ML16239A044

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Seventh Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)
ML16239A044
Person / Time
Site: Oyster Creek
Issue date: 08/26/2016
From: Jim Barstow
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EA-12-051
Download: ML16239A044 (37)
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Text

Exelon Generation (!

Order No. EA-12-051 RS-16-151 RA-16-066 August 26, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Seventh Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

References:

1. NRC Order Number EA-12-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation," dated March 12, 2012
2. NRC Interim Staff Guidance JLD-ISG-2012-03, "Compliance with Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation,"

Revision 0, dated August 29, 2012

3. NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation," Revision 1, dated August 2012
4. Exelon Generation Company, LLC's Initial Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated October 25, 2012
5. Exelon Generation Company, LLC Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated February 28, 2013 (RS-13-033)
6. Exelon Generation Company, LLC First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated August 28, 2013 (RS-13-124)
7. Exelon Generation Company, LLC Second Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated February 28, 2014 (RS 023)
8. Exelon Generation Company, LLC Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated August 28, 2014 (RS-14-201)

U.S. Nuclear Regulatory Commission Integrated Plan Report to EA-12-051 August26,2016 Page 2

9. Exelon Generation Company, LLC Fourth Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated February 27, 2015 (RS-15-031)
10. Exelon Generation Company, LLC Fifth Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051 ), dated August 28, 2015 (RS-15-203)
11. Exelon Generation Company, LLC Sixth Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), dated February 26, 2016 (RS-16-031)
12. NRC letter to Exelon Generation Company, LLC, Oyster Creek Nuclear Generating Station - Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Pool Instrumentation (TAC No. MF0823), dated November 8, 2013 On March 12, 2012, the Nuclear Regulatory Commission ("NRC" or "Commission") issued an order (Reference 1) to Exelon Generation Company, LLC (EGC). Reference 1 was immediately effective and directs EGC to install reliable spent fuel pool level instrumentation. Specific requirements are outlined in Attachment 2 of Reference 1.

Reference 1 required submission of an initial status report 60 days following issuance of the final interim staff guidance (Reference 2) and an overall integrated plan pursuant to Section IV, Condition C. Reference 2 endorses industry guidance document NEI 12-02, Revision 1 (Reference 3) with clarifications and exceptions identified in Reference 2. Reference 4 provided the EGC initial status report regarding reliable spent fuel pool instrumentation. Reference 5 provided the Oyster Creek Nuclear Generating Station overall integrated plan.

Reference 1 requires submission of a status report at six-month intervals following submittal of the overall integrated plan. Reference 3 provides direction regarding the content of the status reports. References 6, 7, 8, 9, 10, and 11 provided the first, second, third, fourth, fifth, and sixth six-month status reports, respectively, pursuant to Section IV, Condition C.2, of Reference 1 for Oyster Creek Nuclear Generating Station. The purpose of this letter is to provide the seventh six-month status report pursuant to Section IV, Condition C.2, of Reference 1, that delineates progress made in implementing the requirements of Reference 1. The enclosed report provides an update of milestone accomplishments since the last status report, including any changes to the compliance method, schedule, or need for relief and the basis, if any. The enclosed report al~o addresses the NRC Interim Staff Evaluation Request for Additional Information Items contained in Reference 12.

This letter contains no new regulatory commitments. If you have any questions regarding this report, please contact David P. Helker at 610-765-5525.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 26th day of August 2016.

U.S. Nuclear Regulatory Commission Integrated Plan Report to EA-12-051 August 26, 2016 Page 3 Respectfully submitted, James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Enclosure:

1. Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation cc: NRC Regional Administrator - Region I NRC Senior Resident Inspector- Oyster Creek Nuclear Generating Station NRC Project Manager, NRR - Oyster Creek Nuclear Generating Station Mr. John D. Hughey, NRR/JLD/JOMB, NRC Manager, Bureau of Nuclear Engineering - New Jersey Department of Environmental Protection Mayor of Lacey Township, Forked River, NJ

Enclosure Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (33 pages)

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation 1 Introduction Oyster Creek Nuclear Generating Station developed an Overall Integrated Plan (Reference 1 in Section 8), documenting the requirements to install reliable Spent Fuel Pool Instrumentation System (SFPIS), in response to Reference 2. This enclosure provides an update of milestone accomplishments since submittal of the sixth six-month status report including any changes to the compliance method, schedule, or need for relief/relaxation and the basis, if any.

2 Milestone Accomplishments The following milestone has been completed since the development of the sixth six-month status report (Reference 11 ), and is current as of August 26, 2016.

  • Complete installation and testing 3 Milestone Schedule Status The following provides an update to the milestone schedule to support the Overall Integrated Plan. This section provides the activity status of each item, and the expected completion date noting any change. The dates are planning dates subject to change as design and implementation details are developed.

Revised Target Completion Activity Target Milestone Date Status Completion Date Submit 60 Day Status Report October 25, 2012 Complete Submit Overall Integrated Plan February 28, 2013 Complete Submit 6 Month Updates:

Update 1 August28,2013 Complete Update 2 February 28, 2014 Complete Update 3 August28,2014 Complete Update 4 February 27, 2015 Complete Update 5 August28,2015 Complete Update 6 February 26, 2016 Complete Complete with Update 7 August28,2016 this submittal Page 1 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 Revised Target Completion Activity Target Milestone Date Status Completion Date Provide Final Safety Evaluation Complete with August 28, April 30, 2016 (SE) Info this submittal 2016 Modifications:

Conceptual Design 302012 Complete Issue Exelon Fleet contract to 202013 Complete procure SFPIS Equipment Begin Detailed Design 102015 Complete Engineering Complete and Issue SFPIS 302015 Complete Modification Package 102016 Begin Installation 202016 Complete (March)

Complete SFPIS Installation 302016 Complete and Testing Compliance 302016 Started 4 Changes to Compliance Method There are no changes to the compliance method as documented in the Overall Integrated Plan (Reference 1).

5 Need for Relief/Relaxation and Basis for the Relief/Relaxation Oyster Creek Nuclear Generating Station expects to comply with the order implementation date and no relief/relaxation is required at this time.

6 Open Items from Overall Integrated Plan and Draft Safety Evaluation The following tables provide a summary of the open items documented in the Overall Integrated Plan or the Draft Safety Evaluation (SE) and the status of each item.

Page 2 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August 26, 2016 Overall Integrated Plan Open Items 01# Description Status 1 Open Item: Complete Continuous level indication (Addressed in Reference 6) will be provided by a guided wave radar system, submersible pressure transducer, or other appropriate level sensing technology that will be determined during the detailed engineering phase of the project.

2 RAI Question: Complete (RAl-1, Please provide a clearly labeled Reference Attachment 1 for a sketch depicting the Ref. 4) sketch depicting the elevation elevation view of the proposed mounting view of the proposed typical arrangement of the mounting brackets and level mounting arrangement for the sensors. Datum points Level 1, Level 2, Level 3, top of fuel racks and level sensor measureable portions of instrument channel range are all depicted in this sketch.

consisting of permanent measurement channel LEVEL 1 - This is the level at which point the equipment (e.g., fixed level reliable suction loss occurs due to uncovering of the weirs. Level 1 corresponds to elevation 117'-

sensors and/or stilling wells, 10 11

  • and mounting brackets).

Indicate on this sketch the LEVEL 2 - This is the level that is adequate to datum values representing provide substantial radiation shielding for a person standing on the spent fuel pool operating deck.

