ML13161A088
| ML13161A088 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/07/2013 |
| From: | St.Onge R Southern California Edison Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML13161A088 (15) | |
Text
SOUTHERN CALIFORNIA
~EDISON An EDISON INTERNATlONAL Company June 7,2013 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 Richard St. Onge Director, Nuclear Regulatory Affairs and Emergency Planning 10 CFR 50.59 10 CFR 72.48 San Onofre Nuclear Generating Station (SONGS) 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 SONGS Units 2 and 3 for the period from February 11, 2011 through January 31, 2013. 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 and implemented for the above time period. Complete facility change documentation is available onsite.
Evaluation N-SE 202026713-0023, completed prior to January 31,2013, was revised by N-SE 202026713-024 on April 19, 2013. Both the orig'inal and revised evaluations are included in Enclosure 1 A.
There are no new commitments made in this letter or enclosures. If you have any additional questions or require additional information, please contact Mark Morgan, Licensing Lead, at (949) 378-6745.
Sincerely,
('\\
Enclosures:
As stated cc:
A. Howell, Regional Administrator, NRC Region IV R. Hall, NRC Project Manager, SONGS Units 2 and 3 B. Benney, NRC Project Manager, SONGS Units 2 and 3 G. G. Warnick, NRC Senior Resident Inspector, SONGS Units 2 and 3 J. C. Staab, NRC Project Manager, SONGS ISFSI P.O. Box 128 San Clemente, CA 92674 SOUTHERN CALIFORNIA
~EDISON An EDISON INTERNATlONAL Company June 7,2013 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 Richard St. Onge Director, Nuclear Regulatory Affairs and Emergency Planning 10 CFR 50.59 10 CFR 72.48 San Onofre Nuclear Generating Station (SONGS) 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 SONGS Units 2 and 3 for the period from February 11, 2011 through January 31, 2013. 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 and implemented for the above time period. Complete facility change documentation is available onsite.
Evaluation N-SE 202026713-0023, completed prior to January 31,2013, was revised by N-SE 202026713-024 on April 19, 2013. Both the orig'inal and revised evaluations are included in Enclosure 1 A.
There are no new commitments made in this letter or enclosures. If you have any additional questions or require additional information, please contact Mark Morgan, Licensing Lead, at (949) 378-6745.
Sincerely,
('\\
Enclosures:
As stated cc:
A. Howell, Regional Administrator, NRC Region IV R. Hall, NRC Project Manager, SONGS Units 2 and 3 B. Benney, NRC Project Manager, SONGS Units 2 and 3 G. G. Warnick, NRC Senior Resident Inspector, SONGS Units 2 and 3 J. C. Staab, NRC Project Manager, SONGS ISFSI P.O. Box 128 San Clemente, CA 92674
ENCLOSURE 1A SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 FACILITY CHANGE REPORT (FCR) 10 CFR 50.59 EVALUATION SUMMARIES FOR THE PERIOD FROM FEBRUARY 11, 2011 THROUGH JANUARY 31, 2013
Page 1 of 11 201104972-60, Post Accident Clean Up (PACU) System (LCS Change)
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Description:==
In the event of a fuel handling accident (FHA), a fuel handling isolation signal (FHIS) from redundant airborne (noble gas) radiation monitors located in the exhaust ducts starts the post-accident cleanup (PACU) units and fuel pool pump room cooling units and isolates the fuel handling building. The proposed activity is to change the LCS 3.7.118 Bases to define a PACU train as OPERABLE regardless of whether it is automatically actuated by a FHIS or placed in service manually. The proposed activity also changes the isolation of the Fuel Handling Building (FHB) Ventilation System exhaust from automatically isolating on a FHIS to automatic or manual isolation, and changes the start of the fuel pool pump room cooling units from automatic to automatic or manual start.
Evaluation Summary:
The design basis FHA radiological consequences dose analysis does not credit operation of the PACU units and assumes that all of the post-FHA radioactive air in the FHB is entirely exhausted to the environment within two hours (per the Updated Final Safety Analysis Report (UFSAR) Section 15.7.3.4). As such, whether the actions are manually or automatically initiated, the radiological consequences are not changed from those reported in the UFSAR.
