RA-16-011, 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)

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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)
ML16057A004
Person / Time
Site: Oyster Creek
Issue date: 02/26/2016
From: David Helker
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EA-12-051, RA-16-011, RS-16-031, TAC MF0823
Download: ML16057A004 (33)
v • d • e


Text

Exelon Generation ,.)

Order No. EA-12-051 RS-16-031 RA-16-011 February 26, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Co.ntrol Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

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)

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 February 26, 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) 1O. 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. 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, and 1O provided the first, second, third, fourth, and fifth 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 sixth 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 also addresses the NRC Interim Staff Evaluation Request for Additional Information Items contained in Reference 11.

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 February 2016.

U.S. Nuclear Regulatory Commission Integrated Plan Report to EA-12-051 February 26, 2016 Page 3 Respectfully submitted, David P. Helker Manager - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Enclosure:

1. Oyster Creek Nuclear Generating Station Sixth 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 G. Lamb, NRR/DORULPL3-2, NRC 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 Sixth Six-Month Status Report for the Implementation of Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (29 pages)

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Oyster Creek Nuclear Generating Station Sixth 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 fifth 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 fifth six-month status report (Reference 10), and is current as of February 26, 2016.

  • Complete and Issue SFPIS Modification Package 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 August 28, 2013 Complete Update 2 February 28, 2014 Complete Update 3 August 28, 2014 Complete Update 4 February 27, 2015 Complete Update 5 August28,2015 Complete Update 6 February 26, 2016 Complete with Page 1of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Revised Target Completion Activity Target Milestone Date Status Completion Date this submittal Update 7 August 28, 2016 Not Started Provide Final Safety Evaluation Complete with April 30, 2016 (SE) Info this submittal 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 Not Started (March)

Complete SFPIS Installation 302016 Not Started and Put Into Service 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 2of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Overall Integrated Plan Open Items 01# Description Status 1 Open Item: Complete Continuous level indication will (Addressed in Reference 6) 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 Ref. 4) sketch depicting the elevation the elevation view of the proposed mounting view of the proposed typical arrangement of the mounting brackets and level sensors. Datum points Level 1, Level 2, Level 3, mounting arrangement for the top of fuel racks and level sensor measureable portions of instrument channel range are all depicted in this sketch.

consisting of permanent measurement channel equipment LEVEL 1 - This is the level at which point the (e.g., fixed level sensors and/or reliable suction loss occurs due to uncovering of the weirs. Level 1 corresponds to elevation stilling wells, and mounting 11 117'-10

  • brackets). Indicate on this sketch the datum values representing LEVEL 2 - This is the level that is adequate to Level 1, Level 2, and Level 3 as provide substantial radiation shielding for a person standing on the spent fuel pool operating well as the top of the fuel racks.

deck. Level 2 corresponds to elevation Indicate on this sketch the 106'-1 1/8 11 , water level 1o* above the highest portion of the level sensor point of any fuel rack seated in the spent fuel measurement range that is pool.

sensitive to measurement of the LEVEL 3 - This is the level where fuel remains fuel pool level, with respect to the covered and actions to implement make-up Level1, Level2,and Level3 water addition may no longer be deferred. This datum points. level corresponds to a water level 7 7/8" above the top of the fuel assembly. Level 3 shall be 1 11 designated at elevation 96 -9

  • The level sensors' measurement range is between elevation 96' -9" and 118' -6 718".

3 RAI Question: Complete (RAl-2, Please provide a clearly labeled Reference Attachment 2 for a sketch depicting sketch or marked-up plant the plan view of the Spent Fuel Pool, equipment Page 3of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Ref.4) drawing of the plan view of locations and cable routing. The cable routing the SFP area, depicting the SFP will extend from the level sensors towards the inside dimensions, the planned level displays.

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

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

loads and dynamic loads.

