ML11161A158

San Onofre, Units 2 & 3 and the Independent Spent Fuel Storage Installation - Facility Change Report

ML11161A158
Person / Time
Site:	San Onofre
Issue date:	06/08/2011
From:	St.Onge R J Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
Download: ML11161A158 (36)

Text

J SOUTHERN CALIFORNIA EDISON An EDISON INTERNATIONAL Company Richard J. St. Onge Director Nuclear Regulatory Affairs 10 CFR 50.59 10 CFR 72.48 June 8, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

Docket Nos. 50-361, 50-362, and 72-41 Facility Change Report San Onofre Nuclear Generating Station Units 2 and 3 and the Independent Spent Fuel Storage Installation

Dear Sir or Madam:

This letter transmits the Facility Change Report required by 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) for San Onofre Nuclear Generating Station Units 2 and 3 for the period from December 19, 2008 through February 10, 2011. Enclosures 1A (for 50.59)and 1 B (for 72.48), San Onofre Nuclear Generating Station Units 2 and 3 Facility Change Report, provide a brief description of any changes, tests, and experiments, including a summary of the evaluation performed in the reporting period for each. The report scope is based on a review of plant records and all evaluations identified for the above time period.Enclosure 2, 50.59 Evaluations Not Previously Submitted, describes some TS Bases change evaluations that had been previously omitted. These Bases changes had been routinely included in Bases reports at that time; they are provided here for completeness.

Complete facility change documentation is available onsite. If you would like any additional information, please contact Ms. Linda T. Conklin at (949) 368-9443.Sincerely,

Enclosures:

As stated cc: E. E. Collins, Regional Administrator, NRC Region IV R. Hall, NRC Project Manager, San Onofre Units 2, and 3 G. G. Warnick, NRC Senior Resident Inspector, San Onofre Units 2 & 3 J. C. Staab, NRC Project Manager, San Onofre ISFSI P.O. Box 128 San Clemente, CA 92674-F-- H 7 OM 5 50 t fAM55 ENCLOSURE lA SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 FACILITY CHANGE REPORT (FCR)10 CFR 50.59 EVALUATION SUMMARIES FOR THE PERIOD FROM DECEMBER 19, 2008 THROUGH FEBRUARY 10, 2011 201083068-02, Auxiliary Building Heating Ventilation Air Conditioning (HVAC)Cooling Unit Operation

Description:

This activity is opening doors on the 50', 70', and 85' elevations of the Auxiliary Control Building and starting and running at least 3 of 4 of the 50' elevation Engineered Safety Features (ESF) Battery Room exhaust fans, 2(3)MA173 and 2(3)MA174, and at least 1 of 4 of the 50' elevation Emergency Cooling Units (ECUs), 2(3)ME255 and 2(3)ME257, in the event that Unit 2(3)ME255, Train A Auxiliary Control Building Elev. 50' ECU, or Unit 2(3)ME257, Train B Auxiliary Control Building Elev. 50' ECU is declared inoperable for any reason (failure to start, failure to pass surveillance criteria, etc.) during normal operation.

Evaluation Summary: This 50.59 evaluated the compensatory measures to be implemented in the event of failure of Class 1 E switchgear and vital power/distribution panel room Emergency Cooling Units (ECUs) during normal operation.

Specifically, the 50.59 evaluated the adverse aspects of this activity identified by the 50.59 Screen in NN 201083068, Task-01: a maximum room temperature that would exceed the design temperature of the spaces, and operator actions that were required outside the control room. For these adverse aspects, the increase in likelihood of occurrence of a malfunction of a Safety System or Component (SSC) important to safety was not more than minimal. This is based on temperature remaining within the design limits for the equipment in the spaces, the actions being proceduralized with no specific time limit for completion during normal operation and provided that the normal cooling system is in service with the increased SSC start cycles and run time is within design assumptions.

It was concluded that the activity may proceed without the need for prior NRC approval.

201286253-17.

Restrict Auto/Manual Transfer of 6.9 kV Buses to Opposite Unit

Description:

Restrict Auto and Manual transfers of non-Class 1 E 6.9 kV buses to the opposite Unit unless the opposite Unit is in Mode 5, 6, or defueled.

Manual transfers will be restricted and automatic transfers will be controlled by maintaining the Cross-tie beakers in manual when the opposite Unit is in Modes 1-4. This will be directed by Operations procedures including S023-6-1.

These changes are being implemented to resolve discrepancies between the existing non-Class 1 E 6.9 kV system design and General Design Criterion 5, "Sharing of Structures, Systems, and Components." Evaluation Summary: This 50.59 evaluated restrictions on automatic and manual transfers of non-Class 1 E 6.9 kV buses powering Reactor Coolant Pumps to the opposite Unit's Reserve Auxiliary Transformer.

The increase in frequency of occurrence of an accident and increase in likelihood of a malfunction of a Safety System or Component (SSC) important to safety are not more than minimal because the Reactor Coolant Pumps (RCPs) will continue to fast transfer to the respective Unit's Reserve Auxiliary Transformer on a plant trip and the non-Class 1 E 6.9 kV backup power from the Reserve Auxiliary Transformer of the opposite Unit is not credited in categorizing the frequency or likelihood.

Likewise, the consequences will not be increased because the non-Class 1 E 6.9 kV backup power from the Reserve Auxiliary Transformer of the opposite Unit is not credited in determining the consequences.

New types of accidents and malfunctions with a different result are not created as the changes only affect forced reactor coolant circulation, which has already been evaluated in the Updated Final Safety Analysis Report (UFSAR), and eliminate the possibility of dual Unit accidents resulting from a malfunction of the shared non-Class 1 E 6.9 kV power system between Unit 2 and Unit 3. There are no design basis limits for fission product barriers exceeded or altered since the facility response remains consistent with the UFSAR Chapter 15 safety analyses.

It was concluded that the activity may proceed without the need for prior NRC approval.

800071304-0380 and 800291130-0380.

Units 2 and 3 Zinc Iniection

Description:

Units 2 and 3 Nuclear Engineering Change Packages 800071304 and 800071305 install a Quality Control (QC) III seismic category Il/I zinc acetate chemical feed system consisting of metering pumps, tanks, valves and controls in the Unit 2 Chemical and Volume Control System (CVCS) piping room and adjacent corridor in the 37' elevation Radwaste building.

Necessary tubing, tube track, hangars, valves, pipe fittings, orifices, and electric power are being installed.

The injection point is a modified vent line on the QC II, Seismic Category I charging pump suction line between the Volume Control Tank (VCT) and the pump suction. The zinc solution is pumped through the CVCS and into the Reactor Coolant System (RCS).Upon completion of the modifications, the new zinc injection systems will be placed in service to inject a soluble zinc acetate di-hydrate solution into the RCS during normal operation.