Level 1, Level 2, and Level 3 as Level 2 corresponds to elevation 106'-1 1/8 11 ,

well as the top of the fuel racks.

water level 1o* above the highest point of any fuel Indicate on this sketch the rack seated in the spent fuel pool.

portion of the level sensor measurement range that is LEVEL 3 - This is the level where fuel remains covered and actions to implement make-up water sensitive to measurement of the addition may no longer be deferred. This level fuel pool level, with respect to corresponds to a water level 7 7/8" above the top the Level 1, Level 2, and Level 3 of the fuel assembly. Level 3 shall be designated datum points. at elevation 96 1-9 11

  • The level sensors' measurement range is between elevation 96'-9" and 118'-6 7/8".

3 RAI Question: Complete (RAl-2, Please provide a clearly labeled See Attachment 2 for a clearly labeled sketch Ref.4) sketch or marked-up plant depicting the plan view of the Spent Fuel Pool, drawing of the plan view of equipment locations and cable routinQ. The cable Page 3 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 the SFP area, depicting the SFP routing will extend from the level sensors towards inside dimensions, the planned the level displays.

locations/placement of the primary and backup SFP The SFP is 27' x 39' and the mounting brackets level sensor, and the proposed will be anchored on the south side of the SFP.

routing of the cables that The level sensors shall be separated by a will extend from the sensors distance comparable to the length of the shortest toward the location of the read- side of the pool.

out/display device.

The level transmitters will be located in the Cable Tray Bridges, adjacent to the Reactor Building.

The level indicators will be located in the Upper Cable Spreading Room above the Main Control Room.

4 RAI Question: Complete (RAl-3, Please provide the following:

a) The design criteria that will a) All SFPIS equipment is designed in Ref.4) accordance with the Oyster Creek Safe Shutdown be used to estimate the total loading on the mounting Earthquake (SSE) design requirements.

device(s), including static weight loads and dynamic The vendor, Westinghouse, has evaluated the loads. Describe the structural integrity of the mounting brackets in methodology that will be used calculations CN-PEUS-15-08. The GTSTRUDL to estimate the total loading, model, used by Westinghouse to calculate the inclusive of design basis stresses in the bracket assembly, considered load maximum seismic loads and combinations for the dead load, live load and the hydrodynamic loads that seismic load on the bracket. The reactionary could result from pool sloshing or other effects that could forces calculated from these loads become the accompany such seismic design inputs to design the mounting bracket forces. anchorage to the refuel floor to withstand a Safe b) A description of the manner Shutdown Earthquake (SSE).

in which the level sensor (and stilling well, if appropriate) will Seismic be attached to the refueling The seismic loads are obtained from the response floor and/or other support structures for each planned spectra curves of Oyster Creek. The following point of attachment of the methodology was used in determining the probe assembly. Indicate in a stresses on the bracket assembly:

schematic the portions of the level sensor that will serve as

  • Frequency analysis, taking into account the dead points of attachment for weight and the hydrodynamic mass of the mechanical/mounting or structure, is performed to obtain the natural electrical connections. frequencies of the structure in all three directions.

c) A description of the manner by which the mechanical *SSE (Safe Shutdown Earthquake) response connections will attach the spectra analysis is performed to obtain member level instrument to permanent stresses and support reactions.

SFP structures so as to Page 4 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 supportthelevelsensor *Modal responses are combined using the Ten assembly. Percent Method per U.S. NRC Regulatory Guide 1.92, Revision 1, "Combining Modal Responses and Spatial Components in Seismic Response Analysis".

  • The seismic loads for each of the three directions are combined by the Square Root of the Sum of Squares (SASS) Method.
  • Sloshing analysis is performed to obtain liquid pressure and its impact on bracket design.
  • The seismic results are combined with the dead load results and the hydrodynamic pressure results in absolute sum. These combined results are compared with the allowable stress values.

Sloshing Sloshing forces were obtained by analysis. The TID-7024, Nuclear Reactors and Earthquakes, 1963, by the US Atomic Energy Commission, approach has been used to estimate the wave height and natural frequency. Horizontal and vertical impact force on the bracket components was calculated using the wave height and natural frequency obtained using TID-7024 approach.

Using this methodology, sloshing forces have been calculated and added to the total reactionary forces that would be applicable for bracket anchorage design. The analysis also determined that the level probe can withstand a credible design basis seismic event. During the design basis event, the SFP water level is expected to rise and parts of the level sensor probe are assumed to become submerged in water. The load impact due to the rising water and submergence of the bracket components has also been considered for the overall sloshing impact.

Reliable operation of the level measurement sensor with a submerged interconnecting cable has been demonstrated by analysis of previous Westinghouse testing of the cable, and the vendor's cable qualification.

The following Westinghouse documents provide information with respect to the design criteria used, and a description of the methodoloQy used Page 5 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 to estimate the total loading on the device.

1) Calculation CN-PEUS-15-08, "Seismic Analysis of the SFP Primary and Backup Mounting Brackets at Oyster Creek Nuclear Generating Station"
2) EQ-QR-269, WNA-TR-03149-GEN Seismic Qualification of other components of SFPIS Oyster Creek Station specific evaluations were performed in ECR OC 14-00389 to address the seismic qualification and structural requirements of other SFPIS equipment. The design criteria in the ECR meet the requirements to withstand a SSE. The methods used in the ECR follow SQUG-GIP, Rev. 2A, IEEE Standard 344-2004 and IEEE Standard 323-2003 for seismic qualification of the instrument.

b) The level sensor, which is one long probe, will be suspended from the launch plate via a coupler/connector assembly. The launch plate is a subcomponent of the bracket assembly, which will be mounted on the refueling floor (Reactor Building EL. 119'-3") via concrete anchors.

Drawing 10067E58, Sheet 1 prepared by Westinghouse shows the connection/mounting details.

c) The bracket assembly that supports the sensor probe and launch plate will be mechanically fastened to the concrete slab of the refueling floor.

Each mechanical connection consists of four concrete expansion anchors that will bolt the bracket assembly to the SFP structure via a base plate. The concrete expansion anchors will be designed to withstand SSE and will meet the Oyster Creek safety related installation requirements. The analysis and details of the Pool-side brackets are provided in Calculation CN-PEUS-15-08 and Drawing 10067E58, Sheet

1. The design and details of the anchorage to the floor are provided in ECR OC 14-00389.

5 RAI Question: Complete (RAl-4, Please provide the following:

a.) Beyond Design Basis Environment -

Ref.4) a) A description of the specific method or combination of Environmental Conditions for SFPI components methods that will be applied to installed in the Spent Fuel Pool area at Oyster Page 6 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 demonstrate the reliability of Creek are bounded by Westinghouse test the permanently installed conditions. Westinghouse qualified SFPI equipment under beyond- instrument to operate in BOB conditions for design-basis (BOB) ambient radiation TIO 1.EO? Rads y. During the BOB event temperature, humidity, shock, vibration, and radiation with water at level 3 (12" above top of fuel rack),

conditions. at Oyster Creek Station radiation Tl D at coupler b) A description of the testing connection will be 8E06 Rads y (per Calculation and/or analyses that will be BYR13-051 - NEI 12-02 Spent Fuel Pool Doses).

conducted to provide The radiation value of 8.E06 R y is lower than assurance that the equipment 1.E07 Ry to which Westinghouse qualified the will perform reliably under the coupler connection and as a result it is bounded.

worst-case credible design Below the coupler connection, the probe is made basis loading at the location of stainless steel. Stainless steel is resistant to where the equipment will be radiation, therefore there are no concerns with the mounted. Include a discussion of this seismic reliability SFPI operation (per Westinghouse letter LTR-demonstration as it applies to SFPIS-13-35). The equipment is qualified by (i) the level sensor mounted in Westinghouse to function during BOB events at the SFP area, and (ii) any 212 deg F and 100% humidity conditions to meet control boxes, electronics, or the NEI 12-02 Revision 1 requirements (EQ-QR-read-out and re-transmitting 269). As a result, SFPI is qualified to function devices that will be employed under BOB environmental conditions of heat, to convey the level humidity and radiation at Oyster Creek in the SFP information from the level area.

sensor to the plant operators or emergency responders. Mild Environment - Westinghouse qualified the c) A description of the specific system components (display panel, transmitter) method or combination of that reside in the mild environment conditions to methods that will be used to determine that the components can satisfactorily confirm the reliability of the perform to those conditions. Westinghouse has permanently installed equipment such that following determined that aging does not have a significant a seismic event the effect on the ability of the equipment to perform instrument will maintain its following a plant design basis earthquake. Exelon required accuracy. has reviewed the documents and found them acceptable. Reference Westinghouse documents EQ-QR-269, WNA-TR-03149-GEN for description of specific qualification methods.