Describe the methodology that The vendor, Westinghouse, has evaluated the will be used to estimate the total structural integrity of the mounting brackets in loading, inclusive of design basis calculations CN-PEUS-15-08 (Reference 15).

maximum seismic loads and the The GTSTRUDL model, used by hydrodynamic loads that could Westinghouse to calculate the stresses in the result from pool sloshing or other bracket assembly, considered load effects that could accompany combinations for the dead load, live load and such seismic forces. seismic load on the bracket. The reactionary b) A description of the manner in forces calculated from these loads become which the level sensor (and the design inputs to design the mounting stilling well, if appropriate) will be bracket anchorage to the refuel floor to attached to the refueling floor withstand a Safe Shutdown Earthquake and/or other support structures (SSE).

for each planned point of attachment of the probe Seismic assembly. Indicate in a schematic The seismic loads are obtained from the the portions of the level sensor response spectra curves of Oyster Creek that will serve as points of attachment for (Reference 16). The following methodology mechanical/mounting or was used in determining the stresses on the electrical connections. bracket assembly:

c) A description of the manner by which the mechanical

  • Frequency analysis, taking into account connections will attach the level the dead weight and the hydrodynamic instrument to permanent SFP mass of the structure, is performed to structures so as to support the obtain the natural frequencies of the level sensor assembly. structure in all three directions.

Page 4 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016

  • Modal responses are combined using the Ten 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 (SRSS)

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

~lashing forces were obtained by analysis.

The TID-7024, Nuclear Reactors and Earthquakes, 1963 (Reference 17), 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 (Reference 17) 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 Page 5of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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 (Reference 18).

The following Westinghouse documents provide information with respect to the design criteria used, and a description of the methodology used 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" (Reference 15)
2) EQ-QR-269 (Reference 19), WNA-TR-03149-GEN (Reference 18)- Seismic Qualification of other components of SFPIS Oyster Creek Station specific evaluations were performed in ECR OC 14-00389 (Reference 13) 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 (Reference 20) prepared by Westinghouse shows the connection/mounting details. This drawing is included in this update as Attachment 3.

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 Page 6 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26 2016 assembly to the SFP structure via a base .

plate. The concrete expansion anch~rs will be designed to withstand SSE and will meet the Oyster Creek safety related installation requirements. The analysis and ?etails of t~e Pool-side brackets are provided in Calculation CN-PEUS-15-08 (Reference 15) and Drawing 10067E58, Sheet 1 (Reference 20). The design and details of the anchorage to the floor are provided in ECR OC 14-00389 (Reference 13).

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

a) Beyond Design Basis Environment -

Ref.4) a) A description of the specific Westinghouse qualified the method or combination of components (probe, connector, cable) methods that will be applied to of the SFPIS located in the SFP area to demonstrate the reliability of the beyond design basis environment.

the permanently installed Components of the system were equipment subjected to beyond design basis under beyond-design-basis conditions of heat and humidity, (BOB) ambient temperature, thermal and radiation aging humidity, shock, vibration, and mechanisms. This testing confirmed radiation conditions.

functionality of these system b) A description of the testing components under these beyond ..

and/or analyses that will be design basis environmental cond1t1ons.

conducted to provide Westinghouse performed testing to assurance that the equipment ensure aging of the components in the will perform reliably under the SFP area will not have a significant worst-case credible design effect on the ability of the equipment to basis loading at the location perform following a plant design basis where the equipment will be earthquake. Exelon has reviewed the mounted. Include a discussion documents and found them acceptable.

of this seismic reliability Reference Westinghouse documents demonstration as it applies to (i)

EQ-TP-351, WNA-TR-03149-GEN, and the level sensor mounted in the EQ-TP-354 for description of specific SFP area, and (ii) any control qualification methods.

boxes, electronics, or read-out and re-transmitting devices that Mild Environment - Westinghouse will be employed to convey the qualified the system components level information from the level (display panel, transmitter) th.~t reside sensor to the plant operators or in the mild environment cond1t1ons to emergency responders.

determine that the components can c) A description of the specific satisfactorily perform to those method or combination of conditions. Westinghouse has methods that will be used to determined that aging does not have a confirm the reliability of the significant effect on the ability of the permanently installed equipment to perform following a plant equipment such that following a desiqn basis earthquake. Exelon has Page 7 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 seismic event the instrument reviewed the documents and found will maintain its required them acceptable. Reference accuracy. Westinghouse documents EQ-QR-269, WNA-TR-03149-GEN for description of specific methods.