The goal is to establish an RCS zinc target concentration of 5 parts per billion (ppb) +/- 5 ppb starting in the last half of San Onofre nuclear Generating Station (SONGS) Unit 2 Cycle 15 after approximately 10.8 full power months of operation for Unit 2. Depleted zinc acetate will be used to avoid generation of activation product Zn-65 and associated increased impact on the plant radiation environment.

At low concentrations zinc replaces radioactive cobalt, nickel, and other activated corrosion products in RCS component oxide layers. Over time, this reduces radiation fields and subsequently worker radiation dose. Additionally, the process will help create a new metal component oxide layer during operation that is less susceptible to general corrosion.

The conditioning of the RCS component oxide layers may also delay the initiation of Primary Water Stress Corrosion Cracking (PWSCC) in RCS components with known susceptibility to the phenomenon.

Evaluation Summary: The Nuclear Steam Supply System (NSSS) vendor evaluations, fuel analyses, and calculations for Zinc addition to the RCS pressure boundary and interconnected systems form the technical basis for the conclusion that addition of a soluble zinc compound within the limitations imposed by the design of the system and associated operating procedures will not require prior NRC review for Cycle 15 operation.

For future cycle zinc injection operation, NRC review is not required if zinc RCS concentrations are limited to 5 ppb +/- 5 ppb concentration, injection rates are limited to 4 ml/min, refueling cycles are limited to 24 months, and fuel related aspects are addressed in the normal core re-load design process when the specifics of the fuel cycle design are known. It was concluded that these changes may be made without prior NRC approval.

800071702-0050, Units 2 and 3 Replacement Design Documentation and Functional Testing for Replacement Steam Generators (RSG) -This Evaluation is applicable to 800175663-0520, Unit 2 Steam Generator Engineering Change Package and 800175664-0170, Unit 3 Steam Generator Engineering Change Packaae

Description:

This activity replaces the design disclosure documentation and reference documentation for the San Onofre Nuclear Generating Station (SONGS) Unit 2 Original Steam Generators (OSGs) with that for the Replacement Steam Generators (RSGs)and performs functional testing of the RSGs. This replacement and testing is required as a result of physically replacing the OSGs with RSGs. Having the OSGs replaced with the RSGs will improve the efficiency and reliability of Unit 2 by replacing a large number of plugged or otherwise degraded heat transfer tubes in each OSG with new tubes made from thermally-treated Alloy 690, which is less susceptible to degradation than the mill annealed Alloy 600 material used for OSG heat transfer tubing.Replacement of the steam generators is a replacement in-kind in terms of an overall fit, form, and function with no, or minimal, permanent modifications to the plant Safety Systems or Components (SSCs).Evaluation Summary: This 50.59 evaluated for impact of Changing from computer program ANSYS to ABAQUS on Structural Integrity the Reactor Coolant System (RCS). ABAQUS and ANSYS were compared, using thermal and stress sample problems.

The results of these sample analyses demonstrated that in all cases the ANSYS and ABAQUS results varied from theoretical solutions by no more than 1 %, and ABAQUS and ANSYS results themselves were also within 1 % of each other. A difference of 2% is considered within the margin of error for this type of analysis; hence ABAQUS produced results that are essentially the same as ANSYS.Seismic Analysis of Reactor Vessel Internals (RVI) was changed from Topical Report CENPD-178 to CENPD-178-P, Revision 1-P. The original seismic analysis of the SONGS Unit 2 and 3 RVI (for the OSGs) was performed with the methodology described in C-E Topical Report CENPD-178 as referenced in the Updated Final Safety Analysis Report (UFSAR) Section 3.7.3.14.1.

Subsequent to submittal of CENPD-178, C-E modified modeling techniques, computer codes, testing methods, and acceptance criteria in response to changes in licensing requirements.

As a result, the entire report was revised and resubmitted to the NRC as CENPD-178-P, Revision 1-P which was approved by the NRC. The Tube Wall Thinning Analyses computer program change from CEFLASH, STRUDL and ANSYS to Manual Calculations and ANSYS was evaluated.

The results of the RSG tube wall thinning analysis are conservative or essentially the same as results from the UFSAR described tube wall thinning analysis for the OSGs. In addition, this approach results in a lower tube wall plugging limit (44%for OSGs versus 35% for RSGs) which is also conservative.

It was concluded that this change may be made without prior NRC approval.

800072665-0060, Units 2 and 3 Containment Temporary Construction Opening for Replacement Steam Generators (RSG)Description:

This activity creates a temporary construction opening in the Units 2 and 3 Containment Buildings at the beginning of the Unit 2 and 3 Cycle 16 Steam Generator Replacement (SGR) outages and restores it at the end of each outage. This activity is required because the existing Containment equipment hatch is too small and its location (at grade elevation) does not permit its use for movement of steam generators (SG). Thus, a construction opening is required to provide a path through the Containment for: (1)Installation and removal of equipment needed to facilitate SG replacement (2) Removal of the Original Steam Generators (OSGs), and (3) Installation of the Replacement SGs.Evaluation Summary: The splicing and repairing of reinforcing steel by welding complies with applicable American Society of Mechanical Engineers (ASME) and American Welding Society (AWS) Code requirements and, therefore, provides an equivalent level of assurance that the restored containment building will continue to perform its design functions.

As such, welded rebar will not create any new failure modes. Because of its inherently robust design and integrity in construction, the containment structure itself is not an initiator of any accident.Per the Updated Final Safety Evaluation Report (UFSAR) Section 3.9.1.2.2.1.11, the ANSYS computer program is a general purpose finite element program for linear and nonlinear structural analysis.

ANSYS, like the UFSAR-described computer program for local analysis, SAP, is "capable of performing static analysis of linear elastic three dimensional structures utilizing the finite element method." To establish the applicability and validity of the new analysis, a benchmark comparison was performed with results of the previous analysis of the Containment structure.

This benchmark comparison consisted of developing an ANSYS model representing the "before" condition of the Containment structure (i.e. prior to the temporary SGR containment opening).

The results were compared to the previous FINEL analysis and found to be within 2% and are, therefore, essentially the same. Based on the comparison of benchmarking results, the use of the ANSYS computer program for local analysis of the Containment structure does not constitute a departure from a method of evaluation described in the UFSAR used in establishing the design basis or in the safety analyses.

It was concluded that this change may be made without prior NRC approval.

800162499-0320, Replacement of Units 2 and 3 Chemical Volume Control System (CVCS) Boric Acid Makeup System Controls (Unit 2 Engineering Change Package NECP 800162890 and Unit 3 NECP 800162499)

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Description:==

Although the new Ovation DCS Boration-Dilution Control System (BDCS) replicates the functionality of the affected analog and discrete Boric Acid Make-Up (BAMU) and Primary Makeup Water (PMW) instrumentation and control components, the 50.59 screens for the activity concluded that the new digital systems could have a potential to adversely affect the Updated Final Safety Analysis Report (UFSAR)-described design functions of the affected components, specifically due to plausible hardware and software failures of the new digital equipment.