Below are Oyster Creek specific details for equipment location and associated qualifications:

Transmitter:

Radiation: Oyster Creek Environmental Qualification report ES-027 Revision 4 states radiation limit "Not Available" where the transmitter is installed. The area is a non-harsh environment. Per the radiological survey map for this location during normal operation, the radiation Page 7 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 in this area (Cable Tunnel above Office Room and Turbine Building) will be <1 mR/hr, and as a result is not expected to exceed 1.E03 Rads.

Temperature: Per SDD-OC-811-A DIV1, Rev 1, Fire Protection New Cable Spreading Room &

Cable Bridge Tunnel document, the cable bridge tunnel area, where the transmitters will be installed, can experience temperatures between O

-130 deg F. For BOB conditions, transmitters are qualified up to 140 deg F. As a result, there is no concern with operating transmitters in cable bridge tunnel area during BOB event conditions. During normal operating conditions, transmitters are qualified by Westinghouse to operate between 50 deg F - 120 deg F per EQ-QR-269. However, upon further investigation, Westinghouse found transmitter components are qualified by manufacturer (Operating Instruction Manual 01/MT 5000 - EN, Revision H 05.2013) (ABB) to operate between -40 deg F to 170 deg F, per Westinghouse letter LTR-SFPIS-16-07 Revision 1.

As a result, Oyster Creek installed location for transmitter is bounded by manufacturer qualified range for operation.

Humidity: Per SDD-OC-811-A DIV1, Rev 1, Fire Protection New Cable Spreading Room & Cable Bridge Tunnel document, the cable bridge tunnel area, where the transmitters will be installed, can experience humidity between O - 100%.

Transmitters are qualified by Westinghouse to operate between 0 - 95% humidity per EQ-QR-269. Transmitter housings are NEMA - 4X, which means they are leak tight. There are two connection points for the transmitter. One for coax cable and one for 4 - 20mA cable connection point. Coax cable connector to transmitter housing is qualified by Westinghouse to be leak tight. This is the same straight connector that Westinghouse qualified to install in SFP area to connect coax cable to probe. For straight connector Westinghouse recommends to install Raychem boot seal to achieve water tight seal (WNA-TR-3149). Oyster Creek installed Raychem boot seal for the straight connector since the transmitter is installed in a location that can experience 100%

Page 8 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 humidity condition to achieve water tight connection. As for 4 - 20mA connection point, Oyster Creek installed water tight seal to prevent condensation intrusion into the connection point.

Display

Enclosures:

Radiation: Oyster Creek Environmental Qualification report ES-027 Revision 4 states radiation limit "Not Available" where the display enclosures are installed. The area is a non-harsh environment. Per the radiological survey map for this location during normal operation, the radiation in this area (Upper Cable Spreading Room) will be

<1 mR/hr, and as a result is not expected to exceed 1.E03 Rads.

Temperature: Per 15050-M4-001 "Calculated Temperature in Upper Cable Spreading Room without Ventilation", the upper cable spreading room area, where the display enclosures will be installed, can experience temperatures between 40 -102 deg F. For BOB conditions, display enclosures are qualified up to 140 deg F. As a result, there is no concern with operating display enclosures in upper cable spreading room area during BOB event conditions. During normal operating conditions, display enclosures are qualified by Westinghouse to operate between 50 deg F - 120 deg F per EQ-QR-269. However, upon further investigation, Westinghouse found components within display enclosures are qualified by the manufacturers to operate between

-4 deg F to 158 deg F, per Westinghouse letter LTR-SFPIS-16-07 Revision 1. As a result, Oyster Creek installed location for display enclosures is bounded by manufacturer qualified range for operation.

Humidity: Per SDD-OC-811-A DIV1, Rev 1, Fire Protection New Cable Spreading Room & Cable Bridge Tunnel document, the upper cable spreading area, where the display enclosures will be installed, can experience humidity between 0 -

100%. Display enclosures are qualified by Westinghouse to operate between 0 - 95%

humidity per EQ-QR-269. Though the SDD-OC-811-A DIV1, Rev 1, Fire Protection New Cable Page 9 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 Spreading Room & Cable Bridge Tunnel document states humidity can be up to 100% in upper cable spreading area, the area is in a building that does not have any sources of water that can lead up to 100% humidity. Also, the display enclosures are a NEMA 4X enclosure per Westinghouse Report WNA-TR-03149. There are two connection points to each display enclosure.

One for 4-20mA cable connection to display enclosure and one for power conduit connection.

To enhance these two connections to make them water intrusion proof when operating in 100%

humidity condition, Oyster Creek added water tight seals to both connection points to display enclosures.

Below table summarizes qualification limits for transmitters and display enclosures for normal operating conditions:

Westinghouse Site Actions Qualified Taken Transmitter Contacted vendor and Temperature 50 -120 deg F 0-vendor 130 data sheet deg F qualified to

-40 to 170 deg F Humidity 0-95% 0- Install 100% water proof seal for 4-20mA connection and Raychem boot seal for straight connector at coax cable connection.

Page 10 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 Display Enclosure Contacted vendor and Temperature 50 - 120 deg F 40 -

vendor 102 data deg F sheets qualified to

-4 to 158 deg F Install water proof Humidity 0-95% 0-seals for 4-100%

20mA signal connection and power conduit connection.

Since the transmitter and display enclosure are installed in a location that is outside Westinghouse qualified range for operation (even though manufacturer individually qualified the components beyond Westinghouse qualified range), Westinghouse recommends performing additional monitoring (once per shift) of the level display when operating outside the qualified environment (50 - 120 deg F), per Westinghouse letter LTR-SFPIS-16-07 Revision 1. Oyster Creek will perform these additional monitoring steps to ensure equipment is functioning properly when operating outside the qualified operating range.

Oyster Creek will obtain display level from both primary and backup channels and will compare it to actual water level in SFP by performing a physical walkdown. This is to ensure both displays are functioning properly against the real water level in the SFP. The acceptance criteria for declaring functionality will be +/- 1 foot when operating outside qualified range for operation.

This acceptance criterion is in accordance with the NEI 12-02 Revision 1 guidance requirements.

In addition to enhanced monitoring, Oyster Creek Page 11of33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 will also reduce time to implement compensatory measures when one channel is found non-functional from 90 days to 45 days, when operating outside the qualified range. This will ensure instruments are returned to a functional status in a timely manner.

Shock and Vibration - SFPIS pool side brackets were analyzed for Safe Shutdown Earthquake design requirements per NRC Order EA 051 and NEI 12-02 guidance. As provided by the NRC Order EA-12-051, the NEI 12-02 guidance and as clarified by the NRC interim staff guidance, the probe, coaxial cable, and the mounting brackets are "inherently resistant to shock and vibration loadings." As a result, no additional shock and vibration testing is required for these components. SFPIS pool side brackets for both the primary and backup Westinghouse SFP measurement channels will be permanently installed and fixed to rigid refuel floors, which are Seismic Category 1 structures. The SFPIS system components, such as level sensor and its bracket, display enclosure and its bracket, were subjected to seismic testing, including shock and vibration test requirements.

