Evaluation :
The equipment is not required to prevent or mitigate any accident scenario and not used for the safe shutdown of the plant. The boron recovery system abandonment and other modifications will not result in more than minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR).
The boron recovery system abandonment and other modifications will not result in more than minimal increase in the likelihood of occurrence of a malfunction of an structures, systems, and components (SSC) important to safety previously evaluated in the UFSAR.
The proposed design change does not prevent or degrade the effectiveness of actions described or assumed in an accident discussed in the UFSAR, alter the assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR, or have a direct role in mitigating the radiological consequences of an accident described in the UFSAR.
Therefore, the new system will not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The proposed design change does not prevent or degrade the effectiveness of actions described or assumed in an accident discussed in the UFSAR, alter the assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR, or have a direct role in mitigating the 3  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 radiological consequences of an accident described in the UFSAR.
Therefore, the new system will not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety, previously evaluated in the UFSAR.
The BA evaporator, concentrates filter, concentrates holding tank, condensate demineralizers, condensate filter, associated pumps and auxiliary steam drain vent condenser are not credited to perform any accident mitigation scenario or required for the safe shutdown of the plant per section 3, 6, and 15 of the UFSAR. The modification will not create a possibility for an accident of a different type other than any previously evaluated in the UFSAR.
The postulated failures are bounded by the existing failure analyses. The design change will not create a possibility for a malfunction of an SSC important to safety with a different result other than any previously evaluated in the UFSAR.
The design functional requirements of the main system are maintained and does not alter, change, or challenge to any fission product barrier (FPB). The design change does not cause a design basis limit for a FPB as described in the UFSAR to be exceeded or altered.
The design change will not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
10-008      Unit 1 Safety Injection (SI) Test Header Project, Revision 1 Reference Document No.:          loop 1000000275, Revision       1 Reference Document

==Title:==
I U1 SI Test Header Valve Optimization Activity

== Description:==

This design change package (DCP) makes changes to the Unit 1 SI System piping and valve network that is used for performing reactor coolant system (RCS) pressure isolation valve (PIV) and emergency core cooling system (ECCS) check valve leak tests. This "safety injection check valve test header" is located primarily inside containment. The changes include:
* Remove ten, 3/4 inch diameter air-operated valves (AOVs) (SI-1-8843,
                    -8877 A-D, -8879A-D, and -8881; these valves are associated with 4
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Enclosure 1 PG&E Letter OCL-12-035 testing the accumulator discharge line PIVs and the charging injection check valves);
* Remove five, 3/4 inch diameter manual va lves (81 2 1, -22 , -23 , -32 ,
and -180);
* Modify 3/4 inch diameter test header piping and its supports inside containment;
* Reclassify a portion of the test flow collection header from Piping Code Class C to Piping Code Class E;
* Modify containment penetration piping for penetrations 34 and 51 B;
* Add four "test stations" at elevation 91 foot inside containment, but do not connect them to plant piping;
* Add 55, 3/4 inch diameter manual valves (one is outside containment in penetration 51 B piping, twenty-nine are added to piping used for the check valve testing, twenty-five are added into the above "test station" piping);
* Install a permanent "travel stop" on the stems on nine of the new, manual, globe valves (302A1B/C/O and 303A1B/C/O, and 310) to limit the opening of these reactor coolant pressure boundary (RCPB) valves
* Modify Control Room Vertical Board 1VB1 to remove ten control switches, de-terminate the related wiring, remove related fuses;
* Modify containment electrical penetrations 2E (2E-IN) and 11 E (11 E-IN) due to the removal of the 10 AOVs;
* Modify the contents of five mechanical panels (PM-29, 66, 69, 71, and 122) inside containment (e.g., determinate wiring, remove solenoid valves (8Vs) , modify instrument air tubing and components); and
* Add isolation valves at the interface between the 81 accumulator fill line and the check valve test header to allow more flexibility for filling the accumulators during testing.
Because of these changes, the following effects will result:
* Any potential leakage from the RCPB through the test lines will be controlled by restricting the opening of the new, manual valves using valve stem travel stops instead of internal ports provided in the AOVs.
* Testing of the PIVs will be restricted to plant operating Modes 4, 5, and below. Only one PIV may be leak-tested at *a time. That is, only one pair of RCPB manual valves may be open for testing at one time.
* How the leak-test valves are controlled is changed. Each AOV position is controlled by a control switch with position indication on Control Room Vertical Board 1VB 1 via electrical circuits, a 8V, and an air supply; each new manual valve position will be controlled locally inside the containment via a T-handle operator.
* One of the alternate flow paths for accumulator fill in the management of severe accidents is eliminated.
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Enclosure 1 PG&E letter DCl-12-035 Why are these changes being made?
* To comply with Diablo Canyon Power Plant (DCPP) commitments made in PG&E letter DCl-08-090, "Nine-Month Response to NRC Generic Letter 2008-01, 'Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,'" dated October 14, 2008, to reduce the probability of gas intrusion and voids formation and collection in ECCS piping;
* To assure more-reliable test boundaries for RCS PIV/ECCS, check valve leakage testing (in part by replacing each AOV with two, normally closed, manual, isolation valves in series);
* The downgrade of Class C piping to Class E was made for economic reasons, since it was determined that maintaining its classification was unnecessary with the new manual valves;
* Restricting the opening of the new manual valves using valve stem travel stops ensures that the flow through the valve remains below the threshold of the small reactor coolant pipe break analysis; and
* To prepare the plant with the eventual use of the four test stations for a more versatile testing configuration with other shutdown activities and added flexibility for the accumulator filling process.
Summary of Evaluation:
A review of the UFSAR described accidents related to this activity determined that this change does not result in a more than minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.
A review of UFSAR described malfunctions related to this activity determined the change does not result in a more than minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
A review of UFSAR described accidents related to this activity determined that this change does not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR.
A review of UFSAR described malfunctions related to this activity determined that this change does not result in a more than minimal increase in the consequences of a malfunction of an SSC previously evaluated in the UFSAR.
The proposed activity does not create a new or different type of accident.
There are no new or existing malfunctions with different results than 6  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 previously evaluated in the UFSAR.
The proposed activity does not result in any design basis limit for a FPB as described in the UFSAR, being exceeded or altered.
The activity does not involve any changes to any method of evaluation.
Based on the negative responses in this evaluation, prior NRC approval is not required to implement this change.
10-009 . Upgrade Safety Parameter Display System (SPDS)
Reference Document No.:           Unit 1 DDP 1000000290, Revision 0, Unit 2 Diablo Canyon Design Change Package NEXIS/SAP Document of Record (DDP) 1000000291, Revision 0 Reference Document

==Title:==
SPDS Replacement Activity

== Description:==

To improve reliability and resolve obsolescence issues, the existing SPDS computing subsystem in the Technical Support Center (TSC) will be removed and the SPDS server application will be hosted on the recently installed
           .transient recording system (TRS) in the Plant Process Computer (PPC)
Room. The SPDS display subsystem in the control room, the TSC and the Emergency Operations Facility (EOF) will be removed and replaced with business grade personal computers (PCs). The new PCs in the control room will be connected to the Plant Data Network (PDN) and receive data directly from the SPDS software running on the TRS. The new SPDS PCs in the TSC and EOF will be connected to the DCPP local area network (LAN) and will receive data from new, redundant demilitarized zone (DMZ) servers located in the T$C that run remote applications over the secure connection. In addition, the Emergency Assessment and Response System (EARS) and Meteorological Information and Dose Assessment System (MIDAS) applications will be re-hosted on the DMZ servers. The redundant DMZ servers will be common to both Units 1 and 2, and all input channels.
To improve security and comply with NRC Regulatory Issue Summary (RIS) 2009-13 recommendations, the modem connection to the NRC will be removed and a secure internet interface provided by the NRC will be installed.,
To enhance security of the PDN and address new NRC RIS 2009-13 Cyber Security recommendations, this project will constrain the PDN to inside the protected area boundary of DCPP. The PDN will no longer extend to the simulator and the EOF. The simulator and the EOF will connect to the PDN 7
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Enclosure 1 PG&E Letter DCL-12-035 via the DCPP LAN and firewalls will provide cyber security. Emergency Response Facility Display System (ERFDS), SPDS, PPC, EARS, and radiation data processor desktop PCs and displays in the EOF will be connected to the DCPP LAN. For consistency, and to facilitate driving displays from the simulator during drills, ERFDS, SPDS, PPC, EARS, and radiation data processor desktop PCs and displays in the TSC will also be connected to the DCPP LAN. The desktop PCs in the TSC and EOF will receive data from remote applications hosted in the DMZ servers.
Independent power feeds from TSC instrument alternating current and from the TSC uninterruptible power system (UPS), the Unit 1 PPC UPS, or the Unit 2 PPC UPS will be provided to PDN and DCPP LAN interface components in the TSC that are required for SPDS, ERFDS, and EARS functionality.
This design change will also install a fiber optic run from the PPC room to the TSC for future use. The following systems are affected by the SPDS Upgrade design change:
* ERFDS
* TRS
* SPDS
* EARS
* MIDAS
* EROS (Emergency Response Data System)
* Meteorological monitoring, radiation data processing
* PDN
* PPC
* Reactor coolant pump (RCP) vibration monitoring inputs to the PDN
* Turbine and feed water supervisory monitoring server
* Radiation Monitors RM-15 and 15R Summary of Evaluation:
The design change to upgrade the SPDS does not introduce the possibility of a change in the frequency of an accident because the affected systems are used for monitoring only with no associated control functions. The affected systems are not initiators of any accident identified in Chapters 3, 6, and 15 of the UFSAR. Per the failure modes and effects analyses performed in the design change evaluation, no new failure modes are introduced.
The likelihood for a malfunction due to environmental, electrical, and seismic interactions will not increase.
The consequences (dose) of an accident previously evaluated in the UFSAR 8  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 will not increase because the affected systems are used for monitoring only and cannot initiate any accidents.
The consequences (dose) of a malfunction previously evaluated in the UFSAR will not increase because no new failure modes are introduced.
The proposed activity does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR because the affected systems cannot initiate an accident.
The failure modes and effects analysis and reliability discussed in the design change evaluation indicate there is less than a minimal increase in the possibility for a malfunction of an SSC important to safety.
The design change does not affect any FPBs or cause any system parameters to change. Upgrading the SPDS does not result in a design basis limit for a fission product barrier (DBLFPB) as described in the UFSAR being exceeded or altered.
There is no change to any method of evaluation as defined in any procedure or licensing bases document.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
10-011      Allow Bypass of P-12 Interlock Reference Document No.:            DDP 1000000332, Revision 0 Reference Document

==Title:==
Bypass P-12 Interlock Outputs at E21 in Mode 3 Activity

== Description:==

This design change documents that it is allowable to bypass the P-12 interlock once borated to cold shutdown conditions in Mode 3. The P-12 interlock is a protection function that blocks steam dumps on low-low Tavg to prevent excessive cooldown due to steam dump control system failure. The existing P-12 can be bypassed with four 40-percent steam dump valves (called Group 1). This change allows increasing the existing bypass to include all 12 40-percent steam dump valves (Group 1 and Group 2).
Operations plans to put the bypass of P-12 in DCPP Operating Procedure OP L-5, "Plant Cooldown from Minimum Load to Cold Shutdown." The bypass is accomplished using the Man Machine Interface of Eagle 21 to 9
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Enclosure 1 PG&E Letter DCL-12-035 perform a software bypass of the TC- 4X2D bi-stables to bypass the P-12 interlock at each protection set.
The bypass of the P-12 interlock to support Plant cooldown has been performed at other Westinghouse Plants (e.g., Wolf Creek, Calloway, and Salem), and allows Operations improved capability to cool down the plant and depressurize within the limits of the pressure and temperature limits report (PTLR).
Summary of Evaluation:
This change does not increase the frequency of occurrence of an accident previously evaluated in the UFSAR.
There is no increase in the likelihood of any malfunctions of SSCs important to safety previously evaluated in the UFSAR.
The impacted accident analysis is the consequence of a steam generator tube rupture (SGTR). This change is bounded by the existing SGTR analysis.
Therefore, there is not a more than minimal increase In the consequences of an accident previously evaluated in the UFSAR.
The consequences from associated malfunctions are bounded by the existing analysis. Therefore, there is not a more than minimal increase in the consequence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
The proposed activity does not create a possibility for an accident of a different type than previously evaluated in the UFSAR.
There has been no change in the failure modes or effects related to the proposed activity. Any failures related to the proposed activity remain bounded by the existing analysis. Therefore, the proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
There is no possibility of a return to criticality associated with the proposed activity. Therefore, there is no impact such that the limiting fuel cladding departure from nucleate boiling ratio (DNBR) results in the UFSAR are exceeded or altered.
There is no evaluation methodology impacted by this change.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
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Enclosure 1 PG&E Letter DCL-12-035 10-013      Maximum Flow from ECCS Reference Document No.:           I Calc 9000006525, Revision 004-02 Reference Document