Evaluation Summary: This evaluation identified the inadvertent boron dilution accident as the only UFSAR-analyzed accident that could credibly be affected by this activity.

The evaluation concluded that the current UFSAR accident analysis for the inadvertent boron dilution accident remains bounding and that the change does not introduce the possibility of any new type of accident.

Although differences exist between the failure modes of this digital BDCS and those of the BAMU and PMW instrumentation and control components to be replaced, the evaluation determined that the results of the failure modes of the digital systems are bounded by the results of the failure modes of the existing equipment.

This change was determined to not require NRC approval prior to its implementation.

800175618-0290, Containment Tendon Pre-Tensioning Sequence Change During Replacement of the Unit 3 Steam Generators

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Description:==

This change modifies the pre-tensioning sequence for tendons replaced during the Unit 3 Steam Generator Replacement Project (SGRP). With the change, tensioning of specified horizontal and vertical replacement tendons in the areas outside the construction opening may begin when the concrete inside the repaired opening reaches a strength of 5000 psi. The remaining horizontal (or hoop) and vertical tendons will be re-tensioned when the repair concrete achieves its prescribed design strength of 6000 psi.Experience gained from the Unit 2 SGRP and recent testing of the qualified concrete mix for San Onofre containments has shown the concrete mix has a characteristic curing pattern in the first two weeks after batching.

The curing concrete reaches 5000 psi relatively quickly (in approximately 3 days) and then slowly increases to 6000 psi after approximately 10 days from batching.

By allowing certain tendons remote from the curing concrete to be re-tensioned once it reaches 5000 psi, this activity allows a reduction of several days in the time required to complete the containment repair.Evaluation Summary: The construction sequence used to tension the tendons will minimize the potential for unbalanced loads or differential stresses in the concrete.

Delaying tensioning of the specified horizontal and vertical tendons until the concrete strength equals 5000 psi meets the design requirements intended to prevent membrane cracks from tension forces as tension stresses are shown to be less than (f'c)A1/2.

Analysis has concluded that the change to creep and relaxation effects from limited tensioning when the concrete strength is at least 5000 psi are sufficiently minimal such that the tendons continue to meet design requirements over the life of the structure.

Hence, use of this revised construction sequence assures compliance with Updated Final Safety Analysis Report (UFSAR) requirements.

The construction sequence change has no impact on the frequency of occurrence or consequence of accidents or malfunctions previously evaluated.

No new accidents of a different type are created. The change has no impact on design basis limits for fission product barriers.

There is no change in the methods of evaluation used in establishing the design bases or in the safety analyses.

Hence, this activity can proceed without seeking NRC approval.

800231042-0050 and 800231043-0040, Unit 2 and 3 Containment Hydrogen Recombiner Removal for Cycles 15 and 16 Operation

Description:

This change removes one Containment Hydrogen Recombiners (E146) from each of Unit 2 and Unit 3 for two cycles of operation (Cycles 15 and 16) to facilitate Steam Generator replacement and other activities.

It was originally planned that Containment Hydrogen Recombiners E146 for both Units 2 and 3 would be restored to their original location following Replacement Steam Generator (RSG) installation in the Cycle 17 refueling outage.Evaluation Summary: The Hydrogen Recombiner is not credited to perform an active function in any safety analyses.

Its passive function, as a containment heat sink, has no impact on the Containment P-T analysis.

It was concluded that these changes may be made without prior NRC approval.

800636071-0140, Unit 2 6.9 kV Reserve Auxiliary Transformer Breaker Interlock Modification

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Description:==

The purpose of this temporary Engineering Change Package (ECP) is to allow a Unit 3 Reactor Coolant Pump (RCP) to be powered from each winding of the Unit 2 Reserve Auxiliary Transformer (RAT) 2XR3 concurrently with two Unit 2 RCPs. (The Unit 2 RCPs are normally powered from the Unit 2 Unit Auxiliary Transformers, but will fast transfer to 2XR3 in the event that Unit 2 were to trip.) This temporary ECP is being implemented to support outage operation of Unit 3 RCPs for sweeping and venting of the Reactor Coolant System prior to having the Unit 3 Unit Auxiliary Transformer 3XU2 or Unit 3 Reserve Auxiliary Transformer 3XR3 available.

To allow this operation, jumpers will be installed to defeat the electrical interlocks that prevent automatic or manual closure of the breakers between the Unit 2 6.9 kV busses (2A01 and 2A02) and the 2XR3 windings when the associated Unit 3 tie breaker is closed.Administrative controls will be implemented to limit the number of Unit 3 RCPs operating on each Unit 3 6.9 kV bus (3A01 and 3A02) to one or less with the Unit 3 bus tie breaker closed to 2XR3. Electrical calculation changes have been issued (E4C-126 ECN D0044288) which support a maximum of 3 RCPs per 2XR3 winding (2 Unit 2 pumps plus 1 Unit 3 pump), for a total of 6 RCPs operating simultaneously from 2XR3.Evaluation Summary: This 50.59 evaluated the temporary modification to jumper the tie breaker interlock between the Unit 2 6.9 kV Reserve Auxiliary Transformer (2XR3) to allow one Unit 3 Reactor Coolant Pump (RCP) to be run on each Unit 3 bus while retaining the capability for 2XR3 to supply all four Unit 2 RCPs subsequent to a Unit 2 trip. Specifically, the 50.59 evaluated the adverse aspects of this activity identified by the 50.59 Screen in Order 800636071 Operation 0130: (1) Operation of RCPs from more than one unit at a time on the same 6.9 kV preferred offsite power source, contrary to the design function specified in UFSAR Section 8.3.1.1.1, and (2) replacing an automatic preventive action (interlock between 6.9 kV bus tie breakers to 2XR3) with a manual action (administrative control) to limit the number of RCPs that are connected to each winding of 2XR3.For these adverse aspects, the increase in frequency of occurrence of an accident and increase in likelihood of a malfunction of a Safety System or Component (SSC)important to safety is not more than minimal because the interlock jumper and administrative controls allow Unit 2's RCPs to fast transfer to 2XR3 on a Unit 2 plant trip. Likewise, this is the rationale for why consequences will not be increased.

Continued...

800636071-0140 (Continued)

New types of accidents and malfunctions with a different result are not created as these changes only affect forced reactor coolant circulation, which has already been evaluated in the Updated Final Safety Analysis Report (UFSAR).There are no design basis limits for fission product barriers exceeded or altered since, with the jumper installed and implementation of the administrative controls, the facility response remains consistent with the Updated Final Safety Analysis Report (UFSAR)Chapter 15 safety analyses.

It was concluded that the activity may proceed without the need for prior NRC approval.