Shock and Vibration - SFPIS pool side brackets were analyzed for Safe Shutdown Earthquake design requirements per NRG Order EA 051 and NEI 12-02 guidance. As provided by the NRG Order EA-12-051, the NEI 12-02 guidance and as clarified by the NRG 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 WNADS-02957, WNA-TR-03149-GEN for shock and vibration.

b) The seismic adequacy of the SFPIS (all components) is demonstrated by vendor testing and analysis in accordance with below listed Page 8 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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 in RAI Sb response below, 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:

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

measurement system meet this requirement so that the potential a) The level probes will be mounted on the south for a common cause event to side of the SFP and will be separated by a adversely affect both channels is distance greater than the span of the shortest minimized to the extent side of the pool. This meets the NEI 12-02, practicable. Revision 1 guidance for channel separation.

b) Further information on how each level measurement system, b) The coax cables from the level instruments in consisting of level sensor the SFP area are routed directly one floor below electronics, cabling, and readout into the reactor building elevation 95' 3" and as a devices will be designed and result the separation between the coax cables in installed to address the refuel floor area is maintained by a distance Page 9of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 independence through the greater than the span of the shortest side of the application and selection of pool. Outside of the refuel floor area, the level independent power sources, the transmitters for the primary and backup use of physical and spatial instrument channels will be installed in each separation, independence of Cable Tray Bridge on elevation 80' with a signals sent to the location{s) of physical separation of approximately 20 feet.

the readout devices, and the The electronics I UPS enclosures for the primary independence of the displays. and backup instrument channels will be installed in the Upper Cable Spreading Room elevation 68' -6" with a physical separation of 8'. Physical and spatial separation of the primary and backup instrument channels is also met by maintaining maximum practical separation between the channel cables and routing the associated instrument channel cables in separate raceways, where possible, while meeting minimum separation requirements for Class 1 E cables per OCGS Specification SP-9000-41-005, Cables &

Raceways. This meets the NEI 12-02, Rev. 1 guidance requirements for instrument channel cable separation. The 120 VAC power to the primary instrument will be provided from safety related Post Accident Instrument Power Panel PDP-733-058. The 120VAC power to the backup level instrument will be provided from safety related Post Accident Instrument Power Panel PDP-733-057. The 120VAC distribution panels for the primary and backup instruments 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 Page 10 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26 2016 ensures, with margin, that the channel will be available to run reliably and continuously following the onset of the BOB event for the minimum duration needed, consistent with the plant mitigation strategies for BOB external events (Order EA 049).

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

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

points.

b) A description of the b) The Westinghouse document WNA-TP-methodology that will be used 04709-GEN describes the methodology for for determining the maximum routine testing/calibration verification and allowed deviation from the calibration methodology. This document also instrument channel design specifies the required accuracy criteria under accuracy that will be employed normal operating conditions. Oyster Creek under normal operating Station calibration and channel verification conditions as an acceptance procedures will follow the guidance and criteria criterion for a calibration provided in this document. Instrument channel procedure to flag to operators calibration will be performed if the level and to technicians that the indication reflects a value that is outside the channel requires adjustment to acceptance band established in the Oyster within the normal condition Creek Station calibration and channel design accuracy.

verification procedures. Calibration will be performed once per 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 planned refueling outage considering normal .

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

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Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26 2016 9 RAI Question: Started (RAl-8, Please provide the following:

a) Westinghouse has provided calibration Ref.4) a) A description of the procedure WNA-TP-04709-GEN and functional capability and provisions the test procedure WNA-TP-04613-GEN proposed level sensing describing the capabilities and provisions equipment will have to enable of SFPIS periodic testing and calibration.

periodic testing and calibration, Westinghouse has provided letter LTR-SFPIS-including how this capability 14-55 detailing the methodology of performing a enables the equipment to be 2 point verification in-situ testing. Oyster Creek tested in-situ.

will review these procedures and letter to ensure b) A description of how such the instructions provided by Westinghouse can testing and calibration will be implemented for calibration/functional enable the conduct of testing/in-situ testing to meet the Order regular channel checks of each requirements. This item will be completed by independent channel against 08/28/2016.

the other, and against any other permanently-installed SFP level b) The level displayed by the channels will be instrumentation. verified per the Oyster Creek Station c) A description of how administrative and operating procedures. If the functional checks will be level is not within the required accuracy per performed, and the frequency at Westinghouse recommended tolerances, which they will be conducted. channel calibration will be performed.

