ML16068A139

Attachment 5 - Technical Information for Proposed Initiating Conditions and Emergency Action Levels

ML16068A139
Person / Time
Site:	Seabrook
Issue date:	02/27/2016
From:	NextEra Energy Seabrook
To:	Office of Nuclear Reactor Regulation
Shared Package
ML16068A128	List: ML16068A130 ML16068A131 ML16068A137 ML16068A139 SBK-L-15120, Independent Spent Fuel Storage Installation (ISFSI) Ics/Eals
References
SBK-L-15120
Download: ML16068A139 (74)

Text

ATTACHMENT 5 Technical Information for Proposed Initiating Conditions and Emergency Action Levels

SEABROOK STATION SPECIFIC VALIDATION DOCUMENTS FOR NEI 99-01 REV. 6 EAL SUBMITrAL Vi- Technical Specifications Table 1.2 Mode Definitions V2 -Spent Fuel Pool Instrumentation Levels Drawing Per EC281289 V3 - EPCALC-06 Effluent Monitor Values for R EALs V4 - ODCM and Technical Specification Basis for Site Boundary Receptor Point V5 - UFSAR Table 12.3 Manipulator Crane and Spent Fuel Building Monitor Ranges V6 - UFSAR Section 9.1.4 Description of Refueling Pathway V7-Table Hi Procedure References V8 - UFSAR Table 12.3-15 WRGM Ranges V9 - 0S1000.09 Procedure References to Refueling Pathway Level Indicators VIO - UFSAR Tab'le 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges Vll - EPCALC-06 RVLIS Values V12 - UFSAR Table 12.3-14 Manipulator Crane Monitor Ranges V13 - UFSAR Section 6.2 Description of Containment Sumps V14-SAMG Calculation Aid CA-3 for Hydrogen Flammability in Containment V15 -Technical Specification Table 1.2 Cold Shutdown Temperature Limit V26 -O0S1000.09 Definition of Reduced Inventory V17 - Site Specific EAL Design Basis for 25 PSlG RCS Pressure Range V18 - UFSAR Section 8.3.2 DCV 105 Limit V29 - NUHOMS HSM Dose Rates Technical Specification V20 - Site Specific CSFST Core Cooling V21 - Site Specific CSFST Integrity V22 - Site Specific CSFST Heat Sink V23 - EPCALC-06 Rad Values for Fission Product Barrier Matrix V24 - Site Specific CSFST Containment V25 - Technical Specification Table 3.3-4 Containment Spray Setpoint V26 - UFSAR Section 8.4.2 SBO Coping Basis V27 - EC282184 Seismic Monitoring System Upgrade Description of OBE Lights V28 - Verification of Fire Alarms V29 - Site Specific CSFST Subcriticality V30 - EPCALC-06 Let Down Monitor Value V3i - Tech nical Specification 3/4.4.8 RCS Specific Activity V32 - UFSAR Table 11.5-1 Effluent Monitor Ranges

CROSS REFERENCE NEI 99-01 REV. 6 INITIATING CONDITIONS! SEABROOK STATION SPECIFIC VALIDATION DOCUMENTS Initiating Condition Applicable Validation Documents ALL V1 RG 1 V3, V4 RG 2 V2 RS1 V3, V4 RS2 V2 RAl V3, V4 RA2 V2, V5, V6 RA3 V7 RU1 V8, V32 RU2 V6, V9, ViO CG 1 VIO, Vll, V12, V13, V14 CS 1 Vll, V12, V13 CA1 Vil, V13 CA3 V15, V16, V17 CU1 V13 CU3 V15 CU4 V18 EU1 V19 FISSION PRODUCT BARRIER POTENTIAL LOSS/LOSS V23 FUEL CLAD BARRIER POTENTIAL LOSS 1.A V20 FUEL CLAD BARRIER LOSS 2.A V20 FUEL CLAD BARRIER POTENTIAL LOSS 2.A V20 RCS BARRIER POTENTIAL LOSS 1.B V2 1 RCS BARRIER POTENTIAL LOSS 2.A V22 CONTAINMENT BARRIER POTENTIAL LOSS 2.A V20 CONTAINMENT BARRIER POTENTIAL LOSS 4.A V24 CONTAINMENT BARRIER POTENTIAL LOSS 4.B V14 CONTAINMENT BARRIER POTENTIAL LOSS 4.C1 V25 HA5 V7 HA6 OPERATIONS ABNORMAL PROCEDURE OS1200.02 HU2 V27 HU4 V28 MG1 V26, V20 MG8 V18 MA5 V29 MU3 V30, V31 M U5 V29 M U7 V25

V1 - Tech Spec Table 1.2 Mode Definitions TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 24 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

hours.

D At least once per W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

SFCP In accordance with the Surveillance Frequency Control Program TABLE 1.2 OPERATIONAL MODES REACTIVITY  % RATED AVERAGE COOLANT MODE CONDITION~keff THERMAL POWER* TEMPERATURE

1. POWER OPERATION > 0.99 > 5% >_350°F

2. STARTUP _>0.99 <*5% _ 350°F

3. HOT STANDBY < 0.99 0 _ 350°F

4. HOT SHUTDOWN < 0.99 0 3500°F > Tavg >2000°F

5. COLD SHUTDOWN < 0.99 0 < 200°F

6. REFUELING** < 0.95 0 < 140°F

• Excluding decay heat.

• • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

SEABROOK - UNIT 1 -

1-8 mnmn 141 Amendment 4

EXECUTIVE

SUMMARY

The purpose of this engineering change (EC) is to install reliable primary and backup level instrumentation to monitor Spent Fuel Pool (SFP) water level in accordance with the requirements of Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Issuance of Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, March 12, 2012. This EC includes several hold points as described in Section 5.1.6.

Revision 001 revised the BC classification from quality Related to Safety Related to ensure that raceway components are procured as Seismic Category I components in accordance with UFSAR Section 3.10(B) and table 3.2-1. Sketch SK-1000 was also revised to correct a typo's on the instrument tag ID's and drawing Safety Classification.

Revision 002 provides for the following:

1. Revision 002 relocates and modifies the Westinghouse designed and fabricated pool-side probe brackets. The original mounting locations on the North and the South sides of the Fuel Storage Building (FSB) spent fuel pool (SFP) are changed to the West side of the pool to facilitate and simplify the installations. The new mounting locations on the West side of the pool will reduce implementation concerns inherent with the original FHM crane rail pocket locations, including welding and modification of the concrete curb. The West location simplifies mounting of the brackets using Hilti concrete anchors and significantly reduces the potential for foreign material entering the pool during installation. Refer to the discussion below and in BC Section 2.5.H for details. See SK 101 570-EC28 1849 located in the EDMS panel for the optional bracket locations.

See SK 11 1576-EC281849, sheets 1-3 for optional bracket assembly details.

2. The modified pool side bracket design includes a davit socket to facilitate installation of a tool to lift the probe assembly for calibration. Hold Point 1 is closed.

3. Since the Fuel Handling Machine cannot physically access the West end of the SFP, exclusion zones will not need to be established around the new sensor locations. Hold Point 2 is closed.

4. Revised the safety class on Sketch SK-lO001 to SC-3.

5. Incorporated Westinghouse Specialty Tool List (FP700485) into the ADL.

6. Incorporated revised Westinghouse spare parts list into FP700484.

7. Revised the channel verification acceptance criteria in BC Section 5.3 to initiate channel maintenance and!/or channel calibration if the deviation between indicated level readings exceeds 3.0 inches between the new SFPIS indicators or 2.0 inches between the new SFPIS indicator and the existingl1-SF-LI-2607 indicator.

EN-AA-205-l1100-F01, Revision 4 ECFORM- 110

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8. Revised the markup of SFPIS Configuration Settings Document (1-NHY-508637) to identify that the level transmitter Low Trim Point, High Trim Point and Sensor Ref. count values are for reference only and vary dependent on transmitter calibration.

9. Replaced FP 700493, SFP15 Compliance Matrix - FPL/ NextEra Seabrook," with new Westinghouse document WNA- DC-00270-NAH Rev. 1. Closed Open Item

10. Added Westinghouse Letter LTR-SFPIS-15-34, Clarification of 10 Year EQ-QR-269 Test Basis -

Supplement to WNA-TR-03149-Gen, Rev. 2," into FP 700483.

11. Added Westinghouse letter for part number clarification into FP 700499 (SFPIS Technical Manual).

12. Added Hold Point 4 for modification SEP lighting on the south west corner of the Spent fuel Pool (Ref. BC Section 2.5.T and 5.1.6).

Revision 003 provides for the following:

1. AWA 01 added FP700489 onto the ADL.

2. AWA 01 revised drawing 11 1576-EC28 1849 sht. i to correct the center line dimension for locating the Davit Socket on the SFPIS mounting bracket. Refer to the markup in the BDMS folder.

3. A 4" minimum embedment could not be achieved for one of the 5/8" diameter HKB3 anchors that are used mount probe 1-SF-LE-26 16. AWA 2 approved installation of the 5/8" diameter HKB3 anchor with a 2-3/4" minimum embedment. The capacity of the 5/8" Hilti KB3 with 2-3/4" vs 4" min. embedment is acceptable for the subject installation. Refer to the markup of SK 111576-EC281849 Shts. 1 & 2 and FP 700468 in EDMS.

4. AWA 03 added a requirement to post signs at the entry points into the Spent Fuel Building and CEVA to exclude the use of portable radios and cell phones during ELAP events (Ref.

AR 02063219). See BC section 5.1.5 for sign requirements.

AWA 03 also added a requirement to post signs on the front of control panel 1-SF-CP-300-A

& B to identify that radio use will interfere with SFPIS indications. BC section 5.1.5 is revised to identify the new posting requirement for panels 1-SF-CP-300 A & B.

5. The scaling values that Westinghouse provided for the SEPIS channels were not referenced to the top of the fuel rack. AWA 04 revised the scaling and corrected the guidance provided in section 5.2 of FP 700499 for level transmitter scaling. An offset value was used to adjust the scaling for the unmeasurable level from the top of the fuel rack to the bottom of the probe. The scaling values for the SEPIS digital display (SF-LI-2616 and 2617) and MPCS analog points (A4220 and A4172) are revised to align the indications with the revised transmitter scaling.

EN-AA-205-1 100-F01, Revision 4 ECFORM- 110

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6. The offset setting that was added to the transmitter scaling by AWA 04 resulted in an unexpected shift in the transmitter output. AWA 05 set the offset setting back to 0.00 inches and revised the transmitter scaling to include the required offset (7.88 inches to 326.41 inches). The revised transmitter parameter settings are shown in the markup of configuration settings document 1-NHY-508637 in the EDMS folder.

7. AWA 06 revised the low and high trim points to provide additional margin to the dead zones at the top and bottom of the probe. This change was recommended by Westinghouse to ensure signal stability as level approaches the upper and lower probe dead zones.

AWA 06 also documented the counts at the revised trim points (high and low) and the sensor reference (fixed and auto). The sensor reference counts for the calibration kit were also recorded for future reference. Refer to the markup of the configuration settings document 1-NHY-508637 in the EDMS folder for the new trim point values and revised counts.

8. AWA 07 was superseded by AWA 08.

9. The field identified that the replacement assembly from inventory (CID 171805) is not the same as the installed assembly. Based on correspondence from Westinghouse that stated that the new battery assembly was fully qualified, AWA 07 approved the use of an alternate battery assembly for Westinghouse P/N. PS20035H05. Further correspondence with Westinghouse revealed that the qualification documentation for the new battery assembly is not ready and will not be available for several weeks. AWA 08 identified that the delivered part (CID 171805) is not acceptable for use at this time. The original style battery assembly is available as a warrantee item and is being shipped to Seabrook with an estimated delivery date of 10/21/15.

AWA 07 also identified that the new battery assembly has a different OEM P/N', than the original assembly (Phoenix Contact P/N UPS-BAT/VRLA-WTR/24DC/26AH-23 20429 vs UPS-BAT/VRLA/24DC/26AH-2320429). Westinghouse identified that the new battery assembly has the same OEM P/N as the o~riginal assembly, but is physically different and has revised ratings. The differences will be evaluated as part of the qualification documentation.

The correct OEM P/N for both configurations is UPS-BAT/VRLA-WTR/24DC/26AH1-2320429. The SFPIS Bill of Materials on FP 700477 is revised to reflect the correct part number for the battery assembly. Refer to the markup in the EDMS folder.

10. Added procedure IS 1682.224 and IS 1682.226 onto the ADL.

11. Added FP700468 onto the ADL.

EN-AA-205-1 100-F01, Revision 4 ECFORM-l 110

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12. As a result of a failure of SFPIS components at Point Beach it was determined that the Westinghouse supplied signal duplicators are digital devices. The parameters fields for tag IDs 1-SF-LY-2616 and 1-SF-LY-2617 are therefore revised to identify that these components are digital devices that contain firmware. The firmware was also classified as level B in accordance with the fleet software control program (IM-AA-205-l1100). Refer to the new IM-AA-202-F0 1software classification form in the EDMS folder the EMRs for tag IDs 1-SF-LY-2616 and 1-SF-LY-2617.

13. The IM Department was not able to load the UJPS-CONF utility software that Westinghouse provided onto a laptop. As part this revision the software classification will be updated to allow use of the newer version of the software that is available from Phoenix Contact as a download (QUINT UPS UPS-CONE_online setup2 2_7_0.exe). Refer to the revised IM-AA-202-F01 software classification form in the EDMS folder.

14. Engineering Evaluation EE-15-016, "CEVA and ESWGR Beyond Design Basis External Event Ventilation Evaluation" was developed to perform a heat up analysis for the CEVA and ESWGR rooms. This evaluation confirmed that area temperatures will remain below the temperature to which Westinghouse qualified the Spent Fuel Pool Instrumentation during an ELAP event. Augmented ventilation will therefore not be required following a beyond Design Basis Event to maintain CEVA and Essential Switchgear Room temperatures below the equipment qualification temperature of 140 deg. F. EC sections 2.3.8, 2.5J and 5.3 are revised to identify that compensatory ventilation will not be required. Specification S-S-i1-E-0227 is also revised to reflect the maximum temperature conditions that could occur in the CEVA and Essential Switchgear Rooms during a beyond design basis extend loss of power event.

15. Since the FLEX equipment storage facility and some of the FLEX equipment will not be available in time to support compliance of NRC Order EA-12-049, an extension has been submitted to the NRC to extend the compliance date for the Order until May of 2016. As a result of the extension, the FSG procedures that are required to support use of the new instrumentation during an ELAP will not be completed in time to support the compliance date for NRC Order EA-12-051. As a remedial action the required guidance for the Spent Fuel Pool Instrumentation will be added to existing plant operating procedure OS 1215.07, "Loss of Spent Fuel Pool Cooling or Level."

16. During the 2015 FLEX audit the NRC inspector noted that the flexible conduit runs from each SFPIS probe to the west wall of the fuel storage building will present a potential trip hazard at the deck of the pool that could render the instrumentation inoperable. It was recommended that the station consider the installation shielding to protect these sections of the cable runs (Ref. AR 02063248). Hold Point 4 is added to address the potential trip hazard associated with the SFPIS sensor cable installations at the south side of the pooi.

Refer to EC section 5.1.6.

EN-AA-205-1 100-F01, Revision 4 ECFORM-l 110

V2 - SFP Instrumentation Levels Per EC28 1849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL Page 5 of 75 INSTRUMENTATION UPGRADE FOR NRC ORDER EA-12-051

17. Revised the channel verification acceptance criteria in EC Section 5.3 to initiate channel maintenance and!/or channel calibration if the deviation between indicated level readings exceeds 3.0 inches between the new SFPIS indicators or the existingl.-SF-LI-2607 indicator.

18. During testing it was determined that the serial numbers for UPS modules 1-SF-XX-2616-1 and l-SF-XX-2617-1 were programmed into the units did not align with the primary and backup channel designations in drawing 1-NHY-508637. The serial numbers were revised to align the panels with field installation and a note was added to identify that the serial numbers are set by the field based the installed unit.

19. The SFPIS technical manual (FP700499) was updated to reflect Westinghouse's revised replacement frequency for the couplers, coaxial cables and coaxial connectors. According to Westinghouse these components are to be replaced every ten years.

Revision 004:

Issued to remove the Critical Attribute for verifying that the Fuel Handling Machine, when in AUTO, will remain 1 ft from probes. This criterion was established in revision 000 when the probe locations were on the North and South pool walls. This attribute no longer applies to the final probe locations on the West end of the pool as installed by Revision 002 to this BC. The Spent Fuel Handling Machine is physically not capable of approaching the installed probe location, refer to details in BC section 2.3.14. For this same reason the MATP has been revised to delete any and all references to testing to verify no interaction between the machine and the installed instrument probes.

Additionally the EDMS folder for Specification S-S-1 -E-0227 has been corrected to eliminate the duplicate page.

This revision affects BC sections 2.5.S, 5.1.1, the MATP and the Specification folder in BDMS. These changes support the installed design as modified by revision 002 and the changes in this revision do not change the Scope, Intent or Requirements of the BC and the original 10 CFR Applicability Determination and 72.48 Pre-Screen along with the BC R/002 10 CFR 50.59 Screen remain applicable.

2.0 DESIGN 2.1 Bases for Current Design The Primary function of the spent fuel storage facilities is to provide for the safe handling and storage of irradiated fuel assemblies under water, new fuel assemblies, and control rods. The safety function of the spent fuel pool and storage racks is to maintain the spent fuel assemblies in a subcritical array during all credible storage conditions, and to provide a safe means of cask loading of the assemblies for dry fuel EN-AA-205-1 100-F01, Revision 4 ECFORM-l 110

V2 - SFP Instrumentation Levels Per EC28 1849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL Page 6 of 75 INSTRUMENTATION UPGRADE FOR NRC ORDER EA-12-051 storage. System and Component safety class, seismic class, code of design and construction, and code classes are listed in UFSAR Table 3.2-2.

The Fuel Storage Building and Spent Fuel Pool are safety-class 3, seismic Category 1 structures designed to the requirements of Regulatory Guide 1.13 (Spent Fuel Storage Facility Design Basis),

Regulatory Guide 1.29 (Seismic Design Classification) and Regulatory Guide 1.117 (Tornado Design Classification) to ensure optimum protection for stored fuel assemblies against the effects of extreme natural phenomena such as safe shutdown earthquake, tornadoes, hurricanes, missiles, and floods.

These design considerations extend to the SFP wall, fuel storage racks, and other critical components whose failure could cause criticality, loss of SFP cooling, or physical damage to stored fuel.

Spent fuel pool water level at the upper elevation of the SFP is indicated by a level transmitter (SF-LT-2607) which causes an audible alarm in the control room on high or low water level. Level from this instrument loop can also be observed as an analog computer point on the MPCS (Al1185) over a 4 foot span (22.875 feet to 26.875 feet, referenced to the reactor flange (-) 1 ft, 10.5 in.).

2.2 Reason for Design Change After the earthquake and tsunami at the Fukushima Daiichi nuclear plant in March, 2011, the lack of indication of Spent Fuel Pool (SFP) water level led to concerns about the potential for loss of inventory. While it was eventually determined that SFP integrity and fuel cooling were not compromised, considerable efforts were made following the event to assure that there was adequate cooling of the fuel in the SFP. These efforts diverted resources and attention from other critical tasks that were required to manage the event, demonstrating that confusion and misapplication of resources can result from beyond-design-basis external events when adequate instrumentation is not available.

The NRC subsequently determined that enhanced spent fuel pool instrumentation represents a substantial increase in protection to public health and safety and issued Order BA- 12-051. The Order requires that licenses install reliable primary and backup level instrumentation to monitor Spent Fuel Pool (SFP) water level. The Nuclear Energy Institute (NEI) provided a document, NEI 12-02, Industry Guidance for Compliance with NRC Order BA- 12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" which provides an approach for complying with Order BA-12-05 1. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order BA-12-051, Reliable Spent Fuel Pool Instrumentation, endorses NEI 12-02, subject to clarifications and exceptions specific to Section 3.4, Qualification, of NEI 12-02, as an acceptable means of meeting the requirements of Order BA-12-051. Design of the Spent Fuel Pool Instrumentation described in this BC will be consistent with the guidelines of JLD-ISG-2012-03 and NBI 12-02.

2.3 Description of Design Change This BC provides for the installation of two new independent level instrument channels (SF-L-26 16 and SF-L-26 17) in the Spent Fuel system. The new instrument channels will be used to monitor spent BN-AA-205-1 100-F01, Revision 4 ECFORM- 110

V2 - SFP Instrumentation Levels Per EC281849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL Page 7 of 75 INSTRUMENTATION UPGRADE FOR NRC ORDER EA-12-051 fuel pooi level during and following beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.

The new Spent Fuel Pool Instrumentation System (SFP1S) was designed and fabricated by Westinghouse Electric Co. Both instrument level channels will be permanently mounted and access into the Spent Fuel Building (SFB) will not be required to operate or obtain level data from the instrument channels.

Each channel will be capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating water level ((-) 1 FT to 24 Ft, Plant Elev.). Each channel will display the indicated SEP water level to reasonably high accuracy when accounting for worse case environmental conditions and instrument uncertainties. Level readings will be provided in feet based on height above the fuel rack, i.e. top of the fuel rack is 0.0 feet. The indicated level range will be from 1.0 to 25.5 feet above the fuel rack.

Three critical spent fuel pool (SEP) water levels within this range are defined in NEJ 12-02 Revision 1.

The Spent Fuel Pool level instrumentation installed per this EC will be designed to provide a reliable means of SEP level measurement at the following critical water levels:

a. Level 1, level adequate to support operation of the normal spent fuel pool cooling system -

This level is defined as:

1. Level at which reliable suction loss occurs due to uncovering the coolant inlet pipe OR any weirs or vacuum breakers associated with suction loss, or

2. The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis.

To determine the higher of the two levels the following was taken into consideration:

(1) The level at which reliable suction loss occurs due to uncovering the coolant inlet pipe or any weirs or vacuum breakers associated with suction loss. This level was established based on nominal suction strainer inlet elevation and conservative estimate for the onset of vortexing.

The actual effect of the strainer on this level has been formally determined by calculation C-S-1-24606, "Spent Fuel Pool Level for Reliable Pump Suction." Based on this calculation the elevation for reliable pump suction is plant elevation 23 ft., 4 inches.