It is noted that the NRC Staff Safety Evaluation for U2/U3 Amendments 208/200 specifically states that the FHB airborne radiation monitors provide for the automatic isolation of the FHB Ventilation System and that the automatic isolation of the FHB Ventilation System is not a required safety function.
The fuel pool pump rooms may experience a peak temperature of 113 degrees Fahrenheit at two hours after a loss of offsite power. However, the most limiting components in the room are Room Temperature Switches which are qualified for a maximum ambient temperature of 140 degrees Fahrenheit. Since the room temperature does not exceed the qualification temperature of the most limiting component in the room, a change in the means of starting the fuel pool pump room cooling units is acceptable.
Page 2 of 11 201627832-005, Calculation M-0027-018, Rev. 0, Component Cooling Water (CCW)
System / Operating Unit to Outage Unit Spent Fuel Pool Heat Exchanger (SFPHX)
Cross-Tie
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Description:==
The proposed change establishes a non-50.54.X version of an inter-Unit CCW cross-tie.
Evaluation Summary:
If a single Saltwater Cooling Pump should fail during a unit outage, the cross-tie can provide backup cooling to the outage units Spent Fuel Pool heat exchanger (under specified requisite conditions). If those requisite conditions are not met, the existing 50.54X version of the procedure would still be available. The proposed change updates the UFSAR, Design Basis Document, and CCW system calculations and demonstrates that the non-50.54X version of the cross-tie satisfies the requirements of General Design Criteria (GDC) 5. GDC 5 states that structures, systems, and components (SSCs) important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cool down of the remaining units.
This evaluation ensures that the CCW flow rates and heat removal rates in the operating unit are unchanged by the cross-tie because new calculation M-0027-018 verifies that the CCW line temperatures, especially the return lines from the outage unit SFPHX(s), remain within their design limit of 135 degrees F. The calculation also confirms that there is no impact to the operating unit CCW flow rates to individual components, to CCW inlet temperatures to those components, to the heat removal capabilities of any of the components cooled by CCW, or to the heat removal capability of the Component Cooling Water Heat Exchanger itself.
Additionally, as discussed in the response to Question 2 of the 50.59 evaluation, the ability of the operating unit to respond to an accident and proceed to an orderly shutdown is not impaired.
In summary, this 10CFR50.59 evaluation has found that the proposed activity satisfies the requirements of GDC 5 and is acceptable for implementation.
Page 3 of 11 202026713-023, Nuclear Engineering Change Package (NECP) 800909421, Revision 1, Allow San Onofre Nuclear Generating Station (SONGS) System Separation Trip Scheme to be placed in "ALARM" but not "OFF".
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Description:==
The proposed activity is to, at the discretion of the Grid Control Center (GCC)/ California Independent System Operator (CAISO), disable the automatic tie-breaker overcurrent scheme of the system separation feature of the SONGS switchyard breakers with the SONGS units in any Mode (Modes 1, 2, 3, 4, 5, 6 or defueled). This will be accomplished by changing the position of switch 8-OL to "ALARM" (switch contacts shown on drawing 5145570; type and vendor drawing shown on legend of drawing 5145571) in the Southern California Edison (SCE) switchyard relay house. While in "AUTO" or "ALARM" positions, the SONGS control room will have manual SCE/San Diego Gas & Electric (SDG&E) system separation control (using control room pushbutton 2/3HS1703). SONGS procedures addressing system separation will also be changed to recognize that the SONGS system separation scheme may be in "ALARM" (manual separation functional).
Evaluation Summary:
Operation of one or both SONGS units in Mode 1, Mode 2, Mode 3 or Mode 4 with the system separation trip scheme "OFF" is not addressed or allowed by this 50.59 evaluation.
The frequency of occurrence of accidents and likelihood of malfunctions are not increased because manual system separation is unaffected and analysis demonstrates that with one or both SONGS units online or offline the overload condition that actuates automatic system separation will not be reached for GDC 17 grid contingency conditions (loss of largest unit, load or line) and the frequency category for Loss of Offsite Power (LOOP) is not affected by the presence or absence of the automatic separation feature.