==Title:==
I Calculates Maximum Flow from ECCS Pumps Activity

== Description:==

The calculation is revised to incorporate a pressure drop penalty for the potential of the sump isolation valve not being fully open. This revision also includes the degassing effect for the sump, and adds a net positive suction head required (NPSHR) table tabulated from the maximum NPSHR values from all residual heat removal (RHR) pumps. Finally, this revision also addresses the differential pressure across the sump screen and plenum based on specific emergency operating procedure (EOP) flow alignment. The calculation revision changes the net positive suction head available (NPSHA) value for the RHR pumps. The NPSHA and NPSHR values are given in UFSAR Table 6.3-11 for the RHR pumps.
Summary of Evaluation:
The calculation revision for RHR pump suction head (input parameter change) does not introduce the possibility of a change in the frequency of an accident because the calculation revision ris not an initiator of any accidents and no new failure modes are introduced.
The proposed activity does not result in more than a minimal increase in the likelihood of occurrence of a malfunction previously evaluated in the UFSAR.
The calculation revision for RHR pump suction head (input parameter change) does not introduce the possibility of a change in the consequences of an accident because the calculation revision is not an initiator of any accidents and no new failure modes are introduced.
The calculation revision for RHR pump suction head (input parameter change) does not introduce the possibility of a change in the consequences of a malfunction because the calculation revision is not an initiator of any malfunctions and no new failure modes are introduced.
The calculation revision for RHR pump suction head (input parameter change) does not introduce the possibility of a new accident because the calculation revision is not an initiator of any accident and no new failure modes are introduced.
The calculation revision for RHR pump suction head (input parameter change) 11 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 does not introduce the possibility for a malfunction of an SSC with a different result because the change does not introduce a new failure mode.
This revision to the calculation for RHR pump suction head (input parameter change) does not result in a design basis limit for a FPB (fuel cladding, RCS boundary, or containment) as described in the UFSAR being exceeded or altered.
The activity does not involve a change to a method of evaluation.
Based on this evaluation, NRC approval is not reql:Jired prior to implementing this activity.
10-015      Unit 2 Containment Fan Cooler Unit (CFCU) Anti-Reverse Rotation Device (ARRD)
Reference Document No.:          I DDP 1000000119,    Revision 0 Reference Document

==Title:==
I CFCU ARRD Activity

== Description:==

The proposed change replaces the existing CFCU motor-to-fan couplings with a combination coupling and ARRD.
Proposed activity:
: 1. Replacement of the existing CFCU motor to fan couplings with ARRD/couplings.
The installation of the ARRD/couplings provides an alternate means to prevent reverse rotation of the CFCU fans and motors. Changing this method of preventing reverse rotation is considered a fundamental change in how this design function is technically performed.
: 2. Disassembly of the backdraft dampers.
Complete disassembly of the backdraft dampers at all five CFCU locations. Parts removed include blades, blade beams, counterweights, counterweight arms, linkage, linkage arms, axles, bearings, and springs. The backdraft damper frame weldment will remain.
The flow distribution of the containment fan cooler system (CFCS) will be affected with the removal of backdraft dampers, allowing backflow through the discharge ducting, past the idle fans.
12 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035
: 3. Closure of the volume control dampers.
Four volume control dampers on a 30 inch diameter duct, downstream of the annular ring will be closed. The flow distribution to containment elevation 199 foot will be affected.
: 4. Conduct a tesUexperiment with CFCU location configured with an ARRD/coupling and the associated backdraft damper secured in the fully opened position to validate the design concept.
Why are these changes being made:
This design change is being implemented because there are two significant design issues with the CFCU backdraft dampers: (1) their performance is unreliable, (2) and they are costly to maintain.
Summary of Evaluation:
The installation of the ARRD/couplings provides an alternate means to prevent reverse rotation that was evaluated to ensure that it does not does adversely affect the system performance of the CFCU. The alternate method is technically appropriate for the intended application and within the bounds of what has been found acceptable under 10 CFR 50.59 evaluation.
Disassembly of the backdraft dampers and closure of the volume control dampers were also found acceptable under 10 CFR 50.59 evaluation.
The activity of installing a single ARRD/coupling, and securing the associated backdraft damper in the fully opened position configures the facility in a situation that was not described in the UFSAR. The 10 CFR 50.59 evaluation concluded that prior NRC approval is not required.
The 10 CFR 50.59 evaluations concluded that no important-to-safety design functions of plant SSCs were compromised with this design change.
The 10 CFR 50.59 evaluations of the proposed activities concluded that they did not result in more than a minimal increase in the frequency of occurrence or consequences of an accident. The proposed activities do not create a possibility for an accident of a different type. The proposed activities do not result in more than a minimal increase in the likelihood of occurrence or consequences of a malfunction. The proposed activities do not result in a DBLFPB as described in the UFSAR being exceeded or altered, and do not involve any changes to any method of evaluation.
13  Contains Security-Related Information - Withhold Under 10-CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 Based on this Evaluation, prior NRC approval is not required to implement this change.
10-016     Unit 1 CFCU ARRD Reference Document No.:          I DDP 1000000108, Revision 0 Reference Document

==Title:==
I CFCU ARRD Activity Description/Summary of Evaluation:
This proposed change replaces the existing CFCU motor-to-fan couplings with a combination coupling and device ARRD.
Proposed activity:
: 1. Replacement of the existing CFCU motor to fan couplings with ARRD/couplings.
The installation of the ARRD/couplings provides an alternate means to prevent reverse rotation of the CFCU fans and motors. Changing this method of preventing reverse rotation is considered a fundamental change in how this design function is technically performed.
: 2. Disassembly of the backdraft dampers.
Complete disassembly of the backdraft dampers at all five CFCU locations. Parts removed include blades, blade beams, counterweights, counterweight arms, linkage, linkage arms, axles, bearings, and springs. The backdraft damper frame weldment will remain.
The flow distribution of the CFCS will be affected with the removal of backdraft dampers, allowing backflow through the discharge ducting, past the idle fans.
: 3. Closure of the volume control dampers.
Four volume control dampers on a 30 inch diameter duct, downstream of the annular ring will be closed. The flow distribution to containment elevation 199 foot will be affected.
: 4. Conduct a tesUexperiment with CFCU location configured with an ARRD/coupling and the associated backdraft damper secured in the fully opened position to validate the design concept.
14 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 Why are these changes being made:
This design change is being implemented because there are two significant design issues with the CFCU backdraft dampers: their performance is unreliable and they are costly to maintain.
Summary of Evaluation:
The installation of the ARRD/couplings provides an alternate means to prevent reverse rotation that was evaluated to ensure that it does not adversely affect the system performance of the CFCU. The alternate method is technically appropriate for the intended application, within the bounds of what has been found acceptable under 10 CFR 50.59 evaluation.
Disassembly of the backdraft dampers, and closure of the volume control dampers were also found acceptable under 10 CFR 50.59 evaluation.
The 10 CFR 50.59 evaluations of the proposed activities concluded that they did not result in more than a minimal increase in the frequency of occurrence or consequences of an accident. The proposed activities do not create a possibility for an accident of a different type. The proposed activities do not result in more than a minimal increase in the likelihood of occurrence or consequences of a malfunction. The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR. The proposed activities do not result in a DBLFPB as described in the UFSAR being exceeded or altered, and do not involve any changes to any method of evaluation.
The 10 CFR 50.59 evaluation concluded that prior NRC approval is not required.
10-017      Backup Spent Fuel Pool (SFP) Cooling System Reference Document No.:           DDP 1000000340, Revision 0 Reference Document

==Title:==
Temporary SFP Cooling System Activity

== Description:==

Currently, the SFP of each unit has a permanent SFP cooling system. The permanent SFP cooling system is described in UFSAR Section 9.1.3. This design change allows the installation of the backup SFP cooling system consisting of two temporary piping and hose systems between Units 1 and 2 SFPs, if needed. Each of the piping and hose systems is connected to one or more submersible pumps located inside the SFP in each unit. The purpose of 15 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information -:- Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 the two pipe lines is to allow recirculation of the water between the two pools such that if the SFP cooling system of one pool is out of service, the Plant can recircu late the water of the SFP not having forced cooling to the SFP of the other unit, which has an operating SFP cooling system. The intent is to allow the use of the inservice SFP cooling system of one unit to cool both pools, if deemed necessary to meet Plant operating limits. The installation of the temporary piping and pumps is to manage the risk associated with fuel pool heat up during the maintenance on the SFP heat exchanger of one Unit. This temporary recirculating system may also be used to provide standby backup cooling during future refueling outages. In the event that the SFP cooling system is out of service for any reason.
The following changes will be implemented:
: 1. Add two sets of temporary piping and hoses to recirculate the water between the SFPs of each unit.
: 2. Add temporary submersible pumps in each SFP to take water from the SFPs and recirculate the water between the two pools. The pumps will be submerged in the SFPs to a depth of approximately 3 inches below the normal pool water surface.
: 3. Add a support structure to temporarily locate the pumps in the pools to maintain the submersion depth and to restrain the pumps to meet Seismically Induced System Interaction requirements.
: 4. Create a temporary breach in the fuel handling building ventilation boundary by opening the doors between the fuel pool areas and the hot machine shop to route the temporary piping and hoses between the two SFPs.
: 5. The input parameter representing the actual heat load to be transferred by the SFP heat exchanger from the SFP to the CCW system will be modified (increased) to reflect that the decay heat from both pools will be transferred to the CCW system of the unit with the operating permanent SFP cooling system. However, the methodology for determining the heat transfer is not changed.
Summary of Evaluation:
The adverse proposed activities (addition of piping and hoses and addition of submersible pumps) will not initiate a fuel handling accident. The proposed activity will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.
The proposed activities of adding piping and hoses, and adding submersible pumps to the SFP will not result in more than a minimal increase in the likelihood of occurrence of malfunctions previously evaluated in the UFSAR.
The new piping, hoses, and pumps are not connected to the permanent 16  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 cooling system.
As evaluated in the UFSAR, the analysis of the consequences of the fuel handling accident assumes the fission product release from one fuel bundle at a water depth of 23 feet. The adverse proposed activities will not cause the water depth to be below 23 feet above the top of the fuel. Therefore, the proposed activities will not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The adverse proposed activities of adding piping and hoses, and submersible pumps are bounded by the discussions and assumptions identified in the UFSAR. Therefore, the proposed activities will not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
The accident associated with the proposed activity is bounded by the fuel handling accident previously evaluated in the UFSAR. Therefore, the adverse proposed activities cannot initiate or create an accident of a different type than any previously evaluated in the UFSAR.
The result of any new malfunctions associated with the proposed activity is bounded by what has been evaluated in the UFSAR. Therefore, the proposed activity will not create a possibility of a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
The adverse proposed activities do not result in a design basis limit for a fission barrier as described in the UFSAR being exceeded or altered .
The proposed activity does not involve a method of evaluation.
Based on the evaluations performed, prior NRC approval is not required to implement the activity.
10-018     Units 1 and 2 Temporary Vital Area Fencing Reference Document No.:          I TMODs 60027442 and 60027595 Reference Document

==Title:==
I 10 Percent Steam Dump Valve Revitalization Activity Description/Summary of Evaluation:
Contains Security related information. See Enclosure 2.
17 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 Summary of Evaluation:
Contains Security related information. See Enclosure 2.
10-019      Pressurizer Level Increase in Modes 3, 4, and 5 UFSAR Reference Document No.:           UFSAR, Revision 19 Reference Document