(2) The level at which the normal spent fuel pool cooling pumps loose required NPSH assuming saturated conditions in the pool. Seabrook Calculation C-S-1-24606, "Spent Fuel Pool Level for Reliable Pump Suction," establishes the point of zero NPSH margin at plant Elevation 22 ft., 4 inches.

EN-AA-205-l1100-F01, Revision 4 ECFORM- 110

V2 - SFP Instrumentation Levels Per EC28 1849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL Page 8 of 75 INSTRUMENTATION UPGRADE FOR NRC ORDER EA-12-051 With the spent fuel pool at 212 degrees F (saturated conditions) the NPSHA is approximately 11.2 ft. The NPSJAR for the spent fuel cooling pump is 10 ft. at 212 degrees F. This results in a ratio of NPSHA!NPSHR value of approximately 1.12. Therefore, the NPSHA is greater than the NPSHR at saturated conditions.

The higher of the above points is the level where the inlet strainer will lose suction (Item (1) above). Therefore, Level 1 has been established at plant elevation 23 ft., 4 inches ('- 24 ft. 10 in.

above fuel racks) for both the primary and backup instrument channels.

b. Level 2, level adequate to provide substantial radiation shielding for a person standing on the spent fuel pooi operating deck - Level 2 is defined as the level that provides substantial radiation shielding for personnel to respond to Beyond-Design-Basis External Events including the initiation of SFP makeup strategies that would require access to the Fuel Storage Building (FSB).

Indicated level on either the primary or backup instrument channel of greater than an elevation of 10 ft., 9-1/2" (-12 ft-3mn above fuel racks) will provide substantial radiation shielding for a person standing on the SFP operating deck. This elevation is approximately 13 feet above the top of the spent fuel assemblies positioned in the spent fuel racks (elev. (-) 2 ft -2.5 in.). With 13 feet of water above the highest fuel element position, the calculated dose rate at the surface of the SFP is less than 2.5 nmrermlhr (Ref. UFSAR Section 12.3.2.l.c). This monitoring level ensures there is adequate water level to provide substantial radiation shielding for personnel to respond to Beyond-Design-Basis External Events including the initiation of SFP makeup strategies that would require access to the Fuel Storage Building (FSB).

c. Level 3, level where fuel remains covered - Level 3 corresponds nominally (e.g. +/-/- 1 foot) to the highest point of any fuel rack seated in the SFP. Level 3 was defined in this manner to provide the maximum range of information to the Operators and emergency response personnel. An indicated level on either the primary or backup instrument channel of less than plant elevation (-) 6 inches. This is the nominal water level approximately 1.5 ft. above the top of the fuel racks. This monitoring level will assure the maximum range of level information is available to the plant Operators and emergency response personnel. This level is also assumed to be the minimum level that assures that adequate water level remains above the top of the stored fuel seated in the SFP (nominal elevation of (-)2 ft - 2.5 in.).

The following figure shows the critical SFP levels described above:

EN-AA-205-l 100-F01, Revision 4 ECFORM- 110

V2 - SFP Instrumentation Levels Per EC281849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL INSTRUMENTATION UPGRADE FOR IPage 9 of 75 NRC ORDER EA-12-051I SEABROOK ELEVATION SPENT FUELview POOL LEVELS OEA1INGFLOOR EL, 25

~~NORMAL. WATER LEVEL EL 23'-4" TO 25 -9'I

-il" TO.

... EVIEL NST RUIME

==--- 'IRMNSMIITER SlENSITIVI T

BAM°%6"

............ LEVEL 1 LOSS 0 r TNPSH0 212"F EL. 22"~V'*/"

4- -TOP OF" STRAIN*ERa.L 21V5"

___________________RELIARLE 4 ITECI'IFCAL SUCTION LEVEL EPECIFICATtON LMllEL..23-4" EL. 2T"l/z *.-

-1 LF.VELFOR SURSTAN~TIALRAOIATON SHIELDINGEL. 1Q-gz 111 "- I-" ..... *' 2 0212" F NPSI. 11.2Ft

-LEVEL WIERE FUEL REMAN$ COVERED 1 0" O " 5 4 -

-- lEVELNSSTRIME .

1"TOP oF FUEL RACI EL.,(-1 r-5s4 "

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°,

2.3.1 Design Inputs:

NRC!/NET Guidance:

a. NRC Order EA-12-051, "Order Modifying Licenses with regard to Reliable Spent Fuel Pool Instrumentation"

b. NEI 12-02, Industry Guidance for Compliance with NRC Order EA- 12-051, to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

c. JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, August 29, 2012, Nuclear Regulatory Commission, Japan Lessons Learned Project Directorate.

Westinghouse Documentation:

a. FP 700481, Spent Fuel Pool Instrumentation System Wire List (WNA-WD-01094-GEN).

b. FP 700498, Spent Fuel Pool Instrumentation System Torque Specification (WNA-IG-00452-GEN).

c. FP 700487, Spent Fuel Pool Instrumentation System (SFPIS) Walk down Report (CIS-WR-13-6).

d. FP 700482, Spent Fuel Pool Instrumentation System (SFPIS) Standard Product Test Strategy (WNA-PT-00 188-GEN).

e. FP 700479, Spent Fuel Pool Instrumentation System Power Consumption (WNA-CN-003 00-GEN).

EN-AA-205-1 100-F0Ol, Revision 4 ECFORM-1 10

V3 - Plant Vent and Main Steam Line Effluent Calculations for RG1/RS1lRA1 0

FPL Energy Seabrook Station SEP# 2006013

SUBJECT:

Calculation of Effluent Release Values Used in NEI EALs FROM: D. L. Young DATE: March 28, 2006 TO: File Attached are the calculations for the effluent release values used in the NEI Emergency Action Levels (EALs). The calculation number is EPCALC-06-02.

DLY/dly cc: 0. Robinson

V3 - Plant Vent and Main Steam Line Effluent Calculationsfor RG1/RS1IRA1 Calculation oV Radiological Effluent EAL Values EPCALC-06-0D2 Plant V.ent-Cffijzen~ Calculation for a Site-Arfe.a Em.f.*rg*nc~y..i...

DFB, Avg Disp mrem-m3 wrem-m 3 Factor mrem-sec Isotope pCi-yr uCi-yr seclm3 uCi-yr (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.83E+03 8.5nDE-07 7.51 E-03 Kr-87 5.92E-03 5.92E+03

• 8.50Eo07 5.03E-03 Kr-88 1.47E-02 1.47E+O4 8.50E-07 1.25E-02 Kr-85m 1.17E-03 1.17E+03

• 8.50E-07 9.95E-04 Xe-135 1.81E-03 1 .81E÷03 *r 8.S0E-07 I .54IE-03 Xe-133 2.94E-04 2.94E+02

• 8.50E 07 2.50E-04 Kr-85 1.6lE-05 1,61 E+01 8.50E-07 1 .37E-05 Xe-131m 9.15E-05 9.15E+O1 8.50E-07 7,78E-05 Xe-133m 2.51 E-04 2.51E÷02 8.50E-07 2.13E-04 Xe-I 35m 3.12E-03 3.12E÷03 8.50E-07 2.65E-03 DF'* (per year) OF' (per hour)

- '32 *"L;rrm.'secJ F"'" 5 I .... iyl~ L (Note 4) 1I(DF'I (per hour)) WRG3M release rate I. .85÷05 uCi-hr ,oI 0E+o2 mre (Note 5) (Note 6)

Note 1 - from ODUM, Appendix B. Section 5.2.1.2 Note 2 - DFBi

• 1E+06 pCiruCi Note 3 - From ODCM, Appendix B, Table B.7-4 Note 4-.11(865*24) or l/hours in a year Note 5 - 1/(3.5IE-06)

Note 6 - Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal 100 mR ReV. 0 Page 1 ReV.0Pae 13127/2008 I II I III II I '1 1 I

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• J!*l~lt*nt and Main Steam Line Effluent Calculationsfor RG1/RS1/RA1 Calculation of Radiological Effluent EAL Values EPCALC-O6-02 Plaint Vent EffluentCalculation for a General Eme~rgency ...

DFB1 DFBL Avg Disp mrem-ma Factor 3 mrein-sec Isotope pci-yr uCi-yr seclmn uCl-yr (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.832+03 8.S0E-O7 7.512-03 Kr-87 5.92E,-03 5.92E÷03 B.50E-07 5,03E-03 Kr-88 1.47E-02 1.47E+04 *k 8.502-07 1.25E-02 Kr-85m 1.172-03 1.17E÷03 8,50E-07 9,952-04

• b Xe-135 1.81 E-03 S81E÷03 8.50*-07 1.54E-03 Xe-133 2.94E-04 2.94E+02 8.50E-07 2.50E-04 Kr-85 1.61 E-05 1,.61E+01 8.50E-07 1.37E-05

• b Xe-I131m 9.15SE-O5 9.15E+01 8,50E-07 7.78E-O5 Xe-133rn 2-511E-04 2Z51E÷02

• 8,50E-07 2.13E-04 Xe-135mn 3.12E-03 3.1225+03 *k 8.50E-07 2.65E-03 rmrem-secI DF'1(per year) IDF', (per hour)

(Note 4) 1/(DF', (per hour)) WRGM release rate uLi (Note 5) (Note 6)

Note 1 - from ODCM, Appendix B, Section 5.2.1.2 Note 2 - DFBi

• 1E+0S pCifuCi Note 3- From ODCM, Appendix B, Table B.7-4 Note 4- 1/(365*24) or 1/hours in a year Note 56- 11(3.51 E-06)

Note 6- Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal I R Rev, 0 0 ~~Page ReV. 232720 312712006

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!3 - Plant Vent and Main Steam Line Effluent Calculationsfor RG1/RS1/RA1 Calculation of Radiological Effluent EAL Values E PCALC,_-06-02 Main steam Line Effluent Calculation for an Unusual Event.

DFB1 DFB1 3 Avg Disp DF'1 ntrem-m 3 mrem-n3 Factor mrarm-sec pCi-yr uCi-.yr sec/m"= uCl-yr Isotope (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.832+03

• 3A02-0B 3,0DE-02 Kr'-87 5.92E-03 5.92E÷03 3.40E-06 2.01 E-02 Kr-a88 1,.47E-02 1.472+04 3..40E-05 5.O0E-02 Kr-8B5m 1.17E-03 1.17E+03

• 3.40E-06 3.982-03 Xe-I135 1.81 E-03 1.812+03 3AOE-06 6.15E-03 Xe--133 2,94E-04 2.94E÷02

• 340E-05 i.00E-03 Kr-B5 1.612-05 1.61E+01 *i 3AOE-05 5.472-05 Xe-131m 9.15SE-05 9.152+01 3.40E-O6 3.112-04 Xe-I133m 2.51 E-04 2.51E÷02 3.402-06 8.53E-04 Xe-I135m 3.122-03 3.12E+03 3.40E-06 I .06E-02 DF'1 (per year) DF'(per hour)

[I~~~omrem-seo Fl~ T~ = "4E0 mrem-sec u-r (Note,;)

lI(DF'1 (per hour)) SI. release rate 7m12e04 -ecl h* = .6+03 uG!

(Note 5) (Note 6)

S.L. release rate S.L. Flowrate Release concentration (Note 7)

This table converts a release concentration to a main steam line monitor dose rate.

Decay MS J(t) Rel Cone Inverse Rounded Time ndi-hr Dose Rate uCi Dose Rate Dose Rate (his) cc-mR hrlmR mPlhr mRlhr 0.5 1.92E-02 1 1.622-03 1.18E+01 0 0

'1 2.16E-0Z / 1.622-03 1.33E+01 0 0 2 2.672-02 1 1.022-O3 1.65E-01 0 0

,5 4.932-02 1 1.52E-03 3.04E÷01 0 0 10 1.27E-01 I 1.62E-03 7.83E+01 0 0 20 4.97E-01 I 1.62E-03 3,06E+02 0 0 (Note 8) (Note 9)

Note I - from ODCM, Appendix B, Section 5.2.1.2 Note 6 - Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal Note 2 - DFBi

• 1E+06 pCi/uCl ODCM total body dose rate of 500 mR/yr Note 3 - From ODCM, Appendix B, Table B.7-6 Note?7- from SBC-709, page 20 - Used SRV Note 4 - 1/(365"24) or 1/hours in a year flow as bounding (ILe., larger than ASDV).

Note 5- 1/(1.412-5) Note 5 -from SBC-709, page 19 Note 9 - lInverse Dose Rate Rev. 0 Page 3 Rev.0 Pae 3312712006

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V3 - Pfa'nf't~Thn* *n*? Main Steam Line Effluent Calculationsfor RG1/RS1IRA1 Calculation of Radiological Effluent EAL Values EPCALC-06-02 Main. Steam Lini* Effiu~ent Calculation for an Alert -

DFB1 DFB1 Avg Diap DF'1 mrem-ma mrem-m3 Factor toremo-sec pCi-yr uCi-yr sec/in3 uCI-yr Isotope (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.83E+03

• 3.40E-05 3,00E-02 Kr-87 5.92E-03 5.92E+03 3.40E-06 2.01 E-02 Kr-88 1.47 E-02 1.47E+04 3,405-0B 5,0OEo02 Kr-85m 1,17E-03 1.17E÷03

• 3.4OE-06 3.9BE-03 Xe-I135 1,81 E-03 1.81E+03 3.40E-06 6.15SE-03 Xe-It33 2.94E-04 2.94E+02 3.40E-06 1.00E-03 1r Kr-85 1.61E-Q5 1.61E÷01 3.4OE-06 5.47E-05 Xe-I131rm 9.15SE-05 9.!5E+01

• 3.40E-06 3.1IE-04 Xe-I133m 2.51 E-04 2.51 E+02 3.40E-06 8.53E-04 Xe-1 35m 3.12E-03 3.12E÷03 3.4OE-06 1.06E-02 DF'I {per year) .... DF'1 (per hour)

F1.3E0 1 mrems I-sCuir I IAi4E-04~Y I = I 41E05 mre'sctui (Note 4) 1I(DF'1 (per hour)) S.L. release rate 7,1E+4 re-seJ .1E+ 1 rI (Note 5) (Note 6)

S.L. release rate SJ.L Flowrate Release concentration

[U~+ sUe I F ~ =i 325 ui I (Note 7)

This table converts a release concentration to a main steam line monitor dose rate.

Decay MS 4(t Rel Conc Inverse Rounded Time uCi-hr uCi Dose Rate Dose Rate Dose Rate (hrs) cc-mR hr/mR mRlhr mRlhr 0.5 i.92E-02 I 3.25E-O1 5.92E-02 17 10 1 2.16E-02 I 3.25E-01 6.66BE-02 15 10 2 2,67E-02 / 3.25E-01 8.23E-02 12 10 5 4.93E-02 / 3.25E-O1 1.52E-01 7 0 10 1.27E-01 / 3.25E-01 3.91E,.01 3 0

20) 4.97E-01 I 3.25E-01 1.53E÷00 1 0 (Note 8) (Note 9)

Appendix B, Section 5.2.1.2 Note 2I -- from Note DFBi ODOM,

• 1E+06 pClluCi NoteS6- Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal 200X ODCM total body dose rate of 500 mPtyr Note 3 - From ODCM, Appendix B, Table B.7-6 Note 7 -from SBC-709, page ~20 - Used SRV Note 4 -I1(365"24) or i/hours in a year flow as bounding (I.e., larger than ASDV).

Note 5- 1I(1.41E-5) Note 8 -from SBC-709, page 19 Note 9 - 1/Inverse Dose Rate RBv, 0 Page 4 Rev.0Pae 43/27/2008

V3 - Plant Vent and Main Steam Line Effluent Calculationsfor RG1IRS1/RAI Calculation of Radiological Effluent EAL Values EPCALC-06-02 Main Steam Line Effluent CalculaUion fora.Site.Area Emergency DFB1 DFBJ Avg Disp DF1R mrem-rn 3 mrem-m* Factor mrem-sec pCi-yr uCi-yr seclm3 u Cl-yr Isotope (Note 1) (Note 2) (Note 3)

Xe-t 38 8.83E-03 8.83E+03

• 3.40E-05 3.00E-02 Kr-87 5.92E-03 5.92E+'03 3.40E-06 2.01 E-02 Kr-88 1,.47E-02 1.47E+04

• 3AO0E--06 5.00E-02 Kr-85m I1,l7E-03 1.17E÷03

• 3AOE-06 3.98E-03 Xe-I135 1.81 E-03 1.812÷03 3.40E-06 6.1 5E-03 Xe-133 2.94E-04 2,94E+02

• 3.40E-06 1,002-03 Kr-85 1.61E-O5 1.61E+01

• 3.40E-06 5.47E-05 Xe-131 m 9.1 5E-05 9.15E+01 a 3.40E-06 3.112-04 Xe-133m 2.51E-04 2.512+02 3.40E-06 8.53E-04 Xe-135m 3.12E-03 3.12E+03

• 3.40E-06 1-06E-02 DFP= (per hour)

DFP1 (per year)

[ 1"iE-5m'rem-s~ec (Note 4) 1I(DF'1 (per hour)) S.L. release rate (Note 5) (Note 6)

S.L. release rate 8.1. Flowrate Release concentration (Note 7) "

This table converts a release concentration to a main steam line monitor dose rate.

Decay MS Ju, Rel Gone lnvelise Rounded Time uCi-hr uCi Dose Rate Dose Rate Dose R~ate (hrs) cc-mR Gc hrlmR mRlhr mPlhr 0.5 1 .922-02 2.852÷00 6.742-03 148 140 I

1 2,162-02 I 7.59E-03 132 130 2 2.672-02  ! 2,85E+00 9,35E-03 107 100 5 4.93E-02 I 2.852+00 1.73E-02 58 50 10- 1.272-01 I 2.85E+00 4.462-02 22 20 20 4.972-01 I 2.852+D0 1.752-01 6 0 (Note 8) (Note 9)

Note 1 - from ODOM, Appendix B, Section 5.2.1.2 Note 100 maR 6 - Dose (per NEIrate99-01 for 1 IC hour needed basis to equal discussion)

Note 2 - DFBi* 1E+06 p~iluCi Note 3 - From O0CM, Appendix B, Table 8.7-6 Note 7 -from SBC-709, page 20 - Used SRV Note 4 - 1/(36_5"24) or l/hours in a year flow as bounding (I~e., larger than ASDV),

Note 5 - 1/(1.41E-5) Note 8-*from SBC-709, page '19 Note 9 - 1/Inverse Dose Rate RoY. 0 Page 5 Rev.0 Pae 5312712006

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!L *W, "J*V.3trC*PM'*!h*,and Main Steam Line Effluent Cacltosfor RGI/RS i/RA I Calculation of Radiological Effluent EAL Values EPCALC-06~.02 Main Steamh Line Effluent Caloulation for a General Emergency .- -

DF81 DFB* Avg Oisp DF' mrem-m* -,Factor torero-sec lsotope p Ci-yr uCi-yr secfma uCi-yr (Note I) (Note 2) (Note 3)

Xe-I 38 8.83a+03 3.402-06 3.0015-02

• i Kr-87 5.92E-03 5.922+03 3.40E-06 2,01E-02 Kr-88 1.47TE-02 1.47E*-04 3.402-06 5.OOE-02 Kr-85m 1.172-03 1.17E+03 ak 3.4(}E-06 3.98E-03 Xe-1 35 1.8lE-03 1.81E+03 3.402-06 6.162-03 Xe-B33 2.94E-04 2.94E'*02 ak 3.40E-06 I .OOE-03 Kr-8$ 1.61 E.05 I1.6IE-i01 3.40E-06 5.47E..05 Xe-tS1m 9.I5E-05 9.152÷01 ** 3A40E-06 3.11IE-04 Xe-I133m 2.512-04 2.5IE-.02 3.402-06 8.632-04 Xe-t 36m 3.12E-03 3.12E+03

• 3.40E-06 1.062-02 OF'1 (per hour)

DF', (per year)

(Note 4) 11I (DPa

'1 prhour)) SL. release rote (Note 5) (Note 6) 3.1. release rate Retease concentration (Note 7)

This table converts a release concentration to a xnain steam line monitor dose rate.

Decay MS Jm Ret Conic hwlers~e Rounded Time uCi-hr uiC! Dose Rate Dose Rate Dose Rate (hra) cc-mR Co hrlmiR mR/hr niRJhr 0.5 1.92E-02 / 2.S5E+O1 6.742-04 1433 1,480 I 2.162-02 I 2.85*-0I 7.59E-04 1318 1,310 2 2.67E-02 / 2.85E+01 9,38E-04 1066 1,060 4.03E.02 I 2.852+01 1.73-0 577 570 10 1,.27E-01 I 2.85E+01 4,462-03 224 220 20 4,97E-flI I 2.85E+01 1.75E-.02 57 80 (Note 8)* (Note 9)

Nate I from O0CM, Appendix B, Section 5.2.1.2 Nate 6 - Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equel Note 2 - DFSi *"12+013 pCiluCi 1000 mR (per NEI 99.01 IC basis dlscussion)

Note 3 - From 00CM, Appendix B, Table 8.7-6 Note?7 - from SBC-709, page 20 - Used SRV Note 4 -1/(365"24) or 1/hours in a year flow as bounding (ILe., larger than ASDV).

Note 5 - 11(1.412-5) Note 8 - from SBC-709, page 19 Note 9 - 1lInverse Dose Rate Rev. 0 312712005

... *- V3 - Pl~'

- =ant , andT il/l'ain Steam Line Effluent Calculationsfor RG1/RS1/RA1 Calculation of Radiological Effluent EAL Values EPCALC-06-02 Prepared by:. Seirlcea*Aayt

.... D te Reviewed by:.

David Robinson Chemistry Supervisor Date Approved by:

EP Mn Rev. 0 Page 7 Rev.0 Pae 73/27/2006

V4 - ODCM and Tech Spec Basis for Site Boundary Receptor Point for RG1/RS1IRAI1 7.0 RADIOACTIVE GASEOUS EFFLUENTS 7.1 Dose Rate CONTROLS C1.7 .1 .. h...... se. .. .T ... .t. ...... .. ,...,.,.. .......... . .... ...........e,1 APPICBILTY At~4alel time s.~~~Ui sT1ia pcfctoJue5~

ACTIONi4: tefo1iW With the dose rate(s) exceeding the above limits, decrease the release rate within 15 minutes to within the above limit(s).

SURVEILLANCE REQOU]REMENTS S.7.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in Part B of the ODCM.

S.7.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table A.7.1-I-.