The results for shock and vibration tests were consistent with the anticipated shock and vibration expected to be seen by mounted equipment. The level sensor electronics are enclosed in a NEMA-4X housing.

The display electronics panel utilizes a NEMA-4X rated stainless steel housing as well. These housings will be mounted to a seismically qualified wall and will contain the active electronics, and aid in protecting the internal components from vibration induced damage. Reference Westinghouse reports WNA-DS-02957, WNA-TR-03149-GEN for Page 12 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 shock and vibration.

b) The seismic adequacy of the SFPIS (all components) is demonstrated by vendor testing and analysis in accordance with below listed standards:

  • IEEE 344-2004, IEEE Recommended Practice for Seismic Qualification of Class 1E Electrical Equipment for Nuclear Power Generating Stations
  • IEEE-323-1974, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations
  • Design Change Package ECR OC 14-00389, "Reliable Spent Fuel Pool Level Instrumentation (Fukushima)"

c) Westinghouse has seismically qualified the SFPIS instrument and its components. CN-PEUS-15-08 describes Pool-side Bracket Seismic Analysis, EQ-QR-269, WNA-TR-03149-GEN, EQ-TP-353 describe remaining seismic qualifications of the instrument components. With the instrument being seismically qualified and installed as described, the instrument is assured to maintain reliable and accurate indication when required.

Westinghouse report WNA-CN-00301-GEN provides the channel accuracy from measurement to display.

6 RAI Question: Complete (RAl-5, Please provide the following:

a) A description of how the two The two channels of the proposed level Ref.4) channels of the proposed level measurement system will be installed such that:

measurement system meet this requirement so that the a) The level probes will be mounted on the south potential for a common cause Page 13 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 event to adversely affect both side of the SFP and will be separated by a channels is minimized to the distance greater than the span of the shortest side extent practicable. of the pool. This meets the NEI 12-02, Revision 1 b) Further information on how guidance for channel separation.

each level measurement system, consisting of level sensor electronics, cabling, and b) The coax cables from the level instruments in readout devices will be the SFP area are routed directly one floor below designed and installed to into the reactor building elevation 95' 3" and as a address independence through result the separation between the coax cables in the application and selection of independent power sources, the refuel floor area is maintained by a distance the use of physical and spatial greater than the span of the shortest side of the separation, independence of pool. Outside of the refuel floor area, the coax signals sent to the location(s) cables maintain a minimum separation of 8ft. The of the readout devices, and the primary coaxial cable will be installed in conduit independence of the displays. and the secondary coaxial cable will be installed in combination of conduit and cable tray. The primary level transmitter is installed on the south Cable Tray Bridge and the secondary level transmitter is installed in the north Cable Tray Bridge on elevation 80 ft. The two level transmitters are physically separated of approximately 15 ft. The indicator electronics I UPS enclosures for the primary and secondary instrument channels are installed in a non-hazardous area, the Upper Cable Spreading Room elevation 68' -6" with a physical separation of 5 ft. The 4-20mA cables that extend from the transmitters to the displays have a minimum separation of 1 inch, maintaining the 1E separation requirements per Oyster Creek Station Specification SP-9000-41-005, Cables &

Raceways. The primary 4-20mA cable will be installed in conduit and the secondary 4-20mA cable will be installed in combination of conduit and cable tray. Distribution panels PDP-733-57 &

PDP-733-58 are located in the Lower Cable Spreading Room. The 120VAC cables from the distribution panels terminate to panel 17R in the Control Room and to the displays in the Upper Cable Spreading Room. The 120VAC power cables are 1 inch apart, maintaining the 1E separation requirements per Oyster Creek Station Specification SP-9000-41-005, Cables &

Raceways. All the equipment and supports will be seismically mounted. The 120VAC distribution panels for the primary and secondary instruments Page 14 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 are powered by different 480V safety buses.

Therefore, the loss of any one bus will not result in the loss of AC power to both instrument channels.

7 RAI Question: Replaced by Interim SE RAI -9.

(RAl-6, Please provide the following:

Ref.4) a) A description of the normal electrical AC power sources and capacities for the primary and backup channels. Describe how these AC sources are independent, and how they may be restored following an extended loss of AC event. b) If the level measurement channels are to be powered through a battery system (either directly or through an Uninterruptible Power Supply (UPS)), please provide the design criteria that will be applied to size the battery in a manner that ensures, with margin, that the channel will be available to run reliably and continuously following the onset of the BDB event for the minimum duration needed, consistent with the plant mitigation strategies for BDB external events (Order EA-12-049).

8 RAI Question: Complete (RAl-7, Please provide the following:

a) An estimate of the expected a) The Westinghouse documents WNA-CN-00301 Ref.4) instrument channel accuracy and WNA-DS-02957-GEN describe the channel performance under both (i) accuracy under both (a) normal SFP level normal SFP level conditions condition and (b) at the Beyond Design Basis (approximately Level 1 or (BOB) condition that would be present if SFP level higher) and (ii) at the BOB were at Level 2 and Level 3 datum points. Each conditions (i.e., radiation, instrument channel will be accurate to within +/-3" temperature, humidity, post- during normal spent fuel pool level conditions. The seismic and post-shock instrument channels will retain this accuracy after conditions) that would be BOB conditions, in accordance with the above present if the SFP level were at the Level 2 and Level 3 Westinghouse documents. This value is within the datum points. channel accuracy requirements of the Order (+/-1 Page 15 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26 2016 b) A description of the foot).

methodology that will be used for determining the maximum b) The Westinghouse document WN~-TP-04709-allowed deviation from the GEN calibration procedure and technical manual instrument channel design describes the methodology for routine accuracy that will be testing/calibration verification and calibr.~tion employed under normal methodology. This document also spec1f1es th~

operating conditions as an required accuracy criteria under normal operating acceptance criterion for a conditions. Oyster Creek Station will follow the calibration procedure to flag guidance and criteria provided in this document. .

to operators and to Instrument channel calibration will be performed if technicians that the channel the level indication reflects a value that is outside requires adjustment to within the acceptance band provided in Westinghouse the normal condition design documents. Calibration will be performed once per accuracy. refueling cycle for Oyster Creek Station. Per Westinghouse document WNA-TP-04709-GEN, calibration on a SFP level channel is to be completed within 60 days of a ~lanned ref~eling outage considering normal testing scheduling .

allowances (e.g., 25%). This is in compliance with the NEI 12-02 guidance for Spent Fuel Pool Instrumentation.

9 RAI Question: Complete (RAl-8, Please provide the following:

a) Westinghouse provided test equipment ~hat Ref.4) a) A description of the provides the capability to calibrate the equipment.

capability and provisions the Westinghouse calibration procedure WNA-TP-proposed level sensing 04709-GEN provide instructions to use the test equipment will have to enable equipment to perform a calibration. T.h~ .

periodic testing and Westinghouse calibration procedure 1s included m calibration, including how this the Oyster Creek's site vendor manual VM-OC-capability enables the 6652. Oyster Creek will perform an in-situ test equipment to be tested in-situ.

during the functional check based on the b) A description of how such Westinghouse Two Point Verification Method, testing and calibration will LTR-SFPIS-14-55.

enable the conduct of regular channel checks of b) The levels displayed by the channels and other each independent channel SFP level instrumentation shall be checked against the other, and against against each other and checked against physical any other permanently- pool level per the Oyster Creek Station processes.

installed SFP level If the level is not within the required accuracy per instrumentation. Westinghouse recommended tolerances, channel c) A description of how calibration will be performed.

functional checks will be c) Oyster Creek will perform a PM to perform a performed, and the frequency functional check 60 days prior to a refuel outage.

at which they will be The PM establishes the current water level by conducted. Describe how measuring the distance to the water referenced calibration tests will be from the bottom of the launch plate. This perlormed,andthefrequency measured distance is then compared to the level at which they will be indication obtaininq the As-Found indication value.