Consequences of accidents and malfunctions are similarly unchanged because manual system separation is unaffected and analysis demonstrates that, with one or both SONGS units online or offline the overload condition that actuates automatic system separation will not be reached for GDC 17 grid contingency conditions, and hence disabling of automatic system separation will not affect the voltage or frequency of offsite power at the SONGS switchyard. As such, system performance is unchanged from that currently described in UFSAR chapter 15. Therefore, there are no changes to the UFSAR chapter 15 safety analyses.
No new accidents or malfunctions are created because the proposed change is limited to repositioning an existing selector switch in the SCE relay house in the SONGS switchyard and changing the associated SONGS procedures. The effects of loss of offsite power are already analyzed in UFSAR Chapter 15. There are no other changes to the switchyard (preferred offsite power source)--in particular, there are no changes to
Page 4 of 11 the existing provisions for fault clearing at the SONGS switchyard and manual system separation capability from the SONGS Control Room (required to maintain the independence of the two required preferred (offsite) power sources for GDC 17) is unaffected. Additionally, no plant systems are affected and there are no additional operator actions.
The change does not involve altering a Design Basis Limit for a Fission Product Barrier (DBLFPB), and does not result in a DBLFPB being exceeded because there are no changes to the voltage or frequency of offsite power at the SONGS switchyard.
Lastly, the change does not involve a change to a method of evaluation.
Page 5 of 11 202026713-024, Proposed Revision to Procedure SO23-13-4, Abnormal Operating Instruction
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Description:==
This revision to the previous evaluation 202026713-023 is performed to address concerns described in SONGS Notification 202261726:
During review of the switchyard control scheme with respect to Standard Review Plan 8.2 (which is in this respect identical to the NUREG 75/087 version to which SONGS was licensed) to determine how to resolve the Federal Energy Regulatory Commission/
CAISO concern with single failure, Design Engineering and Nuclear Regulatory Affairs personnel identified the omission of manual separation direction in pertinent SONGS procedures This revised evaluation supports the proposed revision to procedure SO23-13-4, Abnormal Operating Instruction At the discretion of the Grid Control Center/ CAISO, disable the automatic tie-breaker overcurrent scheme of the system separation feature of the SONGS switchyard breakers with the SONGS Unit 2 in any Mode (Modes 1, 2, 3, 4, 5, 6 or defueled). This will be accomplished by changing the position of switch 8-OL to "OFF" (switch contacts shown on drawing 5145570; type and vendor drawing shown on legend of drawing 5145571) in the SCE switchyard relay house. While switch 8-OL is in the "AUTO" position, the System Separation scheme is enabled for both Units 2 and 3. With switch 8-OL in the OFF position, the System Separation scheme is not available from the Control Room because handswitch 2/3HS1703 is disabled.
Evaluation Summary:
The frequency of occurrence of accidents and likelihood of malfunctions are not increased for Unit 2 because the incoming transmission lines, switchyard buses, and pathways to the onsite Class 1E buses can be isolated by operation of switches in the SONGS Control Room for the individual breakers on: the SCE incoming lines; the bus ties (which isolate the SDG&E incoming lines from the Unit 2 transformer positions); and the Unit 2 transformer positions. Unit 2 thus meets the Standard Review Plan 8.2 paragraph III.d.1 criteria independent of the status of the system separation scheme.
As such, for Unit 2, technical specifications operability of the two required offsite alternating current (AC) sources has not been affected by the position of the System Separation selector switch in the switchyard relay house. Manual system separation for Unit 2 is unaffected and analysis demonstrates that with one or both SONGS units on line or offline the overload condition that actuates automatic system separation will not be reached for GDC 17 grid contingency conditions (loss of largest unit, load or line) and the frequency category for LOOP is not affected by the presence or absence of the automatic separation feature.
Page 6 of 11 For Unit 3, only the incoming SCE transmission lines, switchyard buses, and pathways to the onsite Class 1E buses can be isolated by operation of switches in the SONGS Control Room for the individual breakers. Isolation of the incoming SDG&E lines can be performed from the SONGS Control room only by actuating system separation. Unit 3 thus requires system separation in order to meet the Standard Review Plan 8.2 paragraph III.d.1 criteria. As such, for Unit 3, technical specifications operability of the two required offsite AC sources has been affected by the position of the System Separation selector switch in the switchyard relay house, and switch position (or a modification to provide control of the individual breakers on the incoming SDG&E transmission lines from the SONGS Control Room) is a restraint to Unit 3 for entering mode 4.