A.7-1 A.7-1ODCM Rev. 25

V4 - ODCM and Tech Spec Basis for Site Boundary Receptor Point for RG1/RS1/RA1 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The Low Population Zone shall be as shown in Figure 5.1-2.

MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be shown in Figures 5.1-1 and 5.1-3, respectively.

The definition of UNRESTRICTED AREA used in implementing these Technical Specifications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a.

CONFIGURATION 5.2.1 The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 140 feet.

b. Nominal inside height = 219 feet.

c. Minimum thickness of concrete walls = 4 feet 6 inches.

d. Minimum thickness of concrete dome = 3 feet 6 inches.

e. Minimum thickness of concrete floor pad = 10 feet.

f. Nominal thickness of steel liner =1/4, 3/8, and 1/2 inch for the floor, wall, and dome, respectively.

g. Net free volume = 2.704 X 106 cubic feet.

SEABROOK - UNIT 1 ~ 5-1

" 0

-., . . . miir~ _',+

F-GRE . -jI-- '

SIEAN XLUINARABUNAP

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SEABROK -UNITI 5-

V/5- UFSAR Table 12.3 Manipulator Crane and Spent Fuel Building Monitor Ranges TABLE 12.3-14 AREA RADIATION MONITORS (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr QTY. CLASS REFERENCE Containment Structure 6534 In-Core Instrument Seal Table GM 15 Non IE 12.3-2 (Note 4) 6535A, B Manipulator Crane GM 15 IE 12.3 -3 10-L104 2 (Note 1) (Note 4) 6536-1, 2 Personnel Hatch (Post-LOCA) Ion Chamber 2.5 1+-0 2Non IE 12.3 -3 6576A, B Containment (Post-LOCA) Ion Chamber 25 100-108 r/hr 1E 12.3-3 Primary Auxiliary Building 6537 Sampling Room GM 2.5 10"l.104 2 Non 1E 12.3-6 6538, 6539 RHR Pump Area GM >100 Non 1E 12.3-4 10-1.104 6540 Volume Control Tank Area Ion Chamber 8x104o 2 Non 1E 12.3-7 101-0 6541 Lower Level GM 2.5 Non 1E 12.3-5 10'-1.04 6543 Entrance GM >100 Non 1E 12.3-5 10'-1.104 6544 Entrance GM 2.5 3 Non 1E 12.3-6 10"'-.104 6545, 6546 Charging Pump Room GM 110 Non 1E 12.3-5 6547

V5 - UFSAR Table 12.3-14 - Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE 6508-1,2 PAB-HRAM Ion Chamber >100 10"2-104 r/hr 2 Non 1E 12.3-5 6563-1,2 PAB-HRAM Ion Chamber >100 10-2-104 r/hr 2 Non 1E 12.3-5 6517-1,2 RHR - Pump Vault HRAM Ion Chamber >100 10- -104 r/hr 2 Non 1E Fuel Storage Building 6549 Spent Fuel Pool Area GM 10-1-104 Non 1E 12.3-15 2.5 1 6518 Spent Fuel - HRAM Ion Chamber 2.5 10"2-104 r/hr 1 Non 1E 12.3-15 Control Room 6550 Main Control Board Area GM 0.5 10"1-104 Non 1E 1.2-32 Waste Processing Building 6551 Waste Gas Processing Area GM 2.5 1 Non 1E 12.3-11 10q.0 6552 Truck Loading Area GM 1.5 1 Non 1E 12.3-10 6553 Radwaste Control Room GM 0.5 10-t.104 1 Non 1E 12.3-10 6554 Waste Management Control Panel GM 5 1 Non 1E 12.3-10 10-1.10s Area 6570 Extruder/Evaporator Manifold Area GM (Note 6) 1 Non 1E 12.3-10 10°.0 6571 Compacted Rad Waste Storage Area GM 2.5 1 Non 1E 12.3-10

V5 - UFSAR Table 12.3-14 - Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Administration & Service Buildine 6555 Hot Chemistry Laboratory GM 0.5 10-1-104 1 Nonl1E 12.3-16 6556 Decontamination Room GM 0.5 10"l-104 1 NonlIE 12.3-16 6557 RCA Shlop (Note 2) GM 0.5 1"-0 1 Non 1E 12.3-16 6558 RCA Personnel Decontamination GM 0.5 10"I-104 1 Non lE 12.3-16 Area 6559 RCA Women's Locker Room GM 0.5 10-1-104 1 Non lE 12.3-16 (Note 1) 6535-A and 6535-B will automatically terminate containment purge in the event of high radiation during fuel handling operations.

(Note 2) RCA - Radiologically Controlled Area.

(Note 3) Deleted.

(Note 4) GM - Geiger-Mueller.

(Note 5) Radiation monitoring setpoints are varied during operation to follow station conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the set-points is found in the station operating procedures.

(Note 6) >100

V6 - UFSAR Section 9.1.4 Descriptiion of the Refueling Pathway JSTATJON  :'7;{"::[-i.".". ::;.i*:::z}:". .Fu:l el Storage and Handling .:.. .. ,.:]Sectionr 9.1.:ii.;.;.-*:..

9.1.4 Fuel H:andlina

. . System The Fuel Handling System (FHGS) consists of equipment and structures used for the refueling operation in a safe manner meeting General Design Criteria 61 and 62 of 10 CFR 50, Appendix A.

9.1.4.1 Design Bases

a. The primary design requirement of the equipment is reliability. A conservative design approach is used for all load-bearing parts. Where possible, components are used that have a proven record of reliable service. Throughout the design of equipment in containment, consideration is given to the fact that the equipment will spend long idle periods stored in an atmosphere of 120°F and high humidity.

b. Fuel handling devices have provisions to avoid dropping or jamming of fuel assemblies during transfer operation.

c. Handling equipment used to raise and lower spent fuel has a limited maximum lift height so that the minimum required depth of water shielding is maintained.

d. The Fuel Transfer System (FTS), where it penetrates the containment, has provisions to preserve the integrity of the containment pressure boundary.

e, Criticality during fuel handling operations is prevented by geometrically safe configuration of the fuel handling equipment.

f. Handling equipment will not fail in such a manner as to damage seismic Category 1 equipment or spent fuel in the event of a Safe Shutdown Earthquake.

g. Except as specified otherwise in this document, the crane structures are designed and fabricated in accordance with CMAA Specification No. 70 for Class A-i service.

h. The static design load for the refueling machine crane structure and all its lifting components is normal, dead and live loads, plus three times the fuel assembly weight with a Rod Cluster Control Assembly.

i. The original design allowable stresses for the refueling machine structures and components supporting a fuel assembly are as specified in the ASME Code, Section 111, Subarticle XVII-2200. Allowable stress criteria for rated loads for the spent fuel pool bridge and hoist, cask handling crane, and polar gantry crane are in accordance with CMvAA-70. Modifications to the refueling machine components/structures meet the allowable stress limits per the ATSC Manual of Steel Construction, 9th edition (Allowable Stress Design).

j. The design load on wire rope hoisting cables does not exceed 0.20 times the average breaking str'ength. Two cables are used in the refueling machine and each is assumed to carry one half the load.

V/6- UFSAR Section 9.1.4 Descriptiion of the Refueling Pathway SEABROOK.' A~ Revision 15 STATON Fel Sorag-andHandingSection91 J:.r.A].:, " P-,ge2:4

k. A single finger on the fuel gripper can support the weight of a fuel assembly and Rod Cluster Control Assembly without exceeding the requirements of Item i.

above.

1. All components critical to the operation of the equipment are located so that parts which can fall into the reactor are assembled with the fasteners positively restrained from loosening under vibration.

m. The inertial loads imparted to the fuel assemblies or core components during handling operations are less than the loads Which could cause damage.

n. Physical safety features are provided for personnel operating handling equipment.

Industrial codes and standards used in the design of the fuel handling equipment:

a. Applicable sections of CMAA Specifications No. 70.

b. New Fuel Elevator Hoist: Applicable Sections of HLMI-l00 and ANSI 1330.16.

c. Structural: ASME Code, Section 111, Appendix XVII, Subarticle XVT[.-2200 (Refueling Machine).

d. Electrical: Applicable standards and requirements of the National Electrical Code and NFPA No. 70 are used in the design, installation, and manufacturing of all electrical equipment.

e. Materials: Main load-bearing materials conform to the specifications of the ASTM standard.

f. Safety: OSHA Standards 29 CFR 1910, and 20 CFR 1926 including load testing requirements; the requirements of ANSI Ni18.2, Regulatory Guide 1.29 and General Design Criteria 61 and 62.

Protection of the FHS fr'om wind and tornado effects is discussed in Section 3.3. Flood protection is discussed in Section 3.4. Missile protection is discussed in Section 3.5.

9.1.4.2 System Description The Fuel Handling System (FHS) consists of the equipment needed for the refueling operation on the reactor core. iBasically this equipment is comprised of fuel assembly, core component and reactor component hoisting equipment, handling equipment and a Fuel Transfer System (FTS).

The structures associated with the fuel handling equipment are the refueling cavity, the refueling canal in Containment and in the FSB, and the fuel storage area.

The elevation and arrangements drawings of the fuel handling facilities are shown on Figure 1.2-15, Figure 1.2-16, Figure 1.2-17, Figure 1.2-18, Figure 1.2-19, Figure 1.2-20 and Figure 1.2-21.

V7.- Table HI Procedure References Supporting Plant Areas Identified in RA3/HA5 TARLE Hi - AREAS REOUIRING ACCESS FOR NORMAL OPERATIONS. COOLDOWN OR SHUTDOWN STRUCTURE OR ROOM AREA MODES APPLICABLE PROCEDURE STEP(s)

Primary Auxiliary Buiding 25 ft elevation 1, 2, 3, 4 051000.03, steps 2.3.8, 2.3.9, 7 ft elevation 4.29.4

-26 ft elevation 051000.04, steps 4.1.7, 4.1.11, 4.1.12, 4.1.13, 4.1.16, 4.1.21, 4.1.22, 4.1.23, 4.2.31, 4.2.8.1,

_______________________4.2.16, 4.3.10 Turbine Building 1, 2, 3 051000.03, steps 2.3.3, 2.3.6, 4.2, 4.8, 4.26, 4.27, 4.28, 4.47 051000.04, steps 4.1.6, 4.1.8, 4.1.9, 4.1.15, 4.1.24, 4.1.26, 4.3.45 Switchgear Rooms Essential 1, 2, 3, 4 051000.03, steps 2.3.3, 4.42, 4.41 Non-essential 051000.04, steps 4.1.5, 4.1.28, 4.2.18, 4.2.24 Steam and Feedwater Chases 1, 2, 3 051000.03, steps 2.3.5, 4.20 051000.04, steps 4.1.19, 4.3.29, 4.3.34 Waste Processing Building 25 ft elevation 1, 2, 3 051000.04, step 4.1.2 3 ft elevation

-31 ft elevation Containment 3, 4 0S1000.04, steps 4.1.14, 4.3.29, 4.3.30, 4.3.34 Equipment Vaults 3, 4 0S1000.04, steps 4.2.28, 4.3.16,

___ ____ ___ ___ __ ___ ___ ___ ___ ___ ___ ___ 4.3.30 Procedure references:

051000.03 - Plant Shutdown from Minimum Load to Hot Standby 051000.04 - Plant Cooldown from Hot Standby to Cold Shutdown

V8 - UFSAR Table 12.3-15 WRGM Ranges for RU1 TABLE 12.3-15 AIRBORNE RADIATION MONITORS (SKID MOUNTED DETECTORS)

(Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET LOOP TAG NO. DETECTOR REFERENCE GRD. LOW-HIGH POINT IEEE DIAG. P&ID DETECTOR ENERGY RE- DESCRLIPTION TYPE ISOTOPE mr/hr oCi/cc ttCi/cc OTY. CLASS LEVEL 1-NHY I-NHY 3 37 6526-i Containment Air Beta i' ' Cs' 2.5 Non 1E Note 3 506135 20504 Particulate (Note 4) 33 6526-2 Containment Radiogas Beta Xe' 2.5 i Non 1E Note 1 506135 20504 33 6528-i Plant Vent- WRGM Beta Xe' , I(.r5 2.5 I Non 1E Note 1 506607 20494 (Low Range) 33 6528-2 Plant Vent - WRGM Beta Xe' , Kr85 2.5 1 Non IE Note 1 506607 20494 (Mid Range) 33 3 6528-3 Plant Vent- WRGM Beta Xe' , Kr' 2.5 01-3_7,,r 10 1 Non 1E Note I 506607 20494 (Hi Range) 33 6495 WRGM Backup Ion Chamber Xe' , KI-8 2.5 1 Non 1E Note i 506607 RM-20509 33 506885 6531-2 WPB Radiogas Beta Xe' 2.5 10"7-10-* Non 1E Note I 20498 6532-2 PAB Radiogas Beta 10~-7.0* 506598 20510 2.5 Non lE Note 1 6548 Containment Radiogas Beta 2.5 10*-102 Non 1E Note 1 506136 RMv-20510 Backup

V8 - UFSAR Table 12.3-15 WRGM Ranges for RU1 SEAJIROOK RADIATION PROTECTION Revision: 13 STATION UFSAR TABLE 12.3-15 Sheet: 2 of 2 NOTES Predominant Gamma Energy Max. Beta Energy Note Isotopes 1 Xe13 3 0.346 0.08 1 Xe 13 5 0.92 0.249 Kr85 0.67 0.5 14 Kr85m 0.82 0.150 i131 0.606 0.3 64 i133 1.27 0.53 CS13 4 0.662 0.604 Cs 137 0.5 14 0.662 Co 58 0.474 0.81 Co 60 0.3 14 1.17, 1.33 Same as Note 2, plus:

Rb 88 5.3 1.863 4 Containment air particulate monitor functions as a leakage detector and must survive the SSE (reference Regulatory Guide 1.45 and Standard Review Plan 5.2.5).

5 Radiation monitoring setpoints are varied during operation to follow station operating conditions. Setpoints are maintained within the bounds established in the Technical Specifications.

The methodology for establishing the setpoints is found in the station operating procedures.

V/9 - OS 1000.09 Procedure References to Refueling Pathway Level Indicators for RU2 Figure 4: Monitoring RCS Inventory When in MODEs 5 or 6 (Sheet 2 of 2) 6.2 Changes in any of the above should be noted, as well as any changes in makeup needs. This will enable the operator to detect abnormal makeup once steady state conditions are obtained.

6.3 Changes in containment or fuel storage building pressures can result in sluicing between the two buildings and subsequent level changes. An approximate rule of thumb to determine sluicing is a change of 11/16" in the spent fuel pool level for every 1" change in refueling cavity level

(.690 gallons).

7.0 Preferred and alternate indications for monitoring primary system inventory are listed below.

PARAMETER PREFERRED CHANNEL ALTERNATE CHANNEL Pressurizer level A0390 (RC-LI-462) A0332 (RC-LI-45 9)1 A0333 (RC-LI-460)'

A0334 (RC-LI-461 )'

Reactor Vessel Level A03 82 (RC-LI-9405) TYGON 2 (Non-midloop) RC-XX-73 15-1 RC-XX-73 15-4 RC-LI- 131 1 RC-LI-1321 SHUTDOWN COOLING COLOR GRAPHIC Reactor Vessel Level A0298 (ULTRASONIC) A03 82 (RC-LI-9405)

(Mid-loop) A0299 (ULTRASONIC) TYGON 2 SHUTDOWN COOLING COLOR GRAPHIC VCT Level A0624 CS-LI-185 CS-LR-1 85 CS-LI-i 1122 Refueling Cavity Level A0382 (RC-LI-9405) A 1186 (SF-LI-2629)

(Cavity not full) A0390 (RC-LI-462)

Refueling Cavity Level A0928 (SF-LI-2629- 1) Al 1186 (SF-LI-2629)

(Cavity full) A03 90 (RC-LI-462)

Spent Fuel Pool Level Al1185 (SF-LI-2607) SF-LE-26073 RC-LI-459, 460 and 461 are not cold calibrated. Level changes will appear greater than actual at cold shutdown temperatures.

2 Local indications.

3 Holes on SF-LE-2607 are 3" apart, with the top 1/2" hole (about 1' below floor level) at elevation 25' 10 1/2" (relative to reactor vessel flange). Holes are 3" apart, with 1/2" holes every 1', and 1/4" holes every 3".

OS1000.09 Rev. 29 Page 42 of 48

VlO - UFSAR Table 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges TABLE 12.3-14 AREA RADIATION MONITORS (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Containment Structure 6534 In-Core Instrument Seal Table GM 15 10-'-10 4 INon IE 12.3-2 (Note 4) 6535AB Manipulator Crane GM 15 10'-104 1E 12.3-3 (Note 1) (Note 4) 6536-1, 2 Personnel Hatch (Post-LOCA) Ion Chamber 2.5 1+-0 Non 1E 12.3-3 2

6576A, B Containment (Post-LOCA) Ion Chamber 25 10°-10% /hr 2 IE 12.3-3 Primary Auxiliary Building 2

6537 Sampling Room GM 2.5 Non 1E 12.3-6 6538, 6539 RHR Pump Area GM >100 10-110 4 Non 1E 12.3-4 2

6540 Volume Control Tank Area Ion Chamber 8x10 4 101.107 Non lE 12.3-7 6541 Lower Level GM 2.5 10-lI04 Non 1E 12.3-5 6543 Entrance GM >100 Non IE 12.3-5 6544 Entrance GM 2.5 Non 1E 12.3 -6 10'-l.04 6545, 6546 Charging Pump Room GM 110 3 Non 1E 12.3-5 6547

VIO - UFSAR Table 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE 6508- 1,2 PAB-HRAM Ion Chamber >100 10"2-104 r/hr 2 Non 1E 12.3 -5 6563-1,2 PAB-HRAM Ion Chamber >100 10-2-104 rlhr 2 Non 1E 12.3 -5 65 17-1,2 RHR - Pump Vault HRAM Ion Chamber >100 I10~-2-10 r/hr 2 Non 1E Fuel Storage Building 6549 Spent Fuel Pool Area GM 2.5 10.-104o 1 Non lE 12.3-15 6518 Spent Fuel - HIRAM Ion Chamber 2.5 10-2.104 r/hr 1 Non 1E 12.3-15 Control Room 6550 Main Control Board Area GM 0.5 10-10 1Non 1E 1.2-3 2 Waste Processing Building 6551 Waste Gas Processing Area GM 2.5 10".0 Non 1E 12.3-11 6552 Truck Loading Area GM 1.5 Non IE 12.3-10 6553 Radwaste Control Room GM 0.5 Non 1E 12.3-10 1

6554 Waste Management Control Panel GM 5 10-1.104 Non 1E 12.3-10 Area 6570 Extruder/Evaporator Manifold Area GM (Note 6) 10o.105 Non 1E 12.3-10 6571 Compacted Rad Waste Storage Area GM 2.5 10~-l.04 Non 1E 12.3-10

VIGO - UFSAR Table 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE-- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Administration & Service Building 6555 Hot Chemistry Laboratory GM 0.5 10-1-104 1 Non 1E 12.3-16 6556 Decontamination Room GM 0.5 10"l104 1 Non IE 12.3-16 6557 RCA Shop (Note 2) GM 0.510I0 1 Non IE 12.3-16 6558 RCA Personnel Decontamination 0.5 10"1-104 1 GM Non IE 12.3-16 Area 6559 RCA Women's Locker Room GM 0.5 10-1-104 1 Non 1E 12.3-16 (Note 1) 6535-A and 6535-B will automatically terminate containment purge in the event of high radiation during fuel handling operations.

(Note 2) RCA - Radiologically Controlled Area.

(Note 3) Deleted.

(Note 4) GM - Geiger-Mueller.

(Note 5) Radiation monitoring setpoints are varied during operation to follow station conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the setpoints is found in the station operating procedures.

(Note 6) > 100

CL ~ ii~E!~

IE~ i2 !12! ~ 3 VllI- EP CALC 06-04 RVLIS Values for CG1/CS1I/CA1I 0

FPL- Energy Seabrook Station SEP# 2006021

SUBJECT:

Calculation of RVLIS Values for Emergency Action Levels FROM: D. L. Young DATE: May 8, 2006 TO" File Emergency Initiating Conditions CAl, CS1 and CG1 use Reactor Vessel Level Instrumentation System (RVLIS) readings in their associated Emergency Action Levels (EALs). These RVLIS values were calculated as follows.

Assumptions and Bases

1. These Initiating Conditions are evaluated in Mode 5. It was assumed that the accident was initiated in Mode 5, and that no prior accident had occurred in a higher Mode (i.e.,

there is no adverse containment).

2. Since containment conditions are normal, a RVLIS value without uncertainties was utilized.

3. Per Westinghouse Electric Corporation transmittal to Public Service Company of New Hampshire, Seabrook Station Units 1 and 2, Incore Thermocouples, NAH-3203 S.O. No.

NAH-4755/NAH-2.2.389, dated 01107/87 (film location, roll #715, frame #1081)

• From Attachment 2, a RVLIS full range indication of 66%, without uncertainties, is the mid:dle of the hot legs.

• From Attachment 2, a RVLIS full range indication of 55%, without uncertainties, is the top of the core.

• From Attachment 2, a RVLIS full range value of 35% equals 3.5 feet above the bottom of the core while a RVLIS full range value of 45% equals 8.5 feet above the bottom of the core. Therefore the relationship between RVLIS full range scaling and height in the reactor vessel is 2% per foot.

4. Per Procedure OS1000..12, Operation With RCS At Reduced Inventory/Midloop Conditions

• The radius of the hot legs (i.e., center to top/bottom) is 14.5".

• -71 inches equals the top of the hot leg nozzles.

• -85.5 inches equals the hot leg midplane

• -100 inches equals the bottom of the hot leg nozzles.

I II II II I IIII I III I II I II I I ill l

VII - EP CALC 06-04 RVLIS Values for CG1/CS1/CAI Calculation of RVLIS Values for Emergency Action Levels Page 2 Calculations

1. The RVLIS full range value at the bottom of th hot Jeaqs is 66% - (14.5"/12"ift)

• 2%/ft =

63.6%. This value is rounded to 64%.

2. The RVLIS full range value at 6" below t~he bottom of the hot leecis is 66% -((14.5" +

0 6"Y1 2"/ft)

• 2 0/ ft = 62.6%. This value is rounded to 63%.

3. As stated previously, the RVLUS full range value at the ton of the core is 55%.

The above calculation is EPCALC-06-04.