Page 16 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 conducted. Provide a The probe is then lifted out of the water to a discussion as to how these predetermined mark on the probe. The water level surveillances will be at the predetermined mark is then recorded. The incorporated into the plant probe is lowered back into the water freely surveillance program. suspended from the launch plate. The level d) A description of what indication is recorded. If the As-Found Level preventive maintenance tasks indication is within the tolerance specified, the are required to be performed procedure is exited. If the As-Found value is not during normal operation, and within tolerance the calibration is performed. The the planned maximum PM and calibration steps were taken from surveillance interval that is Westinghouse document WNA-TP-04709-GEN necessary to ensure that the Spent Fuel Pool Instrument System Calibration channels are fully conditioned Procedure and Westinghouse Two Point to accurately and reliably Verification Method, LTR-SFPIS-14-55.

perform their functions when d) Oyster Creek has developed a preventive needed.

maintenance task to perform a functional check 60 days prior to a refuel outage. Oyster Creek has developed a preventive maintenance task to perform an annual visual inspection of the sensor probes (full range) to ensure there is no damage or corrosion.

10 RAI Question: Complete (RAl-9, Please provide the following: a) The Oyster Creek Station primary and backup Ref.4) a) The specific location for the instrument channel displays are located in the primary and backup Upper Cable Spreading Room.

instrument channel displays.

b) If the primary and backup Response for b) and c):

displays are not located in the main control room, please The Oyster Creek Station primary and backup provide a description of the instrument channel displays are located on a floor selected location(s) for the above the control room accessible through a primary and backup displays, nearby stairwell. This location was selected due to including prompt accessibility the proximity to the main control room.

to displays, primary and Radiological habitability, humidity, and alternate route evaluation, temperature at this location are considered habitability at display habitable in Oyster Creek Station's Engineering location(s), continual resource Standard ES-027, "Environmental Parameters availability for personnel Oyster Creek NGS" and Calculation C-1302-822-responsible to promptly read 5360-008, "Reactor Bldg Loss of Ventilation";

displays, and provisions for therefore, this location is not considered a harsh communications with decision environment. Estimated radiological conditions makers for the various SFP inside the Upper Cable Spreading Room for a drain down scenarios and core melt scenario, as well as estimated dose external events. rates from SFP draindown conditions to Level 3 c) The reasons justifying why (Calculation BYR13-187), and exposure to the locations selected will personnel monitoring SFP levels, will remain less enable the information from than emergency exposure limits allowable for these instruments to be emergency responders to perform this action.

Page 17 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 considered "promptly Heat and humidity from SFP boildown conditions accessible". Include have been evaluated for this location. The location consideration of various is several elevations below the SFP operating drain-down scenarios floor and located in a different building physically separated by concrete walls from the SFP such that heat and humidity from a boiling SFP would not compromise habitability at this location.

Spent Fuel Pool Level monitoring will be the responsibility of Operations personnel who will monitor the display periodically once dispatched from the Control Room. Travel time from the Control Room to the primary and secondary displays is approximately 2 minutes based on walkdowns. Radiological habitability, humidity, and temperature for the transit routes were considered habitable in Oyster Creek Station's Engineering Standard ES-027, "Environmental Parameters Oyster Creek NGS" and Calculation C-1302-822-5360-008, "Reactor Bldg Loss of Ventilation"; therefore, these locations are not considered a harsh environment to personnel monitoring the indications. The SFP levels will remain less than emergency exposure limits allowable for emergency responders to perform this action. Heat and humidity from SFP boildown conditions have been evaluated for access to this location, and the access routes are below the SFP operating floor and located in a different building physically separated by concrete walls from the SFP such that heat and humidity from a boiling SFP would not compromise habitability for accessing these displays. Diverse communications are accessible at both display locations. The operators would first employ radio communications or telephone communication. If the radio communications or telephone systems are non-functional, the sound powered phone system is assumed available because no power is required. A sound powered phone jack is located in the Control Room and will be available to dispatch personnel for the monitoring of the indicators. A sound powered phone jack is not located in the Upper Cable Spreading Room; therefore, an operator will physically report the level back to the Control Room periodically.

The display will be accessed on demand. It takes up to 2 minutes to reach the display location, for both the primary and backup channels, when an operator is dispatched from the control room. The Page 18 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 actual time for accessing the display locations is based on walkthroughs in the plant by Operations, Engineering, and Maintenance personnel. The path to access the display locations is within robust seismic category I structures, from the control room to the display locations, located in the Upper Cable Spreading Room.

11 RAI Question: Replaced by Interim SE RAI -13.

(RAI- Please provide the following:

10, a) A list of the operating (both Ref.4) normal and abnormal response) procedures, calibration/test procedures, maintenance procedures, and inspection procedures that will be developed for use of the SFP instrumentation in a manner that addresses the order requirements.

b) A brief description of the specific technical objectives to be achieved within each procedure. If your plan incorporates the use of portable spent fuel level monitoring components, please include a description of the objectives to be achieved with regard to the storage location and provisions for installation of the portable components when needed.

c) Describe how the replacement of an instrument channel component with a commercially available one that may not meet all of the qualifications noted in the OIP submittal would still be considered to be in compliance with the Order requirements. Which qualification provisions described in the OIP would not be followed?

Page 19 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 12 RAI Question: Complete (RAI- Please provide the following:

11, a) Further information a) Performance tests (functional checks) and Ref.4) describing the maintenance Operator performance checks are and testing program the described in detail in the vendor operator's licensee will establish and manual, and the applicable information is implement to ensure that planned to be contained in plant operating regular testing and calibration procedures. Operator performance tests will be is performed and verified by performed periodically as recommended by the inspection and audit to equipment vendor. Channel functional tests per demonstrate conformance Operations Procedures, with limits established in with design and system consideration of vendor equipment specifications, readiness requirements. will be performed at appropriate frequencies Please include a description established in the SFPIS manual. Operator of your plans for ensuring that performance checks will be performed on a necessary channel checks, periodic scheduled basis, with additional functional tests, periodic maintenance on an as-needed basis when flagged calibration, and maintenance by the system's automated diagnostic testing will be conducted for the level features. Channel calibration tests per measurement system and its maintenance procedures, with limits established in supporting equipment. consideration of vendor equipment specifications, b) A description of how the will be performed at frequencies established in guidance in NEl12-02, Section consideration of vendor recommendations.

4.3 regarding compensatory SFPIS channel/equipment maintenance, actions for one or both non- preventative maintenance, and testing program functioning channels will be requirements to ensure design and system addressed. readiness will be established in accordance with c) A description of what Exelon's processes and procedures and in compensatory actions are consideration of vendor recommendations to planned in the event that one ensure that appropriate regular testing, of the instrument channels channel checks, functional tests, periodic cannot be restored to calibration, and maintenance is performed functional status within 90 (and available for inspection and audit). Subject days. maintenance and testing program requirements will be developed during the SFPIS modification design process.

b) and c);

Both primary and backup SFPIS channels incorporate permanent installation (with no reliance on portable, post-event installation) of relatively simple and robust augmented quality equipment. Permanent installation coupled with stocking of adequate spare parts reasonably diminishes the likelihood that a single channel (and greatly diminishes the likelihood that both channels) is (are) out-of-service for an extended period of time. Planned compensatory actions for unlikely extended out-of-service events will be Page 20of33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 controlled by Procedure CC-OC-118, Diverse and Flexible Coping Strategies (Flex) and Spent Fuel Pool Instrumentation Program Implementation, and are summarized as follows:

  1. Required Compensatory Channel(s) Restoration Action if Out-of- Action Required Service Restoration Action not completed within Specified Time 1 Restore Immediately Channel to initiate action functional in accordance status within with note 90 days (or below within 45 days when operating outside qualified range), then proceed to Compensatory Action 2 Initiate action Immediately within 24 initiate action hours to in accordance restore one with note channel to below functional status and restore one channel to functional status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Note: Initiate an Issue Report to enter the condition into the Corrective Action Program.