Describe how calibration tests will be performed, and the c) Functional checks will be performed per frequency at which they will be Westinghouse functionality test procedure at the conducted. Provide a Westinghouse recommended frequency.

discussion as to how these Calibration tests will be performed per surveillances will be Westinghouse calibration procedure at the incorporated into the plant recommended frequency and in accordance surveillance program. with Oyster Creek's maintenance and operating d) A description of what programs. Oyster Creek will develop calibration, preventive maintenance tasks functional test, and channel verification are required to be performed procedures per Westinghouse during normal operation, and recommendations to ensure reliable, accurate the planned maximum and continuous SFPIS functionality. This item surveillance interval that is will be completed by 08/28/2016.

necessary to ensure that the channels are fully conditioned d) Oyster Creek will develop preventive to accurately and reliably maintenance tasks for the SFPIS per perform their functions when Westinghouse recommendations to assure that needed. the channels are fully conditioned to accurately and reliably perform their functions when needed. This item will be completed by 08/28/2016.

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

a) The Oyster Creek Station primary and backup a) The specific location for the instrument channel displays are located in the Page 12 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Ref.4) primary and backup instrument Upper Cable Spreading Room.

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 selected location(s) for the floor above the control room accessible through primary and backup displays, a nearby stairwell. This location was selected including prompt accessibility due to 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-responsible to promptly read 822-5360-008, "Reactor Bldg Loss of displays, and provisions for Ventilation"; therefore, this location is not communications with decision considered a harsh environment. Estimated makers for the various SFP radiological conditions inside the Upper Cable drain down scenarios and Spreading Room for a core melt scenario, as external events. well as estimated dose rates from SFP c) The reasons justifying why draindown conditions to Level 3 (Calculation the locations selected will BYR 13-187), and exposure to personnel enable the information from monitoring SFP levels, will remain less than these instruments to be emergency exposure limits allowable for considered "promptly emergency responders to perform this action.

accessible". Include Heat and humidity from SFP boildown conditions consideration of various drain- have been evaluated for this location. The down scenarios location is several elevations 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 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 emerqencv responders to perform Page 13 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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 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 Page 14 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26 2016 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?

12 RAI Question: Complete (RAI- Please provide the following:

a) Performance tests (functional checks) and 11, a) Further information Operator performance checks are described Ref.4) describing the maintenance and in detail in the vendor operator's testing program the licensee manual, and the applicable information is will establish and implement to planned to be contained in plant operating ensure that regular testing and procedures. Operator performance tests are calibration is performed and planned to be performed periodically as verified by inspection and audit recommended by the equipment vendor.

to demonstrate conformance Channel functional tests per Operations with design and system Procedures, with limits established in readiness requirements. Please consideration of vendor equipment include a description of your specifications, are planned to ~e perfor~ed at plans for ensuring that appropriate frequencies established equivalent necessary channel checks, to or more frequently than existing SFPIS.

functional tests, periodic Manual calibration and operator performance calibration, and maintenance checks are planned to be performed on a will be conducted for the level periodic scheduled basis, with addi~ional measurement system and its maintenance on an as-needed basis when supporting equipment.

flagged by the system's automated diagnostic b) A description of how the testing features. Channel calibration tests per guidance in NEl12-02, Section maintenance procedures, with limits established 4.3 regarding compensatory in consideration of vendor equipment actions for one or both non-specifications, are planned to be performed at functioning channels will be frequencies established in consideration of Page 15of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 addressed. vendor recommendations.

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

Subject maintenance and testing program requirements are planned to be developed during the SFPIS modification design process.