Prepared by:. _______________

David L Young Senior Nuclear Analyst Approved by: *J" Robert L Couture Principal Engineer - Nuclear DLY/dly cc:

IIIIIIII I I II II

Vll - EP CALC 06-04 RVLIS Values for CG1ICS1ICA1 RVLIS LEVEL VESSEL LEVEL

(%) (inches from vessel flange)

-108 119.8" 100 81.3" 90 31.8" 80 -17.7" 70 -67.2"

• 64 -96.9"

• 63 -101.9"

• 60 -116.7"

• 55 -141.5" RC-LI-9405, RC-LIT-9467, and

• 50 -166.2" the Tygon Tube do not indicate reactor vessel level when actual

• 40 -215.7" level is less than -95" due to the 30 -265.2" weir on the RCP discharge.

' 20 -314.7"

• 10 -364.2"

• 0 -413.7"

V12 - Manipulator Crane Monitor Ranges for CG1/CS1 TABLE 12.3-14 AREA RADIATION MONITORS (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Containment Structure 6534 In-Core Instrument Seal Table GM 15 Non IE 12.3 -2 (Note 4) 10'-1.04 6535A, B Manipulator Crane GM 15 1E 12.3-3 2

(Note 1) (Note 4) 6536-1, 2 Personnel Hatch (Post-LOCA) Ion Chamber 2Non IE 12.3-3 Containment (Post-LOCA) Ion Chamber 25 100 -10s /hr 12.3-3 6576A, B 1E Primary Auxiliary Building 6537 Sampling Room GM 2.5 2 Non 1E 12.3 -6 6538, 6539 RHR Pump Area GM >100 Non 1E 12.3-4 6540 Volume Control Tank Area Ion Chamber 8X10 4 101.107 2 Non IE 12.3-7 6541 Lower Level GM 2.5 10-.0 1 Non 1E 12.3-5 4

6543 Entrance GM >100 10-1-10 Non 1E 12.3-5 1

6544 Entrance GM 2.5 10-1.104 Non IE 12.3-6 6545, 6546 Charging Pump Room GM 110 t 0-1.10 Non 1E 12.3-5 6547

V12 - Manipulator Crane Monitor Ranges for CG1/CS1 (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE..._- TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE DESCRIPTION 6508-1,2 PAB-HRAM Ion Chamber >100 10"z- 104 r/hr 2 Non 1E 12.3-5 6563-1,2 PAB-HRAM Ion Chamber >100 10"2-104 r/hr 2 Non IE 12.3-5 6517-1,2 RHR - Pump Vault HRAM Ion Chamber >100 10-2-104 r/hr 2 Non 1E Fuel Storage Building 6549 Spent Fuel Pool Area GM 2.5 1 Non 1E 12.3-15 6518 Spent Fuel - HRAM Ion Chamber 2.5 102-. 10 4 r/hr 1 Non 1E 12.3-15 Control Room 6550 Main Control Board Area GM 0.5 1--0 Non 1E 1.2-32 Waste Processing Building 6551 Waste Gas Processing Area GM 2.5 10-1.104 Non IE 12.3-Il 6552 Truck Loading Area GM 1.5 10-1.104 Non 1E 12.3-10 6553 Radwaste Control Room GM 0.5 10'-1.04 Non 1E 12.3-10 6554 Waste Management Control Panel GM 5 10'-l.04 Non 1E 12.3-10 Area 6570 Extruder/Evaporator Manifold Area GM (Note 6) 10°.105 Non 1E 12.3-10 6571 Compacted Rad Waste Storage Area GM 2.5 10'-l.04 Non 1E 12.3-10

V12 - Manipulator Crane Monitor Ranges for CG1/CS1 (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE_..- DESCRIPTION TYPE mr/hr mr/hr mr/hr CLASS REFERENCE Administration & Service Building 6555 Hot Chemistry Laboratory GM 0.5 10-1.104 Non 1E 12.3-16 1

6556 Decontamination Room GM 0.5 10'-i.10 Non 1E 12.3-16 6557 RCA Shop (Note 2) GM 0.5 10'-.104 1 Non 1E 12.3-16 6558 RCA Personnel Decontamination GM 0.5 10'-l.04 Non 1E 12.3-16 Area 6559 RCA Women's Locker Room GM 0.5 Non 1E 12.3-16 (Note 1) 6535-A and 6535-B will automatically terminate containment purge in the event of high radiation during fuel handling operations.

(Note 2) RCA - Radiologically Controlled Area.

(Note 3) Deleted.

(Note 4) GM - Geiger-Mueller.

(Note 5) Radiation monitoring setpoints are varied during operation to follow station conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the setpoints is found in the station operating procedures.

(Note 6) > 100

V13 - UFSAR Section 6.2 Description of Containment Sumps for CG1/CS 1/CA1/ICU1 SEABROOK ENGINEERED SAFETY FEATURES Revision 15 STATION Containment Systems Section 6.2 UFSAR Page 51

g. Net Positive Suction Head Reqiuirements Adequate net positive suction head is assured by locating the RH-R and CBS pumps at the lowest level in the Auxiliary Building. The RHR and CBS pump available net positive suction heads from the containment sump were determined by assuming the limiting conditions in accordance with NRC Regulatory Guide 1.1 (pressure equal to atmospheric and temperature equal to 21 2°F). The CBS pump available and required net positive suction head is shown in Table 6.2-75.

The RHR pump available and required net positive suction head is shown in Table 6.3-1.

h. Heat Exchanger Surface Fouling The materials used for the CBS heat exchanger are listed in Table 6.1(B)-I. The shell side of the heat exchanger is cooled by the PCCW system which contains a corrosion-inhibiting agent and operates as a closed system. The tube side, which is in contact with the emergency core coolant, is corrosion resistant. The effect of corrosion fouling on heat exchanger surfaces will, therefore, be minimal; however, fouling factors were included in the detailed design of the units to assure the required heat removal capability through conservative design.

i. Heat Exchanger Performance The heat exchanger (residual heat removal, containment spray, primary component cooling water) temperatures have been selected based upon maximum service water (ultimate heat sink) temperatures and the amount of heat removal required. The flows, geometry, and surface area were studied in the evaluation of the containment pressure-temperature analysis. Those parameters selected and tabulated in Table 6.2-76 are those which meet the design basis requirement for containment cooling.

j. Containment Recirculation Sump and Strainer Design The containment recirculation sump collects and strains the water available for supplying the residual heat removal, containment spray, safety injection and high head charging pumps during the recirculation mode of operation following an accident. The sump is designed to meet the intent of Regulatory Guide 1.82.

Two completely independent sumps are located in the containment to maintain the "double train" concept as described in Subsection 6.2.2.2d.

One sump supplies water to Train A and the other sump supplies Train B. The arrangement of these sumps is shown in Figure 6.2-79. The minimum water level in containment during a loss-of-coolant accident is nominally Elevation (-)23.79 ft.

V13 - UFSAR Section 6.2 Description of Containment Sumps for CG1ICS1ICA 1/CU1I SEABROOK ENGINEERED SAFETY FEATURES IRevision 15 STATION Containment Systems Section 6.2 UFSAR jPage 52 A series of debris interceptors are provided on the containment floor within the recirculation flow paths. The debris interceptors reduce the quantity of debris transported to the sumps by trapping debris and allowing the remaining debris more time to settle prior to reaching the sumps.

Heavy particles are prevented from reaching the sumps by sloping the surrounding floor away from the sumps. This facilitates settling of debris on the floor prior to reaching the sump area.

A strainer is installed in each sump. Each strainer consists of rows of vertically oriented strainer panels, consisting of a framework sandwiched between two sets of wire cloth attached to perforated plates. The maximum hole size of the perforated plates and maximum width of a gap between bolted structures is 0.068 inches. The strainer would therefore prevent debris particles 0.068 inches or greater in diameter which may be generated following a large break LOCA from passing through or bypassing the strainer and entering the ECCS system. The minimum physical restriction in the ECCS flow path consists of 0.073 inches, which is the effective opening of the fuel assembly debris filter bottom nozzle in combination with the P-grid. Therefore, the strainer will prevent recirculation of debris particles of sufficient size to impede cooling flow to the core.

The strainer panels are mounted on a plenum structure within the sump. The plenum is sealed to the sump floor and at the sump wall adjacent to the ECCS pipe inlet to ensure that all water entering the sump passes through the strainer panels. Water is drawn through the strainer panels and plenum and into the lower portion of the sump.

The strainer will also act as a vortex preventor to further preclude air intrusion into the ECCS piping.

The strainers have been designed to accommodate the debris generated and transported to the sump during the recirculation phase of a LOCA. The head loss due to debris on the strainer is less than the available NPSH margins for operating ECCS pumps, thereby ensuring that cavitation of the ECCS pumps will not occur.

Therefore, the design meets the intent of Regulatory Guide 1.82.

The potential for clogging of the sump strainers by equipment and piping insulation or loose insulation in the containment is minimized by the type of insulation used.

V13 - UFSAR Section 6.2 Description of Containment Sumps for CG1/CS 1/CA 1ICU1 SEABROOK ENGINEERED SAFETY FEATURES Revision 15 STATION Containment Systems Section 6.2 UFSAR Page 53 The thermal insulation inside the containment for piping and equipment except the reactor pressure vessel is fiberglass blanket insulation of the type commercially known as Nukon, manufactured by Owen's-Coming Fiberglass. The outside surface of the insulation blankets is covered with a stainless-steel jacket or is encapsulated in stainless steel wire mesh. Nukon is consistent with the recommendations of Regulatory Guide 1.36. The reactor pressure vessel is insulated with stainless-steel reflective insulation or fiberglass blanket.

Clogging of the strainers by nonsafety-related equipment is unlikely due to the remote location of the sumps relative to the NNS equipment and physical barriers separating the sumps from other areas in the containment. The supplementary neutron shielding around the reactor vessel which could be displaced by blowdown forces during an accident is designed to remain anchored and intact; hence, it is not a potential source of strainer blockage during an accident.

The design of the sump suction piping ensures that adequate flow and net positive suction head are available to all pumps under the most limiting containment conditions, as required by Regulatory Guide 1.1. The two sumps and the pumps they service are designed so that any single active or passive failure will not cause the loss of both A and B Train components.

The sumps are visually inspected on a periodic basis to assure that they are clean, free of debris and that all strainers are intact and in position. The containment sump line isolation valves are exercised periodically to assure operability within Technical Specification requirements.

k. Periodic Testing The provisions for periodic testing and inspection of the containment spray system are discussed in Subsection 6.2.2.4.

Applicable Codes, Standards and Guides The codes, standards and guides applicable to the containment spray system are summarized in Table 6.2-79.

m. Remote Manual Operation of the Containment Building Spray System The CBS system is designed to function completely automatically under accident conditions, hence there are no operations which must be performed manually by the operator from the main control board to initiate the proper function of the system during an accident. After the suctions of the CBS pumps are automatically switched over from the RWST to the containment recirculation sumps, the isolation valves (CBS-V2, CBS-V5) in the discharge line from the RWST will be closed by the operator.

V14 - SAMG Calculation Aid CA-3 for Hydrogen Flammability for CG1.2clContainment Potential Loss Number Title Revision CA-3 HYDROGEN FLAMMABILITY IN CONTAINMENT Rev. 0 Purpose The purpose of CA-3 is:

• Define whether the hydrogen in the containment atmosphere is flammable, and

• Estimate the hydrogen concentration in the containment atmosphere based on an estimated oxidation percentage.

Assumptions

• This aid is valid for scenarios where the containment fans are on or off.

• Hydrogen released into containment is created with the reactor core through zirconium oxidation.

• 6% hydrogen concentration by volume is defined as the minimum concentration that will sustain a global burn.

• The hydrogen burn is based on a simplified adiabatic isochoric complete combustion for hydrogen.

• No hydrogen ignitors are considered and no previous bumns are assumed to have occurred.

• All energy released from the bumn is absorbed by the containment atmosphere as increased internal energy.

• The aid assumes the air, steam, and hydrogen within the containment atmosphere are thoroughly mixed at 100% humidity and all behave as ideal gases.

• The air volume is based on initial conditions, plus 10% to account for instrument air leakage, and remains constant throughout each scenario.

• During the core/concrete scenario, the core is assumed to be ex-vessel and the containment atmosphere is superheated.

• During the core/concrete scenario, no additional noncondensable, noncombustible gases are released while additional hydrogen is released.

• For the venting scenarios, the air, steam, and hydrogen are released in the same ratio as they exist in containment when venting takes place.

• When combustion occurs, all hydrogen is assumed to be consumed.

• Expected containment failure has been defined at the pressure at which there is a 5%

probability of containment failure, minus 10 psi.

Additional Data The figures displayed in this aid are for a wet hydrogen measurement only. Wet measurement is chosen because the post accident sampling is performed with steam in the sample. For sampling cases above 300 °F, or cases where heat tracing is lost, condensation in the analyzer will occur and conditions will approach that of a dry measurement where a higher concentration is read resulting in the appearance of a more limiting condition.

Page 2 of 8

V15 - Tech Spec Table 1.2 Cold Shutdown Temp Limit for CU3/CA3.1I TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

SFCP In accordance with the Surveillance Frequency Control Program TABLE 1.2 OPERATIONAL MODES REACTIVITY  % RATED AVERAGE COOLANT MODE CONDITION~keff THERMAL POWER* TEMPERATURE

1. POWER OPERATION > 0.99 > 5% > 350°F

2. STARTUP > 0.99 _<5% > 350°F

3. HOT STANDBY < 0.99 0 > 35 0oF

4. HOT SHUTDOWN < 0.99 0 3500 F > Tavg >2000°F

5. COLD SHUTDOWN < 0.99 0 < 2000 F

6. REFUELING** < 0.95 0 < 140°F

• Excluding decay heat.

• • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

SEABROOK - UNIT 11-Amnet14 1-8 Amendment 141

Vlti - i-roceciure U:57000.CMJ De~im~tIon 0? Kecluced Inventory tar CA3 Figure 3: Shutdown and Refueling Related Definitions (Sheet 1 of 2)

1. CORE ALTERATIONS A CORE ALTERATION is defined in Technical Specifications as definition 1.9.

Conservation decision making during refueling equipment problems must be used to ensure Technical Specification compliance is maintained.

OE34388 - Non-conservative Implementation of a CORE ALTERATION During core offload activities, a fuel handling equipment failure resulted in a fuel assembly being suspended, motionless in the core region for several days time. The station initially concluded a CORE ALTERATION was not in progress based on no assembly movement and secured the SRO position providing oversight of fuel handling activities. Subsequently, the station determined the subject condition should have been treated as a CORE ALTERATION since the fuel assembly remained in the core region and the subject move sheet had been started but not completed. In this case, SRO oversight of the bundle should have been maintained per 10CFR50.54.

2. RCS Loops Filled Technical Specifications T.S. 3.4.1.4.1 and T.S. 3.4.1.4.2 contain different restrictions in MODE 5 depending on RCS loops filled or RCS loops not filled. See Technical Specification Bases 3/4.4.1, Reactor Coolant Loops and Coolant Circulation for all conditions needed to declare the RCS "Loops filled."

3. Reactivity Additions Transferring water to the RCS, refueling cavity, refueling canal, transfer canal, or spent fuel pool that is lower in boron concentration is acceptable provided that the boron concentration is greater than the refueling boron concentration requirement. Likewise, transferring water to the RCS, refueling cavity, refueling canal, transfer canal, or spent fuel pool that is lower in temperature (down to 50°F) than the water contained in those volumes is also acceptable.

4. RCS Reduced Inventory Condition As discussed in NRC Generic Letter 88-17, an RCS reduced inventory condition exists whenever the RCS water level is lower than negative 36 inches (as measured from the reactor vessel flange).

OS1000.09 Rev. 29 Page 39 of 48

V17 - Site Specific EAL Design Basis for 25 PSIG RCS Pressure Range for CA3.2 Per the EAL design Basis, The wide-range RCS pressure transmitters have *arange of 0 to 3,000 psig. The main control boards have two post-accident monitoring qualified meters, one for each wide-range RCS pressure transmitter. These meters have major divisions at 100 psig intervals and minor divisions at 50 psig intervals. Since it is possible to read the approximate mid-point between minor divisions, the value is set to 25 psig.

V18 - UFSAR Section 8.3.2 - DCV 105 Limit for MGS/MS8/CU4 ISTATION. .5;*~{:**:i{:.. - {* i' On{:i:

site Power Systems .:;:;i<>{-g"*;{ .*!:- ::: ISection 8.3 2:,

UF-A .:::*}i<::t * .:ii.. -'... -**:;:::!!i.t*-::j" Page 68#!{}.::.*!;:

8.3.2 DC Power System 8.3.2.1 Description The station DC power system is comprised of the battery chargers, station batteries and 125V Distribution System. It provides the sources of power for direct current load groups, vital control and instrumentation systems, and control and operation of Class 1E and non-Class I1E electrical equipment. It is a two-wire ungrounded system.

The battery chargers (rectifiers) provide the normal steady-state DC power; the station batteries provide for normal transient loads and also act as the reserve source upon failure of the rectifier or the AC supply to it. Figure 8.3-2, Figure 8.3-37 and Figure 8.3-3 8 present the one-line diagrams for the station DC electric power system, and show the connections to the AC Vital Instrumentation and Control Power System.

The safety-related portion of the station DC power system shown on Figure 8.3-37 consists of four 125-volt batteries, chargers and DC buses. The loads supplied from the buses include inverters for redundant vital instrument buses, distribution panels for power to the Class lE direct current loads, power for control and operation of the Class lB systems for Engineered Safety Features, and power for selected non-Class lB loads.

Each DC bus consists of metal-enclosed 125-volt DC switchgear consisting of vertical sections housing buses, circuit breakers, instruments and accessory equipment. The breakers are low voltage manual power circuit breakers. Figure 8.3-37 shows that the safety-related DC system incorporates mechanically interlocked manual circuit breakers which will permit the connection of two DC supply buses within the same train to a single battery, but prevents paralleling the two batteries in the train.

The nonsafety-related portion of the station DC systems shown on Figure 8.3-38 consists of two 125-volt batteries, chargers and DC buses. The loads supplied fr'om the buses include inverters for the computer and auxiliary power panels feeding nonvital equipment requiring constant supply and control power feeders to nonvital equipment (13 .8-ky switchgear, turbine generator emergency oil pump, etc.).

a. System Separation, Ventilation and Redundancy Four safety-related 125-volt batteries are supplied. Each battery is housed in an individual room in the seismic Category I Control Building. Separate ventilating systems are provided for the battery rooms of each train (see Subsection 9.4.10).

The batteries are seismically qualified and are mounted on seismic Category I racks. The safety-related battery chargers and DC buses are also seismically qualified.

V18 - UFSAR Section 8.3.2 -. DCV 105 Limit for MG8/MS8/CU4

S"I:'AB:<**doK<*: FL':.CTR**IC POWER

.<:"..J~":"( . ".-:!~* !;i§Revision 16.:i;.:.i.)

Each battery has its own charger and DC bus. The battery chargers and DC buses of each train are located in an area adjacent to their associated battery rooms and are physically separated from the char gers and buses associated with the redundant train (see Figure 8.3-27 and Figuie 8.3-36).

Four DC supplies are provided for the four NSSS inverteis for vital instrument buses and the power and control requirements of the two .engineered safety features trains (see Subsection 8.3,.1~d). Equipment is located and cables are routed in a manner to assure continued independence and separation so that the loss of DC supply to either train does not prevent the minimum safety function of the other train from being performed.

One nonsafety-related inverter for the station computer is powered from the Train A DC system through a Class lE breaker on Bus lilC. One nonsafety-related DC power panel is powered from the Train B DC system through a subfeed from safety-related DC power panel PP-i111B. All remaining nonsafety-related loads (DC motors, other nonsafety-related inverters, nonvital control panels) are connected to the nonsafety-related batteries (Figure 8.3:.38).

b. Station Battery Capacity The safety-related station batteries are lead-calcium, power station type. Each battery consists of 59 cells, and has a nominal 8-hour rating of 2280-ampere hours.

Each safety-related battery is sized to supply its safety-related and nonsafety-related loads for the durations indicated in Table 8.3-5. Battery B-1C is capable of providing power to the nonvital computer inverter, 1-2A, for 15 minutes while supplying its safety-related loads; the inverter load is automatically disconnected from the DC system after the 15-minute period. This disconnection is accomplished by a safety-related trip circuit on the Class lE breaker feeding inverter I-2A. This cir'cuit, which monitors the time the inverter draws power from the battery, is testable.

Tn addition, each safety-related battery is sized to have sufficient capacity to serve as the source, for the duration indicated in Table 8.3-5, for two load groups of the same train during the period when one battery is out of service (see Figure 8.3-37). Figure 8.3-51 shows the separate and combined load profiles for the safety-related batteries.

The safety-related station batteries also have sufficient capacity for the four-hou" Station Blackout coping duration. The Station Blackout battery sizing evaluation includes the one battery/two bus configuration (see Section 8.4.4.2).

V18 - UFSAR Section 8.3.2 - DCV 105 Limit far MG8IMSBICU4

• STATION Onsite. Power Systems Section 8.3it There aic two nonsafety-related batteries (13-2A and B-2B) provided in the Turbine Building. The nonsafety-related station baffezies are lead calcium, power station type, consisting of 59 cells. Battery B-2A supplies various DC motors for the turbine auxiliaries, various control panels and the Turbine Building DC lighting. Battery B-2B supplies the computer inverter T-2B, the nonvital instrument inverter 1-4, Control Building DC lighting and various control panels.

Each Class lE battery was sized in accordance with the recommended practices in IE:EE Standard 485-1978. These practices were applied as follows:

1. The system maximum voltage (140 volts) and maximum equalizing cell voltage (2.33V per cell) were selected. This resulted in a selection of 59 cells which include margin between the equalizing voltage (137.5 V),

and the system's maximum voltage.

2. A duty cycle diagram was developed, based upon the combined known and anticipated loads for both DC buses of the same train (see Figure 8.3-5 1).

3. The battery capacity data were selected from the manufacturer's data, based upon the minimum cell voltage (1.78V per cell permitted by the system minimum voltage of 105V).

4. The calculated minimum required cell size was increased by 25 percent for end-of-life compensation.

5. Temperature correction factors were applied to the calculated minimum required cell size, to allow for operation at the minimum design temperature (65°F for batteries B-lA and B-1C, and 60°F for batteries B-lB and B3-1D). These temperatures also apply to Station Blackout.