Page 7 of 11 800162499-0320, Expansion of the Ovation Distributed Control System (DCS) to include Boric Acid Makeup (BAMU) and Primary Makeup Water Systems (PMW)
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Description:==
This 50.59 evaluation pertains to the aspects of Unit 2 NECP 800162890 and Unit 3 NECP 800162499 that were determined to be adverse in 50.59 screen assignments 800162890-0620 and 800162499-0380, respectively. The proposed activity will expand the Emerson Ovation distributed control system (DCS) in each unit to include the non-safety-related indication and control functions of the BAMU and PMW systems. The BAMU and PMW systems are subsystems of the Reactor Coolant Systems Chemical and Volume Control System (CVCS). The proposed activity will replace selected analog and discrete BAMU and PMW instrumentation and control components with a new DCS controller drop in the control room cabinet area and a new DCS control workstation on main control board 2(3)CR58. Because it will handle the CVCS functions of boration and dilution, the new Ovation control system in each unit will be referred to as the boration-dilution control system (BDCS).
Evaluation Summary:
This evaluation identified the inadvertent boron dilution accident as the only UFSAR-analyzed accident that could credibly be affected by the proposed 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 the proposed 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 proposed digital systems are bounded by the results of the failure modes of the existing equipment.
Page 8 of 11 800172197-0410, Unit 2 Replacement Reactor Vessel Head (RRVH), Replacement Control Element Drive Mechanisms (CEDMs), and Simplified Head Assembly (SHA)
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Description:==
Replace the Reactor Vessel Head (RVH) and CEDMs with new components that incorporate materials and methods of fabrication that will improve resistance to primary water stress corrosion cracking, eliminate the reactor vessel head dome and flange welds, and reduce the number of CEDM welds. Concurrently, Rapid Refueling Modifications are being installed to simplify the removal and installation of the head assembly during outages. The RRVH assembly for SONGS Unit 2 consists of one (1) piece of integrated vessel head (dome and flange), ninety-one (91) CEDM nozzles including guide cones, ten (10) In-core Instrument nozzles, one (1) vent pipe up to and including the vent flange, and fifteen (15) lifting lugs welded to the outside of the RVH dome. The SHA consists of 3 individual shrouds (lower, middle, and upper). The reactor head shroud is defined as the structure that bolts into the reactor vessel head lifting lugs and extends all the way to the connection pins for the tripod.
Evaluation Summary:
The RRVH is a single forged piece with fewer welds that is otherwise identical to the existing reactor vessel head. The changes in the CEDM design are focused at five areas: pressure housing materials; the interface between the ball seal housing and the upper pressure housing; the design of the bimetallic weld; the CEDM motor stationary lift magnet; and the geometry of the connection between the upper pressure housing and motor housing.
As discussed in the responses to Questions 1 and 2 of the 50.59 evaluation, the differences in the CEDM design are acceptable and satisfy applicable design criteria.
The structural evaluations for the SHA evaluated the adequacy of the new arrangement for missile impact, seismic loads, and heavy load drops. The analyses demonstrate that the calculated displacements for replacement SHA and its interfaces remain within design limits. Therefore, it is concluded that prior NRC review is not required.
Page 9 of 11 800873488-0124, Unit 2 Steam Generator (SG) Return to Service - 70% Power
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Description:==
The proposed activity is to establish an administrative operating Reactor Power limit of 70% for Unit 2 during the next operating cycle. The two specific aspects of the proposed activity that screened in are:
- 1. Adverse impacts of narrow-range SG water level measurement bias of -2.0% with respect to high water level in the SGs, and
- 2. Adverse impacts of SG blowdown flow measurement bias of -0.41% with respect to blowdown processing system operation and to secondary calorimetric calculation by Core Operating Limit Supervisory System (COLSS).