ENCLOSURE 1B SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 FACILITY CHANGE REPORT (FCR)SUMMARIES OF 10 CFR 72.48 EVALUATIONS PERFORMED BY TRANSNUCLEAR IN SUPPORT OF THE SAN ONOFRE INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)FOR THE PERIOD FROM DECEMBER 19, 2008 THROUGH FEBRUARY 10, 2011 200235652-07, Unit 2 Fuel Debris -Dry Cask Storage

Description:

During pre Unit 2 Cycle 16 dry storage inspections debris was found on five Spent Fuel Assemblies (SFAs). The debris was successfully removed from four of these but a polystyrene acrylonitrile bristle was not recoverable from fuel assembly$2F234. This evaluation addresses the impact of leaving this debris in the assembly when it is placed into a dry storage cask for long term storage. This 72.48 evaluation is specific to this piece of debris and only applies to this loading.Evaluation Summary: Leaving the polystyrene acrylonitrile bristle in the assembly when it is placed into a 24PT4 Dry Shielded Canister (DSC) for long term storage does not adversely affect the design function of the DSC for confinement, shielding, criticality control, structural and thermal safety. The conclusion was based on the Transnuclear evaluation of the bristle and its potential impact on the pressure boundary, the SFA's stored, the DSC inert environment, the DSC drying and sealing operations, the decay heat load, the design basis source term, K effective, DSC weight and confinement boundary.Since the responses to all the eight 72.48 evaluation questions were "No," SFA$2F234 can be loaded and stored without an amendment to the General License.

200655692-11, Unit 3 Fuel Debris -Dry Cask Storage

Description:

SCE proposes to allow the storage of Dry Shielded Casks (DSCs) containing small rust flakes which originated from the reactor top flange during fuel movements.

These small flakes are basically iron oxide that become lodged in the peripheral fuel assemblies in the reactor core and are difficult to remove.Evaluation Summary: The existence of small particles of iron oxide on stored Spent Fuel Assemblies (SFA's) in a 24PT4 DSC was determined to have a negligibly small impact on the behavior of the fuel assemblies and no measurable affect on the DSC basket, shell or inert atmosphere.

As such, the rust flakes do not affect any of the supporting analyses that have been completed.

The quantity of material was determined to be insignificant in relation to the overall size and volume of the DSC and its contents.

It was concluded that this change may be made without prior NRC approval.

200984377-05, Fuel Storage Canister Pressurization

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Description:==

A Dry Shielded Cask (DSC) was backfilled with water after completion of the inner cover welding, as required to avoid the potential for air to be in contact with the fuel.While the DSC was being aligned for blowdown, the DSC vent and siphon isolation valves were closed for 23 minutes. This allowed pressure to increase in the DSC.The siphon valve was opened followed by opening the vent valve. This resulted in the helium gas bottle rack relief valves (set at 18 psig) to open and water to flow out.A relief valve set at 50 psig did not lift. Previous canister loads were done by opening the vent valve first followed by the siphon valve.Evaluation Summary: This evaluation determined that the maximum material stress increased due to the over pressure condition remained below American Society of Mechanical Engineers (ASME) code allowables and had no impact or consequences on the design function and capabilities of the DSC. It was concluded that this change may be made without prior NRC approval.

800133840-0380, Dry Cask Storage Canister Pressurization Excessive Loading Ram Force

Description:

This activity evaluates surface scratches found on a dry cask storage canister after recording an excessive loading ram force. Note: This evaluation was completed in the previous reporting window but the applicable Order was not completed until June 5, 2009 [during the current reporting window]Evaluation Summary: The nonconformance resulting from the scratches was determined to not adversely affect the features, function or performance of the transfer cask. This conclusion was based on an evaluation of the reduction of the inner linear thickness that resulted from the gouges. Design calculations evaluated to demonstrate that the scratches did not adversely affect the system design. The transfer cask was found to be within the stress allowables and was accepted without need for any repair work. It was concluded that this nonconformance was acceptable without prior NRC approval.

ENCLOSURE 2 50.59 Evaluations Not Previously Submitted The following Technical Specification (TS) Bases Change 50.59 evaluations were discovered during the currently ongoing San Onofre Units 2 and 3 TS Conversion review. They were not included in the 1998 or a subsequent San Onofre Units 2 and 3 Facility Change Report but were routinely included in Bases reports at that time. These evaluations pre-date the current evaluation methodology from Regulatory Issue Summary (RIS) 2005-20 issued on September 26, 2005 and are included for completeness.

It should be noted that the San Onofre Units 2 and 3 TS Bases will be replaced in their entirety coincident with TS conversion implementation.

B 96-001(i)

Change to Bases B 3.3.1 "Shutdown Margin (SDM)"

Description:

This change revises the Bases to list separate functions for Reactor Coolant Flow -Low Steam Generator 1 and Reactor Coolant Flow -Low Steam Generator 2 and revises the Bases to clarify that the core power indication can be adjusted one channel at a time.Safety Evaluation:

The change clarifies that the LCO for Reactor Coolant Flow -Low is applicable to each steam generator and provides background information for the Reactor Protective System Instrumentation.

It was concluded that this change may proceed without prior NRC approval.B 96-001(ii)

Change to Bases B 3.3.12 "Remote Shutdown System"

Description:

This change deletes an erroneous reference to the Component Cooling Water System and Service Water System.Safety Evaluation:

This editorial Bases change supports the requirements for the Remote Shutdown System specified in Technical Specification 3.3.12. It was concluded that this change may proceed without prior NRC approval.B 96-001(iii)

Change to Bases B 3.3.13 "Source Range Monitoring Channels"

Description:

Limiting Condition for Operation (LCO) 3.3.13 requires two channels of source range monitoring instrumentation to be operable.

The Background section of Bases 3.3.13 states that the source range (startup) monitoring channels provide neutron flux count rate level indication from 0.1 cps to 500,000 cps. The neutron flux count rate level however, ranges from 0.1 cps to 100,000 cps. The proposed change revises the Background section of the Bases to correct the count rate level indication.

Safety Evaluation:

The change clarifies statements in the Bases regarding the neutron flux count rate. This change supports requirements for the Source Range Monitoring Channels specified in Technical Specification 3.3.13. It was concluded that this change may proceed without prior NRC approval.

B 96-001(iv)

Change to Bases B 3.1.10 "Boration Systems -Shutdown"

Description:

This change identified the amount of borated water required in either the Boric Acid Make Up (BAMU) tanks or Refueling Water Storage Tank (RWST) to be 4150 gallons (per the current Technical Specification), and added the as-indicated values on Control Room instrumentation for this volume.Safety Evaluation:

The revised Bases is consistent with the current Technical Specification values. It was concluded that this change may proceed without prior NRC approval.B 96-001(v)

Change to Bases B 3.7.10 "Emeraency Chilled Water (ECW) System"

Description:

The proposed change addresses the application of LCO 3.7.10 to emergency space cooler and/or ECW System inoperability on systems served by the ECW System. The Background Section of Bases B 3.7.10 is being divided into the several subsections.

Safety Evaluation:

The redundant cooling capacity of these systems is consistent with single failure criteria and the assumptions used in the accident analyses.

The changes were determined to not alter the plant configuration or modify the operating practice.