Responses for 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 controlled by Procedure CC-OC-118-1001, 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 Page 16 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 90 days (or if below channel restoration not expected within 90 days, 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 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 In the refuel floor area, the two sensors will be mounted in

-2, information describing how different locations of the SFP and separated by a distance Ref. the final arrangement of the comparable to the shortest side of the pool. Outside of the

5) SFP instrumentation and refuel floor area, the coaxial cables that extend from the two routing of the cabling sensors toward the location of the transmitters (sensors between the level electronics) will be installed using separate routes. The primary instruments, the electronics coaxial cable will be installed in conduit and the backup coaxial Page 17of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 and the displays, meets the cable will be installed in cable tray. The Reactor Building and Order requirement to Cable Tray Bridges are seismic I structures and will remain arrange the SFP level operable during and after a beyond design basis event. The instrument channels in a conduits and tray system (in which the coax cables are manner that provides installed) will be separated to achieve maximum practical reasonable protection of the separation while meeting separation for Class 1E cables in level indication function accordance with current plant licensing bases criteria for against missiles that may electrical separation as defined in OCGS Specification SP-result from damage to the 9000-41-005, Cables & Raceways. The minimum separation structure over the SFP. between the coax cables from the spent fuel pool to the transmitters is 8 feet. The 4-20mA cables that extend from the transmitters located in the Cable Tray Bridges to the electronics enclosure (display) located in the Upper Cable Spreading Room will be installed in seismic I structures, installed in separate conduits and will be routed separately with a minimum distance of 8 feet. The 120VAC power cables will be installed in a seismic I building, in separate conduits and will be routed separately. They will be routed to achieve maximum practical separation in accordance with NEI 12-02, Rev. 1 requirements while meeting Class 1E separationrequirements 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 Ref. to verify the design criteria methodology for seismic testing of the SFP instrumentation

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

maximum seismic loads a. Calculation CN-PEUS-15-08, "Seismic Analysis of the and the hydrodynamic SFP Primary and Backup Mounting Brackets at Oyster loads that could result from Creek Nuclear Generating Station" (Reference 15) pool sloshing or other b. EQ-QR-269 (Reference 19), WNA-TR-03149-GEN effects that could (Reference 18)- Seismic Qualification of other accompany such seismic components of SFPIS forces.

Oyster Creek Station specific evaluations were performed in ECR OC 14-00389 (Reference 13) 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.

Page 18of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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, (Reference 13).

and the methodology that was used to qualify the Design Inputs include, but not limited to, the following:

structural integrity of the

1. Component weights and dimensions, core hole locations affected structures/equipment. and support details.
2. The capability of concrete expansion anchors.
3. The loads (dynamic and static) for the probe mounting 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 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 Page 19of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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. Environmental Conditions Ref. selected methods, tests and for SFPIS Components installed in the Spent Fuel Pool Area at

5) analyses used to OYS are bounded by below test conditions, except for radiation demonstrate the Tl D 12" above top of fuel rack for beyond design basis qualification and reliability conditions (BOB). The BOB radiation TIO, 12" above top of of the installed equipment fuel rack for OYS Station is 4.E07 Ry, per Calculation BYR13-in accordance with the 051 - NEI 12-02 Spent Fuel Pool Doses. (Calculation BYR13-187, "Radiation Doses in the vicinity of the Spent Fuel Pool at Order requirements.

Reduced Water Level" proves that Calculation BYR 13-051 bounds the Exelon Fleet for these values.) The BOB radiation value to which the Westinghouse equipment is qualified to is 1.E07 Ry, per Section 5.1.1 of WNA-TR-03149-GEN. The radiation value of 4.E07 Ry is higher than 1.E07 Ry to which Westinghouse qualified the instrument to. However, this value of 4.E07 Ry is applicable only when the water is at Level 3. At Level 2 the TIO reduces to 2.E07 Ry and it further reduces to 8.E06 at Level 1 and above. With SFP water level at Level 3 the only components of SFPI that are exposed to high radiation are the stainless steel probe and the stainless steel anchor.

The materials with which the probe and the anchor are manufactured are resistant to radiation effects. The stainless steel anchor and stainless steel probe can withstand 40 year dose. Westinghouse updated the design specification (WNA-DS-02957-GEN (Reference 21)) and LTR-SFPIS-13-35, Revision 1 (Reference 22) documentation to include the above technical justification.