6. Sizing calculations were performed using methods similar to Figure 3 of IEEE 485-1978 to determine the minimum required cell size.

7. A minimum design mar'gin of 15 percent was included in the original battery purchase specification calculated cell size to allow capacity for Tuture loads.

V19 - NUHOMS HSM Technical Specification Dose Rates for EAL EU1 HSM-H Dose Rate Evaluation Program 5.4 5.4 HSM-H Dose Rate Evaluation Program This program provides a means to help ensure that the cask (DSC) is loaded properly and that the facility will meet the off-site dose requirements of 72.104(a).

1. As part of its evaluation pursuant to 10 CFR 72.212, the licensee shall perform an analysis to confirm that the limits of 10 CFR Part 20 and 10 CER 72.104 will be satisfied under the actual site conditions and configurations considering the planned number of HSMs to be used and the planned fuel loading conditions.

2. On the basis of the analysis in TS 5.4.1, the licensee shall establish a set of HSM-H dose rate limits which are to be applied to 32PTH DSCs used at the site. Limits shall establish peak dose rates for:

a. HSM-H front surface,

b. HSM-H door centerline, and

c. End shield wall exterior.

3. Notwithstanding the limits established in TS 5.4.2, the dose rate limits may not exceed the following values as calculated for a content of design basis fuel as follows:

a. 800 mremlhr at the front bird screen,

b. 2 mremlhr at the door centerline, and

c. 2 mrelr at the end shield wall exterior.

4. If the measured dose rates do not meet the limits of TS 5.4.2 or TS 5.4.3, whichever are lower, the licensee shall take the following actions:

a. Notify the U.S. Nuclear Regulatory Commission (Director of the Office of Nuclear Material Safety and Safeguards) within 30 days,

b. Administratively verify that the correct fuel was loaded,

c. Ensure proper installation of the HSM-H door,

d. Ensure that the DSC is properly positioned on the support rails, and

e. Perform an analysis to determine that placement of the as-loaded DSC at the ISFSI will not cause the ISFSI to exceed the radiation exposure limits of 10 CER Part 20 and 72 and/or provide additional shielding to assure exposure limits are not exceeded.

NUHOMS HD System Technical Specifications5- 5-9

V20 - Site Specific CSFST Care Coaling far MG1/MS5/Fuel Clad and CantainmentPatentialLass Number Title Revision F-0.2 CORE COOLING (C) 20 (RED) GO TO FR-C.1 GO TO FR-C.1 GO TO FR-C.2 GO TO FR-C.2 GO TO FR-C.3 GO TO FR-C.2

• • .*

• GO TO FR-C.3

C)CSF SAT G:\Word\lmagesP\lmagesP.OP\F02.ds4 1 of 1

INumber Title F-0.4 V21 - Site Specific CSFST Integrity INTEGRITY (P)

IRevision 21 (FIGURE 1)

W 4In 1W o

T1 T2 COLD LEG TEMPERATURE GO TO FR-P.1

-mmI-------- (ORANGE) GO T GOFR.P.1T N (* GO TO (YELLOW)@ w@*

• FR-P.2

~'ES I ,

IDECREASE IN NO (GREEN) L

~ )SAT CSF ENTER D-. ALL COLD LEGS TEMPERATURE LESS THAN 1000 F GOTO IN THE LAST YE SIXTY MINUTES (ORANGE) E1-l -ol FR-P.1 ATURES R THAN YE

• GO TO (YELLOW)0 0@* FR-P.2 FL (GREEN)

O SAT CSF (GREEN)

G:\Word~lmagesP~mages_.P.OP\FO4.ds4 FL O SAT CSF 1 of 3

V21 - Site Specific CSFST Integrity F-0.4 INTEGRITY (P) 21 FIGURE 1 2500 2800 2560 2500 226°F/

2050PS 2000 (5

U) 03.

uJ 1500 rDr w) 1000 n) 500 RED ORANGE -- GREEN

-J 0

4 - 1 100 194°/ 200 300 400 T1 T2 RCS COLD LEG TEMPERATURE (OF)

F04-1 2 of 3

V21 - Site Specific CSFST Integrity Number Title Revision F-0.4 INTEGRITY (P) 21 FIGURE 2 LTOP PRESSURE LIMITS Reactor Coolant System Pressure vs Temperature Curve 2800 Max. PORV i Setpoint 2385 PSIG ':

2600 I 1600 PSID 2400 (wiRtC SP-zSRG Liub itl j* "*/

2200 2000 1800 1 II LTOP Pr~Q'~I irn I I I I!1 I I 1I 1I 1 1 1rl-* l IJ l I t I I I I I /] I I ,I I l I II I i51600 Li**t711-iJ i ii i i i*lq i ii i iJ"i ii i i ii i ii:i i i i*"i i iii Yi ii ;ii S1400

. .. . .. . .. . [...i-... . .. . .. .. . .. . ... .... ... -

2000F Cs~1200 For RCP Operation

• L 100°F 1000 (Admin. 325 PS*

800 LTOP Min.

Setpoint 426 PSIG Lsub ooio0 50OF Subcooling 6O00 RCP Required NPSH 400 200 0

0 50 100 150 200 250 300 350 400 450 500 550 600 650 RCS Temperature (F) 3 of 3

V22 - Site Specific CSFST Heat Sink for MS5IRCS PotentialLoss SNumber Title Revision F-O.3 HEAT SINK (H) 21 GO TO FR-H.1 0

(YELLOW) (1h KY***** GO TO FR-H.2 ENTER--I (YELLOW) (

S GO TO

• ******' FR-H.3 (YELLOW) ,*GO TO 0 ** FR-H.4 GO TO FR-H.5 G:\Word'lmagesP\lmagesP.O P\F03.ds4 K__*Q (GREEN)

SAT CSF 1 of 1

V23 - EPCALC- 06-01 Rad Values for Fission ProductBarrierLoss and PotentialLoss FPL Energy Seabrook Station SEP# 2006012

SUBJECT:

Calculation of Containment Dose Rate Values Used in FPB EALs FROM: D. L. Young DATE: March 28, 2006 TO: File Attached are the calculations forthe containment dose rate values used in the Fission Product Barrier (FPB) Emergency Action Levels (EALs). The calculation number is EPCALC-06.01.

Senior Nuw DLY/dly cc: W. Cash

• ' I I II II " .... ... I IIII I I 'I I' I I III

Calculation of Fission Product Barrier Matrix EPCALC-06-O1 EAL Values for Containment Dose Rates Containment BarrierPotential Loss - Containment Ra*diation-Monitor Reading 'm ..

ra,,

NEIE 99-01. Rev. 4, Basis: Unless there is a (site-specific) analysis justifying a higher value, it is re*:ommended "

that a radiation monitor reading corresponding to 20% fuel clad damage be specified here. '-

Calculation Basis: SBC-1023, Update of METPAC Data Affected by Power Uprate and AST Implementation, O The calculation was perforated using an assumed value of 20% fuel clad failure.

Time NG Core NG NG Cone uCi-hr Men tinventory (assumed) I .

(Note 1) (Note 2) (Note 3) (Note 4) (Note 5) "fuel clad damager for EAL l CD

.... o....................................... ......... ..... ........................

2 3,622+08 1.002-02 4,732+01 4.712-02 "1,003 -

4 5 3,27E+08 1.00E-02 4.262+01 8,93E-02 615 Free volume of conta~nment 2,702÷06 ft per UJFSAR Sec. 15B,2,11,A 10 2.982+08 1,00OE-02 3.882+01 1 ,05B.-01 389 Free volum~e of containment 7.066E+10 ml [2.7E+06 *28315.846592 ml/ft3j 20 2.53E+08 1.002-02 3.3'tE+0t 1,552-01 212 F~ree volume of containment 7,602+10 cc per UFSAR Sec. iSB.2.Ul.A

• Free volume of containment 7,665+04 mu per UFSAR Sec. 15B.2,ll.A5 Note I -Values taken from SBC-1023, Table 3,10"1 Note 2-5%

• 2O% 1%

Note 3 - [(Core NG

• 12+6 uCllCi) *.01*]/7.662+10 cc .

Nlote 4.. Values taken from SBCo1 023, Tables 4.5 and 4.6 Nlote 5 - Cont NG ConcI'J(t) '

0)3 m'-

oD

-I

-o

_ _ a Hey. 0 Page1 3/2412006 IIII I I II I II I I III II I I ,ill i ,i ,, , ,,,,,,,,

Calculation of Fission Product Barrier Matrix EPCALC-06-01 EAL Values for Containment Dose Rates CA~

iRCe:Barrioi Loss - Contalnment Radiation Monitor Reading -o NEt 99-01, Rev. 4, Basis: The reading should be calculated assuming the instantaneous release and dispersal 0 of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e.. I-wvithin T/S) into the containment atmosphere.

Oatculatlon Basis: SBC-t023, Update of METPAC Vata Affected by Power Uprate and AST Implementation. 9 The calculation was performed using RCS actlvity valufes at Technical Specification limits.

0 Decay RCS ROB Cnmt J(t) Cnmt Volume of water inRiCS 1.152+04 UFSAR Table 5.1-i*

Time NG Cone NG NG Cons uCI-hi Man Gallons of watter In RCS 8,622+04 gal [12100 ft3

• 7.48 gal/ft3]

(his) (uCilgm) (uCI) uCflco cc-R Rlhr lbs of waler In RCS 7,t95+05 lbs 190505 gal

• 8.345 lbs/gal]

(Note 1) (Note 1) (Note 2) (Note 3) (Note 4) (Note 5) gins or cc of waler In RCS 3,275+08 gin [755269 lbs *454 gnislIb] 0) 0.5 1.51tE÷02 4.95E+t0 l8o47E-01 3.305.032 20 CD 1 1.465+02 4,788+10 6,245-01 3,DIE-02 16 Free volume of conlalnrnnett 2.70E+00 ftJ per UFSAR Sac, 15B.2,11,A C,)

2 1.395+02 4.585+10 5,94E-0I 4,715.02 13 Froe volume of contulnmenl 7.665÷10 ml [2,75+05

• 283'16,846592 ml/ftS] 0 5 1.265+02 4,';3E+l0 5,395-01 8,935-02 8 Free volume of containinont 7,1305+10 cc 10 1,152÷02 3,70E÷10 4,915-01 1.05E-01 5 Free volunio of containment 7.600E04 per UFSAR Sec. 15B.2.I.,A C,)

20 1,035+02 3.385+10 4.42E-01 1.5135-01 3 C,)

Note I - Values ftom 850-1 023, Table 3.2 and 3.9, 0 Also see UFSAR Table 150..2 Note 2. ((uCi/gm)*3.272+8 gin) -a Note 3 - (u~i)/7.60E+10 cc 0 Nlote4- Values taken from SBC-1023, Tables 4,5 and 4.6 Nlote 5- Cent NG Conc/J(t) w 0)

CD I-.

0 Co C,)

0) 0 e.g.

CD 0) 0 C,)

C,)

R*. 0 Page2 Rev, 0Page 23/24/2001 O

I i il Iml i i Ill I i III I IIII II I

11111 I Calculation of Fission PrOduct Barrier Matrix EPCALC-tJ6-01 EAL Values for Containment Dose Rates 3~l Fuel Clad Barrior: Loss - Containment Radiation Monitor Reading M21 99-01, Rev. 4, Basis: The reading should be caicu~ated assuming the instantaneous release and dispersal of

'he reactor coolant noble gas and iodine inventory associated with a concentration of 300 uCi/gm dose equivalent 1-131 into the containment atmosphere.

3alouiatron Basis: $BC-1023, Update of METPAC Data Affected by Power Uprate and AST Implementation. The atculatlon was performed using ROS activity' values associated wilh a concentration of 300 uCi/gm dose 6 7~4

• quivalenl 1-131.

Total Ci of DE1-131 in ROS=

f300 uCi/cc *3.27E+8 gml~l E+6 uCi/CI = 9.81 E+4 Cl 0.

Decay DEI..131I NG DEl131t NG In NG cone J~t) Cnmt Time in RCS in Core in Core Cant in Conlt uGi.hr Posl:.LOOA CD (hrs) (Cl) (CI) (Cl) (Ci) (uCl/coo) cc.R ... (Rlhr) CD (Note I) (Note 2) (Note 3) (Note 4) (Note 5) (Note 5) (Note?7) CA 0.5 9,81E+04 4,33E+08 "1.32E+08 3,21E+05 4.19E+D0 3,301*.02 127 0 I 9.81B+04 3,90P=÷8 '].32Eq÷08 2.912+08 3.79E+-00 3.912.02 97 2 9,81E+04 3,22E÷08 1.302+05 2.73E+05 3.662+00 4.712-02 78 5 9,81E+04 3,27E+a8 1,25E.08 2.5132+05 3.342400 6..932.02 46 (A CA 10 9.812+04 2,96g+08 '1.IBE+08 2,4,52+05 3.202÷00 1.062-01 31 20 9.812+04 2.532+08 1.08E+08 2.31E+05 3.012+00 1,562-01 19 Note 1- See caiculation above Notoe5- (NG In Cent Ci/7.652+1O cc) 1E+6 uCi/Ci -I, Note 5- Values taken from SBC-1 023, Tables 4.6 and 4.6 0 Note 2- Values taken from SB C-I023, Note 3.. Values taken from SBC-1023, Table 5.5 Table 3.1 0 Note 7 - NO cone in ContlJ~t} 0 Nate 4 -9.81 E+4" (NG in core/DEl-131 in core) p.,.

Free volume of containment Volume of RCS 7.66E+i0 cc 3.27E+0& gn, w

CD

• '1

~*5 CD I-.

0 CA CA 0

CD CD 0

CR CR Roy. 0 Rev.O~~Page 331406 312412006 II I III II II I IIII I I I . al II

IIII . J "

Calculation of Fission Product Barrier Matrix EPCALC-06-01 EAL Values for Containment Dose Rates Prepared by; Reviewed by: ________

Li~L Bill Gash Radlallon Proteotion Manager Dete "Co 0o Approved by: 3J&O(o Date

-o 0=

Rev. 0 Page 4 Rtv.IIIPanIIIIII24II00

V24 - Site Specific CSFST Containmentfor Containment PotentialLoss I Number Title Revision F-0.5 CONTAINMENT (Z) 20 (RED) GO TO ENTER FR-Z.1 (RED) GO TO FR-Z.1

' u (ORANGE) (* GO TO J* GO TO

---

• FR-Z.2 (YELLOW) (h GO TO ALL CONTAINMENT N PHSPENETRATIONS YS (YELLOW) GO TO ISOLATED * *

• FR-Z.1 (YELLOW)GOT I (YELOW* GO TO 0* FR-Z.3 ENO YSS I/Hr YE G:\Word~lmagesP'lmages_P.OP\F05.ds4 liQ CSF SAT 1 of 1

V25 - Tech Spec Containment Spray Setpoint for MU7lContainment Potential Loss TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) _z s TRIP SETPOINT ALLOWABLE VALUE

1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generator, Phase "A" Isolation, Containment Ventilation Isolation, and Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency Fan/Filter Actuation, and Latching Relay).

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

c. Containment Pressure-Hi-I 4.2 0.71 1.67 <*4.3 psig < 5.3 psig

d. Pressurizer Pressure-Low N.A. N.A. N.A. _>1800 psig _>1786 psig

e. Steam Line Pressure-Low 13.1 10.71 1.63 _>585 psig >_568 psig*

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. N.A, N.A,

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Containment Pressure--H i-3 0.71 1.67 3.0 _ 18.0 psig < 18.7 psig SEABROOK - UNIT 13/324AedntN.3 3/4 3-24 Amendment No. 33

V26 - UFSAR Section 8.4.2 SBO Coping Basis for MG1 SEABROOK STATION ELECTRIC POWER Compliance with 10 CFR 50.63, Loss of All Alternating IRevision 14 Section 8.4 UFSAR Current Power (Station Blackout) Page 1 8.4 COMPLIANCE WITH 10 CFR 50.63. LOSS OF ALL ALTERNATING CURRENT POWER (STATION BLACKOUT) 8.4.1 Basic Requirements This section describes Seabrook's compliance with 10 CFR 50.63 which requires that each light-water-cooled nuclear power plant be able to withstand and recover from a loss of all alternating current power or station blackout (loss of both offsite power and onsite emergency power). Regulatory Guide 1.155 (RG 1.155), "Station Blackout," provided a method for complying with 10 CFR 50.63. RG 1.155 stated that NUMARC 87-00, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," also provided acceptable guidance for meeting the requirements of 10 CFR 50.63.

Seabrook followed NUMARC 87-00 except where the Regulatory Guide took precedence.

Station Blackout is considered a non-design basis accident. No other active or passive failures or design basis events are required to be considered during Station Blackout. Safe shutdown for Station Blackout (see 10 CFR 50.2 and NUMARC 87-00) means bringing the plant to a hot shutdown or hot standby condition. The Seabrook analysis and procedures proceed with plant cooldown until secondary side pressure is reduced to about 250 psig and the plant is in a hot standby condition.

Seabrook responds to Station Blackout as an AC Independent plant relying only on the station batteries as a source of electrical power for the coping duration specified in Section 8.4.2. When the Station Blackout analysis was initially performed there were no alternate AC power sources to support response as an Alternate AC (AAC) plant. The Supplemental Emergency Power System (SEPS) was subsequently installed and may be used as a source of power during a station blackout event but will not be credited as an alternate AC power source.

8.4.2 Station Blackout Duration Seabrook Station's Blackout coping duration is four hours. This is based on evaluation of the offsite power design characteristics, emergency AC power system configuration and emergency diesel generator (EDG) reliability. The offsite power design characteristics included the expected frequency of grid-related loss of offsite power, the estimated frequency of loss of offsite power from severe and extremely severe weather, the number of switchyards and the type of bus transfers. Site-specific weather data were used to evaluate the reliability of offsite power relative to weather-caused outages. One out of two emergency diesel generators is required to operate safe shutdown equipment following a loss of offsite power. A target EDG reliability of 0.975 will be maintained by implementation of the Maintenance Rule EDG performance criteria.

8.4.3 Procedures Station procedures address the action necessary to cope with a Station Blackout (loss of all AC power) including actions such as opening cabinet doors. Also, as required by RG 1.155, procedures address AC power restoration and severe weather conditions.

V27- EC282 184 Seismic Monitoring System Upgrade Description of OBE Lights for HU2 signal proportional to the magnitude of the motion to the three recorders located in SM-CP-5 8.AThe recorders monitor the motion signals and if a level exceeds preset trigger - - -- -

values, then the recorders trigger to record the event, a red indicating light is illuminated on SM-CP-58, and a main plant computer alarm is actuated indicating that a seismic event is in progress. After the event recording ends, the central controller is notified and the data is retrieved, analyzed, and reports are generated. This process typically takes 2-3 minutes. Jf the magnitude on the free field accelerometer exceeds OBE levels, then a - - --

yellow indicating light is illuminated on SM-CP-58. The Operators then take the actions specified in the applicable procedures.

EN-AA-205-1 100-F01, Revision 4 ECFORM- 110

V28 - Verification of Fire Alarms for IC HU4 Verification of Fire Alarms for NiEJ 99-0 1 Rev. 6 IC I{U4 Per FP62296, Seabrook Station Pre-Fire Strategies Manual Table 112 Location Control Room Alarm Location Control Room Alarm Condensate Storage Tank Fuel StorageBuilding FP-.CP-454 Enclosure*

Containment FP-CP*-376 Primary Auxiliary Building FP-CP-446/FP-CP-4531FP-CP-559 Control Building FP-CP-558 Service Water Pump House FP-CP-3801FP-CP-474 Cooling Tower FP-CP-381 Steam and Feedwater Pipe FP-CP-45 1 EAST

__________Chases FP-CP-452 WEST Diesel Generator Buildings FP-CP-561(A) & FP- North Tank Farm*

___________________ CP-562(B) ____________ __________

Emergency Feed water FP-CP-560 Startup Feedwater Pump Pump House Area*___________

RHIR Equipment Vaults ..... FP-CP-475 __________

• Fire Protection Design Basis Document (DBD-FP-02), Rev. 0, documents NRC approved deviations from 10 CFR 50, Appendix R, fire protection guidelines for location of fire detection systems in the indicated locations.

V29 - Site Specific CSFST Subcriticalityfor MS5IMA5IMU5 Number Title IRevision F-0.1 SUBCRITICALITY (S) 20 (RED) GO TO FR-S.I (ORANE)EE E GO TO mn m un n u FR-S.I ENTER, (YELLOW)

GO TO FR-S.2 CSF:

SAT (GREEN)

GO TO (YELLOW) (*

OOOOO FR-S.2 CSF NOTE: USE GAMMA-METRICS SAT (GREEN)

G:\Word~lmagesP\lmagesP.OP\FO01.ds4 1 of 1

V/30 - EPCALC-06-01 Letdown Monitor Value for MU3 0 FPL Energy Seabrook Station SEP# 2006020

SUBJECT:

Calculation of Value for Initiating Condition SU4, EAL #1 FROM: 0. L. Young DATE: April 17, 2006 TO: File Initiating provides Condition indication SU4, of fuelEAL clad#1degradation addresses greater a letdown radiation than monitor Technical reading that Specification allowable limits. This value was calculated as follows.

• The equipment tag identifier for the Letdown Monitor is 1-RM-6520-1

• ROS noble gas activity at 10OIEb, = 1.887E+2 uCi/grn (per SBC-1023, Table 3.2)

• Noble gas activity will govern the monitor reading, not iodine

• Letdown Monitor Conversion Factor = 14.148 mR/hr/uCi/ml (per ER 4.3, Figure 12) therefore, The 1-RM-6520-1 reading at the Technical Specification allowable limit =

1.887E+2 *iig

• 14.148 mR/hr/u*fm = 2,669.73 mR/hr or 2,670 mR/hr The above calculation is EPCALC-06-03.

Prepared by:.

David SeniorL. Young Analyst Nuclear Approved by:

William T. Cash Radiation Protection Manager DLY/dly CC:

V31 - Tech Spec 3/4.4.8 Specific Activity for MU3 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microCurie per gram DOSE EQUIVALENT 1-131, and

b. Less than or equal to 100/E microCuries per gram of gross radioactivity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

• ---------NOTE...------..............

LCO 3.0.4.c is applicable to DOSE EQUIVALENT 1-131.