It was concluded that this change may proceed without prior NRC approval.B 96-001(vi)

Change to Bases B 3.7.11 "Control Room Emergency Air Cleanup System (CREACUS)"

Description:

The purpose of this change is to provide guidance to Operations on the applicability of Technical Specification 3.7.11, "Control Room Emergency Air Cleanup System" regarding the emergency ventilation supply unit's test requirements.

Safety Evaluation:

This change does not affect plant configuration or plant operations.

The change clarifies the requirement for filtration efficiency testing based on the results of dose calculations.

It was concluded that this change may proceed without prior NRC approval.

B 96-001(vii)

Change to Bases B 3.7.7.1 "Component Cooling water (CCW) Safety Related Makeup System"

Description:

This editorial change corrects the value which specifies Tank T-055 and Tank T-056 water levels in percent as indicated in the Control Room.Safety Evaluation:

Since the proposed change was determined to be editorial it was concluded that this change may proceed without prior NRC approval.B 96-003 Change to Bases B 3.1.10 "Boration Systems -Shutdown"

Description:

This change revises Bases 3.1.10 upper limit of the boron concentration specified in both the Limiting Condition for Operation (LCO) section of the Bases, and the section that discusses the surveillance requirements.

The upper limit was chosen based on the temperature dependence for the solubility of boric acid.Safety Evaluation:

The requirements are already established, the proposed change will correct the upper limit for the boron concentration in the Boric Acid Makeup (BAMU) Tank specified in the Bases. It was concluded that this change may proceed without prior NRC approval.B 96-004 Change to Bases B 3.8.4 "DC Sources -Operating" and B 3.8.9 "Distribution Systems -Operating"

Description:

The changes provide clarification to allow support systems to be taken out of service without their associated systems being declared inoperable under LCO 3.8.9. Specifically, on-line maintenance of the 1 E chargers may be performed under the Required Actions of LCO 3.8.4.C.Safety Evaluation:

The proposed changes only clarify and do not change the defined technical specification design bases for any of the systems, structures, or components which they affect. It was concluded that this change may proceed without prior NRC approval.

B 96-006 Change to Bases B 3.3.1 "Reactor Protective System Instrumentation

-Operating"

Description:

This change revises the Bases to Surveillance Requirement (SR) 3.3.1.11.

The changes are based on the Cycle Independent Shape Annealing Matrix (CISAM) Report. The proposed change is necessary to clarify the bases to make it clear that the Shape Annealing Matrix (SAM) values do not need to be remeasured after each refueling (i.e., a verification, not a re-measurement, will be performed).

In addition, the bases would be revised to state that verification of the SAM is necessary to confirm that refueling did not produce a significant change in the Core Protection Calculators (CPC) axial shape synthesis.

Safety Evaluation:

The CISAM methodology involves implementation of a SAM generated from a more accurate mid-cycle test. The CISAM is then verified during a reload cycle and if acceptable will continue to be used for the upcoming fuel cycle. A core follow monitoring program will continue to monitor CPC axial shape performance during the cycle. Therefore, the CPC calculations will continue to remain conservative at a 95/95 probability/confidence level for the CISAM methodology.

Since the CPC calculations that provide the trip function will continue to, conservatively provide the reactor trip to mitigate the consequences on an Anticipated Operational Occurrence or accident when required by the safety analyses, the margin to safety will not be reduced by implementing the CISAM methodology.

It was concluded that this change may proceed without prior NRC approval.B 96-007 Change to Bases B 3.3.1 "Reactor Protective System Instrumentation

-Operatingf" Bases B 3.3.2 "Reactor Protective System Instrumentation

-Shutdown," Bases B 3.3.4 "RPS Logic and Trip Initiation," and Bases B 3.3.6"Engineered Safety Features Actuation System (ESFAS) Logic and Manual Trip."

Description:

A recent Plant Protection System (PPS) failure left the condition of a channel bypass relay indeterminable.

A conservative position had been taken by Plant Operations that the matrix was inoperable, and hence, TS 3.3.4, Condition A, was applicable.

The Required Action required restoration of the inoperable channel to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This incident has highlighted the need for clarification of the Technical Specification requirements for PPS Trip Channel Bypass, and what is required if the Trip Channel Bypass status is indeterminable.

In this instance, the Trip Channel Bypass push button was depressed to place the channel function in Bypass and no indication of Bypass was received, even though the push button remained depressed.

Safety Evaluation:

The proposed change revises the Bases to clarify requirements for the PPS Trip Channel Bypass. If a channel does not actually go into Bypass when the push button is depressed, the channel is free to process any trip signal, which is conservative.

By having the channel in Bypass, even if the Bypass contacts do not close, the electrical interlock and administrative controls are implemented, meeting the requirement to have an inoperable channel in Trip or Bypass. This prevents any other channel from going into Bypass, preserving the function's ability to trip with any other single channel failure. It was concluded that this change may proceed without prior NRC approval.B 96-008 Change to Bases B 3.8.1 "AC Sources -Operating"

Description:

This change revises the Bases discussion that testing the critical trips will not result in any increase in Diesel Generator (DG) failures or spurious DG trips.Safety Evaluation:

The surveillance will continue to verify that the non-critical DG trips are bypassed when required for emergency service of the DGs, ensuring their availability as assumed in the accident analyses.

It was concluded that this change may proceed without prior NRC approval.B 96-014 Change to Bases B 3.3.8 "Containment Purge Isolation Signal (CPIS)"

Description:

These modifications are part of the licensee's efforts to delete or upgrade/replace existing radiation monitoring equipment with modern state-of-the-art microprocessor-based equipment.

Safety Evaluation:

Changes to the design and operation of CPIS are addressed in the NRC approved Units 2 and 3 Amendments 132 and 121. The changes to the applicable TS Bases do not result in any plant configuration changes. This modification of the TS Bases does not result in any new interactions with systems or components than those previously identified in these Amendments.

The proposed wording of the Bases incorporates these issued Amendments.

It was concluded that based on the prior NRC approval that was provided by License Amendments 132 and 121, this change may be implemented.

B 97-001 Change to Bases B 3.8.1.16 "AC Sources -Operating"

Description:

This change clarifies Bases 3.8.1.16 to correctly state that the timers are not reset until the Engineered Safety Feature (ESF) Safety Injection Actuation Signal (SIAS) is removed, or a Loss of Voltage Signal (LOVS)/Sustained Degraded Voltage Signal (SDVS)/Degraded Grid Voltage with SIAS Signal (DGVSS) occurs, thereby de-energizing and resetting the timers.When LOVS/SDVS/DGVSS occurs the LOVS reset contacts open, de-energizing the load sequence timers.Safety Evaluation:

The proposed change does not affect the consequences of a Diesel Generator (DG) failure when called upon due to loss of offsite power with or without a concurrent SIAS. The surveillance will continue to verify the capability of the DGs to synchronize and transfer loads to the offsite power supply and maintain their readiness (running at rated speed and voltage) to connect to the bus and re-sequence the ESF loads should a subsequent LOVS/SDVS/DGVSS signal occur while the ESF actuation signal is still present. It was concluded that this change may proceed without prior NRC approval.B 97-002 Change to Bases B 3.3.7.1 "DG-Under Voltage Start"

Description:

This change clarifies the guidelines for performing a channel check to explicitly state a comparison is performed between output of the potential transformers that feed the Loss of Voltage Signal (LOVS) undervoltage relays.Safety Evaluation:

The proposed change clarifies the specific surveillances performed.