Identify the equipment out of service time is greater than the specified allowed out of service time, develop and implement an Page 21 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 alternate method of monitoring, determine the cause of the non-functionality, and the plans and schedule for restoring the instrumentation channel(s) to functional status.

Draft Safety Evaluation Open Items 01# Description Status 1 RAI Question: Complete (RAI Please provide additional The Reactor Building, Control Room, Cable Tray Bridges,

-2, information describing how Lower Cable Spreading Room and Upper Cable Spreading Ref. the final arrangement of the Room (in which the equipment and cabling are located) are

5) SFP instrumentation and seismic I structures and will remain operable during and after a routing of the cabling beyond design basis event. All equipment and cabling are between the level installed to achieve maximum practical separation in instruments, the electronics accordance with NEI 12-02 Rev.1 requirements while meeting and the displays, meets the 1E separation requirements per Oyster Creek Station Order requirement to Specification SP-9000-41-005, Cables & Raceways. In the RB arrange the SFP level refuel floor area, the two sensors will be mounted in different instrument channels in a locations of the SFP and separated by a distance comparable manner that provides to the shortest side of the pool. The coaxial cables are reasonable protection of the protected in conduit and are in the refuel floor for a short level indication function distance. The coaxial cables that extend from the two sensors against missiles that may toward the location of the transmitters (sensors electronics) result from damage to the have a minimum separation of 8 feet, exceeding the 1E structure over the SFP. separation requirements per Oyster Creek Station Specification SP-9000-41-005, Cables & Raceways. The primary coaxial cable will be installed in conduit and the secondary coaxial cable will be installed in combination of conduit and cable tray.

The primary transmitter is located in the south Cable Tray Bridge and the secondary transmitter is located 15 feet away in the north Cable Tray Bridge. The electronic enclosures (displays) located in the Upper Cable Spreading Room are mounted 5 feet apart. The 4-20mA cables that extend from the transmitters to the displays have a minimum separation of 1 inch, maintaining the 1E separation requirements per Oyster Creek Station Specification SP-9000-41-005, Cables &

Raceways. The primary 4-20mA cable will be installed in conduit and the secondary 4-20mA cable will be installed in combination of conduit and cable tray. Distribution panels PDP-733-57 & PDP-733-58 are located in the Lower Cable Spreading Room. The 120VAC cables from the distribution panels terminate to panel 17R in the Control Room and to the displays in the Upper Cable Spreading Room. The 120VAC Page 22 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 power cables are 1 inch apart, maintaining the 1E separation requirements per Oyster Creek Station Specification SP-9000-41-005, Cables & Raceways. All the equipment and supports will be seismically mounted.

2 RAI Question: Complete (RAI For RAI 3(a) above, please The following Westinghouse documents provide the analyses

-4, provide the analyses used used to verify the design criteria and describe the methodology Ref. to verify the design criteria for seismic testing of the SFP instrumentation and electronics

5) and methodology for units, inclusive of design basis maximum seismic loads and seismic testing of the SFP hydrodynamic loads that could result from pool sloshing and instrumentation and the other effects that could accompany such seismic forces:

electronics units, a. Calculation CN-PEUS-15-08, "Seismic Analysis of the SFP including, design basis Primary and Backup Mounting Brackets at Oyster Creek maximum seismic loads Nuclear Generating Station" (Reference 15) and the hydrodynamic b. EQ-QR-269, WNA-TR-03149-GEN - Seismic Qualification loads that could result from of other components of SFPI S pool sloshing or other Oyster Creek Station specific evaluations were performed in effects that could ECR OC 14-00389 to address the seismic qualification and accompany such seismic structural requirements of other SFPIS equipment. The design forces. criteria in the ECR meet the requirements to withstand a SSE.

The methods used in the ECR follow SQUG-GIP, Rev. 2A, IEEE Standard 344-2004 and IEEE Standard 323-2003 for seismic qualification of the instrument.

3 RAI Question: Complete (RAI For each of the mounting The structural integrity and mounting of SFP level equipment

-5, attachments required to was based on formal configuration change process, plant Ref. attach SFP level equipment drawings, and approved work plans per Exelon procedures and

5) to plant structures, please processes. Details are provided in ECR OC 14-00389.

describe the design inputs, Design Inputs include, but not limited to, the following:

and the methodology that 1. Component weights and dimensions, core hole locations was used to qualify the and support details.

structural integrity of the 2. The capability of concrete expansion anchors.

affected 3. The loads (dynamic and static) for the probe mounting structures/equipment. bracket.

4. Concrete properties.
5. Seismic accelerations requirements for electrical equipment.
6. Allowable stresses for structural bolts.

Methodology to qualify the safety related structural integrity includes, but were not limited to, the following:

1. Structural Weldments - Qualifying the weld design entails the selection of a weld's physical attributes, such as type, configuration and size, which will make it suitable for transferring the prescribed loads within appropriate limits.

This process involves determining the maximum unit forces on the weld and comparing them with the weld capacity.

The methodology determines weld design forces by assuming nominal linear stress/strain distribution. For each Page 23 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 design, the engineer must confirm that the distribution of stiffness within the joint is consistent with this assumption.

In some cases more refined techniques may be required to predict appropriate distribution of weld forces.

2. Concrete Expansion Anchors - The design methodology of concrete expansion anchor assemblies involves 1) application of component attachment loads to the mounting plate, 2) analysis of the assembly to determine the resultant tension and shear forces on individual anchors, 3) evaluation of the anchor forces relative to anchor allowables, and 4) computation and evaluation of bending stresses in the mounting plate. Reactions for the attached component (applied to the plate at the centroid of the attachment weld) were resolved into moments, shears and axial loads (about the major axes of the mounting plate).
3. Existing Main Control Room Panel 17R- SQUG-GIP, "Generic Implementation Procedure for Seismic Verification of Nuclear Plant Equipment", Rev. 2A was used for its evaluation.
4. Existing Cable Trays - SQUG-GIP, "Generic Implementation Procedure for Seismic Verification of Nuclear Plant Equipment", Rev. 2A was used for their evaluations.
5. Conduit and Conduit Supports - Conduit, conduit supports, and pull box supports for the SFPLI modification utilized pre-engineered designs per Oyster Creek Procedure EP-059, Rev. 2, "Conduit Support Design and Installation", and Procedure 2400-GME-3780.52, Rev. 14, "Installation, Testing and Termination of Wire and Cable".

4 RAI Question: Complete (RAI For RAI 6 above, please Below is a summary of the test conditions used by

-7, provide the results for the Westinghouse to qualify the SFPIS.

Ref. selected methods, tests and Environmental Conditions for SFPIS Components in the

5) analyses used to Spent Fuel Pool Area demonstrate the qualification and reliability Level sensor probe, coax coupler and connector assembly, of the installed equipment launch plate and pool side bracket assembly, coax cable are in accordance with the designed and qualified to operate reliably in the below Order requirements. specified environmental conditions.