==Title:==
Pressurizer Level Increase in Modes 3, 4, and 5 UFSAR Activity Description/Summary of Evaluation:
This proposed activity implements two UFSAR revisions related to the control of pressurizer level in shutdown modes of operation that are not currently described in the UFSAR. There is no change to the UFSAR described pressurizer level control during power operation modes.
The first revision to UFSAR Section 5.5.9.3.2 adds a description that pressurizer level is maintained between greater than or equal to 22 percent and ,less than or equal to 90 percent of indicated level in the shutdown Modes 3, 4, and 5. The maximum pressurizer level of less than or equal to 90 percent indicated level is consistent with the current Technical Specification (TS) Limiting Condition for Operation 3.4.9, which is applicable in Modes 1, 2, and 3. Currently, the UFSAR only describes the pressurizer level as being controlled at 60 percent at full load and reduced to 22 percent at zero power level, and does not specifically discuss pressurizer level in shutdown modes of operation. This change also adds a statement clarifying that when the low temperature overprotection (LTOP) system is enabled, the administrative controls and requirements of the PTLR take precedence.
There has been no change in LTOP operation or the PTLR and this statement is only being added to clarify the existing precedence of plant configuration control by the PTLR when LTOP is enabled.
This second UFSAR change to Section 15.1.1 clarifies that it is acceptable to control pressurizer between greater than or equal to 22 percent and less than or equal to 90 percent in shutdown Modes 3, 4, and 5 based on evaluations of the accident analysis documented in the new Reference 28 added to Section 15.5.10. These evaluations are summarized in a safety assessment. This evaluation process of changes in plant control system values is consistent with UFSAR Section 15.1.1 which states, "Since initial startup, setpoints and control system components have been maintained to optimize performance.
When changes are made, the accident analyses are reviewed and revised as necessary."
18 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter OCL-12-035 Summary of Evaluation:
The proposed activity was evaluated per 10 CFR 50.59 as a change to the facility as described in the UFSAR by responding to questions one through seven as summarized below.
The increased pressurizer level in the shutdown modes, and change in RCS mass, volume, and pressurizer steam space with respect to that currently described in the UFSAR, have been technically evaluated as documented in a safety assessment. The 10 CFR 50.59 responses as summarized below are based on the evaluation conclusions that the structural and thermal hydraulic analyses currently described in the UFSAR conservatively bound the plant conditions in the shutdown Modes 3, 4, and 5 with the increased pressurizer level.
Pressurizer level is an initial plant operating condition assumed in the accident analysis and is not an initiator of any accident previously evaluated in the UFSAR. Therefore, the proposed activity does not result in an increase in the probability of occurrence of an accident previously evaluated in the UFSAR.
Since the UFSAR accident analyses continue to bound the plant thermal response during the shutdown conditions of this proposed activity, there is no increased challenge to the potential malfunction of any plant SSC. The proposed activity does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
Since the current UFSAR accident analyses continue to meet all applicable fuel design limits, RCS overpressure limits, containment integrity limits, and secondary mass release assumptions, there is no impact on the offsite dose (00) and control room dose calculations in UFSAR Section 15.5. The proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The limiting single failures of the SSCs assumed in the UFSAR accident analyses and which form the basis for the calculated ODs and control doses in Section UFSAR 15.5 remain bounding and are not impacted . The pressurizer and all other SSCs will continue to operate within design and analysis limits such that no new system interactions or new failure modes have been created. The proposed activity does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the U FSAR.
The proposed activity does not alter or place any SSC in a configuration 19  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 outside design or analysis limits and does not create any new accident scenarios. The proposed activity does not create a possibility for an accident of a different type than previously evaluated in the UFSAR.
There is no change in the characteristic thermal plant response of any accident or assumed SSC malfunction as assumed in the UFSAR. The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
The UFSAR accident analyses continue to meet the currently described fuel design limits (DNBR, peak cladding temperature (PCT)), RCS overpressure limits, and containment integrity limits; and all evaluations have been performed with respect to these current UFSAR design basis limits such that no changes to any limits are required. The proposed activity does not result in a DBLFPB as described in the UFSAR being exceeded or altered.
Per licensing basis impact evaluation (LBIE) screen response 2.c, this proposed activity does not involve a change that adversely revises or replaces an UFSAR described evaluation methodology that is used in establishing the design bases or that is used in the safety analysis. Therefore, this 50.59 evaluation question is not applicable for this proposed activity.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
10-020      Discontinue Use of Diablo Creek Water Reference Document No.:           I DDP 1000000379 Reference Document

==Title:==
I Discontinue Use of Diablo Creek Water Activity Description/Summary of Evaluation:
What is being changed:
This document-change-only design change removes credit for Diablo Creek as a source of water for use by the Plant. The DCPP and Independent Spent Fuel Storage Installation (ISFSI) design and licensing bases are reviewed and revised to acknowledge no further capability, credit, or use of Diablo Creek water for Plant purposes. There is no physical work associated with the scope of this DCP.
Specifically, the changes will be:
20 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 Th is page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035
* Revise the DCPP UFSAR content to eliminate Diablo Creek as a source of long term cooling water (LTCW) for the Plant (DCPP UFSAR Section 6.5.2.1 .1 and Figure 6.5-2). The creek is listed as the 7th of 8 possible water sources (in terms of water quality).
* Revise Plant design criteria documents (e.g., Design Criteria Memorandums (DCMs) T-17, S-3B, S-16, S-18) to reflect not using Diablo Creek water for any Plant purposes.
* Revise the ISFSI UFSAR content that describes Diablo Creek as one of three sources of water for the Plant (ISFSI UFSAR Sections 2.4.1 and 2.4.9).
* Revise Plant Security Calculation EQP-822, to no longer rely on Diablo Creek as a water source (see also DCL-95-046, "Response to 10 CFR 73.55. 'Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage"').
* Other Plant documentation (e.g., Plant procedures, drawings, functional locations (FLOCs)) will also *be revised to remove references to Diablo Creek as a Plant water source.
In addition, revise all the necessary documents to reflect the abandonment of Water Well No.1, the breaching of the creek dam, and the removal of the permanent creek-water pump system. This includes revision to ISFSI UFSAR Sections 2.4.1, 2.5.1, and 2.6.4.4.
Why is the change made:
The December 14, 2006, Coastal Development Permit for the DCPP Steam Generator Replacement Project included a special provision to cease withdrawing water from Diablo Creek. Also, an NRC Component Design Basis Inspection audit in February 2010 identified the lack of Plant procedures to implement use of Diablo Creek water for the LTCW system as described in DCPP UFSAR Section 6.5.2.1.1.
Describe how the change may interface with the Plant licensing basis:
The DCPP UFSAR (Section 6.5.2.1.1, Figure 6.5-2) and ISFSI UFSAR (Sections 2.4.1 and 2.4.9) describe Diablo Creek as a possible/actual source of water to the Plant. A Plant security-related evaluation in 1995 also credited the availability of creek water to mitigate a security challenge.
Summary of Evaluation:
The use or non-use of Diablo Creek water is not the initiator of any previously evaluated accident. Therefore, the proposed activity does not result in more than a minimal increase in the frequency of occurrence of an accident 21  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 previously evaluated in the UFSAR.
Loss of an adeq uate water supply is not specifi cally identified as a malfunction for LTCW. An adequate water supply remains available. There is not more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
Diablo Creek water is not a cause of any accident previously evaluated and is not credited for the mitigation of any accident previously evaluated.
Therefore, the radiological consequences of accidents are not affected by this change.
Diablo Creek water is not a cause of any malfunction previously evaluated and is not credited for the mitigation of any malfunction previously evaluated.
Also, the consequences of malfunctions that may require the LTCW system do not change since the creek is not a credited source of LTCW. Therefore, the consequences of malfunctions are not affected by this change.
The use or non-use of Diablo Creek water is not the initiator of any accident of a type different from those evaluated in the DCPP UFSAR.
Discontinuing use of creek water does not affect the credited water sources and does not create the possibility for any malfunction with a different result than previously evaluated.
The design basis limits for the three DCPP FPBs were reviewed. The proposed activity does not result in a design basis limit for a FPB as described in the UFSAR being exceeded or altered.
This design change does not involve any method of evaluation described in the DCPP UFSAR.
Based on this evaluation, prior NRC approval is not required to implement this change.
10-021      Integrated Head Assembly (IHA)
Reference Document No.:          I DDP 1000000078, Revision 1 Reference Document

==Title:==
I Diablo Canyon Unit 1 IHA Activity Description/Summary of Evaluation:
The design of the DCPP IHA, a new structure that attaches to the reactor vessel closure head (RVCH), involves replacement of the existing control rod 22 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 drive mechanism (CRDM), seismic support structure, the CRDM ventilation cooling system, RVCH lift rig, reactor vessel (RV) head vent and level indication piping, valves, and other components associated with the existing RVCH. The IHA and the replacement RVCH (RRVCH) were installed in the Unit 1 sixteenth refueling outage. This 10 CFR 50.59 evaluation supersedes original Unit 1 10 CFR 50.59 documentation (which credits Unit 2 LBIE 2008-026) and supplemental documentation (LBIE 2009-021) as they apply to the IHA.
Summary of Evaluation:
Two aspects of the IHA design are screened in for determination under 10 CFR 50.59 of the impact on evaluation methodologies described in the UFSAR. These determinations are summarized below:
Use of BWSPAN Modal Combination Method The BWSPAN computer code was used for piping analyses of the RV Head Vent System and RV Level Indication System to assess impacts of the IHA design change. The BWSPAN method combines (using square root of the sum of squares (SRSS) method) all the modes first for each earthquake and then combines the modal resultant for each earthquake (x + y, y + z) by absolute sum to determine loads. The analysis of record (AOR) method combines the earthquakes (x + y and y + z) by absolute sum on the modal level and then the responses for all modes are combined by SRSS. These two combination methods were compared algebraically in a derivation that demonstrates that the BWSPAN modal combination method will always produce terms (loads) that are equal to or greater than (the conservative direction, i.e., design margins are reduced) the DCPP method.
Therefore, since the change in element of the method is essentially the same or conservative, there is no departure from a method of evaluation described in the UFSAR and prior NRC approval is not required for this activity.
Use of ANSYS The ANSYS computer code was used to perform dynamic analyses to evaluate IHAlRRVCH impacts on the RV supports, nozzles, internals, fuel, reactor coolant loop piping and equipment supports. The AOR used the WECAN computer code. As described in a PG&E calculation:
* ANSYS and WECAN are both finite element codes that use the same modeling methods - beam/pipe elements connected with gap or spring elements, accounting for fluid-solid interactions for the seismic case in the same way.
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Enclosure 1 PG&E Letter DCL-12-035
* The techniques used to solve the equations of motion are different.
The WECAN code uses non-linear superposition with composite modal damping , whereas ANSYS uses direct integration with proportional damping. Since the use of proportional damping results in a nonlinear relationship between damping and frequency, the structure is underdamped over a wide frequency.
* Benchmarking has been performed that compares WECAN and ANSYS for similar dynamic analyses. The results obtained from this comparison for ANSYS are shown to be essentially the same as those forWECAN.
Therefore, since the change in element of the method is considered essentially the same, there is no departure from a method of evaluation described in the UFSAR and prior NRC approval is not required for this activity.
10-022      Prompt Operability Assessment (POA) Update per SAP Notification (SAPN) 50301167-7 for MMD M000085 Unit 1 Cycle 17 Reload Evaluation Reference Document No.:           MMD M000085, Revision 1 Reference Document

==Title:==
Unit 1 Cycle 17 Reload Activity Description/Summary of Evaluation:
What is being changed This proposed activity evaluates increased engineered safety features (ESF) delay times associated with the first level undervoltage relay (FLUR) I second level undervoltage relay (SLUR) nonconforming condition per a POA for potential impact on the UFSAR accident analyses that support the Unit 1 Cycle 17 core reload design change. The Westinghouse evaluations of the increased ESF delay times documented in Westinghouse Letter PGE-1 0-54 are not being incorporated into the DCPP design basis at this time but provide the technical basis for future updates and eventual resolution of the FLURISLUR nonconforming condition. Therefore, only those evaluations of the nonconforming condition that support the Unit 1 Cycle 17 core reload must be reviewed per 10 CFR 50.59 to ensure they are consistent with the approved methodologies listed in TS 5.6.5.b that are used to verify the Unit 1 Cycle 17 core operating limits report (COLR) limits remain bounded.
The temporary modification (T-Mod) to the FLUR implemented as a compensatory measure per the POA only ensures no ESF equipment is damaged due to degraded voltage and does not impact any ESF delay times or core reload analyses . Therefore, the FLUR T-Mod is not within the scope of this LBIE review.
24 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 The increased ESF delay times that bound the nonconforming FLURISLUR condition associated with a 230 kV degraded voltage event and that are being evaluated in support of Unit 1 Cycle 17 core reload are summarized below:
Increased ESF Delay Times to Bound 230 kV Degraded Voltage Event ECCS Injection Flow                                          27 sec to 42 sec CFCU Heat Removal                                            48 sec to 52 sec Auxiliary Feedwater (AFW) Flow                               60 sec to 65 sec Containment Spray (CS) Flow                                   80 sec to 100 sec The current ESF delay times listed are with respect to the ESF initiating signal credited for accident mitigation in the DCPP safety analysis. The increased ESF delay times listed for the 230 kV degraded voltage event are with respect to reaching a SI setpoint, which initiates the automatic transfer of onsite power from the 500 kV to the 230 kV system and results in the additional SLUR delay time occurring as established in a Design Input Transmittal (DIT).
Summary of Evaluation:
Evaluation Scope The increased ESF delay times are not physically related to the cause of any accident and do not introduce the possibility of a change in the frequency of an accident previously evaluated in the UFSAR.
All ESF functions will continue to perform in the same manner as described in the UFSAR. Therefore, the increased ESF delay times do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
Since the applicable UFSAR accident analyses which support the Unit 1 Cycle 17 core reload continue to meet all applicable fuel design limits, RCS overpressure limits, containment leakage limits, secondary mass release assumptions, and 10 CFR 50 .46 criteria, there is no impact on the calculated offsite doses (ODs) or control room doses in UFSAR Section 15.5. Therefore, the increased ESF delay times do not result in more that a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The limiting single failures of the SSCs assumed in the UFSAR accident analyses which support the Unit 1 Cycle 17 core reload and which form the basis for the calculated ODs and control room doses in Section 15.5 remain bounding and are not impacted. Therefore, the increased ESF delay times do not result in an increase in the consequence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
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Enclosure 1 PG&E Letter DCL-12-035 The increased ESF delay times do not alter or place any SSC in a configuration outside design or analysis limits and do not create any new accident scenarios. The proposed activity does not create a possibility for an accident of a different type than previously evaluated in the UFSAR.
The increased ESF delay times do not change the characteristic thermal plant response of any accident or assumed SSC malfunction as assumed in the UFSAR. The SSCs important to safety will continue to perform their safety function as assumed in the safety analysis and will continue to operate within design and analysis limits such that no new system interactions or new failure modes have been created. The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
The UFSAR core reload accident analyses continue to meet the applicable fuel design limits (DNBR, PCT), RCS overpressure limits, and containment integrity limits as described in the UFSAR. All evaluations have been performed with respect to these current UFSAR design basis limits and no changes to any limits are required to establish that the Unit 1 Cycle 17 core limits remain bounded. Therefore, this proposed activity does not result in a design basis limit for a FPB as described in the UFSAR being altered or exceeded.
The increased ESF delay times evaluated in support of Unit 1 Cycle 17 core reload are input parameter changes that do not adversely revise or replace an UFSAR described methodology. Therefore, this 50.59 question is not applicable for this proposed activity.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
10-023     UFSAR 15.4.3 Steam Generator Tube Rupture (SGTR) Revised Analysis for Margin to Overfill (MTO)
Reference Document No.:          I UFSAR 15.4.3, Revision   19 Reference Document