This change does not modify any current testing or the response or performance of installed equipment.

It was concluded that this change may proceed without prior NRC approval.B 97-003 Change to Bases B 3.1.10 "Boration Systems -Shutdown"

Description:

The Applicable Safety Analyses described in the Bases for TS 3.1-10 requires administrative controls to rack out two of the three Charging Pumps to prevent an inadvertent dilution.

This is determined to be an unnecessary requirement.

The Applicable Safety Analysis is being modified to clarify that the restriction only applies during reduced inventory conditions.

Safety Evaluation:

The proposed change clarifies the assumptions made to the Applicable Safety Analysis.

This change does not modify any current operations or the response or performance of installed equipment.

It was concluded that this change may proceed without prior NRC approval.B 97-004 Change to Bases 3.8.1.18 "AC Sources -Operating"

Description:

The Bases for the surveillance requirement associated with Emergency Diesel Generator (EDG) loading sequencing (Surveillance Requirement (SR) 3.8.1.18) requires revision to indicate that the Emergency Cooling Unit (ECU) start time is dependent upon the actual Component Cooling Water (CCW) pump start time plus the start time tolerance to meet the+/-10% interval tolerance requirements of the Technical Specification.

Safety Evaluation:

With the proposed modification, the overall actuation time for the ECU fans, with or without loss of offsite power, will increase.

This increase in starting time of the ECU fan is still within the bounding limits of the time actually used in the containment pressure-temperature (P-T)analyses for the design basis Loss Of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) events. Changing the start time within the bounding limits of the assumed start time in these P-T analyses has no impact on the results of the analyses.

Therefore, the proposed modification, which moves the ECU fan start time to after the CCW pump start time, assures CCW system continued operability.

It was concluded that this change may proceed without prior NRC approval.B 97-007 Change to Bases B 3.3.8 "Containment Purge Isolation Signal (CPIS)"

Description:

This change clarifies that, if both monitors are in service, the CROSS CHANNEL CHECK should be performed between two channels (noble gases "C" channels) because containment noble gases monitors are identical and a gas distribution should be the same for all the areas of the containment.

Safety Evaluation:

This change does not alter existing tests already performed on the critical protective functions not specified in this Surveillance Requirement (SR). The proposed change clarifies the specific surveillance performed.

This change does not modify any current testing or the response or performance of installed equipment.

It was concluded that this change may proceed without prior NRC approval.

B 97-008 Change to Bases B 3.1.4 "Moderator Temperature Coefficient (MTC)," B 3.5.1"Safety Injection Tanks (SITs)," B 3.8.5 "DC Sources -Shutdown," B 3.9.5 "Shutdown Cooling (SDC) and Coolant Circulation

-Low Water Level," and B 3.9.6 "Refueling Water Level."

Description:

This change updates the Bases and adds additional information to the Bases to clarify the original intent. This change also removes incorrect information.

Safety Evaluation:

B 3.1.4 Moderator Temperature Coefficient (MTC)The deletion of the superfluous statement, "The same characteristic is true when the MTC is positive and coolant temperature decreases occur" is editorial.

Changing the term "original" to"reference" is editorial.

The reference accident analysis is correct because it allows for updating analyses as appropriate for each fuel cycle. Deleting the sentence: "The variation of the MTC, with temperature assumed in the safety analysis," is accepted as valid once the Beginning of Cycle (BOC) and Middle Of Cycle (MOC) measurements are used for normalization.

B 3.5.1 Safety Injection Tanks The minimum boron concentration requirement for the Safety Injection Tanks (SITs) is changed from 1850 parts per million (ppm) to 2200 ppm by Unit 2 and 3 License Amendments 135 and 124.B 3.8.5 DC Sources -Shutdown Adding information to the Bases to ensure the surveillances are performed does not change the meaning or intent of the Bases.B 3.9.5 Shutdown Cooling (SDC) and Coolant Circulation

-Low Water Level Deleting the last sentence from the background corrects an error. The condition described in that sentence is appropriately written and contained in B 3.9.4.B 3.9.6 Refueling Water Level The Bases to B 3.9.6 is revised to be consistent with the Applicability statement as approved by the NRC in License Amendment 134 for Unit 2 and 123 for Unit 3.These changes are editorial changes to the Bases or approved by NRC License Amendment.

They do not change how the plant is operated.

It was concluded that with NRC License Amendments 134 and 123 these changes may proceed.B 97-009 Change to Bases B 3.3.13 "Source Range Monitoring Channels"

Description:

This change permits normal temperature fluctuations associated with maintaining the plant status, provided they remain within limits established for the plant conditions.

These normal fluctuations would not cause the plant to change Modes.Safety Evaluation:

The proposed change allows for the normal temperature fluctuations, within a Mode, associated with maintaining plant status. These temperature changes are bounded by the boron dilution accident analysis.

It was concluded that this change may proceed without prior NRC approval.B 97-010 Change to Bases B 3.3.1 "RPS Instrumentation

-Operating"

Description:

This change revises information in the Bases of Surveillance Requirement (SR) 3.3.1.1,"Perform a Channel Check of each RPS Instrument Channel" to remove the statement "A CHANNEL CHECK must be performed on all inputs." The reason for this change is to revise the Bases requirements to be consistent with the definition of a channel check.Safety Evaluation:

Either, checking the input signals to the Core Protection Calculators (CPCs), the output signals, or status from the CPCs provides assurance that the Reactor Protection System is operating properly.

Thus, the CHANNEL CHECK is being performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as specified by Surveillance Requirement (SR) 3.3.1.1. The safety function of the CPC system is to provide a trip if required during Updated Final Safety Analysis Report (UFSAR) chapter 15 accident events. The CPC system will continue to provide this protection since no new failure modes have been introduced.

It was concluded that this change may proceed without prior NRC approval.

B 97-011 Change to Bases B 3.1.5 "Control Element Assembly (CEA) Alignment," B 3.3.1"Reactor Protective System (RPS) -Operating," B 3.4.12 "Low Temperature Overpressure Protection (LTOP) System," B 3.5.2 "Emergency Core Cooling System (ECCS) -Operating," B 3.7.14 "Fuel Handling Building Post-Accident Cleanup Filter System," B 3.8.1 "AC Sources -Operating," and B 3.9.3 "Containment Penetrations"

Description:

A review of the Revised Technical Specifications, the Bases to the Technical Specifications, the Licensee Controlled Specifications and their Bases has been performed to ensure that these new documents were fully and correctly implemented.