Parameter Normal BOB Temperature 50-140°F 212°F Pressure Atmospheric Atmospheric Humidity 0-95% RH 100% (saturated Page 24 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 steam)

Radiation TIDy 1E03 Rads 1 E07 Rads (above pool)

Radiation 1 E07 Rads 1 E09 Rads TIDy (probe and (12" above weight only) top of fuel rack)

Environmental Conditions Outside of the Spent Fuel Pool Area The level sensor transmitter and bracket, electronics display enclosure and bracket are designed and qualified to operate reliably in the below specified environmental conditions for an extended period defined in the "Duration" row in the table below.

Parameter Normal BOB BOB (Level Sensor Electronics Only)

Temp 50-120°F 140°F 140°F Pressure Atmos Atmos Atmos 0-95%

Humidity 0-95% RH 0-95% RH RH (non-cond (non-cond)

Duration 3 days 3 days 3 days Radiation s1 E03 R S1 E03 Ry TIDy Y S1 E03 Ry Thermal and Radiation Aging - organic components in Page 25 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 SFP Area Westinghouse documents EQ-QR-269, EQ-TP-354, WNA-TR-03149-GEN provide thermal and radiation aging program details for the SFPIS components. Westinghouse completed their thermal and radiation aging testing programs to qualify the SFPIS components to 1O years. Exelon has reviewed the documents and found them acceptable.

Seismic Category I Testing Seismic qualification testing performed by Westinghouse along with the technical evaluations performed by Westinghouse confirms that the SFPIS meets the seismic requirements of the vendor's design specification. Westinghouse's design specification satisfies the Oyster Creek Station installation requirements to withstand a SSE.

Vibration Justification Components of the system (i.e., bracket, transmitter, display enclosure) will be permanently installed to meet the requirements to withstand a SSE and will meet the Oyster Creek Station safety related installation requirements.

Westinghouse has analyzed the pool side bracket to withstand design basis SSE. Other components of the SFPIS were subjected to shock and vibration during the seismic testing and met the requirements necessary for mounted equipment.

Sloshing Justification The sloshing calculation performed by Westinghouse was reviewed for a design basis seismic event and found acceptable. Sloshing forces were taken into consideration for the anchorage design of the pool side bracket to ensure the bracket is rigidly mounted to include sloshing affects.

5 RAI Question: Complete Please provide the following:

(RAI a) A description of the a) The primary SFPLI instrument channel

-9, electrical ac power sources will be normally powered from safety related 120VAC, Post Ref. and capacities for the Accident Instrument Panel PDP-733-58. The backup SFPLI

5) primary and backup channels. instrument channel will be normally powered from the safety b) Please provide the results related 120VAC, Post Accident Instrument Panel PDP-733-57.

of the calculation depicting The panels are supplied by different safety buses, which the battery backup duty maintain power source independence, from 4160V buses to cycle requirements the instrument channels. The channel powered by Panel PDP-demonstrating that its 733-58 is powered from 4160V bus D and the channel capacity is sufficient to maintain the level indication powered by Panel PDP-733-57 is powered from 4160V bus C.

Upon loss of normal AC power, individual batteries installed in Page 26 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 function until offsite resource each channel's electronics I UPS enclosure will automatically availability is reasonably maintain continuous channel operation for at least 3 days.

assured.

Before reaching the 3-day battery life, the Post Accident Instrument Panels will be restored using the FLEX diesel generator. Spatial and physical separation of the power cables will be maintained and comply with Oyster Creek Station Specification SP-9000-41-005, Cables & Raceways.

b) Westinghouse Report, WNA-CN-00300-GEN, provides the results of the calculation depicting the battery backup duty cycle. This calculation demonstrates that battery capacity is 4.22 days to maintain the level indicating function to the display location. The displays are located in the Upper Cable Spreading Room elevation 63'-6". The results of the calculation meet the NEI 12-02 requirements.

6 RAI Question: Complete (RAI Please provide the a) The Oyster Creek Station primary and backup instrument

-12, following: channel displays are located in the Upper Cable Spreading Ref. a) The specific location for Room which is a robust seismic category I structure.

5) the primary and backup instrument channel display. b) The Oyster Creek Station primary and backup instrument b) For any SFP level channel displays are located in the Turbine Building on instrumentation displays elevation 63'-6", the floor above the Control Room. The Upper located outside the main Cable Spreading Room is accessible through a staircase control room, please adjacent to the Control Room. The Upper Cable Spreading describe the evaluation Room, Control Room and the path between the two locations used to validate that the are considered non-harsh environments per Engineering display location can be accessed without Standard ES-027, "Environmental Parameters Oyster Creek unreasonable delay NGS"; therefore, the radiological exposure at these locations following a BOB event. shall remain less than emergency exposure limits allowable for Include the time emergency responders to perform this action. Heat and available for personnel to humidity from SFP boildown conditions are not applicable in access the display as these locations. The locations are several floors below the SFP credited in the evaluation, operating floor and located in a different building physically as well as the actual time separated by concrete walls, such that heat and humidity from (e.g., based on walk-a boiling SFP would not compromise habitability at these throughs) that it will take for personnel to access the locations.

display. Additionally, please During ELAP conditions radiation due to boiling of SFP is not include a description of the an issue for Oyster Creek because the SFP is located in a radiological and different building compared to where the displays are located environmental conditions (SFP is in reactor building and displays are in Turbine on the paths personnel might take. Describe Building). Also, during ELAP there is no LOCA assumed whether the display location simultaneously. During normal operating conditions the dose in Page 27 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 remains habitable for the traverse path and display location is nominal(< 1mRem/hr) radiological, heat and which should not change as LOCA is not assumed during humidity, and other ELAP.

environmental conditions following a BOB event. The temperature in the traverse path and in the upper cable Describe whether personnel spreading room (where the display enclosures will be installed) are continuously stationed will be maximum 99 deg F without ventilation, per Cale 15050-at the display or monitor the M4-001, Revision 2.

display periodically.

The humidity in the traverse path and in the upper cable spreading room can be up to 100% per ES-027, Revision 4.

The path to display enclosures from MGR is very short (2 minutes or less) and operators should take very minimal time to read out the display as it's readily available. As a result, humidity should not be a concern for operators to traverse through this path.

Spent Fuel Pool Level monitoring will be the responsibility of Operations personnel who will monitor the display periodically once dispatched from the Control Room. Travel time from the Control Room to the primary and secondary displays is approximately 2 minutes based on walkdowns performed.

Diverse communications are accessible at both display locations. The operators would employ radio communications or the telephone communication. If the radio communications or telephone systems are nonfunctional, the sound powered phone system shall be available. A sound powered phone jack is located in the Control Room; however, a sound powered phone jack is not located in the Upper Cable Spreading Room.

The sound powered phone will be available to dispatch personnel for the monitoring of the indicators. The operator can then physically report back the level to the Control Room. The display will be accessed on demand.

7 RAI Question: Started (RAI Please provide a list of the The Westinghouse calibration procedure and technical manual

-13, procedures addressing are embedded into the Oyster Creek SFPIS vendor manual Ref. operation (both normal and VM-OC- 6652. Oyster Creek has developed a program

5) abnormal response), document CC-OC-118, "Oyster Creek Implementation of calibration, test, Diverse and Flexible Coping Strategies (FLEX) and Spent Fuel maintenance, and inspection Pool Instrumentation Program" defining the requirements of the procedures that will be SFPIS.

developed for use of the SFP instrumentation. The licensee Oyster Creek developed Operator Round guidance to perform is requested to include a channel checks against each other and checks against brief description of the physical pool level.

specific technical objectives The program document CC-OC-118, "Oyster Creek to be achieved within each procedure. Implementation of Diverse and Flexible Coping Strategies (FLEX) and Spent Fuel Pool Instrumentation Program" Page 28 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 provides guidance for taking compensatory actions when either one or both channels are out of service. This guidance will be incorporated into implementing operating procedures.