Evaluation Summary:
The increase in frequency of previously evaluated accidents is less than minimal because: the narrow-range steam generator level instrument bias does not affect the Feedwater Control System or inadvertent actuation of the Auxiliary Feedwater System; the blowdown system would still be operated within its design limits; COLSS is not an accident initiator; and no other SSCs are affected.
The increase in likelihood of previously evaluated malfunctions of SSCs important to safety is less than minimal because: the SG water level would be unchanged from that assumed for the high level alarm and reactor trip in existing analyses for increase in feedwater flow, so that flooding of the level taps providing alarm and trip would not occur prior to actuation; the blowdown system would still be operated well within its design limits; and COLSS system/components and other SSCs are unaffected.
The increase in consequences of previously evaluated accidents is less than minimal because: the SG water level and timing of the high level alarm and reactor trip are unchanged from that assumed in existing analyses for increase in feedwater flow; the mass/energy and hence radiological releases from postulated blowdown system piping ruptures are independent of the initial system flow rate; the impact on the COLSS calorimetric is insignificant; and no other SSCs are affected.
No accidents of a different type are created because: the affected level channels are not initiators of any accidents and no new failure modes are introduced.
No malfunctions of SSCs important to safety with a different result are created because no new failure modes are introduced.
No design basis limit for a fission product barrier is exceeded or altered because:
narrow-range steam generator high water level and high blowdown flow rate are not associated with any fission product barrier.
Page 10 of 11 There is no departure from a method of evaluation described in the UFSAR because the activity does not involve a change to a method of evaluation.
Page 11 of 11 800876795-0030, Emergency Diesel Generator (EDG) Non Critical Trip Bypass Modification
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Description:==
This proposed activity changes the method of control of the EDG described in the UFSAR for the scenario of an emergency start/run during a loss of voltage signal (LOVS) condition. Each EDG is supported by protection features which are classified as either critical or noncritical as described in UFSAR 8.3.1.1.4.3. Presently, during a Safety Injection Actuation Signal (SIAS), all EDG noncritical protective trips are bypassed for the duration of the SIAS condition until operator action re-enables them by means of the SIAS override feature. This differs from the scenario of the LOVS condition in which the EDG will start and run with all respective protective trips, both critical and noncritical, enabled and available to trip the EDG should any of the respective trip setpoints be reached. This modification will add an automatic bypass circuit for the EDG noncritical trips during a LOVS condition. Additionally, this modification provides the capability for manual reset of the noncritical trip bypass circuitry upon restoration of offsite power by means of a new pushbutton added to control room panel CR63.
Evaluation Summary:
This modification adds a bypass of the noncritical protective features for the emergency condition of LOVS and does not alter the already existing bypass of these features for the accident condition. Bypassing these features for the emergency condition is consistent with the regulatory position to avoid interfering with the successful operation of the EDG when it is most needed. As described in the answers to the evaluation criteria, no new accidents or malfunctions are created because the proposed change is limited to a new bypass circuit for the LOVS condition. Also, the new manual reset pushbutton to re-enable the noncritical protective features upon restoral of offsite power meets the Reg Guide 1.9, section 1.8 requirement that Capacity for automatic reset (of the trip bypass function) is not acceptable. The new LOVS reset pushbutton will have no adverse effect on the existing design function or operation of the EDG SIAS automatic start and manual override feature.
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 FEBRUARY 11, 2011 THROUGH JANUARY 31, 2013
Page 1 of 1 200877882-0006, 10CFR72.48 Evaluation of Damage to the OS197-3 Transfer Cask Inner.
SO23-207-16-M175 (TN NCR 2011-018 Rev 1, LR 721004-913)
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Description:==
Evaluation of new scratches found on the inside surface (the inner liner) of the OS197-3 Transfer Cask owned by Southern California Edison while being used at Cooper Nuclear Station.
Evaluation Summary:
The interior area between the rails of the Transfer Cask meets the minimum required wall thickness of 0.44 inches, and The gouges located along the edge of the liner plate at the cask opening are dispositioned as Use-as-Is by Transnuclear and have no effect upon the structural strength of the inner liner or affect the thermal or shielding performance.
The Transfer Cask is acceptable without repair (a Use-As-Is disposition) with no impact on the license or certificate of compliance.