The editorial Changes proposed in this document are to address questions that came from that review and provide clear, correct information.

Safety Evaluation:

These are editorial changes without impact. It was concluded that this change may proceed without prior NRC approval.B 97-012 Change to Bases 3.9.3 "Containment Penetrations"

Description:

This change allows use of the containment mini purge system during MODE 6 as an alternative to the main purge.Safety Evaluation:

This revision does not change how the plant mini purge system is normally operated [with the Core Protection Isolation System (CPIS) operable]

and operation of either the purge or mini purge system introduces any new safety concerns or contributes to the cause of an accident in the safety analyses.

It was concluded that this change may proceed without prior NRC approval.B 97-015 Change to Bases B 3.1.10 "Boration Systems -Shutdown," B 3.3.3 "Control Element Assembly Calculators (CEACs)," B 3.3.5 "Engineered Safety Features Actuation System Instrumentation," and B 3.7.6 "Condensate Storage Tank (CST T-120 and T-121)"

Description:

Proposed change 1 clarifies the Bases for Surveillance Requirements (SRs) 3.1.10.1 and 3.1.10.4 to resolve an inconsistency.

Proposed change 2 clarifies the Bases for SR 3.3.3.4 which clarifies how the surveillance is to be performed.

There is no change in the way in which the surveillance is performed.

Proposed change 3 corrects an error in the Bases for SR 3.3.5.4 which requires that a channel calibration of the Recirculation Actuation Signal (RAS), including the bypass removal function, be performed.

However, a bypass removal function is not part of the RAS. The proposed change will revise the SR Bases discussion to delete reference to the bypass removal function.Also, the term "detector" is changed to "sensor." These changes are editorial.

Proposed change 4 changes B 3.7.6. The only accidents associated with the Condensate Storage Tanks are flooding events due to postulated random tank wall ruptures, as described in the Updated Final Safety Analysis Report (UFSAR). The proposed change to the Bases does not affect the quality class or seismic category of the Condensate Storage Tank.Safety Evaluation:

These changes are primarily clarification and editorial except change 4 which was determined to not affect the quality class or seismic category of the Condensate Storage Tank. It was concluded that these changes may proceed without prior NRC approval.B 97-016 Change to Bases B 3.3 "Instrumentation"

Description:

Revise the Bases to incorporate surveillance requirement frequencies approved by the NRC in Units 2 and 3 License Amendments 133 and 122. This Technical Specification Bases change is an editorial change to update surveillance frequencies, section numbers, and references in Bases 3.3.1, 3.3.2, 3.3.3, 3.3.4, 3.3.5, and 3.3.6.Safety Evaluation:

This change is editorial to update the surveillance intervals and section numbers in the Bases in accordance with the NRC approved Technical Specifications and to provide correct references in the Bases. It was concluded that with NRC License Amendments 133 and 122 these changes may proceed.B 97-017 Change to Bases B 3.3.7 "Diesel Generator (DG) -Undervoltaqe Start"

Description:

This change provides clarification of the necessary actions to complete a channel check for the Loss of Voltage (LOV) Function channels and the Degraded Voltage Function channels.A statement is being added to the Bases to indicate observing the flags for the channel relays and the 4kV Bus undervoltage alarm window is an acceptable method for comparison to detect malfunctions of the LOV and Degraded Voltage channels.

Safety Evaluation:

This change will document additional guidance in the Bases for performing a channel check of the LOV and Degraded Voltage channels.

The channel check will consist of an observation of the channel behavior and observing the relay flags and the 4kV bus undervoltage alarm. This change does not modify or remove any requirements.

This is a clarification of actions that are already performed.

It was concluded that this change may proceed without prior NRC approval.B 97-018 Change to the TS Bases to Provide Various Clarifications

==

Description:==

This change adds information to the Bases to clarify the original intent and correct information in the Bases. The changes are to Bases B 3.7.1 Background, B 3.7.4.2 Surveillance Requirements, B 3.7.7.4 Surveillance Requirements, and B 3.7.8.3 Surveillance Requirements to correct erroneous information which was inadvertently incorporated in the Bases by the Technical Specification Improvement Program (TSIP).Safety Evaluation:

These changes are not related to equipment modifications or changes to operating procedures.

No operational limits are being changed. It was concluded that this change may proceed without prior NRC approval.B 97-020 Change to Bases B 3.3.11 "Post Accident Monitoring Instrumentation (PAMI)"

Description:

The change deletes reference to a control room alarm on Condensate Storage Tank level as part of the Bases. The alarm is not required by Regulatory Guide 1.97 and it is not part of plant design. It would move background wording in the Limiting Condition for Operation (LCO)back into the background portion of the bases. The change clarifies the wording on performing surveillances Safety Evaluation:

The change deletes reference to the control room alarm on the Condensate Storage Tank which is not required by Reg. Guide 1.97. Other wording changes are clarifications.

It was concluded that this change may proceed without prior NRC approval.

B 97-022 Changes to Bases B 3.4.1 "RCS DNB (Pressure, Temperature, and Flow)Limits" and B 3.8.4 "DC Sources -Operating"

Description:

This change corrects two Bases sections.

The B 3.4.1 applicable safety analyses temperature ranges. The safety analyses were performed over the temperature range of 520°F through 560 0 F, not 522 0 F through 558 0 F, as was stated in B 3.4.1. The power ranges are being corrected consistent with the safety analyses.

Additionally, this change corrects B 3.8.4 to reference the correct issuance date of Regulatory Guide 1.129, Revision 1.Safety Evaluation:

These changes are limited to correcting the above described information in the Bases. It was concluded that these changes may proceed without prior NRC approval.B 97-024 Change to Bases B 3.4.13 "RCS Operational Leakage"

Description:

This change deletes the condition of being near operating pressure for the performance of a Reactor Coolant System (RCS) water inventory balance in Modes 3 and 4.Safety Evaluation:

This change removes the phrase "near operating pressure" from the Bases of TS 3.4.13. This provides consistency in the wording of the Technical Specifications and the Bases for the performance of a water inventory balance during Modes 3 and 4. Performing a water inventory balance is one method to identify leakage from the RCS during normal plant operation.

It was concluded that this change may proceed without prior NRC approval.B 97-025 Change to Bases B 3.3.7 "Diesel Generator (DG) -Undervoltage Start"

Description:

This change clarifies that the Potential Transformers (PTs) are not part of the channel calibration.

Safety Evaluation:

The PTs are used for a relay application and therefore calibration is not necessary.

This change does not modify any equipment of the channel calibration.

It was concluded that this change may proceed without prior NRC approval.

B 97-026 Change to Bases B 3.7.5 "Auxiliary Feedwater System"

Description:

This change deletes the statement of minimum flow capacity.Safety Evaluation:

The ability to provide minimum flow capacity of 500 gallons per minute (gpm) is verified through a combination of Inservice Testing, Operations alignment verification, and system modeling in design calculations.