==Title:==
I SGTR Activity Description/Summary of Evaluation:
What is being changed The UFSAR Section 15.4.3 is being updated to incorporate a revised SGTR Analysis for MTG case as documented in Westinghouse Letter PGE-1 0-56, 26 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter OCL-12-035 that no longer credits the time critical operator action (TCOA) to stop the turbine driven auxiliary feedwater (TOAFW) pump at 5.54 minutes after reactor trip. All AFW flow is now assumed to be isolated at 10 minutes after the tube rupture occurs. The SGTR 00 case as described in UFSAR 15.4.3 remains bounding since no longer crediting this TCOA would provide a benefit to the 00 results currently reported in Section 15.5.20. Although no reanalysis for the SGTR 00 case is required, this TCOA at 5.54 minutes will no longer be described in the UFSAR for this case.
Analysis Input Parameter Changes The revised SGTR MTO analysis incorporates three input assumption changes compared to the current AOR as described in the UFSAR.
: 1. OCPP no longer credits the manual operator action to stop the TOAFW pump at 5.54 minutes after the reactor trip occurs. All AFW flow is now assumed to be isolated at 10 minutes after the tube rupture occurs.
: 2. OCPP has revised the assumed AFW flow rates as documented in OIT-50275213-1-0 to more accurately model the AFW flow as a function of the ruptured steam generator (SG) pressure.
: 3. OCPP has revised the assumed ECCS flow rates as documented in 0IT-50275213-2-0 to reduce the margin of conservatism in order to minimize the loss of analysis margin due to no longer crediting the TCOA to stop the TOAFW pump early.
UFSAR Revisions The following UFSAR revisions will be incorporated as part of this proposed activity.
UFSAR Section 15.4.3 will be revised to include a description of the revised SGTR MTO analysis and the overfill results.
UFSAR Table 15.4-13 will be updated to no longer list the stopping of the TOAFW pump at 5.54 minutes in the sequence of events for the SGTR 00 case and will be retitled 15.4-138.
New UFSAR Table 15.4-13A will be added to list the sequence of events for the revised SGTR MTO analysis.
Note: The only SGTR MTO figure currently in the UFSAR is Figure 15.4.3-7A for SG liquid volume. The corresponding figure for the 00 case is titled 15.4.3-78. This change will add several new figures for the MTO case and relabel the corresponding figures for the 00 case.
27  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 Add new results Figures 15.4.3-1 A through 15.4.3-6A for SGTR MTO analysis. The existing Figure 15.4.3-7A for MTO SG volume will be updated to reflect the revised plot.
Renumber Figures 15.4.3-1 through 15.4.3-6 to 15.4.3-1 8 through 15.4.3-68 to differentiate from comparable SGTR MTO Analysis figure numbers. Revise header text on Figures 15.4.3-1 8 to 15.4.3-78 and 15.4.3.8 through 15.4.3-11 to identify these results are for SGTR 00 analysis.
Summary of Evaluation:
The revised SGTR MTO analysis which no longer credits the TDAFW pump TCOA at 5.54 minutes, and the associated input parameter changes to ECCS and AFW flow only evaluate the plant response to this accident and are not an initiator of a SGTR or any other UFSAR accident. Therefore this proposed activity does not introduce the possibility of a change in the frequency of an accident previously evaluated in the UFSAR.
The AFWs and ECCS SSCs continue to perform and mitigate the SGTR accident in the same manner as described in the UFSAR. The revised SGTR MTO analysis evaluates the same limiting single failures of SSCs important to safety as currently described in the UFSAR. The revised SGTR MTO analysis continues to demonstrate acceptable MTO to ensure there is no liquid relief challenge to the SG atmospheric dump valves (ADVs) or main steam safety valves (MSSVs). Therefore, the revised SGTR MTO analysis does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
Since the revised SGTR MTO analysis still demonstrates that the ruptured SG does not overfill with liquid, the UFSAR 00 results in Table 15.5-71 and control room dose results in Table 15.5-74 have not changed, and continue to
* meet the 10 CFR 100 and General Design Criterion (GDC) 19 limits, respectively. Therefore, the revised SGTR MTO analysis does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The revised SGTR MTO analysis continues to demonstrate acceptable MTO to ensure there is no liquid relief challenge to the SG atmospheric dump valves (ADVs) or main steam safety valves (MSSVs). Since SG overfill does not occur, the SGTR 00 case is not impacted and the limiting single failure of the power operated relief valve to stick open on the ruptured SG remains bounding. Therefore, the UFSAR 00 results in Table 15.5-71 and control room dose results in Table 15.5-74 have not changed, and continue to meet the 10 CFR 100 and GDC 19 limits, respectively. The revised SGTR MTO 28  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 analysis does not result in an increase in the consequence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
The revised SGTR MTO analysis does not alter or place any SSC in a configuration outside design or analysis limits and does not create any new accident scenarios. The proposed activity does not create a possibility for an accident of a different type than previously evaluated in the UFSAR.
The characteristic thermal plant response has not changed for the revised SGTR MTO analysis and the same limiting single failures of SSCs important to safety have been evaluated as currently described in the UFSAR. The revised SGTR MTO analysis continues to demonstrate acceptable MTO to ensure there is no liquid relief challenge to the SG ADVs or MSSVs. The SSCs important to safety will continue to perform their safety function as assumed in the SGTR analysis and will continue to operate within design and analysis limits such that no new system interactions or new failure modes have been created. The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
The SGTR MTO analysis continues to demonstrate that the SGTR accident can be successfully mitigated and the tube rupture flow terminated without challenging the fuel, RCS or SG integrity. The SGTR MTO analysis has been evaluated with respect to the same design basis limits currently described in the UFSAR and no changes to any limits were required to establish acceptable SG overfill margin. Therefore, this proposed activity does not result in a design basis limit for a FPB as described in the UFSAR being altered or exceeded.
The SGTR MTO analysis only involves input parameter changes that do not adversely revise or replace an UFSAR described methodology. Therefore, per NEI 96-07, Section 4.3.8, this 10 CFR 50.59 question is not applicable for this proposed activity.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
10-024     Unit 1 Replace Reactor Coolant Drain Tank (RCDT) Concrete Hatch Covers with Grating Reference Document No.:          I DDP 1000000397, Revision 0 Reference Document

==Title:==
I Install Grating Over RCDT 29 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 Activity Description/Summary of Evaluation:
This Evaluation addresses the replacement of the Containment RCDT concrete hatch covers with a grating to provide a flow path for reactor coolant to the ECCS sump for a postulated RCS pipe break (Ioss-of-coolant (LOCA))
inside the biological shield wall that is consistent with the existing design analysis. This modification corrects a nonconforming condition by providing an adequate coolant flow path to replace the now blocked personnel access opening at the 91 foot containment elevation .
What is being changed This design change makes the following changes to the Unit 1 Containment internal structure:
* Removes the concrete maintenance hatch covers above the RCDT compartment and permanently stores them on the 91 foot elevation close to the biological shield wall south of the maintenance hatch.
* Replaces the concrete maintenance hatch covers with removable stainless steel grating.
* Installs three light gage stainless steel covers over the grating to block airflow during normal plant operation.
Because of these changes, the following effects will result:
* During a postulated LOCA inside the biological shield wall, the coolant will fill the reactor cavity, displace the stainless steel cover plates, and exit through the new steel grating covering the maintenance hatch opening providing a flow path that is consistent with that modeled in the ECCS sump analysis.
* During normal operation, the air flow to the reactor cavity will not be affected since the stainless steel cover plates on top of the maintenance hatch grating preclude any significant bypass flow and maintain the integrity of the existing air flow path.
* The permanent storage of the maintenance hatch concrete covers in Containment will maintain floor loading on the 91 foot elevation within design allowables, adequately addresses seismic interaction considerations, and has no significant impact on flood height and debris transport.
Why are these changes being made To provide a flow path for reactor coolant to reach the ECCS sump following a 30  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 postulated RCS pipe break (inside the biological shield wall) consistent with that modeled in the existing design analysis . This modification corrects a nonconforming condition.
Summary of Evaluation:
10 CFR 50.59 evaluations were performed on the following proposed activities:
Proposed Activity Number 1: Replacement of an unobstructed opening with a grating having a removable cover plate to provide a flow path from the lower reactor cavity to the 91 foot Containment elevation.
Proposed Activity Number 2: Modification of the cavity ventilation supply air flow boundary at the RCDT Maintenance Hatch permitting small amount of bypass flow.
Proposed Activity Number 3: Modification of the cavity ventilation air boundary at the RCDT Maintenance Hatch such that it will not maintain ventilation boundary integrity following a seismic event.
The 10 CFR 50.59 evaluations concluded that:
* The proposed activities do not result in more than a minimal increase in the frequency of occurrence or consequences of an accident.
* The proposed activities do not create a possibility for an accident of a different type.
* The proposed activities do not result in more than a minimal increase in the likelihood of occurrence or consequences of a malfunction.
* The proposed activity does not create the possibility of a malfunction of an SSC important to safety with a different result than previously evaluated in the UFSAR.
* The proposed activities do not result in a DBLFPB as described in the UFSAR being exceeded or altered.
* The proposed activities do not involve any changes to any method of evaluation.
Based on the negative responses in this Evaluation, prior NRC approval is not required to implement this change 10-025     Auxiliary Building Control Board Replacement - Phase 4B Reference Document No.:          I DDP 1000000321, Revision 0 Reference Document