MODES 1, 2, and 3*:

a. With the specific activity of the reactor coolant greater than 1 microCurie per gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 5000 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and

b. With the specific activity of the reactor coolant greater than 100/IE microCuries per gram, be in at least HOT STANDBY with Tavg less than 5000 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than 1 microCurie per gram DOSE EQUIVALENT 1-131 or greater than 100/I: microCuries per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-3 until the specific activity of the reactor coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-3.

• With Tavg greater than or equal to 5000 F.

SEABROOK- UNIT 1 3/4 4-19 SEABOOK-NIT 3/44-19Amendment No. 4-44, 115

V32 - UFSAR Table 11.5 Monitor Values for RU1 TABLE 11.5-1 PROCESS AND EFFLUENT RADIATION MONITORS (Note 5)

Det Back Range Alarm Instrument Tag Detector Grd Low-High Set Point Reference Detector Safety Energy* Loop Diag. P&ID No. Re- Deescription Typ mr/hr (ucilcc) Isotooe_ Qtv Class Level 1-NH-Y I-NHY 58 13 1 137 6454 St*term Drains Gamma 0.5 10-6 10.2 Co ,! ,CS Non lE Note 2 506765 SD-20404 Scint 6502 WTaste Gas Inlet to Gamma 15.0 10-2 10+2 Xe 133 Non 1E Note 1 506897 20772 Caarbon Delay Beds Scint 6503 [aste Gas Gamma 15.0 10-3 10k' Krs 5 Note 1 506898 20770 ompressor Inlet Scint 6504 H 2Gas Compressor Gamma 15.0 10-3 10+1 Note 1 506899 20773 isch. Scint 133 6505 C'ondenser Air Evac Beta 0.5 Note 3 Xe Note 1 506055 20774 Scint 6515,6516 Primary Component Gamma 2.5 10-7 10-3 co 5 s,11'3 1,C5 137 2 Note 2 506190, 20211, Cooling Water Scint 506194 20205 6509 Liquid Waste Test Tk 10-6 10"2 C58,I131 ,S137 Gamma 2.5 Note 2 506927 20831 Disch to CWS Scint c58,I131,C 137 6510, 6511, Steam Gen B lowdown Gamma 2.5 10-6 10-2 4 Note 2 506815 20521 6512, 6513 Sample Loops 1,2,3,4 Scint 6519 Steam Gen Blowdown Gamma 2.5 10-7 10-3 Co 58 , 1131,CS 137 Note 2 506734 20626 Flash Tank Drain Scint 6520 Reactor Coolant Gross GM 15 10-1 10+4 mr/hr Co 58 ,I 13 1,CS 137 Note 2 506269 20722 Activity Monitor See Table 11.5-1 (Sheet 3) for notes.

V32 - UFSAR Table 11.5 Monitor Values for RU1 (Note 5)

Det Back Range Alarm Detector Grd Low-High Set Point Reference Detector Safety Energy* Loop Diag. P&ID Instrument Tag No. Re- Tye mr/hr (iclc (ui/c Isotone Oty Class Level 1-NHY 1-NHY Description 33 6481-1, 6482-1 Main Steam Line Gamma 2.5 10-° 10+Smrihr Xe' - 506551 20580 6481-2, 6482-2 Monitor Scint -2, -3, -4 20581 6490 Aux Steam Cond Gamma 0.5 10- 7 10"3 Co5 s,i'3 , CS' 37 1 Nonl1E Note 2 507165 20908 Scint 6521 Turb. Bldg. Sump Liq. Gamma 2.5 10-6 10-2 COss, 1131, CS' 3 7 20195 Non lE Note 2 506713, Monitor Scint 506716 6527A1, A2 COP Monitors GM 2.5 10' -l0 6 cpm Xe 133 4 1E Note 1 506211 20504 B1, B2 33 6560 Resin Sluice Line GM 100 Note 4 Xe1 1 Non 1E 506694 20252 33 6561 Resin Transfer Line GM 100 Note 4 Xe' 1 Non lE 506694 20735 33 6564 Sluice Pump Line GM 100 Note 4 Xe1 1 Non lE 586692 20252 58 37 6473 Water Treatment Gamma 2.5 10 Co , i131. CS' 1 Non lE Note 2 506976 20040 Liquid Effluent Scint Radiation Monitor

1~

V32 - UFSAR Table 11.5 Monitor Values for RU1 SEABROOK RADIOACTIVE WASTE MANAGEMENT Revision: 13 STATION UFSAR TABLE 11.5-1 Sheet: 3 of 3 Max. Beta Energy Predominant Gamma Energy (Mev)

Note Isotopes 1 Xe' 33 0.346 0.08 1 Xe' 3 5 0.92 0.249 Kr85 0.67 0.5 14 KrsSm 0.82 0.150 2 ~ 131 0.606 0.364 1133 1.27 0.53 CS'3 4 0.662 0.604 Cs' 3 7 0.514 0.662 Co58 0.474 0.81 Co6 ° 0.3 14 1.17, 1.33 3 Condenser Air Evacuation Monitor to have output in counts per min (cpm) (10' to 106).

4 Monitors 6560, 6561, 6564 have output in mr/hr (100 to 105).

5 Radiation monitoring setpoints are varied during operation to follow station operating conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the setpoints is found in the Station Offsite Dose Calculation Manual (ODCM) and/or station operating procedures.

A

ATTACHMENT 5 Technical Information for Proposed Initiating Conditions and Emergency Action Levels

SEABROOK STATION SPECIFIC VALIDATION DOCUMENTS FOR NEI 99-01 REV. 6 EAL SUBMITrAL Vi- Technical Specifications Table 1.2 Mode Definitions V2 -Spent Fuel Pool Instrumentation Levels Drawing Per EC281289 V3 - EPCALC-06 Effluent Monitor Values for R EALs V4 - ODCM and Technical Specification Basis for Site Boundary Receptor Point V5 - UFSAR Table 12.3 Manipulator Crane and Spent Fuel Building Monitor Ranges V6 - UFSAR Section 9.1.4 Description of Refueling Pathway V7-Table Hi Procedure References V8 - UFSAR Table 12.3-15 WRGM Ranges V9 - 0S1000.09 Procedure References to Refueling Pathway Level Indicators VIO - UFSAR Tab'le 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges Vll - EPCALC-06 RVLIS Values V12 - UFSAR Table 12.3-14 Manipulator Crane Monitor Ranges V13 - UFSAR Section 6.2 Description of Containment Sumps V14-SAMG Calculation Aid CA-3 for Hydrogen Flammability in Containment V15 -Technical Specification Table 1.2 Cold Shutdown Temperature Limit V26 -O0S1000.09 Definition of Reduced Inventory V17 - Site Specific EAL Design Basis for 25 PSlG RCS Pressure Range V18 - UFSAR Section 8.3.2 DCV 105 Limit V29 - NUHOMS HSM Dose Rates Technical Specification V20 - Site Specific CSFST Core Cooling V21 - Site Specific CSFST Integrity V22 - Site Specific CSFST Heat Sink V23 - EPCALC-06 Rad Values for Fission Product Barrier Matrix V24 - Site Specific CSFST Containment V25 - Technical Specification Table 3.3-4 Containment Spray Setpoint V26 - UFSAR Section 8.4.2 SBO Coping Basis V27 - EC282184 Seismic Monitoring System Upgrade Description of OBE Lights V28 - Verification of Fire Alarms V29 - Site Specific CSFST Subcriticality V30 - EPCALC-06 Let Down Monitor Value V3i - Tech nical Specification 3/4.4.8 RCS Specific Activity V32 - UFSAR Table 11.5-1 Effluent Monitor Ranges

CROSS REFERENCE NEI 99-01 REV. 6 INITIATING CONDITIONS! SEABROOK STATION SPECIFIC VALIDATION DOCUMENTS Initiating Condition Applicable Validation Documents ALL V1 RG 1 V3, V4 RG 2 V2 RS1 V3, V4 RS2 V2 RAl V3, V4 RA2 V2, V5, V6 RA3 V7 RU1 V8, V32 RU2 V6, V9, ViO CG 1 VIO, Vll, V12, V13, V14 CS 1 Vll, V12, V13 CA1 Vil, V13 CA3 V15, V16, V17 CU1 V13 CU3 V15 CU4 V18 EU1 V19 FISSION PRODUCT BARRIER POTENTIAL LOSS/LOSS V23 FUEL CLAD BARRIER POTENTIAL LOSS 1.A V20 FUEL CLAD BARRIER LOSS 2.A V20 FUEL CLAD BARRIER POTENTIAL LOSS 2.A V20 RCS BARRIER POTENTIAL LOSS 1.B V2 1 RCS BARRIER POTENTIAL LOSS 2.A V22 CONTAINMENT BARRIER POTENTIAL LOSS 2.A V20 CONTAINMENT BARRIER POTENTIAL LOSS 4.A V24 CONTAINMENT BARRIER POTENTIAL LOSS 4.B V14 CONTAINMENT BARRIER POTENTIAL LOSS 4.C1 V25 HA5 V7 HA6 OPERATIONS ABNORMAL PROCEDURE OS1200.02 HU2 V27 HU4 V28 MG1 V26, V20 MG8 V18 MA5 V29 MU3 V30, V31 M U5 V29 M U7 V25

V1 - Tech Spec Table 1.2 Mode Definitions TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 24 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

hours.

D At least once per W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

SFCP In accordance with the Surveillance Frequency Control Program TABLE 1.2 OPERATIONAL MODES REACTIVITY  % RATED AVERAGE COOLANT MODE CONDITION~keff THERMAL POWER* TEMPERATURE

1. POWER OPERATION > 0.99 > 5% >_350°F

2. STARTUP _>0.99 <*5% _ 350°F

3. HOT STANDBY < 0.99 0 _ 350°F

4. HOT SHUTDOWN < 0.99 0 3500°F > Tavg >2000°F

5. COLD SHUTDOWN < 0.99 0 < 200°F

6. REFUELING** < 0.95 0 < 140°F

• Excluding decay heat.

• • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

SEABROOK - UNIT 1 -

1-8 mnmn 141 Amendment 4

EXECUTIVE

SUMMARY

The purpose of this engineering change (EC) is to install reliable primary and backup level instrumentation to monitor Spent Fuel Pool (SFP) water level in accordance with the requirements of Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Issuance of Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, March 12, 2012. This EC includes several hold points as described in Section 5.1.6.

Revision 001 revised the BC classification from quality Related to Safety Related to ensure that raceway components are procured as Seismic Category I components in accordance with UFSAR Section 3.10(B) and table 3.2-1. Sketch SK-1000 was also revised to correct a typo's on the instrument tag ID's and drawing Safety Classification.

Revision 002 provides for the following:

1. Revision 002 relocates and modifies the Westinghouse designed and fabricated pool-side probe brackets. The original mounting locations on the North and the South sides of the Fuel Storage Building (FSB) spent fuel pool (SFP) are changed to the West side of the pool to facilitate and simplify the installations. The new mounting locations on the West side of the pool will reduce implementation concerns inherent with the original FHM crane rail pocket locations, including welding and modification of the concrete curb. The West location simplifies mounting of the brackets using Hilti concrete anchors and significantly reduces the potential for foreign material entering the pool during installation. Refer to the discussion below and in BC Section 2.5.H for details. See SK 101 570-EC28 1849 located in the EDMS panel for the optional bracket locations.

See SK 11 1576-EC281849, sheets 1-3 for optional bracket assembly details.

2. The modified pool side bracket design includes a davit socket to facilitate installation of a tool to lift the probe assembly for calibration. Hold Point 1 is closed.

3. Since the Fuel Handling Machine cannot physically access the West end of the SFP, exclusion zones will not need to be established around the new sensor locations. Hold Point 2 is closed.

4. Revised the safety class on Sketch SK-lO001 to SC-3.

5. Incorporated Westinghouse Specialty Tool List (FP700485) into the ADL.

6. Incorporated revised Westinghouse spare parts list into FP700484.

7. Revised the channel verification acceptance criteria in BC Section 5.3 to initiate channel maintenance and!/or channel calibration if the deviation between indicated level readings exceeds 3.0 inches between the new SFPIS indicators or 2.0 inches between the new SFPIS indicator and the existingl1-SF-LI-2607 indicator.

EN-AA-205-l1100-F01, Revision 4 ECFORM- 110

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8. Revised the markup of SFPIS Configuration Settings Document (1-NHY-508637) to identify that the level transmitter Low Trim Point, High Trim Point and Sensor Ref. count values are for reference only and vary dependent on transmitter calibration.

9. Replaced FP 700493, SFP15 Compliance Matrix - FPL/ NextEra Seabrook," with new Westinghouse document WNA- DC-00270-NAH Rev. 1. Closed Open Item

10. Added Westinghouse Letter LTR-SFPIS-15-34, Clarification of 10 Year EQ-QR-269 Test Basis -

Supplement to WNA-TR-03149-Gen, Rev. 2," into FP 700483.

11. Added Westinghouse letter for part number clarification into FP 700499 (SFPIS Technical Manual).

12. Added Hold Point 4 for modification SEP lighting on the south west corner of the Spent fuel Pool (Ref. BC Section 2.5.T and 5.1.6).

Revision 003 provides for the following:

1. AWA 01 added FP700489 onto the ADL.

2. AWA 01 revised drawing 11 1576-EC28 1849 sht. i to correct the center line dimension for locating the Davit Socket on the SFPIS mounting bracket. Refer to the markup in the BDMS folder.

3. A 4" minimum embedment could not be achieved for one of the 5/8" diameter HKB3 anchors that are used mount probe 1-SF-LE-26 16. AWA 2 approved installation of the 5/8" diameter HKB3 anchor with a 2-3/4" minimum embedment. The capacity of the 5/8" Hilti KB3 with 2-3/4" vs 4" min. embedment is acceptable for the subject installation. Refer to the markup of SK 111576-EC281849 Shts. 1 & 2 and FP 700468 in EDMS.

4. AWA 03 added a requirement to post signs at the entry points into the Spent Fuel Building and CEVA to exclude the use of portable radios and cell phones during ELAP events (Ref.

AR 02063219). See BC section 5.1.5 for sign requirements.

AWA 03 also added a requirement to post signs on the front of control panel 1-SF-CP-300-A

& B to identify that radio use will interfere with SFPIS indications. BC section 5.1.5 is revised to identify the new posting requirement for panels 1-SF-CP-300 A & B.

5. The scaling values that Westinghouse provided for the SEPIS channels were not referenced to the top of the fuel rack. AWA 04 revised the scaling and corrected the guidance provided in section 5.2 of FP 700499 for level transmitter scaling. An offset value was used to adjust the scaling for the unmeasurable level from the top of the fuel rack to the bottom of the probe. The scaling values for the SEPIS digital display (SF-LI-2616 and 2617) and MPCS analog points (A4220 and A4172) are revised to align the indications with the revised transmitter scaling.

EN-AA-205-1 100-F01, Revision 4 ECFORM- 110

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6. The offset setting that was added to the transmitter scaling by AWA 04 resulted in an unexpected shift in the transmitter output. AWA 05 set the offset setting back to 0.00 inches and revised the transmitter scaling to include the required offset (7.88 inches to 326.41 inches). The revised transmitter parameter settings are shown in the markup of configuration settings document 1-NHY-508637 in the EDMS folder.

7. AWA 06 revised the low and high trim points to provide additional margin to the dead zones at the top and bottom of the probe. This change was recommended by Westinghouse to ensure signal stability as level approaches the upper and lower probe dead zones.

AWA 06 also documented the counts at the revised trim points (high and low) and the sensor reference (fixed and auto). The sensor reference counts for the calibration kit were also recorded for future reference. Refer to the markup of the configuration settings document 1-NHY-508637 in the EDMS folder for the new trim point values and revised counts.

8. AWA 07 was superseded by AWA 08.

9. The field identified that the replacement assembly from inventory (CID 171805) is not the same as the installed assembly. Based on correspondence from Westinghouse that stated that the new battery assembly was fully qualified, AWA 07 approved the use of an alternate battery assembly for Westinghouse P/N. PS20035H05. Further correspondence with Westinghouse revealed that the qualification documentation for the new battery assembly is not ready and will not be available for several weeks. AWA 08 identified that the delivered part (CID 171805) is not acceptable for use at this time. The original style battery assembly is available as a warrantee item and is being shipped to Seabrook with an estimated delivery date of 10/21/15.

AWA 07 also identified that the new battery assembly has a different OEM P/N', than the original assembly (Phoenix Contact P/N UPS-BAT/VRLA-WTR/24DC/26AH-23 20429 vs UPS-BAT/VRLA/24DC/26AH-2320429). Westinghouse identified that the new battery assembly has the same OEM P/N as the o~riginal assembly, but is physically different and has revised ratings. The differences will be evaluated as part of the qualification documentation.

The correct OEM P/N for both configurations is UPS-BAT/VRLA-WTR/24DC/26AH1-2320429. The SFPIS Bill of Materials on FP 700477 is revised to reflect the correct part number for the battery assembly. Refer to the markup in the EDMS folder.

10. Added procedure IS 1682.224 and IS 1682.226 onto the ADL.

11. Added FP700468 onto the ADL.

EN-AA-205-1 100-F01, Revision 4 ECFORM-l 110

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12. As a result of a failure of SFPIS components at Point Beach it was determined that the Westinghouse supplied signal duplicators are digital devices. The parameters fields for tag IDs 1-SF-LY-2616 and 1-SF-LY-2617 are therefore revised to identify that these components are digital devices that contain firmware. The firmware was also classified as level B in accordance with the fleet software control program (IM-AA-205-l1100). Refer to the new IM-AA-202-F0 1software classification form in the EDMS folder the EMRs for tag IDs 1-SF-LY-2616 and 1-SF-LY-2617.

13. The IM Department was not able to load the UJPS-CONF utility software that Westinghouse provided onto a laptop. As part this revision the software classification will be updated to allow use of the newer version of the software that is available from Phoenix Contact as a download (QUINT UPS UPS-CONE_online setup2 2_7_0.exe). Refer to the revised IM-AA-202-F01 software classification form in the EDMS folder.

14. Engineering Evaluation EE-15-016, "CEVA and ESWGR Beyond Design Basis External Event Ventilation Evaluation" was developed to perform a heat up analysis for the CEVA and ESWGR rooms. This evaluation confirmed that area temperatures will remain below the temperature to which Westinghouse qualified the Spent Fuel Pool Instrumentation during an ELAP event. Augmented ventilation will therefore not be required following a beyond Design Basis Event to maintain CEVA and Essential Switchgear Room temperatures below the equipment qualification temperature of 140 deg. F. EC sections 2.3.8, 2.5J and 5.3 are revised to identify that compensatory ventilation will not be required. Specification S-S-i1-E-0227 is also revised to reflect the maximum temperature conditions that could occur in the CEVA and Essential Switchgear Rooms during a beyond design basis extend loss of power event.

15. Since the FLEX equipment storage facility and some of the FLEX equipment will not be available in time to support compliance of NRC Order EA-12-049, an extension has been submitted to the NRC to extend the compliance date for the Order until May of 2016. As a result of the extension, the FSG procedures that are required to support use of the new instrumentation during an ELAP will not be completed in time to support the compliance date for NRC Order EA-12-051. As a remedial action the required guidance for the Spent Fuel Pool Instrumentation will be added to existing plant operating procedure OS 1215.07, "Loss of Spent Fuel Pool Cooling or Level."

16. During the 2015 FLEX audit the NRC inspector noted that the flexible conduit runs from each SFPIS probe to the west wall of the fuel storage building will present a potential trip hazard at the deck of the pool that could render the instrumentation inoperable. It was recommended that the station consider the installation shielding to protect these sections of the cable runs (Ref. AR 02063248). Hold Point 4 is added to address the potential trip hazard associated with the SFPIS sensor cable installations at the south side of the pooi.

Refer to EC section 5.1.6.

EN-AA-205-1 100-F01, Revision 4 ECFORM-l 110

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17. Revised the channel verification acceptance criteria in EC Section 5.3 to initiate channel maintenance and!/or channel calibration if the deviation between indicated level readings exceeds 3.0 inches between the new SFPIS indicators or the existingl.-SF-LI-2607 indicator.

18. During testing it was determined that the serial numbers for UPS modules 1-SF-XX-2616-1 and l-SF-XX-2617-1 were programmed into the units did not align with the primary and backup channel designations in drawing 1-NHY-508637. The serial numbers were revised to align the panels with field installation and a note was added to identify that the serial numbers are set by the field based the installed unit.

19. The SFPIS technical manual (FP700499) was updated to reflect Westinghouse's revised replacement frequency for the couplers, coaxial cables and coaxial connectors. According to Westinghouse these components are to be replaced every ten years.

Revision 004:

Issued to remove the Critical Attribute for verifying that the Fuel Handling Machine, when in AUTO, will remain 1 ft from probes. This criterion was established in revision 000 when the probe locations were on the North and South pool walls. This attribute no longer applies to the final probe locations on the West end of the pool as installed by Revision 002 to this BC. The Spent Fuel Handling Machine is physically not capable of approaching the installed probe location, refer to details in BC section 2.3.14. For this same reason the MATP has been revised to delete any and all references to testing to verify no interaction between the machine and the installed instrument probes.

Additionally the EDMS folder for Specification S-S-1 -E-0227 has been corrected to eliminate the duplicate page.

This revision affects BC sections 2.5.S, 5.1.1, the MATP and the Specification folder in BDMS. These changes support the installed design as modified by revision 002 and the changes in this revision do not change the Scope, Intent or Requirements of the BC and the original 10 CFR Applicability Determination and 72.48 Pre-Screen along with the BC R/002 10 CFR 50.59 Screen remain applicable.

2.0 DESIGN 2.1 Bases for Current Design The Primary function of the spent fuel storage facilities is to provide for the safe handling and storage of irradiated fuel assemblies under water, new fuel assemblies, and control rods. The safety function of the spent fuel pool and storage racks is to maintain the spent fuel assemblies in a subcritical array during all credible storage conditions, and to provide a safe means of cask loading of the assemblies for dry fuel EN-AA-205-1 100-F01, Revision 4 ECFORM-l 110

V2 - SFP Instrumentation Levels Per EC28 1849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL Page 6 of 75 INSTRUMENTATION UPGRADE FOR NRC ORDER EA-12-051 storage. System and Component safety class, seismic class, code of design and construction, and code classes are listed in UFSAR Table 3.2-2.