Environmental Conditions for SFPIS Components in the Spent Fuel Pool Area Level sensor probe, coax coupler and connector assembly, launch plate and pool side bracket assembly, coax cable are Page 20 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 designed and qualified to operate reliably in the below specified environmental conditions.

Parameter Normal BOB Temperature 50-140°F 212°F Pressure Atmospheric Atmospheric 100%

Humidity 0-95% RH (saturated steam)

Radiation TIO y 1 E03 Rads 1E07 Rads (above pool)

Radiation TIO y 1 E09 Rads 1 E07 Rads (12" above (probe and top of fuel weight only) 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.

BOB Parameter Normal BOB (Level Sensor Electronics Only)

Temperature 50-120°F 140°F 140°F Pressure Atmospheric Atmospheric Atmospheric 0-95% RH 0-95% RH Humidity 0-95% RH (non- (non-condensing) condensing)

Duration 3 days 3 days 3 days Radiation

51 E03 Ry :51 E03 Ry :51 E03 Ry TIDy Thermal and Radiation Aging - organic components in SFP Area Westinghouse documents EQ-QR-269, EQ-TP-354, WNA-TR-Page 21of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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 confirm 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 a 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 effects.

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 instrument channel will be normally powered from the safety channels. related 120VAC, Post Accident Instrument Panel PDP-733-57.

b) Please provide the results These are on different safety buses, which maintains power of the calculation depicting source independence. Upon loss of normal AC power, the battery backup duty individual batteries installed in each channel's electronics I cycle requirements UPS enclosure will automatically maintain continuous channel demonstrating that its capacity is sufficient to operation for at least 3 days. Before reaching the 3-day battery maintain the level indication life, the Post Accident Instrument Panels will be restored using function until offsite resource the FLEX diesel generator.

Page 22 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 availability is reasonably assured. 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, 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.

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 on elevation 63' -6", the floor instrumentation displays above the control room. The Upper Cable Spreading Room is located outside the main accessible through two doors and a staircase. This location control room, please was selected due to proximity to the main control room. The describe the evaluation Upper Cable Spreading Room is not applicable in Engineering used to validate that the Standard ES-027, "Environmental Parameters Oyster Creek display location can be NGS"; therefore, the radiological habitability at this location will accessed without remain less than emergency exposure limits allowable for unreasonable delay emergency responders to perform this action. Heat and following a BOB event. humidity from SFP boildown conditions have been evaluated Include the time for this location.. The location is at an elevation several floors available for personnel to below the SFP operating floor and located in a different access the display as building physically separated by concrete walls, such that heat credited in the evaluation, and humidity from a boiling SFP would not compromise as well as the actual time habitability at this location.

(e.g., based on walk- Spent Fuel Pool Level monitoring will be the responsibility of throughs) that it will take for Operations personnel who will monitor the display periodically personnel to access the once dispatched from the Control Room. Travel time from the display. Additionally, please Control Room to the primary and secondary displays is include a description of the approximately 2 minutes based on walkdowns. Radiological radiological and habitability for the transit routes to both displays is not available environmental conditions in Engineering Standard ES-027, "Environmental Parameters on the paths personnel Oyster Creek NGS"; therefore, the radiological habitability at might take. Describe this location will remain less than emergency exposure limits whether the display location allowable for emergency responders to perform this action.

remains habitable for Heat and humidity from SFP boildown conditions have been radiological, heat and evaluated for access to this location, and the access routes are humidity, and other below the SFP operating floor and located in a different environmental conditions building such that heat and humidity from a boiling SFP would following a BOB event. not compromise habitability for accessing these displays.

Describe whether personnel Diverse communications are accessible at both display are continuously stationed locations. The operators would employ radio communications at the display or monitor the or telephone communications. If the radio communications or display periodically. telephone systems are nonfunctional, the sound powered phone system is assumed available because no power is Page 23of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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 actual time for accessing the display locations is based on a walkthrough from the Control Room to the Upper Cable Spreading Room. The path to access the display locations is within the robust seismic category I structures, from the Control Room to the display locations, in the Upper Cable Spreading Room. Upon obtaining the indicated SFP level from the display location, the operators will use the radio communication system to provide the information to the Control Room immediately.