==Title:==
I Auxiliary Board Phase 4B Digital upgrade 31 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 Activity Description/Summary of Evaluation:
This design change (Phase 4B of the Auxiliary Board design) is a digital upgrade and system integration of the Auxiliary Control Board CVCS boards POCV1, POCV2, miscellaneous boards POB1, POB2, PTM01 and associated remote panel instruments to improve equipment reliability and availability.
The upgrade is a continuation of the Auxiliary Building control board panel upgrade and relies upon completion of previous design change phases. The new system will be designed with redundant networking systems and independent power sources. The human factors design of the new digital control system enhances the ability of operators to control and monitor the process. The new control system consolidates needed information onto a display that provides a more effective view of system operation. The new system provides a more time efficient manner for an operator to seek meter readings or other indications.
The Auxiliary Building Control Board for DCPP provides an Operator Station to control the receiving, storage, treatment and discharge of liquid radwaste products generated by both Units 1 and 2. Many systems are common to both units. The systems are constantly in use to reduce and concentrate radioactive waste for off-site disposal or on-site storage.
Many of the panel instruments and controllers are air operated and no longer available. Due to their age and lack of adequate drawings these components are difficult to maintain and troubleshoot. The existing panel configurations and indications are poorly located and labeled. Controls and related indications are not always adjacent to one another, and in some cases the indications only exist remotely in the field. There are several components that were installed for systems that were never made functional and other components are no longer used.
The electrical and instrument control schemes being replaced or modified provide nonsafety-related (Design Class II) functions and are not relied upon for mitigation or safe shutdown. The design change does not change existing alarm and control.
Summary of Evaluation:
The LBIE Screen identified the following adverse activities:
* The new digital control system fundamentally alters the wayan SSC performs or controls a design function. The new digital control system changes the manner in which the system information is presented and controlled by the operator. This represents a change in the Human 32  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 Machine Interface (HMI).
* The ambient air temperature monitoring system will be supplied by two redundant sources of Class II nonvital power whereas UFSAR Section 9.4 states that the ambient air temperature monitoring system is to be supplied by Class IE power.
* The new digital control system represents a change to a procedure that adversely affects how UFSAR described SSC functions are performed or controlled. The new digital control system processes and displays the system parameters differently and the operator uses an HMI to interface and control the system.
The 10 CFR 50 .59 evaluation evaluates the above adverse aspects of the design change.
The new digital control system will be designed with redundant networking systems and independent power sources. The loss of one power supply does not cause a system failure. This provides greater system diversity and reliability. The failure modes for a process high, low or power failure will remain the same as before. The existing system failure modes are not affected. This activity will not cause any new failure modes. This system is not relied upon for accident mitigation and will not cause any new accidents.
The system will not increase the consequence of any accident. This activity does not directly or indirectly interface with a FPB. Therefore all answers are no for the adverse digital control aspects of this design change.
Supplemental Safety Evaluation Report (SSER) 7, Section 7.8, states the Equipment Temperature Monitoring system is of high quality, testable and powered from a reliable power source. Supplying the system with two redundant non-vital power sources does not change or create any new failure modes for this Class II system. This issue has been reviewed and found to meet the commitment for reliable power specified in SSER 7. This system is not relied upon for accident mitigation and will not cause any new accidents.
The system will not increase the consequence of any accident. This activity does not directly or indirectly interface with a FPB. Therefore all answers are no for the adverse power supply aspect of this design change.
The evaluation concludes that these adverse aspects of the design change are acceptable and a license amendment is not required. Prior NRC approval is not required to implement the design change.
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Enclosure 1 PG&E Letter DCL-12-035 10-026      Unit 2 Safety Injection (SI) Test Header Project and Unit 2 SI Travel Stop Project Reference Document No.:            DDP 1000000171, Revision 0 and DDP 1000000367, Revision 0 Reference Document

==Title:==
Provide Isolation Valves for SI Test Line Valves and U2 SI Test Valve Travel Stops Activity Description/Summary of Evaluation:
This LBIE Screen addresses the installation of valve stem travel stops as provided by this modification package issued for apparent cause evaluation (ACE) corrective action on travel stop deficiency. This LBIE also addresses the entire scope of the Unit 2 SI test header project implemented during Unit 2 fifteenth refueling outage in order to complete an ACE corrective action related to LBIE deficiencies.
What is being changed This DCP makes changes to the Unit 2 piping and valve network that is used for performing RCS PIV/ECCS check valve leak tests. This "SI test header" is located primarily inside Containment. The changes include:
* Remove 19 AOVs and 6, 3/4 inch diameter manual valves;
* Modify 3/4 inch diameter test header piping and its supports inside Containment;
* Reclassify a portion, of the test flow collection header from Piping Code Class C to Piping Code Class E;
* Modify Containment penetration piping for penetrations 24, 25, 33, 34, 51 B, and 75 (inside Containment) and 51 B (outside Containment);
* Install and connect four "test stations" at elevation 91 foot inside Containment to the SI test header piping;
* Add 56, 3/4 inch diameter manual, globe valves in the reconfigured test header piping system (one is outside Containment in penetration 51 B piping);
* Install a permanent "travel stop" on the stems of 13 of the new manual, globe valves to limit the opening of these RCPB valves.
* Modify Control Room 2VB1 to remove nineteen control switches, determinate the related wiring, remove related fuses; locally relocate three, other, 2VB 1 control switches;
* Modify Containment electrical penetrations 2E (2E-IN) and 11 E (11 E-IN) associated with the wiring for the nineteen AOVs; I~
* Modify the contents of six mechanical panels inside Containment (e.g.,
34 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 determinate wiring, remove SVs, modify instrument air tubing and components); and
* Modify the location where the leak-test valves are controlled (AOV positions were controlled at control switches on Control Room Vertical Board 2VB 1; new manual valve positions are controlled locally inside the Containment).
How does it differ from the present condition
* Repeated AOV seat leaks have occurred during PIV/ECCS check valve seat leak testing. These existing, Control Room-operated AOVs in the Sl test header inside Containment are replaced with manual isolation valves.
* Any (potential) leakage from the RCPB through the test lines will be controlled by restricting the opening of the new, manual valves using valve stem travel stops instead of internal ports provided in the AOVs.
* Testing of the PIVs will be restricted to plant operating Modes 4 and 5.
* In Mode 4, only one RCPB Sl test flow path may be established at a time. That is, only one pair of RCPB manual valves may be open for testing at one time.
* In Mode 5, up to four RCPB Sl test flow paths may be established simultaneously by opening associated RCPB manual valves.
* How the leak-test valves are controlled is changed. Currently, each AOV position is controlled by a control switch with position indication on Control Room vertical board 2VB 1 via electrical circuits, a SV, and an air supply. Whereas in the modified design, each new manual valve position is controlled locally inside the Containment via aT-handle operator.
* Additional, manual isolation valves are added at strategic locations in the Sl test header piping to have added assurance that "voids" cannot collect/migrate in test header piping and accumulate in ECCS piping.
* One of the alternate flow paths for accumulator fill in the management of severe accidents is eliminated.
Why are these changes being made
* To assure more reliable test boundaries for RCS PIV/ECCS check valve leakage testing (in part by replacing each AOV with normally-closed, manual isolation valve(s) - Note: RCPB AOVs are replaced with 2 manual valves. Other Sl Test AOVs are replaced with a single manual valve);
* To restrict the opening of the new, manual valves using valve stem travel stops, ensuring that the flow through the valve remains below the threshold of the small reactor coolant pipe break analysis; 35  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035
* To configure the plant, by using the four, new test stations, for optimized testing performance with other shutdown activities and added flexibility for the accumulator filling process;
* To comply with DCPP commitments to NRC Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and CS Systems," to reduce the probability of ECCS gas intrusion and "voids" formation and collection;
* The downgrade of Class C piping to Class E was made for economic reasons, since it was determined that maintaining its classification was unnecessary with the new, manual valves;
* To eliminate maintenance/spare parts/testing of associated removed components.
Summary of Evaluation:
An in-Containment test header is used to periodically test the RCPB PIVs and other ECCS check valves for seat leakage. Previously air-operated globe valves controlled from the Control Room Vertical Board were used.
These air-operated globe valves were replaced with local, manual test/isolation valves. With the installation and use of the manual valves for future testing, all testing will only be performed in plant Modes 4 and 5. The 10 CFR 50.59 evaluation concluded that:
* The interfaces of this test header and its valves with the RCPB and Containment Isolation System and SI accumulator fill function were evaluated and found acceptable.
* The local manipulation of the new valves inside Containment for these tests was determined to be acceptable.
* The use of valve stem travel stops for restricting flow through the RCPB valves was also found acceptable.
* The reclassification of a portion of the test header outboard of the normally-closed RCPB test valves from Piping Code Class C to Piping Code Class E was determined to be acceptable.
The proposed activity does not result in a more than minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.
The proposed activity does not result in a more than minimal increase in the likelihood of occurrence of a malfunction previously evaluated in the UFSAR.
The proposed activity does not result in a more than minimal increase in the consequences of an accident or the malfunction of an SSC important to safety previously evaluated in the UFSAR.
36  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 The proposed activity does not create the possibility for an accident of a different type than any previously evaluated .in the UFSAR.
The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
The proposed activity does not result in a DBL for a FPB being exceeded or altered.
The proposed activity does not involve a method of evaluation.
Based on the negative responses in this Evaluation, prior NRC approval is not required to implement this change.
11-003      Blast and Bullet Resistant Enclosure (BBRE) Guard Towers Reference Document No.:            DDP 1000000403, Revision 0 Reference Document

==Title:==
Force on Force (FOF) Mods Install BBRE Towers 1 thru 6 Activity Description/Summary of Evaluation:
Six BBRE guard towers are being installed at strategic locations within the protected area perimeter to enhance the plant's security system. The BBRE is a heavy, armored structure mounted on the tower 25 - 65 ft above the foundation. These changes are evaluated under 10 CFR 50.59 and 10 CFR 72.48 for impacts to SSCs important to safety.
Summary of Evaluation:
The BBRE guard towers are not initiators of any UFSAR described events and therefore have no impact on the frequency of occurrence of accidents previously evaluated in the DCPP or ISFSI UFSAR.
The guard towers do not introduce the possibility of a change in the likelihood of a malfunction of an SSC important to safety because their installation is not an initiator of any malfunctions and no new failure modes associated with important to safety SSCs are introduced. Therefore, there is no effect on the likelihood of occurrence of a malfunction previously evaluated in the UFSAR.
Installation of the BBRE guard towers does not result in more than a minimal increase in consequences of an accident previously evaluated in the DCPP 37 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 or ISFSI UFSARs. Guard Towers 3, 4, and 5 are not located in proximity to SSCs that are important to safety or that can result in any consequences in the event of tower collapse. Guard Towers 1 and 6 are designed to withstand the design basis seismic and tornado events (will remain intact and not collapse) so that important to safety SSCs in proximity are not affected and there are no consequences. Guard Tower 2 is located in proximity to the Radwaste Storage Building. The building is a Design Class II structure and is itself not qualified for such events beyond California Building 'Code requirements . Consequences due to collapse of all or part of the building are enveloped by previously evaluated events (e.g., off-gas decay tank rupture, dropping a spent fuel bundle). Tornado generated missiles (from the guard tower or stairway structures) are bounded by the DCPP and Diablo Canyon ISFSI UFSAR described missiles and do not result in consequences.
Installation of BBRE Guard Towers 1 - 6 does not introduce the possibility of a change in the consequences of a malfunction of an SSC important to safety because the BBRE guard towers are not initiators of any malfunctions and no new failure modes are introduced.
No possibility of an accident of a different type is introduced because installation of BBRE Guard Towers 1 - 6 does not create an initiator of any accident and no new failure modes are introduced.
The installation of BBRE Guard Towers does not introduce the possibility for a malfunction of an SSC with a different result because the activity does not introduce a new failure mode. Therefore, the proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result from any previously evaluated in the FSARU.
The proposed activity does not involve UFSAR-described design basis limits for FPBs, so none are exceeded or altered.
The proposed activity also does not involve any UFSAR-described changes to methods of evaluation, so there are no departures from methods of evaluation.
Based on this evaluation, prior NRC approval is not required to implement this activity.
11-005     Unit 2 Replace RCOT Concrete Hatch Covers with Grating Reference Document No.:          I DDP 1000000453 Rev 0 Reference Document