The Fuel Storage Building and Spent Fuel Pool are safety-class 3, seismic Category 1 structures designed to the requirements of Regulatory Guide 1.13 (Spent Fuel Storage Facility Design Basis),

Regulatory Guide 1.29 (Seismic Design Classification) and Regulatory Guide 1.117 (Tornado Design Classification) to ensure optimum protection for stored fuel assemblies against the effects of extreme natural phenomena such as safe shutdown earthquake, tornadoes, hurricanes, missiles, and floods.

These design considerations extend to the SFP wall, fuel storage racks, and other critical components whose failure could cause criticality, loss of SFP cooling, or physical damage to stored fuel.

Spent fuel pool water level at the upper elevation of the SFP is indicated by a level transmitter (SF-LT-2607) which causes an audible alarm in the control room on high or low water level. Level from this instrument loop can also be observed as an analog computer point on the MPCS (Al1185) over a 4 foot span (22.875 feet to 26.875 feet, referenced to the reactor flange (-) 1 ft, 10.5 in.).

2.2 Reason for Design Change After the earthquake and tsunami at the Fukushima Daiichi nuclear plant in March, 2011, the lack of indication of Spent Fuel Pool (SFP) water level led to concerns about the potential for loss of inventory. While it was eventually determined that SFP integrity and fuel cooling were not compromised, considerable efforts were made following the event to assure that there was adequate cooling of the fuel in the SFP. These efforts diverted resources and attention from other critical tasks that were required to manage the event, demonstrating that confusion and misapplication of resources can result from beyond-design-basis external events when adequate instrumentation is not available.

The NRC subsequently determined that enhanced spent fuel pool instrumentation represents a substantial increase in protection to public health and safety and issued Order BA- 12-051. The Order requires that licenses install reliable primary and backup level instrumentation to monitor Spent Fuel Pool (SFP) water level. The Nuclear Energy Institute (NEI) provided a document, NEI 12-02, Industry Guidance for Compliance with NRC Order BA- 12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" which provides an approach for complying with Order BA-12-05 1. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order BA-12-051, Reliable Spent Fuel Pool Instrumentation, endorses NEI 12-02, subject to clarifications and exceptions specific to Section 3.4, Qualification, of NEI 12-02, as an acceptable means of meeting the requirements of Order BA-12-051. Design of the Spent Fuel Pool Instrumentation described in this BC will be consistent with the guidelines of JLD-ISG-2012-03 and NBI 12-02.

2.3 Description of Design Change This BC provides for the installation of two new independent level instrument channels (SF-L-26 16 and SF-L-26 17) in the Spent Fuel system. The new instrument channels will be used to monitor spent BN-AA-205-1 100-F01, Revision 4 ECFORM- 110

V2 - SFP Instrumentation Levels Per EC281849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL Page 7 of 75 INSTRUMENTATION UPGRADE FOR NRC ORDER EA-12-051 fuel pooi level during and following beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.

The new Spent Fuel Pool Instrumentation System (SFP1S) was designed and fabricated by Westinghouse Electric Co. Both instrument level channels will be permanently mounted and access into the Spent Fuel Building (SFB) will not be required to operate or obtain level data from the instrument channels.

Each channel will be capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating water level ((-) 1 FT to 24 Ft, Plant Elev.). Each channel will display the indicated SEP water level to reasonably high accuracy when accounting for worse case environmental conditions and instrument uncertainties. Level readings will be provided in feet based on height above the fuel rack, i.e. top of the fuel rack is 0.0 feet. The indicated level range will be from 1.0 to 25.5 feet above the fuel rack.

Three critical spent fuel pool (SEP) water levels within this range are defined in NEJ 12-02 Revision 1.

The Spent Fuel Pool level instrumentation installed per this EC will be designed to provide a reliable means of SEP level measurement at the following critical water levels:

a. Level 1, level adequate to support operation of the normal spent fuel pool cooling system -

This level is defined as:

1. Level at which reliable suction loss occurs due to uncovering the coolant inlet pipe OR any weirs or vacuum breakers associated with suction loss, or

2. The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis.

To determine the higher of the two levels the following was taken into consideration:

(1) The level at which reliable suction loss occurs due to uncovering the coolant inlet pipe or any weirs or vacuum breakers associated with suction loss. This level was established based on nominal suction strainer inlet elevation and conservative estimate for the onset of vortexing.

The actual effect of the strainer on this level has been formally determined by calculation C-S-1-24606, "Spent Fuel Pool Level for Reliable Pump Suction." Based on this calculation the elevation for reliable pump suction is plant elevation 23 ft., 4 inches.

(2) The level at which the normal spent fuel pool cooling pumps loose required NPSH assuming saturated conditions in the pool. Seabrook Calculation C-S-1-24606, "Spent Fuel Pool Level for Reliable Pump Suction," establishes the point of zero NPSH margin at plant Elevation 22 ft., 4 inches.

EN-AA-205-l1100-F01, Revision 4 ECFORM- 110

V2 - SFP Instrumentation Levels Per EC28 1849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL Page 8 of 75 INSTRUMENTATION UPGRADE FOR NRC ORDER EA-12-051 With the spent fuel pool at 212 degrees F (saturated conditions) the NPSHA is approximately 11.2 ft. The NPSJAR for the spent fuel cooling pump is 10 ft. at 212 degrees F. This results in a ratio of NPSHA!NPSHR value of approximately 1.12. Therefore, the NPSHA is greater than the NPSHR at saturated conditions.

The higher of the above points is the level where the inlet strainer will lose suction (Item (1) above). Therefore, Level 1 has been established at plant elevation 23 ft., 4 inches ('- 24 ft. 10 in.

above fuel racks) for both the primary and backup instrument channels.

b. Level 2, level adequate to provide substantial radiation shielding for a person standing on the spent fuel pooi operating deck - Level 2 is defined as the level that provides substantial radiation shielding for personnel to respond to Beyond-Design-Basis External Events including the initiation of SFP makeup strategies that would require access to the Fuel Storage Building (FSB).

Indicated level on either the primary or backup instrument channel of greater than an elevation of 10 ft., 9-1/2" (-12 ft-3mn above fuel racks) will provide substantial radiation shielding for a person standing on the SFP operating deck. This elevation is approximately 13 feet above the top of the spent fuel assemblies positioned in the spent fuel racks (elev. (-) 2 ft -2.5 in.). With 13 feet of water above the highest fuel element position, the calculated dose rate at the surface of the SFP is less than 2.5 nmrermlhr (Ref. UFSAR Section 12.3.2.l.c). This monitoring level ensures there is adequate water level to provide substantial radiation shielding for personnel to respond to Beyond-Design-Basis External Events including the initiation of SFP makeup strategies that would require access to the Fuel Storage Building (FSB).

c. Level 3, level where fuel remains covered - Level 3 corresponds nominally (e.g. +/-/- 1 foot) to the highest point of any fuel rack seated in the SFP. Level 3 was defined in this manner to provide the maximum range of information to the Operators and emergency response personnel. An indicated level on either the primary or backup instrument channel of less than plant elevation (-) 6 inches. This is the nominal water level approximately 1.5 ft. above the top of the fuel racks. This monitoring level will assure the maximum range of level information is available to the plant Operators and emergency response personnel. This level is also assumed to be the minimum level that assures that adequate water level remains above the top of the stored fuel seated in the SFP (nominal elevation of (-)2 ft - 2.5 in.).

The following figure shows the critical SFP levels described above:

EN-AA-205-l 100-F01, Revision 4 ECFORM- 110

V2 - SFP Instrumentation Levels Per EC281849 for NRC Order EA-12-051 for RG2/RS2/RA2 EC - 281849 Rev. 004 SPENT FUEL POOL INSTRUMENTATION UPGRADE FOR IPage 9 of 75 NRC ORDER EA-12-051I SEABROOK ELEVATION SPENT FUELview POOL LEVELS OEA1INGFLOOR EL, 25

~~NORMAL. WATER LEVEL EL 23'-4" TO 25 -9'I

-il" TO.

... EVIEL NST RUIME

==--- 'IRMNSMIITER SlENSITIVI T

BAM°%6"

............ LEVEL 1 LOSS 0 r TNPSH0 212"F EL. 22"~V'*/"

4- -TOP OF" STRAIN*ERa.L 21V5"

___________________RELIARLE 4 ITECI'IFCAL SUCTION LEVEL EPECIFICATtON LMllEL..23-4" EL. 2T"l/z *.-

-1 LF.VELFOR SURSTAN~TIALRAOIATON SHIELDINGEL. 1Q-gz 111 "- I-" ..... *' 2 0212" F NPSI. 11.2Ft

-LEVEL WIERE FUEL REMAN$ COVERED 1 0" O " 5 4 -

-- lEVELNSSTRIME .

1"TOP oF FUEL RACI EL.,(-1 r-5s4 "

4~L~[HEfl1~R

• ux-*-u-*`*`=*-u-u*.*-*-u-=T*-*.*-*-=y`=-*-*.*-*=`v`*u*a*.*=*=*.*-=* ml

"'-B'TOM 41~K AhK Or` SEP EL. :-l 15'~ -1Z I::: -

°,

2.3.1 Design Inputs:

NRC!/NET Guidance:

a. NRC Order EA-12-051, "Order Modifying Licenses with regard to Reliable Spent Fuel Pool Instrumentation"

b. NEI 12-02, Industry Guidance for Compliance with NRC Order EA- 12-051, to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

c. JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, August 29, 2012, Nuclear Regulatory Commission, Japan Lessons Learned Project Directorate.

Westinghouse Documentation:

a. FP 700481, Spent Fuel Pool Instrumentation System Wire List (WNA-WD-01094-GEN).

b. FP 700498, Spent Fuel Pool Instrumentation System Torque Specification (WNA-IG-00452-GEN).

c. FP 700487, Spent Fuel Pool Instrumentation System (SFPIS) Walk down Report (CIS-WR-13-6).

d. FP 700482, Spent Fuel Pool Instrumentation System (SFPIS) Standard Product Test Strategy (WNA-PT-00 188-GEN).

e. FP 700479, Spent Fuel Pool Instrumentation System Power Consumption (WNA-CN-003 00-GEN).

EN-AA-205-1 100-F0Ol, Revision 4 ECFORM-1 10

V3 - Plant Vent and Main Steam Line Effluent Calculations for RG1/RS1lRA1 0

FPL Energy Seabrook Station SEP# 2006013

SUBJECT:

Calculation of Effluent Release Values Used in NEI EALs FROM: D. L. Young DATE: March 28, 2006 TO: File Attached are the calculations for the effluent release values used in the NEI Emergency Action Levels (EALs). The calculation number is EPCALC-06-02.

DLY/dly cc: 0. Robinson

V3 - Plant Vent and Main Steam Line Effluent Calculationsfor RG1/RS1IRA1 Calculation oV Radiological Effluent EAL Values EPCALC-06-0D2 Plant V.ent-Cffijzen~ Calculation for a Site-Arfe.a Em.f.*rg*nc~y..i...

DFB, Avg Disp mrem-m3 wrem-m 3 Factor mrem-sec Isotope pCi-yr uCi-yr seclm3 uCi-yr (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.83E+03 8.5nDE-07 7.51 E-03 Kr-87 5.92E-03 5.92E+03

• 8.50Eo07 5.03E-03 Kr-88 1.47E-02 1.47E+O4 8.50E-07 1.25E-02 Kr-85m 1.17E-03 1.17E+03

• 8.50E-07 9.95E-04 Xe-135 1.81E-03 1 .81E÷03 *r 8.S0E-07 I .54IE-03 Xe-133 2.94E-04 2.94E+02

• 8.50E 07 2.50E-04 Kr-85 1.6lE-05 1,61 E+01 8.50E-07 1 .37E-05 Xe-131m 9.15E-05 9.15E+O1 8.50E-07 7,78E-05 Xe-133m 2.51 E-04 2.51E÷02 8.50E-07 2.13E-04 Xe-I 35m 3.12E-03 3.12E÷03 8.50E-07 2.65E-03 DF'* (per year) OF' (per hour)

- '32 *"L;rrm.'secJ F"'" 5 I .... iyl~ L (Note 4) 1I(DF'I (per hour)) WRG3M release rate I. .85÷05 uCi-hr ,oI 0E+o2 mre (Note 5) (Note 6)

Note 1 - from ODUM, Appendix B. Section 5.2.1.2 Note 2 - DFBi

• 1E+06 pCiruCi Note 3 - From ODCM, Appendix B, Table B.7-4 Note 4-.11(865*24) or l/hours in a year Note 5 - 1/(3.5IE-06)

Note 6 - Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal 100 mR ReV. 0 Page 1 ReV.0Pae 13127/2008 I II I III II I '1 1 I

• IIIII IIII

trJitt*, ,"E,'*,~*2i

• J!*l~lt*nt and Main Steam Line Effluent Calculationsfor RG1/RS1/RA1 Calculation of Radiological Effluent EAL Values EPCALC-O6-02 Plaint Vent EffluentCalculation for a General Eme~rgency ...

DFB1 DFBL Avg Disp mrem-ma Factor 3 mrein-sec Isotope pci-yr uCi-yr seclmn uCl-yr (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.832+03 8.S0E-O7 7.512-03 Kr-87 5.92E,-03 5.92E÷03 B.50E-07 5,03E-03 Kr-88 1.47E-02 1.47E+04 *k 8.502-07 1.25E-02 Kr-85m 1.172-03 1.17E÷03 8,50E-07 9,952-04

• b Xe-135 1.81 E-03 S81E÷03 8.50*-07 1.54E-03 Xe-133 2.94E-04 2.94E+02 8.50E-07 2.50E-04 Kr-85 1.61 E-05 1,.61E+01 8.50E-07 1.37E-05

• b Xe-I131m 9.15SE-O5 9.15E+01 8,50E-07 7.78E-O5 Xe-133rn 2-511E-04 2Z51E÷02

• 8,50E-07 2.13E-04 Xe-135mn 3.12E-03 3.1225+03 *k 8.50E-07 2.65E-03 rmrem-secI DF'1(per year) IDF', (per hour)

(Note 4) 1/(DF', (per hour)) WRGM release rate uLi (Note 5) (Note 6)

Note 1 - from ODCM, Appendix B, Section 5.2.1.2 Note 2 - DFBi

• 1E+0S pCifuCi Note 3- From ODCM, Appendix B, Table B.7-4 Note 4- 1/(365*24) or 1/hours in a year Note 56- 11(3.51 E-06)

Note 6- Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal I R Rev, 0 0 ~~Page ReV. 232720 312712006

.... . . L

• iLUt .iI:' .L*.;J .eJ V

!3 - Plant Vent and Main Steam Line Effluent Calculationsfor RG1/RS1/RA1 Calculation of Radiological Effluent EAL Values E PCALC,_-06-02 Main steam Line Effluent Calculation for an Unusual Event.

DFB1 DFB1 3 Avg Disp DF'1 ntrem-m 3 mrem-n3 Factor mrarm-sec pCi-yr uCi-.yr sec/m"= uCl-yr Isotope (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.832+03

• 3A02-0B 3,0DE-02 Kr'-87 5.92E-03 5.92E÷03 3.40E-06 2.01 E-02 Kr-a88 1,.47E-02 1.472+04 3..40E-05 5.O0E-02 Kr-8B5m 1.17E-03 1.17E+03

• 3.40E-06 3.982-03 Xe-I135 1.81 E-03 1.812+03 3AOE-06 6.15E-03 Xe--133 2,94E-04 2.94E÷02

• 340E-05 i.00E-03 Kr-B5 1.612-05 1.61E+01 *i 3AOE-05 5.472-05 Xe-131m 9.15SE-05 9.152+01 3.40E-O6 3.112-04 Xe-I133m 2.51 E-04 2.51E÷02 3.402-06 8.53E-04 Xe-I135m 3.122-03 3.12E+03 3.40E-06 I .06E-02 DF'1 (per year) DF'(per hour)

[I~~~omrem-seo Fl~ T~ = "4E0 mrem-sec u-r (Note,;)

lI(DF'1 (per hour)) SI. release rate 7m12e04 -ecl h* = .6+03 uG!

(Note 5) (Note 6)

S.L. release rate S.L. Flowrate Release concentration (Note 7)

This table converts a release concentration to a main steam line monitor dose rate.

Decay MS J(t) Rel Cone Inverse Rounded Time ndi-hr Dose Rate uCi Dose Rate Dose Rate (his) cc-mR hrlmR mPlhr mRlhr 0.5 1.92E-02 1 1.622-03 1.18E+01 0 0

'1 2.16E-0Z / 1.622-03 1.33E+01 0 0 2 2.672-02 1 1.022-O3 1.65E-01 0 0

,5 4.932-02 1 1.52E-03 3.04E÷01 0 0 10 1.27E-01 I 1.62E-03 7.83E+01 0 0 20 4.97E-01 I 1.62E-03 3,06E+02 0 0 (Note 8) (Note 9)

Note I - from ODCM, Appendix B, Section 5.2.1.2 Note 6 - Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal Note 2 - DFBi

• 1E+06 pCi/uCl ODCM total body dose rate of 500 mR/yr Note 3 - From ODCM, Appendix B, Table B.7-6 Note?7- from SBC-709, page 20 - Used SRV Note 4 - 1/(365"24) or 1/hours in a year flow as bounding (ILe., larger than ASDV).

Note 5- 1/(1.412-5) Note 5 -from SBC-709, page 19 Note 9 - lInverse Dose Rate Rev. 0 Page 3 Rev.0 Pae 3312712006

. .... I II III I II I I IIII III Ill .... l,

V3 - Pfa'nf't~Thn* *n*? Main Steam Line Effluent Calculationsfor RG1/RS1IRA1 Calculation of Radiological Effluent EAL Values EPCALC-06-02 Main. Steam Lini* Effiu~ent Calculation for an Alert -

DFB1 DFB1 Avg Diap DF'1 mrem-ma mrem-m3 Factor toremo-sec pCi-yr uCi-yr sec/in3 uCI-yr Isotope (Note 1) (Note 2) (Note 3)

Xe-138 8.83E-03 8.83E+03

• 3.40E-05 3,00E-02 Kr-87 5.92E-03 5.92E+03 3.40E-06 2.01 E-02 Kr-88 1.47 E-02 1.47E+04 3,405-0B 5,0OEo02 Kr-85m 1,17E-03 1.17E÷03

• 3.4OE-06 3.9BE-03 Xe-I135 1,81 E-03 1.81E+03 3.40E-06 6.15SE-03 Xe-It33 2.94E-04 2.94E+02 3.40E-06 1.00E-03 1r Kr-85 1.61E-Q5 1.61E÷01 3.4OE-06 5.47E-05 Xe-I131rm 9.15SE-05 9.!5E+01

• 3.40E-06 3.1IE-04 Xe-I133m 2.51 E-04 2.51 E+02 3.40E-06 8.53E-04 Xe-1 35m 3.12E-03 3.12E÷03 3.4OE-06 1.06E-02 DF'I {per year) .... DF'1 (per hour)

F1.3E0 1 mrems I-sCuir I IAi4E-04~Y I = I 41E05 mre'sctui (Note 4) 1I(DF'1 (per hour)) S.L. release rate 7,1E+4 re-seJ .1E+ 1 rI (Note 5) (Note 6)

S.L. release rate SJ.L Flowrate Release concentration

[U~+ sUe I F ~ =i 325 ui I (Note 7)

This table converts a release concentration to a main steam line monitor dose rate.

Decay MS 4(t Rel Conc Inverse Rounded Time uCi-hr uCi Dose Rate Dose Rate Dose Rate (hrs) cc-mR hr/mR mRlhr mRlhr 0.5 i.92E-02 I 3.25E-O1 5.92E-02 17 10 1 2.16E-02 I 3.25E-01 6.66BE-02 15 10 2 2,67E-02 / 3.25E-01 8.23E-02 12 10 5 4.93E-02 / 3.25E-O1 1.52E-01 7 0 10 1.27E-01 / 3.25E-01 3.91E,.01 3 0

20) 4.97E-01 I 3.25E-01 1.53E÷00 1 0 (Note 8) (Note 9)

Appendix B, Section 5.2.1.2 Note 2I -- from Note DFBi ODOM,

• 1E+06 pClluCi NoteS6- Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equal 200X ODCM total body dose rate of 500 mPtyr Note 3 - From ODCM, Appendix B, Table B.7-6 Note 7 -from SBC-709, page ~20 - Used SRV Note 4 -I1(365"24) or i/hours in a year flow as bounding (I.e., larger than ASDV).

Note 5- 1I(1.41E-5) Note 8 -from SBC-709, page 19 Note 9 - 1/Inverse Dose Rate RBv, 0 Page 4 Rev.0Pae 43/27/2008

V3 - Plant Vent and Main Steam Line Effluent Calculationsfor RG1IRS1/RAI Calculation of Radiological Effluent EAL Values EPCALC-06-02 Main Steam Line Effluent CalculaUion fora.Site.Area Emergency DFB1 DFBJ Avg Disp DF1R mrem-rn 3 mrem-m* Factor mrem-sec pCi-yr uCi-yr seclm3 u Cl-yr Isotope (Note 1) (Note 2) (Note 3)

Xe-t 38 8.83E-03 8.83E+03

• 3.40E-05 3.00E-02 Kr-87 5.92E-03 5.92E+'03 3.40E-06 2.01 E-02 Kr-88 1,.47E-02 1.47E+04

• 3AO0E--06 5.00E-02 Kr-85m I1,l7E-03 1.17E÷03

• 3AOE-06 3.98E-03 Xe-I135 1.81 E-03 1.812÷03 3.40E-06 6.1 5E-03 Xe-133 2.94E-04 2,94E+02

• 3.40E-06 1,002-03 Kr-85 1.61E-O5 1.61E+01

• 3.40E-06 5.47E-05 Xe-131 m 9.1 5E-05 9.15E+01 a 3.40E-06 3.112-04 Xe-133m 2.51E-04 2.512+02 3.40E-06 8.53E-04 Xe-135m 3.12E-03 3.12E+03

• 3.40E-06 1-06E-02 DFP= (per hour)

DFP1 (per year)

[ 1"iE-5m'rem-s~ec (Note 4) 1I(DF'1 (per hour)) S.L. release rate (Note 5) (Note 6)

S.L. release rate 8.1. Flowrate Release concentration (Note 7) "

This table converts a release concentration to a main steam line monitor dose rate.