7 RAI Question: Complete (RAI Please provide a list of the Appropriate quality measures will be selected for the SFPIS

-13, procedures addressing required by Order EA-12-051, consistent with Appendix A of Ref. operation (both normal and NEI 12-02. Site procedures will be developed for system

5) abnormal response), inspection, calibration and test, maintenance, repair, operation calibration, test, and normal and abnormal responses, in accordance with maintenance, and inspection Exelon's procedure control process. Technical objectives to be procedures that will be achieved in each of the respective procedures are described developed for use of the SFP below:

instrumentation. The licensee is requested to include a Procedure Objectives to be achieved:

brief description of the

1. System Inspection: To verify that system components are in specific technical objectives to be achieved within each place, complete, and in the correct configuration, and that the procedure. sensor probe is free of significant deposits.
2. Calibration and Test: To verify that the system is within the specified accuracy, is functioning as designed, and is appropriately indicating SFP water level.
3. Maintenance: To establish and define scheduled and preventive maintenance requirements and activities necessary to minimize the possibility of system interruption.
4. Repair: To specify troubleshooting steps and component repair and replacement activities in the event of system malfunction.
5. Operation: to provide sufficient instructions for operation and use of the system by plant operation staff.
6. Responses: To define the actions to be taken upon observation of system level indications, including actions to be taken at the levels defined in NEI 12-02.

Page 24 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 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 August 28, 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)
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. C-1302-822-5360-008, Rev. 2, "Reactor Bldg Loss of Ventilation" Page 25of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016

12. ES-027, Rev. 004, "Environmental Parameters Oyster Creek NGS"
13. ECR OC 14-00389, Rev. 001, "Reliable Spent Fuel Pool Level Instrumentation (Fukushima)"
14. Letter LTR-SFPIS-14-55, Rev 0, "SFPIS 2 Point Verification Methodology"
15. Calculation CN-PEUS-15-08, Rev. 1, "Seismic Analysis of the SFP Primary and Backup Mounting Brackets at Oyster Creek Nuclear Generating Station"
16. "Design Basis Seismic Response Analyses for the Oyster Creek Nuclear Generating Station Reactor, Intake and Turbine Buildings," Report Number 50069-R-001, Rev. 0, EQE International, September 1995
17. TID-7024, Nuclear Reactors and Earthquakes, 1963, by the US Atomic Energy Commission (Appendix F - Dynamic Analysis of Fluids in Containers Subjected to Acceleration)
18. Westinghouse Document, WNA-TR-03149-GEN, "SFPIS Standard Product Final Summary Design Verification Report," Revision 2
19. Westinghouse Document, EQ-QR-269, "Design Verification Testing Summary Report for the Spent Fuel Pool Instrumentation System," Revision 5
20. 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
21. Westinghouse Document, WNA-DS-02957-GEN, "Spent Fuel Pool Instrumentation System System Design Specification", Revision 4
22. 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 26 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Attachment 1 Elevation View of Spent Fuel Pool Instrumentation System

J:+

I 119' -3" Refueling Floor II 117'-10" Level 1 f 118'-6 7/8" Upper Range Value 106'-1 1/B" Level 2 ---+-

96'-9" Level 3 96'-9" Lower 96'-1 1/8" Top of Fuel .D.AAAAA.D..a Range Value SPENT FUEL POOL 23'-6" Ground Floor REACTOR BUILDING ELEVATION VIEW Page 27 of 29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Attachment 2 Plan View of Spent Fuel Pool Instrumentation System lORTH TWISTED CON DARY SHIELDED TRANSMITTER PAIR

--~~~ag*~~~--

SECONDARY INDICATI RI MARY 1RANSMITIER RI MARY INDICATOR REACTOR BUILDING MAIN OFFICE BUILDING TURBINE BUILDING PLAN Page 28of29

Oyster Creek Nuclear Generating Station Sixth Six-Month Status Report for the Implementation of SFPIS February 26, 2016 Attachment 3 Mounting Bracket Drawing 10067E58

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