==Title:==
I Install Grating at RCDT Hatch 38 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Contains Security-Related Information ---:- Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 Activity Description/Summary of Evaluation:
This LBIE addresses the replacement of the Units 1 and 2 Containment RCDT concrete hatch covers with grating to provide a flow path for reactor coolant to the ECCS sump for a postulated LOCA inside the biological shield wall that is consistent with the existing design analysis. These modifications correct a nonconforming condition identified in the DCPP corrective action program (CAP) by providing a coolant flow path to replace the now-blocked personnel access opening at the 91 foot containment elevation.
Notes:
: 1. The DCPP CAP identifies that the LBIE for DDP 1000000397, which modified the Unit 1 Containment RCDT concrete 'hatch covers during 1R 16, was deficient in addressing the reactor cavity sub-compartment pressurization analysis.
: 2. The DCPP CAP identifies that the UFSAR Chapter 6.2 description of the subcompartment pressurization analysis for the lower reactor cavity does not correctly reflect an AOR for Units 1 and 2. The license basis for subcompartment pressurization consists of two analyses; an initial analysis 2
described in detail in the UFSAR consisting of 426 and 369 square inch ( in )
2 break sizes, and an asymmetric loading analysis constituting 115 and 76 in break sizes provided by the AOR. The latter analysis was specifically performed to address NRC concerns over asymmetric loading, and the approved plant design provided pipe displacement restraints to limit the 2
break sizes to 115 and 76 in . However, the larger break analysis remained 2
as a bounding analysis in the UFSAR. The UFSAR 426 and 369 in break analysis did not take credit for the RCDT hatch as a vent opening; however, the AOR did take credit.
Enhancements to the UFSAR descriptions are not be addressed by this modification. However, since the RCDT hatch opening is not credited in the 2
UFSAR 426 and 369 in break analysis, and the design change does not adversely affect the vent opening area credited in the AOR (or the results of the analysis), the evaluation of this modification under the 10 CFR 50.59 process was found to be acceptable without requiring closure of all related corrective actions.
What is being changed These DCPs make the following changes to the Units 1 and 2 Containment internal structures:
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Enclosure 1 PG&E Letter DCL-12-035
* Remove the concrete maintenance hatch covers above the RCDT compartment and permanently stores them on the 91 foot elevation close to the biological shield wall south of the maintenance hatch.
* Replace the concrete maintenance hatch covers with removable stainless steel grating.
* Install three light gage stainless steel covers over the grating to block airflow during normal plant operation.
Why are these changes being made The concrete hatch covers are replaced with grating to provide a flow path for reactor coolant and refueling water storage tank (RWST) water to reach the ECCS sump from the Reactor Cavity following a postulated RCS pipe break (inside the biological shield wall) that is consistent with the flow path assumed in the existing design analyses. This modification corrects a nonconforming condition documented in the DCPP CAP.
The steel covers are installed so that during normal plant operation the air flow to the reactor cavity will not be significantly affected since bypass flow through the maintenanc~ hatch grating will be minimized. The steel covers are not fixed to the grating, but tethered to the grating to permit the up-swell of water following a LOCA to push the covers out of the flow path. The tethers limit the steel covers travel so that they can not impact other equipment.
Summary of Evaluation:
10 CFR 50.59 evaluations were performed on the following proposed activities:
Proposed Activity Number 1, Replacement of an unobstructed opening with a grating having a removable cover plate to provide a flow path from the lower reactor cavity to the 91'-0" containment elevation.
Proposed Activity Number 2, Modification of the cavity ventilation supply air flow boundary at the RCDT maintenance hatch permitting small amount of bypass flow.
Proposed Activity Number 3, Modification of the cavity ventilation air boundary at the RCDT maintenance hatch such that it will not maintain ventilation boundary integrity following a seismic event.
The 10 CFR 50.59 evaluations concluded that:
* The proposed activities do not result in more than a minimal increase 40  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 in the frequency of occurrence or consequences of an accident.
* The proposed activities do not create a possibility for an accident of a differe nt type .
* The proposed activities do not result in more than a minimal increase in the likelihood of occurrence or consequences of a malfunction.
* The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in FSARU.
* The proposed activities do not result in a DBLFPB as described in the U FSAR being exceeded or altered.
* The proposed activities do not involve any changes to any method of evaluation.
Based on the negative responses in this evaluation, prior NRC approval is not required to implement this change.
11-006      230 kV System Dual Unit Trip Licensing and Design Bases Change (document only change)
Reference Document No.:           I DDP 1000000470, Revision 0 Reference Document

==Title:==
I Change License Basis Dual Unit Trip Activity Description/Summary of Evaluation:
What is being changed:
Change license and design bases of 230 kV system/immediate offsite power circuit. The term "Startup offsite power circuit" is used . This design change documents that the Startup offsite power circuit has the capacity to operate the ESF for dual unit trips and the shared components have adequate capacity to preclude overload and the subsequent loss of function for the worst case loading condition. The specific scenarios evaluated are listed in UFSAR, Section 8.2.2.1 for clarification and completeness.
Other UFSAR changes: (1) clarify in the UFSAR Section 8.2.1.1 that cross-tie of the startup busses makes both Units 1 and 2 Startup offsite power circuit inoperable, (2) change Section 8.2.2.1 "Operability is based on the ability to transfer to the 230 kV system without loading the EDGs and provide adequate voltage to safety-related loads" to "Operability is based on maintaining the minimum voltage requirements of the onsite Class 1E electrical distribution system (See UFSAR Section 8.3.1.1.8.2(1)(a))," (3) add a clarification to Section 8.2.2.2 (editorial in nature), (4) add references to Section 8.2.3, and (5) change the Transmission Operations Center (TOC) to what it is now called, the Grid Control Center (GCC) .
41 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 TS Bases change: The TS Bases describe the Startup offsite power circuit (in the TS Bases, the te rm "Offsite Circu it #1" is used). The shared components in the 230 kV switchyard are not listed . Since the capacity of these components is analyzed by DCPP, it is prudent to list them.
Why is this change made The NRC has said in correspondence dated December 14, 2009, and July 7, 2010, that a dual unit trip is the DCPP design bases. They also took exception to the wording, that the 230 kV system has "sufficient capacity and capability to operate the ESF for a design bases accident (or unit trip) on one unit, and those system required for an orderly shutdown of the second unit."
The term "orderly shutdown of the second unit" is changed to "concurrent safe shutdown of the second unit" per this change. This clarifies that the intent is to use the preferred 230 kV system for shutdown of the second unit versus the delayed 500 kV system for the second unit that the NRC said was implied by the term "orderly shutdown."
The NRC also disagreed with the statement, "Operability is based on the ability to transfer to the 230 kV system without loading the EDGs, and provide adequate voltage to safety-related loads."
Summary of Evaluation:
This change does not increase the frequency of Io.ss of offsite power events.
This change makes no hardware change to the Plant and is not an initiator of a loss of offsite power. The proposed activity does not result in a more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSARU.
The proposed activity does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
This change does not impact the consequences of any accident.
Since there is no impact on the availability of the startup offsite power circuit per this design change, there is no affect on the radiological consequences of a malfunction previously evaluated.
This change to the startup offsite power circuit does not create the possibility of an accident of a different type. There are no new accident initiators per this design change.
42  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 The failure mechanisms of the startup offsite power circuit are not changed and there is no adverse impact of this change on the important to safety function of the EDGs. Therefore, the proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR.
This design change does not alter or exceed any FPB value and does not alter any associated system parameter.
No method of evaluation is impacted by this design change.
The 10 CFR 50.59 evaluation shows that prior NRC approval is not required in order to implement the proposed activity.
11-007      Unit 1 and Unit 2 First Level Undervoltage Load Shed Relay Setpoint Change Reference Document No.:          I SAPO 60024240 Reference Document

==Title:==
I Setpoint Change to 27HFT1 & 2 U1 Activity Description/Summary of Evaluation:
This evaluation is being performed to correct the deficiency identified by NRC (IR 05000275/2010007 and 05000323/2010007) in the screen that supported the Unit 1 and Unit 2 first level undervoltage load shed relay.
What is being changed:
The FLUR load shed relay setpoints are input parameters to engineering calculation. These setpoints associated with the ESF load shed are temporarily raised as a compensatory measure such that upon loss or degraded 4 kV bus voltage, the ESF loads will be shed from the class 1E 4 kV bus in preparation to transfer to EDGs at higher voltage and in shorter time delay. This change is being made to protect ESF equipment and prevent the loss of their ESF function. The following setpoint change description mentions First Level Protection and Second Level Protection.
This is not to specifically imply FLURs and SLURs. Rather first level protection is protection against loss of voltage and second level protection is protection against degraded voltage. The old and new setpoints are:
A. 27HxT1 Time Delayed Undervoltage Relay The time delayed undervoltage protection relay (27HxT1) operates 43 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 using an internal inverse time/voltage relationship that is set using two device settings.
The old 27HxT1 device settings were:
Pickup Voltage: 85 Vac (2972 Vac at the bus)
Time Dial Setting: 25 The new 27HxT1 device settings are:
Pickup Voltage: 106 Vac (3707 Vac at the bus)
Time Dial Setting: 22 The following demonstrates the old and new time delayed relay settings against the TS values:
First Level Protection Old Setpoint - First Level Protection: At 0 percent bus voltage trip in less than 3.95 sec (2.98 sec setpoint + 0.97 sec uncertainty)
New Setpoint- First Level Protection: At 0 percent bus voltage trip in less than 2.96 sec (2.24 sec setpoint + 0.72 sec uncertainty)
Second Level Protection Old Setpoint - Second Level Protection: At 2583 Vac (62.1 percent) bus voltage trip in less than 17.5 sec (16.43 sec setpoint + 0.97 sec uncertainty)
New Setpoint- Second Level Protection: At 2583 Vac (62.1 percent) bus voltage trip in less than 6.6 sec (5.87 sec setpoint + 0.72 sec uncertainty)
B. 27HxT2 Instantaneous Undervoltage Relay Old Setpoint - First and Second Level Protection: Trip instanteously at voltages equal to or greater than 2924 Vac (70.3 percent) bus voltage (2971 setpoint - 47 V uncertainty)
New Setpoint - First and Second Level Protection: Trip instanteously at voltages equal to or greater than 3330 Vac (80.8 percent) bus voltage (3408 setpoint - 78 V uncertainty)
A calculation has been prepared in order to demonstrate that the new 44  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 undervoltage relay setpoints allow the relays to meet their design function.
In the previous screen, this calculation was considered to be an engineering calculation. In this screen, this calculation is considered a safety analysis per NEI 96-07 (definition 3.12). In addition to the new setpoints, the calculation evaluates new events not formerly part of the DCPP design and licensing basis. The evaluation for this particular change is excluded from the scope of this LBIE screen and is instead addressed in the LBIE screen and Evaluation for DDP 1000000470, Revision o.
Why is this being changed:
DCPP CAP identified that with pre T-Mod setpoints, under certain postulated scenarios, the safety function of a number of systems designed to respond to a design basis accident will be challenged. The FLUR setpoints are being raised by aT-Mod as a compensatory measure to protect the safety-related equipment from sustained degraded voltage on 4 kV vital bus.
Summary of Evaluation:
The LBI E was performed because three adverse impacts were identified in the LBIE screen. Those three adverse impacts and the summary of the evaluation is as follows:
(1) Moving the T1 and T2 setpoint in the direction of separation from offsite power.
(2) The higher setpoint may cause higher actuation of the relay and accelerated mechanical fatigue causing component failure.
(3) Engineering calculation is considered a safety analysis.
The undervoltage protection relay is not the initiator of an accident.
Therefore, the change in the undervoltage protection setpoint will not increase the frequency of the accidents previously evaluated in the UFSAR.
By increasing the voltage setpoints on T1 and T2 relays, the potential exists that they may actuate more frequently and fail due to mechanical fatigue.
Analysis shows a 0.6 percent increase in the number of cycles as compared to the expected relay actuations. According to NEI 96-07 Revision 1, Section 4.3.2, the threshold of minimal increase in likelihood of occurrence of a malfunction is a factor of two (Example 8). Since, the reduction in life of the relay is 0.6 percent (which is less than the threshold of a factor of two); the increase in probability of malfunction of the relay is not more than minimal.
An untimely early separation (in terms of voltage) from offsite power by 45  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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This page is decontrolled when separated from Enclosure 2.
Enclosure 1 PG&E Letter DCL-12-035 undervoltage protection is another failure mode of these relays. Analysis has
* concluded that with worst case instrument uncertainty and variation on 230 kV system, the new setpoint will not cause unnecessary separation of ESF loads from 230 kV system.
The change in the undervoltage protection setpoint will not affect the transfer time of the ESF loads to the EDGs as assumed in the accident analysis and dose calculations. Therefore, this change will not affect the consequences of design basis accidents previously evaluated in UFSAR.
The undervoltage protection setpoint was changed to protect the ESF equipment. Therefore, it will not affect the failure of the emergency safeguard system and the consequences of their failure. The failure of relay to shed loads is bounded by a single failure assumed in the UFSAR 6.3.3.2.12. Therefore, there is no increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
The undervoltage protection relays and their setpoints are not the initiator of any accident of a different type than any previously evaluated in the UFSAR.
The change in undervoltage protection setpoint will not create the possibility for an accident different than those previously evaluated in UFSAR.
A single failure of the undervoltage protection will not create a malfunction with a different result that was not previously evaluated in UFSAR. Analysis has concluded that with worst case instrument uncertainty and variation on 230 kV system, the new setpoint will not cause unnecessary separation of ESF loads from 230 kV system. The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
The change in the undervoltage protection setpoint does not involve an SSC that could alter the design basis limits for the FPBs. Any failure in the undervoltage protection is bounded by the existing accident analysis demonstrating the design basis limits for FPB is not exceeded.
The proposed activity does not involve a change to the method of evaluation previously described in the UFSAR.
The LBIE concludes that prior NRC approval is not necessary for the T-MOD implementation.
111-008     POA Update per SAPN 50301167-7 for MMD M000084 Unit 2 Cycle 46 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 Reload Evaluation Reference Document No.:          I MMD M000084,     Revision 5 Reference Document