These requirements are unaffected by this change. The Auxiliary Feedwater (AFW) system will still be demonstrated as being capable of supplying the required flow. The proposed surveillance method verifies the flowpath between the Condensate Storage Tank (CST) and the Steam Generators (SGs) is OPERABLE.

It was concluded that this change may proceed without prior NRC approval.B 97-028 Change to Bases B 3.3.11 "Post-Accident Monitoring Instrumentation" and Licensee Controlled Specification (LCS) Bases 3.3.102 "Radiation Monitoring Instrumentation"

Description:

This change will include the basis for not cross checking the containment high range area monitors in the Bases of the Technical Specification and the LCS. These instruments are part of the post accident monitoring of the containment atmosphere.

Safety Evaluation:

The CHANNEL CHECK of an instrument is a method for assessing OPERABILITY by observation of the instrument and comparison when possible to other instruments measuring the same parameters.

Comparison of the containment high range area monitors is not meaningful.

OPERABILITY of these monitors for a CHANNEL CHECK will be based on observation of the monitors only. It was concluded that these changes may proceed without prior NRC approval.B 97-029 Change to Bases B 3.1.14 "Special Test Exceptions (STE) -Reactivity Coefficient Testing"

Description:

This Bases currently permits performance of specific physics tests and is being revised to remove an erroneous reference to Technical Specification (TS) 3.1.8.

Safety Evaluation:

The change revises the Bases to be consistent with Technical Specifications.

It was concluded that this change may proceed without prior NRC approval.B 97-030 Change to Bases B 3.6.3 "Containment Isolation Valves"

Description:

This change makes two changes. The first change clarifies the actuation signals for Section B and Section E valves. The second change revises safety analysis parameters to reflect the current analyses of record.Safety Evaluation:

Part 1 of this change clarifies what is meant by an "actuation signal" and part 2 of this change updates analysis parameters and references.

There is no change to plant design, construction, or operation as a result of this Bases change.It was concluded that this change may proceed without prior NRC approval.B 97-031 Change to Bases B 3.8.3 "Diesel Fuel Oil, Lube Oil, and Starting Air"

Description:

This Bases Surveillance Requirement (SR) 3.8.3.2 states that the SR ensures sufficient lube oil inventory is available for 7 days of operation for each Diesel Generator (DG). This statement is being removed from B 3.8.3.2 because it does not apply to San Onofre Units 2 and 3.Safety Evaluation:

This change removes information from the Bases that is not applicable to San Onofre Units 2 and 3. The Bases for SR 3.8.3.2 is to ensure the Diesel Generators (DGs) will run for 7 days with a full load. The DG lube oil sump contains sufficient capacity to supply lube oil for a 7 day run with a full load. It was concluded that this change may proceed without prior NRC approval.B 97-032 Change to Bases B 3.1.10 "Boration Systems -Shutdown"

Description:

The proposed change revises the Bases to correct indicated levels to allow for instrument Total Loop Uncertainty (TLU) to ensure adequate amount of borated water to mitigate an inadvertent boron dilution event. The change is supported by a formal calculation that includes TLU.

Safety Evaluation:

This change provides conservatism in the required value to ensure adequate amount of borated water to mitigate an inadvertent boron dilution event. It was concluded that this change may proceed without prior NRC approval.B 97-035 Change to Bases B 3.7.8 "Saltwater Cooling System"

Description:

This change revises the Bases to clarify that portions of Surveillance Requirements (SRs)3.7.8.2 and 3.7.8.4 may be completed during normal operations by adding the phrase,"Although testing of some of the components of this circuit may be accomplished during normal operations." Also, changed the "perform" to "complete" in one instance.Safety Evaluation:

This change is a clarification.

The pumps and valves will still be verified to actuate on receipt of an actual or simulated actuation signal. The testing frequency remains 24 months. There is no change to the level of assurance that the Salt Water Cooling (SWC) system will perform as designed.

It was concluded that this change may proceed without prior NRC approval.B 97-036 Change to Bases B 3.8.1 "AC Sources -Operating"

Description:

This change implements NRC Amendments 136 and 128. The change revises the Bases for SR 3.8.1.8 to be consistent with the change to the Technical Specifications.

This Bases change provides a detailed discussion on the verification of the capability to transfer power sources from the normal preferred power source to the alternate preferred power source. This change is strictly editorial.

Safety Evaluation:

This change revises the TS Bases to be consistent with NRC License Amendments 136 and 128. This change is an administrative update of the Bases. It was concluded that with these License Amendments this change may proceed.

B 97-037 Change to Bases B 3.3.1 "Reactor Protective System (RPS) Instrumentation

-Operating," B 3.3.2 "Reactor Protective System (RPS) Instrumentation

-Shutdown, B3.3.3 "Control Element Assembly Calculators (CEAC)." B 3.3.4 "RPS Logic and Trip Initiation," and B 3.3.5 "Engineered Safety Features Actuation System (ESFAS)Instrumentation."

Description:

A review of the revised Technical Specifications, the Bases to the Technical Specifications, and Licensee Controlled Specifications and their Bases was performed to ensure that these new specifications and bases had been fully and correctly implemented.

The changes proposed in this document are to clarify information contained in the Technical Specification Bases.Safety Evaluation:

These changes clarify information in the Bases. This information does not impact operating limits, procedures, or equipment.

The changes provided are format and editorial changes. It was concluded that these changes may proceed without prior NRC approval.B 97-038 Change to Bases B 3.7.5 "Auxiliary Feedwater System"

Description:

Clarifies the portions of Surveillance Requirements (SRs) 3.7.5.3 and 3.7.5.4 which may be completed during normal operations.

Safety Evaluation:

This change is a clarification.

The pumps and valves will still be verified to actuate on receipt of an actual or simulated actuation signal. The testing frequency remains 24 months. There is no change to the level of assurance that the AFW system will perform as designed.

Those portions of the SR that could increase the risk of a plant transient will still be performed during outage conditions.

It was concluded that these changes may proceed without prior NRC approval.B 97-040 Change to Bases B 3.3.7 "Diesel Generator (DG) -Undervoltage Start"

Description:

Clarifies that the time delay of the Loss of Voltage (LOV) function applies to a step change from nominal voltage to zero volts at the relay and changes the identification of the channel to be calibrated in SR 3.3.7.4 from the undefined "DG -LOVS" to the clearer and more descriptive "Degraded Voltage and Loss of Voltage."

Safety Evaluation:

Loss of power (a moderate frequency incident) is evaluated in the Updated Final Safety Analysis Report (UFSAR) section 15.2.1.4.

The loss of power is followed by automatic startup of the standby diesel generators, the power output of which is sufficient to supply electrical power to all necessary engineered safety features systems and to provide the capability of maintaining the plant in a safe shutdown condition.

It was concluded that this change may proceed without prior NRC approval.-18-