7 Potential Draft Safety Evaluation Impacts There are no potential impacts to the Draft Safety Evaluation identified at this time.

8 References The following references support the updates to the Overall Integrated Plan described in this enclosure.

1. Oyster Creek Nuclear Generating Station, "Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051 )," dated February 28, 2013 (RS 033)
2. NRC Order Number EA-12-051, "Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation," dated March 12, 2012
3. USNRC letter to Exelon Generation Company, LLC, Request for Additional Information Regarding Overall Integrated Plan for Reliable Spent Fuel Pool Instrumentation, dated August28,2013
4. Exelon Generation Company, LLC, letter to USNRC, "Response to Request for Additional Information - Overall Integrated Plan in Response to Commission Order Modifying License Requirements for Reliable Spent Fuel Pool Instrumentation (Order No. EA 051 )", dated September 18, 2013 (RS-13-212)
5. USNRC letter to Exelon Generation Company, LLC, "Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Pool Instrumentation", dated November 8, 2013
6. First Six-Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, dated August 28, 2013 (RS-13-124)
7. Second Six Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, dated Feburary 28, 2014 (RS-14-023)
8. Third Six Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, dated August 28, 2014 (RS-14-201)
9. Fourth Six Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, dated Feburary 27, 2015 (RS-15-031)

Page 29 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016

10. Fifth Six Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, dated August 28, 2015 (RS-15-203)
11. Sixth Six Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, dated February 26, 2016 (RS-16-031)
12. C-1302-822-5360-008, Rev. 2, "Reactor Bldg Loss of Ventilation"
13. ES-027, Rev. 004, "Environmental Parameters Oyster Creek NGS"
14. ECR OC 14-00389, Rev. 001, "Reliable Spent Fuel Pool Level Instrumentation (Fukushima)"
15. Letter LTR-SFPIS-14-55, Rev 0, "SFPIS 2 Point Verification Methodology"
16. Calculation CN-PEUS-15-08, Rev. 1, "Seismic Analysis of the SFP Primary and Backup Mounting Brackets at Oyster Creek Nuclear Generating Station"
17. "Design Basis Seismic Response Analyses for the Oyster Creek Nuclear Generating Station Reactor, Intake and Turbine Buildings," Report Number 50069-R-001, Rev. O, EOE International, September 1995
18. TID-7024, Nuclear Reactors and Earthquakes, 1963, by the US Atomic Energy Commission (Appendix F - Dynamic Analysis of Fluids in Containers Subjected to Acceleration)
19. Westinghouse Document, WNA-TR-03149-GEN, "SFPIS Standard Product Final Summary Design Verification Report," Revision 2
20. Westinghouse Document, EQ-QR-269, "Design Verification Testing Summary Report for the Spent Fuel Pool Instrumentation System," Revision 5
21. Drawing 10067E58, Sheet 1, Rev. 0, "Oyster Creek Generating Station Spent Fuel Pool Primary and Backup Mounting Bracket Plan, Sections, and Details", prepared by Westinghouse
22. Westinghouse Document, WNA-DS-02957-GEN, "Spent Fuel Pool Instrumentation System System Design Specification", Revision 4
23. Letter LTR-SFPIS-13-35, "Basis for Radiation Dose Requirement and Clarification of Production", Revision 1 9 Attachments
1. Elevation View of Spent Fuel Pool Instrumentation System
2. Plan View of Spent Fuel Pool Instrumentation System
3. Mounting Bracket Drawing 10067E58 Page 30 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 Attachment 1 Elevation View of Spent Fuel Pool Instrumentation System I

119' -3" Refueling Floor II 117'-10" Level 1 =J--t f 118'-6 7/8" Upper Range Value 106'-1 1/B" Level 2 96'-1 1/8" Top of Fuel I 96' - 9" Level 3 I: AAAAAA AA AAA.a.

I 96'-9" Lower Range Value SPENT FUEL POOL 23'-6" Ground Floor REACTOR BUILDING ELEVATION VIEW Page 31of33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 Attachment 2 Plan View of Spent Fuel Pool Instrumentation System lORTH TWISTED CDNDARY OAXIAL SHIELDED TRANSMITIER CABLE PAIR i--~~~39'~~~-

RI MARY TRANSMITTER REACTOR BUILDING MAlf\J OFFICE BUILDING TURBINE BUILDING PLAN Page 32 of 33

Oyster Creek Nuclear Generating Station Seventh Six-Month Status Report for the Implementation of SFPIS August26,2016 Attachment 3 Mounting Bracket Drawing 10067E58

=--=*==a=-====" 7 ~~~~~~--'L_~~~~~~~6=--~-~~~~.....L..~~~~~~--"5'--~~~~~~-"<-~~~~~~-4-=---~~~~~~-'--~~~~~~-'=3'--~~~--.,,---~""'-~~~~~~~2 =---;;;em1LL "OF"'"'MA

--='TE

,..,,;R

~IAl..S;a';"'.

.,;-~~~~~~~~~~-,

PART DEaCAIPT10N H

1, MJ. WELIJll'G 9HAlJ,.

  • I N ~NC& wrnt AW8 D1*..JIJIR ORAllME BDILliR a PRl!SllUREVIEml.

II COIDll ll!le:CTICll'lil Dt.C-ta). wm.D191S MAY 1111 ~BD TOAWS Dt.e O"AaM:B. 9QIUlblll & ~

II

-.. -----.~+..------------ ::

~cooa sacT10N IX PDta).

2. WELD MAT'muAL SHAU. a& 'IHB llAMB AS nm. BASE Ml!TAL OR COMPA.Ta.E MA.Tl!RIALAS PE!R TAl!IL.S II G II G
a. MA"IEAIAL ~ BHALL llE WITHIN llD..02" TDLERAHCI!.
4. ~ T"OL.ERANCE. RJR ITEM 719 s: z.a*.
a. NO. 2 Nl!Ol.U8I!. LU'BfUCAHTTO 81! APPl.20 ON THI! THREADS.

BASEPLATE (8YOTHERS)

F F E 2 E

T.O.C.

El.EV. 118'.S-1.6 6116-1& TAPPED HOLE.D.INl"DEEP (FOR. OROUNDINCI) 0 FACE OIF D LINER 2 .12&

c c PLAT&: TAPPED oza (THREADED) TO L .....

DETAIL3

~

CAPT\JR& BOLT PLATI; DRILl.ED era.ax o.r FOR BOLT TiiREAD CLEAAANC£ (HO THREAD)

BEENOTEB PlATI!. DRJLU!D - - -

THRU WITH 0. tr 1s C~ TO Hl!l..PTHI!

BOLT FIND HOLE CENTER. I B B MOUNTING PLATE: NOTES:

1~~~~Pl!RANS1112..1-18118. '1'1EENDOF COUPLER BHAl.J.. BE *1 PITCH FROM THE BOTIOM FACE OF THE HALF INCH THICK PLATE, CJ COUPLER BHALL BE IN8TALLED TO A 8NUO TIGHT CONDITION AND THEN TIGHTENED TI> AT LSAST tM REVOLUTION, NOT EXCEEDING 112 REVOLUTION.

D) NEOUJBI! NO. 2 WBfUCANT SHALL IE INSTALLED ON THE THREADS Bl!FORI! INSTALLING THI! COUP'Ll!R TO TIE MOUNTING PL.A'TE..

O>>t*~~~U..C ~n.:~l"OR.CONSTRUCTION ~~~ ~-=-~...:.-

~

Page 33 of 33

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