==Title:==
I Activity Description/Summary of Evaluation:
What is being changed This proposed activity evaluates increased ESF delay times associated with the FLURISLUR nonconforming condition per the POA for potential impact on the UFSAR accident analyses that support the Unit 2 Cycle 17 core reload design change. The Westinghouse evaluations of the increased ESF delay times documented in PGE-10-54 are not being incorporated into the DCPP design basis at this time but provide the technical basis for future updates and eventual resolution of the FLURISLUR nonconforming condition. Therefore, only those evaluations of the nonconforming condition that support the Unit 2 Cycle 17 core reload must be reviewed per 10 CFR 50.59 to ensure they are consistent with the approved methodologies listed in TS 5.6.5.b that are used to verify the Unit 2 Cycle 17 COLR limits remain bounded.
The T-Mod to the FLUR implemented as a compensatory measure per the POA only ensures no ESF equipment is damaged due to degraded voltage and does not impact any ESF delay times or core reload analyses.
Therefore, the FLUR T-Mod is not within the scope of this LBIE review.
The increased ESF delay times that bound the nonconforming FLURISLUR condition associated with a 230 kV degraded voltage event and that are being evaluated in support of Unit 2 Cycle 17 core reload are summarized below:
Increased ESF Delay Times to Bound 230 kV Degraded Voltage Event ECCS Injection Flow                                           27 sec to 42 sec CFCU Heat Removal                                            48 sec to 52 sec AFW Flow                                                       60 sec to 65 sec CS Flow                                                       80 sec to 100 sec The current ESF delay times listed are with respect to the ESF initiating signal credited for accident mitigation in the DCPP safety analysis. The increased ESF delay times listed for the 230 kV degraded voltage event are with respect to reaching a SI setpoint which initiates the automatic transfer of onsite power from the 500 kV to the 230 kV system and results in the additional SLUR delay time occurring as established in the DIT.
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Enclosure 1 PG&E Letter DCL-12-035 Summary of Evaluation:
The increased ESF delay times are not physically related to the cause of any accident and do not introduce the possibility of a change in the frequency of an accident previously evaluated in the UFSAR.
All ESF functions will continue to perform in the same manner as described in the UFSAR. Therefore, the increased ESF delay times do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
Since the applicable UFSAR accident analyses which support the Unit 2 Cycle 17 core reload continue to meet all applicable fuel design limits, RCS overpressure limits, containment leakage limits, secondary mass release assumptions, and 10 CFR 50.46 criteria, there is no impact on the calculated ODs or control room doses in UFSAR Section 15.5. Therefore, the increased ESF delay times do not result in more that a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The limiting single failures of the SSCs assumed in the UFSAR accident analyses which support the Unit 2 Cycle 17 core reload and which form the basis for the calculated ODs and control room doses in Section 15.5 remain bounding and are not impacted. Therefore, the increased ESF delay times do not result in an increase in the consequence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
The increased ESF delay times do not alter or place any SSC in a configuration outside design or analysis limits and do not create any new accident scenarios. The proposed activity does not create a possibility for an accident of a different type than previously evaluated in the UFSAR.
The increased ESF delay times do not change the characteristic thermal plant response of any accident or assumed SSC malfunction as assumed in the UFSAR. The SSCs important to safety will continue to perform their safety function as assumed in the safety analysis and will continue to operate within design and analysis limits such that no new system interactions or new failure modes have been created. The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
The UFSAR core reload accident analyses continue to meet the applicable fuel design limits (DNBR, PCT), RCS overpressure limits, and containment integrity limits as described in the UFSAR. All evaluations have been performed with respect to these current UFSAR design basis limits and no 48  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 changes to any limits are required to establish that the Unit 1 Cycle 17 core limits remain bounded. Therefore, this proposed activity does not result in a design basis limit for a FPB as described in the UFSAR being altered or exceeded.
Per LBIE screen response 2.c, the increased ESF delay times evaluated in support of Unit 2 Cycle 17 core reload are input parameter changes that do not adversely revise or replace an UFSAR described methodology.
Therefore, this 50.59 question is not applicable for this proposed activity.
11-009      Replacement of LBIE Review - Gross Failed Fuel Detector (GFFD)
Removal Reference Document No.:            DDP N049369, Revision 0 and DDP N050369, Revision 0 Reference Document

==Title:==
Elimination of Gross Failed Fuel Monitor Activity Description/Summary of Evaluation:
The LBIE review for the removal of the GFFD per the design change process circa 1998 has been reperformed in response to questions raised as a part of the corrective action process during the License Basis Verification Project.
The design function of the GFFD was to continuously monitor primary coolant neutron radiation to detect failed fuel (i.e., cladding breach) and provide indication via control room annunciation and recorder (count rate in cpm). [UFSAR Sections 7.7.1.11 and 9.3.5 (Revision 10) and Figure 7.7-15 (Revision 6)].
Summary of Evaluation:
The methods to detect fuel clad failure included the GFFD in combination with other means (e.g., RCS sample analysis). The GFFD and the means of detection remaining following its removal (i.e., the change):
a) Perform the function in response to an accident, so the proposed changes can not be an initiator of UFSAR described accidents or have an effect on frequency of occurrence of UFSAR described accidents.
b) Are for monitoring fuel clad failure only, so the proposed changes are not initiators of related malfunctions or introduce new failure modes.
Hence, there are no impacts on the likelihood of occurrence of UFSAR described malfunctions.
c) Have no automatic safety or control functions and none of the 49 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 accidents that result in fuel clad failure credit operator response to a related annunciation for mitigating consequences, so there is no more than a minimal increase on consequences of UFSAR evaluated accidents due to the proposed changes.
d) Involve only monitoring functions and are not involved in causing UFSAR described malfunctions. Hence, there is no more than a minimal increase on consequences of UFSAR malfunctions due to the proposed changes.
e) Are not initiators of accidents, do not introduce any new failure modes nor affect existing failure modes, so no accidents of a different type or malfunctions with a different result are created due to the proposed changes.
f)  Do not affect any system parameter related to DBLFPBs, so no DBLFPBs are altered or exceeded due to the proposed changes.
g) Do not affect UFSAR described methods of evaluation, so the proposed changes do not result in a departure from a method of evaluation.
The 10 CFR 50.59 evaluation concludes that prior NRC approval is not required for removal of the GFFD.
11-010     Replace Unit 1 7100 Process Controls System (PCS)
Reference Document No.:           loop 1000000237, Revision 0 Reference Document

==Title:==
I Replace Unit 1 7100 PCS Activity Description/Summary of Evaluation:
The PCS converts plant parameters such as temperature, pressure, level and flow into electrical signals. The electrical signals then undergo signal conditioning, processing and setpoint measurement. Equipment control signals are generated to operate system equipment such as pumps, valves and heaters. System indications are provided in the main control room, hot shutdown panel and Plant Process Computer (PPC). Alarm input signals are provided to the main annunciator system .
This design change replaces the Westinghouse 7100 PCS equipment with a new programmable logic controller (PLC) based system that is reliable and easily maintained. The design of the new PCS is in accordance with the Conceptual Design Document (COD) and Functional Requirement 50 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 Specifications (FRS).
Summary of Evaluation:
Digital upgrade elements were evaluated per 10 CFR 50.59 using the guidance of NEI 01-01, Guideline on Licensing Digital Upgrades TR-1 02348, Revision 1. The 10 CFR 50.59 evaluation concludes the following:
: 1. Based on the evaluation including a qualitative assessment and Failure Modes and Effects Analysis (FMEA) performed for the design change, the design change does not more than minimally increase the frequency of an accident previously evaluated in the UFSAR.
: 2. Based on the evaluation including a qualitative assessment and FMEA performed for the design change, the design change does not more than minimally increase the likelihood of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
: 3. The evaluation concludes replacing the PCS does not alter the performance of systems used to mitigate the consequences of an accident, therefore the design change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
: 4. The evaluation concludes replacing the PCS will not result in a malfunction of SSCs used to mitigate the consequences of an accident. Therefore, the design change does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
: 5. The evaluation concludes all PCS accident initiators are bounded by the accidents previously evaluated in the UFSAR. Therefore, the design change does not create the possibility for an accident of a different type than any previously evaluated in the UFSAR.
: 6. Based on the evaluation including a qualitative assessment and the FMEA performed for the design change, the design change does not create the possibility for a malfunction of and SSC important to safety with a different result than previously evaluated in the UFSAR.
: 7. The evaluation concluded none of the systems or parameters controlled or indicated by the PCS affect a design basis limit for a FPB. There are no systems or controlling parameter numerical values changed that were used in the UFSAR to determine the integrity of a FPB. Therefore, the design change does not result in design basis 51  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled w hen separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 limit for a FPB as described in the UFSAR being exceeded or altered.
: 8. The evaluation concluded the design change to replace the PCS with a PLC based system does not involve a method of evaluation as defined in the UFSAR. Therefore, the design change does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design basis or in the safety analysis.
The 10 CFR 50.59 evaluation concludes that the design change may be implemented without prior NRC approval.
11-013      Component Cooling Water (CCW) Inventory Control Procedures Reference Document No.:            PEP M-246, Revision 6; OP F-2:111 Unit 1, Revision 28; OP F-2:111 Unit 2, Revision 24; STPV-18S Unit 1, Revision 6; and STPV-18s Unit 2, Revision 2 Reference Document

==Title:==
Feed and Bleed of the CCW System, Component Cooling Water System - Shutdown and Clearing, and Nonintrusive Test of MU-1-971 Activity Description/Summary of Evaluation:
Evaluate procedures PEP M-246, STP V-18S and OP F-2:111 for the steps to manually drain and fill the CCW system while it is fully operational. The CCW system has no convenient designed-in removal capability of excess inventory that is above the systems' water return headers to the CCW pumps. Draining and filling will impact both trains simultaneously.
Additionally two of these procedures will also manually bypass the normal automatic makeup valves to the CCW surge tank. This LBIE will determine if crediting operator actions to control the manual draining and filling are acceptable.
Summary of Evaluation:
Loss of the CCW system is not a UFSAR evaluated accident, so there is no increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.
Both a nonmechanistic leak of 200 gpm and the overfill of the CCW system are malfunctions of the CCW described in the UFSAR. Manual operator actions to keep the CCW system full are credited in accordance with 52 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.

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Enclosure 1 PG&E Letter DCL-12-035 Information Notice 97-78/American National Standard Institute 58.8 and considered acceptable. There is no more than a minimal increase in the likelihood of occurrence of a malfunction of the CCW system.
The CCW system draining and filling has no impact on the consequence of any accident previously evaluated in the UFSAR.
The CCW system draining and filling has no more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
Draining and filling the CCW system while operating in the normal surge tank level band of the CCW surge tank does not create the possibility of a different result in the malfunction of the CCW system.
The CCW system malfunctions of a 200 gpm leak and a 250 gpm in-leakage have been evaluated in the UFSAR. These activities are within the bounds of the UFSAR malfunctions and therefore there is no possibility of a malfunction with a different result.
These activities do not affect the design basis limit for a FPB.
These activities do not involve a method of evaluation.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
11-014     Residual Heat Removal (RHR) System Pressurization Due to Small Break Loss of Coolant Accident (SBLOCA)
Reference Document No.:            Calculation 9000019736, Rev. 0 Reference Document

==Title:==
STA-220 RHR System Pressurization Due to SBLOCA Activity Description/Summary of Evaluation:
What is being changed This proposed activity establishes a new post-LOCA design condition determined in calculation STA-220 that the RHR pumps should not run on recirculation flow for more than,30 minutes without initiating CCW cooling to the RHR heat exchangers (HX). This requirement resulted in the addition of a TCOA as credited in calculation STA-220. STA-220 credits this TCOA to establish the maximum RHR system heatup and pressurization that could 53 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 occur during certain SBLOCA scenarios in which the RCS pressure remains above the RHR pump shut off head.
This proposed activity includes the revision to Procedures EOP E-1 (Revision 28 for Unit 1 and Revision 20 for Unit 2) which implement the new TCOA by adding the following caution statement before Step 8 of EOP E-1 as follows:
CAUTION: An RHR Pump should not be run at shutoff head for more than 30 minutes without CCW cooling to its RHR heat exchanger.
Why is it being changed DCPP CAP identified that the LBIE screens performed for design calculation STA-220 and the associated procedure change to EOP E-1 were inadequate since they implicitly credited a new operator action that was not evaluated per 10 CFR 50.59 . This evaluation per 10 CFR 50.59 is being retroactively performed for the new TCOA 10 as implemented in STA-220 and EOP E-1.
Summary of Evaluation:
The SBLOCA is a Condition III event, which is classified as an infrequent fault as described in the UFSAR (Section 15.3). The potential operator action to establish CCW flow to the RHR heat exchanger can only occur after a SBLOCA and is not an initiator of a SBLOCA or any other accident.
Therefore, this proposed activity does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.
The SSC malfunction important to safety for the SBLOCA as described in UFSAR Section 6.2 is the failure of one train of ECCS injection flow. The credited operator actions ensure that the RHR system pressure remains within the design capability of the 8982A1B valves. Therefore, there is no impact on the likelihood of one train of ECCS failing to start or the likelihood of the sump recirculation valves failing to open as detailed in the UFSAR Table 6.3A-1. Per the guidance in NEI 96-07 Section 4.3.2, this new time critical operator action has been evaluated with respect to successful completion to ensure that it does not result in any increase in the likelihood of occurrence of a malfunction of an SSC important to safety. Therefore, it does not result in any increase in the likelihood of occurrence of a malfunction of an SSC important to safety than previously evaluated in the UFSAR This proposed activity ensures the RHR system performance during a 54  Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DGL-12-035 SBLOGA remains as described in the UFSAR. There is no impact on post-LOGA EGGS recirculation leakage, SBLOGA long term core heatup, or any other EGGS design function that could impact the SBLOGA radiological consequences presented in the UFSAR. Therefore, this proposed activity does not result in an increase to the consequences of an accident previously evaluated in the UFSAR.
The UFSAR Section 15.5.11 radiological consequences for a SBLOGA assume fuel failures that bound the limiting single failure of one EGGS train during the injection and recirculation phases. This proposed activity and new operator action ensure that the EGGS continues to perform within the design constraints described in the UFSAR such that the radiological consequences associated with the failure of one EGGS train are not impacted and remain bounding. Therefore, this proposed activity does not result in an increase in the consequences of a malfunction of an SSG important to safety previously evaluated in the UFSAR.
This proposed activity adds a similar cautionary note and potential operator action that already exists in the EOPs and does not alter or place any SSG in a configuration outside design or analysis limits and does not create any new accident scenarios. Therefore, this proposed activity does not create an accident of a different type than any previously evaluated in the UFSAR.
This proposed activity and credited operator action ensure that the RHR system pressure remains within the design capability of the 8982A1B valves, such that there is no impact on the SSG malfunction evaluations or results described in UFSAR Table 6.3A-1. The 8982A1B valves are within the scope of the Generic Letter (GL) 89-10 MOV program, which establishes that the basis for the maximum MOV differential design pressure does not to consider the potential for valve mispositioning due to operator error. The EGGS SSGs important to safety will continue to perform their safety function as described in the UFSAR SBLOGA analysis and will continue to operate within design and analysis limits such that no new system interactions or new failure modes have been created. Therefore, this proposed activity does not create a possibility of a malfunction of an SSG with a different result than previously evaluated.
The SBLOGA analysis results presented in the UFSAR continue to be bounding such that there is no impact on the FPBs associated with the RGS pressure boundary and fuel integrity. Therefore, this proposed activity does not result in exceeding or altering the design basis limits of a FPB.
Since this proposed activity does not involve any change in methodology described in the UFSAR, this question is not applicable for this evaluation.
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Enclosure 1.
PG&E Letter DCL-12-035 Based on this evaluation, NRC approval is not required prior to implementing this activity.
11-015      Unit 1 Polar Crane Modification Reference Document No.:           I DDP 1000000390, Revision 0 Reference Document

==Title:==
I Unit 1 Polar Crane Modification Activity Description/Summary of Evaluation:
The existing Unit 1 Polar crane is an antiquated design. The reliability of this crane has decreased considerably with every successive outage. Numerous problems have been encountered with the operation of the crane. Several relays, contactors, and control components have failed, resulting in several forced repair windows for the crane and a significant impact on crane availability during previous outages. Controls and motors for all functions need to be upgraded.
Summary of what is being changed:
This design change involves the following modifications to the Unit 1 Polar Crane drive system:
* Removes existing Main and Auxiliary Hoist M/G set and associated hardware.
* Installs new Main and Auxiliary Hoist Variable Frequency Drive (VFD) controls, Motor, Encoder and Disc Brake.
* Replaces existing Gantry and Trolley controls and associated hardware/enclosures with new VFD Control Systems. Note: Removes the temporary Gantry controls installed under DCN SE-49872 in Unit 1 fourteenth refueling outage.
* Installs new Trolley Motor and Disc Brake.
* Installs a new Power-Trak, cables, Guide Tray and Tow Arm to the Trolley.
* Installs power transfer switch to switch between main and alternate power.
* Installs new load cell pin.
* Installs new scoreboard.
* Installs a permanent Isolation Transformer for isolating the crane power Supply system from rest of the plant power supply and remove the temporary 51 kVA Isolation Transformer installed under DCN SE-49872 during 1R14 outage
* Installs new Control Console with all components attached or installed 56 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 including the Operator's Chair and a Jib Crane to facilitate VFD handling.
* Replaces both Main and Auxiliary Hoist gearbox mounted drum brakes with new Dynamic Braking System.
* Installs one E-stop near captain's chair and another E-stop accessible from the 140 foot deck.
* Modifies the electrical circuits for the Main Hoist upper limit switch (LS) (weighted LS) and provides fault indication and a reset selector switch.
* Installs a new Girder Jib Crane.
* Installs new Slip Ring Assembly, including support structure attached to Containment Dome Liner.
* Modifies two Containment Spray (CS) Piping Supports (176-149G and 176-151 G) at Elevation 209 foot to provide adequate seismic gap with the Polar Crane and preclude contact during design bases earthquakes.
As a result of the above modification, indirect changes are involved in the following areas:
* The seismic analysis of the Unit 1 Polar Crane is updated using ANSR model to address accumulative weight changes resulting from the "1982 modifications" up to completion of this DCP. The "1982 modifications" are as described in DCM S-42B Appendix A.
* Impact on heat removal capability/post-accident Containment pressure due to decrease in Containment Free Volume.
* The impact on heat removal capability due to increase in Heat Sink.
* Increase in post-accident combustibles gases inside of Containment due to increase of zinc and aluminum quantities inside Containment.
* Increase of Combustible Loading in Containment Fire Area 1, Zone 1 C.
Summary of Evaluation:
The 10 CFR 50.59 evaluations were performed on the following proposed activities:
Proposed Activity Number 1: the addition of weight to the Polar Crane resulting in increased seismic response forces and moments and higher critical member stresses.
Proposed Activity Number 2: the addition of weight to the Polar Crane resulting in increased seismic response displacements causing higher wheel lift and reduced seismic gap between the Polar Crane and Containment.
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Enclosure 1 PG&E Letter DCL-12-035 Proposed Activity Number 3: use of a personal computer (PC) based computer program of the ANSR model in lieu of the original mainframe ANSR-1 computer model for performing the seismic analysis of the Polar Crane.
Proposed Activity Number 4: use of a later version of the SAP computer program (SAP 2000) for generating modal properties, extracting frequencies and participation factors for the Polar Crane seismic analysis in lieu of the original SAP IV computer program.
The 10 CFR 50.59 evaluations concluded that:
* The proposed activities do not result in more than a minimal increase in the frequency of occurrence or consequences of an accident.
* The proposed activities do not create a possibility for an accident of a different type.
* The proposed activities do not result in more than a minimal increase in the likelihood of occurrence or consequences of a malfunction.
* The proposed activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in FSARU.
* The proposed activities do not result in a DBLFPB as described in the UFSAR being exceeded or altered.
* The proposed activities do not involve any changes to any method of evaluation.
Based on the negative responses in this Evaluation, prior NRC approval is not required to implement this change.
11-017      Units 1 and 2 Auxiliary and Fuel Handling Building Ventilation System (AFHBVS) Control System Reference Document No.:           DDP J-1 0000001 06, Revision 0 and DDP J-1 000000107, Revision 0 Reference Document

==Title:==
Unit 1 AFHBVS Control System and Unit 2 AFHBVS Control System Activity Description/Summary of Evaluation:
This LBIE Section 0 supersedes the LBIE associated with DDP J-1 0000001 06 (Unit 1) and J-1 0000001 07 (Unit 2).
Replace the obsolete logic control system for the AFHBVS. The new system 58 Enclosure 2 Contains Security-Related Information - Withhold Under 10 CFR 2.390 This page is decontrolled when separated from Enclosure 2.
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Enclosure 1 PG&E Letter DCL-12-035 will be based on a redundant PLC. The new AFHBVS control system will be designed to replicate the operation and logic functions of the existing control system. On ly the AF HBVS control system is being rep laced. There are no changes to system fans or dampers.
As with the existing system the new AFHBVS control system will be comprised of two subsystems; the Auxiliary Building Ventilation System (ABVS) and the Fuel Handling Building Ventilation System (FHBVS). The new AFHBVS control system equipment will be mounted in existing panels POV1 and POV2 in the main control room.
The replacement AFHBVS control system will be based on a PLC platform developed and supplied by Triconex, a division of Invensys Systems. The product line of the PLC referred to as Tricon. The Tricon is a state-of-the art logic controller that provides fault tolerance by means of Triple-Modular Redundant (TMR) architecture. The new PLCs use three identical processors each of which independently execute the control program in parallel with the other two processors. Specialized hardware/software voting mechanisms qualify and verify all digital inputs and outputs from the field.
Each processor is isolated from the others, no single-point of failure in any processor can pass to another. If a hardware failure occurs on one processor, the other processors override it. A faulted card can be removed and replaced while the controller is online without interrupting the system operation and safety-related functions.
The new system increases the reliability and dependability of the AFHBVS, provides diagnostic tools, and increases ease of maintenance and troubleshooting.
For ease of determining the system status, the new AFHBVS control system replaces the existing mimic bus LED status lights at POV1 and POV2 with graphic displays on touch screen monitors. As an operator convenience and to ease operator burden an additional touch screen monitor is installed at Vertical Board VB4 .
The new control system allows the operator to select Safeguards Only (SGO) mode from the VB4 ABVS mode selector switch whereas previously SGO mode could be entered only if a fan failed with the system operating in Building and Safeguards mode.
In the new control system if 125 Vdc control power is lost to a Tricon controller the associated dampers will go to their fail-safe position and the redundant fans will start from the other train. Previously, operator action was required to manually insert an "S" signal on both panels to align the unaffected dampers with the dampers affected by the loss of 125 Vdc power.
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Enclosure 1 PG&E Letter DCL-12-035 In the new system, insertion of the "S" signal on both panels will be automatic.
Physical modifications are required to POV1, POV2 and VB4. These panels are seismically qualified and therefore, their seismic qualification is affected.
Power to POV1 will be supplied from vital Class 1E 125 Vdc Distribution Panel 11 (21) and vital Class 1E 120 Vac Distribution Panel PY11A (PY21A).
Power to POV2 will be supplied from vital Class 1E 125 Vdc Distribution Panel 13 (23) and vital 1E 120 Vac Distribution Panel PY13 (PY23). The design Class II monitors and their associated CPUs in each POV panel will be powered from a common Class II 120 V AC Distribution Panel PY17N (PY27N). This represents a change in electrical bus loading.
Summary of Evaluation:
The LBIE Screen determined the following elements associated with the digital upgrade are adverse:
2.a.1 - The new system relies on software to perform t~e logic functions previously performed by discrete components.
2.a.2 - Multiple AFHBVS functions are being combined into one digital device.
2.a.3 - The design change replaces the existing mimic bus LED status lights with graphic displays on touch screen monitors .at POV1 and POV2. An additional touch screen monitor is also installed at Vertical Board VB4 as an operator convenience. These changes fundamentally alter the existing means of performing or controlling design functions.
For the purpose of evaluation, the above digital upgrade elements were linked as allowed by NEI 96-07 Section 4.2. These elements were evaluated per 10 CFR 50.59 using the guidance of NEI 01-01 Guideline on Licensing Digital Upgrades TR-1 02348 Revision 1.
The 10 CFR 50.59 evaluation was performed by answering the NEI 01-01 Appendix A supplemental questions. The 10 CFR 50.59 evaluation concludes:
The AFHBVS is not the initiator of any accident previously evaluated in the UFSAR. Therefore, the design change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.
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Enclosure 1 PG&E Letter DCL-12-035 Based on the NEI 01-01 Appendix A supplemental evaluation questions, the adverse activities associated with the digita l upgrade will not resu lt in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
Based on the NEI 01-01 Appendix A supplemental evaluation questions, the adverse activities associated with the digital upgrade will not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The design change will not:
* Prevent or degrade the effectiveness of actions described or assumed in an accident discussed in the UFSAR.
* Alter the assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR.
* Prevent or degrade mitigating the radiological consequences of an accident described in the UFSAR.
Based on the NEI 01-01 Appendix A supplemental evaluation questions, the adverse activities associated with the digital upgrade will not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
The proposed design change will not:
* Prevent or degrade the effectiveness of actions described or assumed in an accident discussed in the UFSAR.
* Alter the assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR.
* Prevent or degrade mitigating the radiological consequences of an accident described in the UFSAR.
Based on the NEI 01-01 Appendix A supplemental evaluation questions, the adverse activities associated with the digital upgrade will not create possibility for an accident of a different type than any previously evaluated in the UFSAR.
Based on the NEI 01-01 Appendix A supplemental evaluation questions, the adverse activities associated with the digital upgrade will not create possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.
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Enclosure 1 PG&E Letter DCL-12-035 Based on the NEI 01-01 Appendix A supplemental evaluation questions, the adverse activities associated with the digital upgrade will not affect any FPBs.
Therefore, the design change does not result in a design basis limit for a FPB as described in the UFSAR being exceeded or altered.
The LBIE Screen concluded that the design change did not adversely revise or replace any UFSAR described evaluation methodologies used in establishing the design bases or were used in the safety analyses.
Therefore, this evaluation question is not applicable.
Based on this evaluation, NRC approval is not required prior to implementing this activity.
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