Decay MS Ju, Rel Gone lnvelise Rounded Time uCi-hr uCi Dose Rate Dose Rate Dose R~ate (hrs) cc-mR Gc hrlmR mRlhr mPlhr 0.5 1 .922-02 2.852÷00 6.742-03 148 140 I

1 2,162-02 I 7.59E-03 132 130 2 2.672-02  ! 2,85E+00 9,35E-03 107 100 5 4.93E-02 I 2.852+00 1.73E-02 58 50 10- 1.272-01 I 2.85E+00 4.462-02 22 20 20 4.972-01 I 2.852+D0 1.752-01 6 0 (Note 8) (Note 9)

Note 1 - from ODOM, Appendix B, Section 5.2.1.2 Note 100 maR 6 - Dose (per NEIrate99-01 for 1 IC hour needed basis to equal discussion)

Note 2 - DFBi* 1E+06 p~iluCi Note 3 - From O0CM, Appendix B, Table 8.7-6 Note 7 -from SBC-709, page 20 - Used SRV Note 4 - 1/(36_5"24) or l/hours in a year flow as bounding (I~e., larger than ASDV),

Note 5 - 1/(1.41E-5) Note 8-*from SBC-709, page '19 Note 9 - 1/Inverse Dose Rate RoY. 0 Page 5 Rev.0 Pae 5312712006

.,. , . l, Ill IIIII III I Ill ..... . .. . ,, ,. n lU I I q II

• r -- -J* i... . ....

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!L *W, "J*V.3trC*PM'*!h*,and Main Steam Line Effluent Cacltosfor RGI/RS i/RA I Calculation of Radiological Effluent EAL Values EPCALC-06~.02 Main Steamh Line Effluent Caloulation for a General Emergency .- -

DF81 DFB* Avg Oisp DF' mrem-m* -,Factor torero-sec lsotope p Ci-yr uCi-yr secfma uCi-yr (Note I) (Note 2) (Note 3)

Xe-I 38 8.83a+03 3.402-06 3.0015-02

• i Kr-87 5.92E-03 5.922+03 3.40E-06 2,01E-02 Kr-88 1.47TE-02 1.47E*-04 3.402-06 5.OOE-02 Kr-85m 1.172-03 1.17E+03 ak 3.4(}E-06 3.98E-03 Xe-1 35 1.8lE-03 1.81E+03 3.402-06 6.162-03 Xe-B33 2.94E-04 2.94E'*02 ak 3.40E-06 I .OOE-03 Kr-8$ 1.61 E.05 I1.6IE-i01 3.40E-06 5.47E..05 Xe-tS1m 9.I5E-05 9.152÷01 ** 3A40E-06 3.11IE-04 Xe-I133m 2.512-04 2.5IE-.02 3.402-06 8.632-04 Xe-t 36m 3.12E-03 3.12E+03

• 3.40E-06 1.062-02 OF'1 (per hour)

DF', (per year)

(Note 4) 11I (DPa

'1 prhour)) SL. release rote (Note 5) (Note 6) 3.1. release rate Retease concentration (Note 7)

This table converts a release concentration to a xnain steam line monitor dose rate.

Decay MS Jm Ret Conic hwlers~e Rounded Time uCi-hr uiC! Dose Rate Dose Rate Dose Rate (hra) cc-mR Co hrlmiR mR/hr niRJhr 0.5 1.92E-02 / 2.S5E+O1 6.742-04 1433 1,480 I 2.162-02 I 2.85*-0I 7.59E-04 1318 1,310 2 2.67E-02 / 2.85E+01 9,38E-04 1066 1,060 4.03E.02 I 2.852+01 1.73-0 577 570 10 1,.27E-01 I 2.85E+01 4,462-03 224 220 20 4,97E-flI I 2.85E+01 1.75E-.02 57 80 (Note 8)* (Note 9)

Nate I from O0CM, Appendix B, Section 5.2.1.2 Nate 6 - Dose rate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> needed to equel Note 2 - DFSi *"12+013 pCiluCi 1000 mR (per NEI 99.01 IC basis dlscussion)

Note 3 - From 00CM, Appendix B, Table 8.7-6 Note?7 - from SBC-709, page 20 - Used SRV Note 4 -1/(365"24) or 1/hours in a year flow as bounding (ILe., larger than ASDV).

Note 5 - 11(1.412-5) Note 8 - from SBC-709, page 19 Note 9 - 1lInverse Dose Rate Rev. 0 312712005

... *- V3 - Pl~'

- =ant , andT il/l'ain Steam Line Effluent Calculationsfor RG1/RS1/RA1 Calculation of Radiological Effluent EAL Values EPCALC-06-02 Prepared by:. Seirlcea*Aayt

.... D te Reviewed by:.

David Robinson Chemistry Supervisor Date Approved by:

EP Mn Rev. 0 Page 7 Rev.0 Pae 73/27/2006

V4 - ODCM and Tech Spec Basis for Site Boundary Receptor Point for RG1/RS1IRAI1 7.0 RADIOACTIVE GASEOUS EFFLUENTS 7.1 Dose Rate CONTROLS C1.7 .1 .. h...... se. .. .T ... .t. ...... .. ,...,.,.. .......... . .... ...........e,1 APPICBILTY At~4alel time s.~~~Ui sT1ia pcfctoJue5~

ACTIONi4: tefo1iW With the dose rate(s) exceeding the above limits, decrease the release rate within 15 minutes to within the above limit(s).

SURVEILLANCE REQOU]REMENTS S.7.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in Part B of the ODCM.

S.7.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table A.7.1-I-.

A.7-1 A.7-1ODCM Rev. 25

V4 - ODCM and Tech Spec Basis for Site Boundary Receptor Point for RG1/RS1/RA1 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The Low Population Zone shall be as shown in Figure 5.1-2.

MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be shown in Figures 5.1-1 and 5.1-3, respectively.

The definition of UNRESTRICTED AREA used in implementing these Technical Specifications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a.

CONFIGURATION 5.2.1 The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 140 feet.

b. Nominal inside height = 219 feet.

c. Minimum thickness of concrete walls = 4 feet 6 inches.

d. Minimum thickness of concrete dome = 3 feet 6 inches.

e. Minimum thickness of concrete floor pad = 10 feet.

f. Nominal thickness of steel liner =1/4, 3/8, and 1/2 inch for the floor, wall, and dome, respectively.

g. Net free volume = 2.704 X 106 cubic feet.

SEABROOK - UNIT 1 ~ 5-1

" 0

-., . . . miir~ _',+

F-GRE . -jI-- '

SIEAN XLUINARABUNAP

=0*

SEABROK -UNITI 5-

V/5- UFSAR Table 12.3 Manipulator Crane and Spent Fuel Building Monitor Ranges TABLE 12.3-14 AREA RADIATION MONITORS (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr QTY. CLASS REFERENCE Containment Structure 6534 In-Core Instrument Seal Table GM 15 Non IE 12.3-2 (Note 4) 6535A, B Manipulator Crane GM 15 IE 12.3 -3 10-L104 2 (Note 1) (Note 4) 6536-1, 2 Personnel Hatch (Post-LOCA) Ion Chamber 2.5 1+-0 2Non IE 12.3 -3 6576A, B Containment (Post-LOCA) Ion Chamber 25 100-108 r/hr 1E 12.3-3 Primary Auxiliary Building 6537 Sampling Room GM 2.5 10"l.104 2 Non 1E 12.3-6 6538, 6539 RHR Pump Area GM >100 Non 1E 12.3-4 10-1.104 6540 Volume Control Tank Area Ion Chamber 8x104o 2 Non 1E 12.3-7 101-0 6541 Lower Level GM 2.5 Non 1E 12.3-5 10'-1.04 6543 Entrance GM >100 Non 1E 12.3-5 10'-1.104 6544 Entrance GM 2.5 3 Non 1E 12.3-6 10"'-.104 6545, 6546 Charging Pump Room GM 110 Non 1E 12.3-5 6547

V5 - UFSAR Table 12.3-14 - Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE 6508-1,2 PAB-HRAM Ion Chamber >100 10"2-104 r/hr 2 Non 1E 12.3-5 6563-1,2 PAB-HRAM Ion Chamber >100 10-2-104 r/hr 2 Non 1E 12.3-5 6517-1,2 RHR - Pump Vault HRAM Ion Chamber >100 10- -104 r/hr 2 Non 1E Fuel Storage Building 6549 Spent Fuel Pool Area GM 10-1-104 Non 1E 12.3-15 2.5 1 6518 Spent Fuel - HRAM Ion Chamber 2.5 10"2-104 r/hr 1 Non 1E 12.3-15 Control Room 6550 Main Control Board Area GM 0.5 10"1-104 Non 1E 1.2-32 Waste Processing Building 6551 Waste Gas Processing Area GM 2.5 1 Non 1E 12.3-11 10q.0 6552 Truck Loading Area GM 1.5 1 Non 1E 12.3-10 6553 Radwaste Control Room GM 0.5 10-t.104 1 Non 1E 12.3-10 6554 Waste Management Control Panel GM 5 1 Non 1E 12.3-10 10-1.10s Area 6570 Extruder/Evaporator Manifold Area GM (Note 6) 1 Non 1E 12.3-10 10°.0 6571 Compacted Rad Waste Storage Area GM 2.5 1 Non 1E 12.3-10

V5 - UFSAR Table 12.3-14 - Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Administration & Service Buildine 6555 Hot Chemistry Laboratory GM 0.5 10-1-104 1 Nonl1E 12.3-16 6556 Decontamination Room GM 0.5 10"l-104 1 NonlIE 12.3-16 6557 RCA Shlop (Note 2) GM 0.5 1"-0 1 Non 1E 12.3-16 6558 RCA Personnel Decontamination GM 0.5 10"I-104 1 Non lE 12.3-16 Area 6559 RCA Women's Locker Room GM 0.5 10-1-104 1 Non lE 12.3-16 (Note 1) 6535-A and 6535-B will automatically terminate containment purge in the event of high radiation during fuel handling operations.

(Note 2) RCA - Radiologically Controlled Area.

(Note 3) Deleted.

(Note 4) GM - Geiger-Mueller.

(Note 5) Radiation monitoring setpoints are varied during operation to follow station conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the set-points is found in the station operating procedures.

(Note 6) >100

V6 - UFSAR Section 9.1.4 Descriptiion of the Refueling Pathway JSTATJON  :'7;{"::[-i.".". ::;.i*:::z}:". .Fu:l el Storage and Handling .:.. .. ,.:]Sectionr 9.1.:ii.;.;.-*:..

9.1.4 Fuel H:andlina

. . System The Fuel Handling System (FHGS) consists of equipment and structures used for the refueling operation in a safe manner meeting General Design Criteria 61 and 62 of 10 CFR 50, Appendix A.

9.1.4.1 Design Bases

a. The primary design requirement of the equipment is reliability. A conservative design approach is used for all load-bearing parts. Where possible, components are used that have a proven record of reliable service. Throughout the design of equipment in containment, consideration is given to the fact that the equipment will spend long idle periods stored in an atmosphere of 120°F and high humidity.

b. Fuel handling devices have provisions to avoid dropping or jamming of fuel assemblies during transfer operation.

c. Handling equipment used to raise and lower spent fuel has a limited maximum lift height so that the minimum required depth of water shielding is maintained.

d. The Fuel Transfer System (FTS), where it penetrates the containment, has provisions to preserve the integrity of the containment pressure boundary.

e, Criticality during fuel handling operations is prevented by geometrically safe configuration of the fuel handling equipment.

f. Handling equipment will not fail in such a manner as to damage seismic Category 1 equipment or spent fuel in the event of a Safe Shutdown Earthquake.

g. Except as specified otherwise in this document, the crane structures are designed and fabricated in accordance with CMAA Specification No. 70 for Class A-i service.

h. The static design load for the refueling machine crane structure and all its lifting components is normal, dead and live loads, plus three times the fuel assembly weight with a Rod Cluster Control Assembly.

i. The original design allowable stresses for the refueling machine structures and components supporting a fuel assembly are as specified in the ASME Code, Section 111, Subarticle XVII-2200. Allowable stress criteria for rated loads for the spent fuel pool bridge and hoist, cask handling crane, and polar gantry crane are in accordance with CMvAA-70. Modifications to the refueling machine components/structures meet the allowable stress limits per the ATSC Manual of Steel Construction, 9th edition (Allowable Stress Design).

j. The design load on wire rope hoisting cables does not exceed 0.20 times the average breaking str'ength. Two cables are used in the refueling machine and each is assumed to carry one half the load.

V/6- UFSAR Section 9.1.4 Descriptiion of the Refueling Pathway SEABROOK.' A~ Revision 15 STATON Fel Sorag-andHandingSection91 J:.r.A].:, " P-,ge2:4

k. A single finger on the fuel gripper can support the weight of a fuel assembly and Rod Cluster Control Assembly without exceeding the requirements of Item i.

above.

1. All components critical to the operation of the equipment are located so that parts which can fall into the reactor are assembled with the fasteners positively restrained from loosening under vibration.

m. The inertial loads imparted to the fuel assemblies or core components during handling operations are less than the loads Which could cause damage.

n. Physical safety features are provided for personnel operating handling equipment.

Industrial codes and standards used in the design of the fuel handling equipment:

a. Applicable sections of CMAA Specifications No. 70.

b. New Fuel Elevator Hoist: Applicable Sections of HLMI-l00 and ANSI 1330.16.

c. Structural: ASME Code, Section 111, Appendix XVII, Subarticle XVT[.-2200 (Refueling Machine).

d. Electrical: Applicable standards and requirements of the National Electrical Code and NFPA No. 70 are used in the design, installation, and manufacturing of all electrical equipment.

e. Materials: Main load-bearing materials conform to the specifications of the ASTM standard.

f. Safety: OSHA Standards 29 CFR 1910, and 20 CFR 1926 including load testing requirements; the requirements of ANSI Ni18.2, Regulatory Guide 1.29 and General Design Criteria 61 and 62.

Protection of the FHS fr'om wind and tornado effects is discussed in Section 3.3. Flood protection is discussed in Section 3.4. Missile protection is discussed in Section 3.5.

9.1.4.2 System Description The Fuel Handling System (FHS) consists of the equipment needed for the refueling operation on the reactor core. iBasically this equipment is comprised of fuel assembly, core component and reactor component hoisting equipment, handling equipment and a Fuel Transfer System (FTS).

The structures associated with the fuel handling equipment are the refueling cavity, the refueling canal in Containment and in the FSB, and the fuel storage area.

The elevation and arrangements drawings of the fuel handling facilities are shown on Figure 1.2-15, Figure 1.2-16, Figure 1.2-17, Figure 1.2-18, Figure 1.2-19, Figure 1.2-20 and Figure 1.2-21.

V7.- Table HI Procedure References Supporting Plant Areas Identified in RA3/HA5 TARLE Hi - AREAS REOUIRING ACCESS FOR NORMAL OPERATIONS. COOLDOWN OR SHUTDOWN STRUCTURE OR ROOM AREA MODES APPLICABLE PROCEDURE STEP(s)

Primary Auxiliary Buiding 25 ft elevation 1, 2, 3, 4 051000.03, steps 2.3.8, 2.3.9, 7 ft elevation 4.29.4

-26 ft elevation 051000.04, steps 4.1.7, 4.1.11, 4.1.12, 4.1.13, 4.1.16, 4.1.21, 4.1.22, 4.1.23, 4.2.31, 4.2.8.1,

_______________________4.2.16, 4.3.10 Turbine Building 1, 2, 3 051000.03, steps 2.3.3, 2.3.6, 4.2, 4.8, 4.26, 4.27, 4.28, 4.47 051000.04, steps 4.1.6, 4.1.8, 4.1.9, 4.1.15, 4.1.24, 4.1.26, 4.3.45 Switchgear Rooms Essential 1, 2, 3, 4 051000.03, steps 2.3.3, 4.42, 4.41 Non-essential 051000.04, steps 4.1.5, 4.1.28, 4.2.18, 4.2.24 Steam and Feedwater Chases 1, 2, 3 051000.03, steps 2.3.5, 4.20 051000.04, steps 4.1.19, 4.3.29, 4.3.34 Waste Processing Building 25 ft elevation 1, 2, 3 051000.04, step 4.1.2 3 ft elevation

-31 ft elevation Containment 3, 4 0S1000.04, steps 4.1.14, 4.3.29, 4.3.30, 4.3.34 Equipment Vaults 3, 4 0S1000.04, steps 4.2.28, 4.3.16,

___ ____ ___ ___ __ ___ ___ ___ ___ ___ ___ ___ 4.3.30 Procedure references:

051000.03 - Plant Shutdown from Minimum Load to Hot Standby 051000.04 - Plant Cooldown from Hot Standby to Cold Shutdown

V8 - UFSAR Table 12.3-15 WRGM Ranges for RU1 TABLE 12.3-15 AIRBORNE RADIATION MONITORS (SKID MOUNTED DETECTORS)

(Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET LOOP TAG NO. DETECTOR REFERENCE GRD. LOW-HIGH POINT IEEE DIAG. P&ID DETECTOR ENERGY RE- DESCRLIPTION TYPE ISOTOPE mr/hr oCi/cc ttCi/cc OTY. CLASS LEVEL 1-NHY I-NHY 3 37 6526-i Containment Air Beta i' ' Cs' 2.5 Non 1E Note 3 506135 20504 Particulate (Note 4) 33 6526-2 Containment Radiogas Beta Xe' 2.5 i Non 1E Note 1 506135 20504 33 6528-i Plant Vent- WRGM Beta Xe' , I(.r5 2.5 I Non 1E Note 1 506607 20494 (Low Range) 33 6528-2 Plant Vent - WRGM Beta Xe' , Kr85 2.5 1 Non IE Note 1 506607 20494 (Mid Range) 33 3 6528-3 Plant Vent- WRGM Beta Xe' , Kr' 2.5 01-3_7,,r 10 1 Non 1E Note I 506607 20494 (Hi Range) 33 6495 WRGM Backup Ion Chamber Xe' , KI-8 2.5 1 Non 1E Note i 506607 RM-20509 33 506885 6531-2 WPB Radiogas Beta Xe' 2.5 10"7-10-* Non 1E Note I 20498 6532-2 PAB Radiogas Beta 10~-7.0* 506598 20510 2.5 Non lE Note 1 6548 Containment Radiogas Beta 2.5 10*-102 Non 1E Note 1 506136 RMv-20510 Backup

V8 - UFSAR Table 12.3-15 WRGM Ranges for RU1 SEAJIROOK RADIATION PROTECTION Revision: 13 STATION UFSAR TABLE 12.3-15 Sheet: 2 of 2 NOTES Predominant Gamma Energy Max. Beta Energy Note Isotopes 1 Xe13 3 0.346 0.08 1 Xe 13 5 0.92 0.249 Kr85 0.67 0.5 14 Kr85m 0.82 0.150 i131 0.606 0.3 64 i133 1.27 0.53 CS13 4 0.662 0.604 Cs 137 0.5 14 0.662 Co 58 0.474 0.81 Co 60 0.3 14 1.17, 1.33 Same as Note 2, plus:

Rb 88 5.3 1.863 4 Containment air particulate monitor functions as a leakage detector and must survive the SSE (reference Regulatory Guide 1.45 and Standard Review Plan 5.2.5).

5 Radiation monitoring setpoints are varied during operation to follow station operating conditions. Setpoints are maintained within the bounds established in the Technical Specifications.

The methodology for establishing the setpoints is found in the station operating procedures.

V/9 - OS 1000.09 Procedure References to Refueling Pathway Level Indicators for RU2 Figure 4: Monitoring RCS Inventory When in MODEs 5 or 6 (Sheet 2 of 2) 6.2 Changes in any of the above should be noted, as well as any changes in makeup needs. This will enable the operator to detect abnormal makeup once steady state conditions are obtained.

6.3 Changes in containment or fuel storage building pressures can result in sluicing between the two buildings and subsequent level changes. An approximate rule of thumb to determine sluicing is a change of 11/16" in the spent fuel pool level for every 1" change in refueling cavity level

(.690 gallons).

7.0 Preferred and alternate indications for monitoring primary system inventory are listed below.

PARAMETER PREFERRED CHANNEL ALTERNATE CHANNEL Pressurizer level A0390 (RC-LI-462) A0332 (RC-LI-45 9)1 A0333 (RC-LI-460)'

A0334 (RC-LI-461 )'

Reactor Vessel Level A03 82 (RC-LI-9405) TYGON 2 (Non-midloop) RC-XX-73 15-1 RC-XX-73 15-4 RC-LI- 131 1 RC-LI-1321 SHUTDOWN COOLING COLOR GRAPHIC Reactor Vessel Level A0298 (ULTRASONIC) A03 82 (RC-LI-9405)

(Mid-loop) A0299 (ULTRASONIC) TYGON 2 SHUTDOWN COOLING COLOR GRAPHIC VCT Level A0624 CS-LI-185 CS-LR-1 85 CS-LI-i 1122 Refueling Cavity Level A0382 (RC-LI-9405) A 1186 (SF-LI-2629)

(Cavity not full) A0390 (RC-LI-462)

Refueling Cavity Level A0928 (SF-LI-2629- 1) Al 1186 (SF-LI-2629)

(Cavity full) A03 90 (RC-LI-462)

Spent Fuel Pool Level Al1185 (SF-LI-2607) SF-LE-26073 RC-LI-459, 460 and 461 are not cold calibrated. Level changes will appear greater than actual at cold shutdown temperatures.

2 Local indications.

3 Holes on SF-LE-2607 are 3" apart, with the top 1/2" hole (about 1' below floor level) at elevation 25' 10 1/2" (relative to reactor vessel flange). Holes are 3" apart, with 1/2" holes every 1', and 1/4" holes every 3".

OS1000.09 Rev. 29 Page 42 of 48

VlO - UFSAR Table 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges TABLE 12.3-14 AREA RADIATION MONITORS (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Containment Structure 6534 In-Core Instrument Seal Table GM 15 10-'-10 4 INon IE 12.3-2 (Note 4) 6535AB Manipulator Crane GM 15 10'-104 1E 12.3-3 (Note 1) (Note 4) 6536-1, 2 Personnel Hatch (Post-LOCA) Ion Chamber 2.5 1+-0 Non 1E 12.3-3 2

6576A, B Containment (Post-LOCA) Ion Chamber 25 10°-10% /hr 2 IE 12.3-3 Primary Auxiliary Building 2

6537 Sampling Room GM 2.5 Non 1E 12.3-6 6538, 6539 RHR Pump Area GM >100 10-110 4 Non 1E 12.3-4 2

6540 Volume Control Tank Area Ion Chamber 8x10 4 101.107 Non lE 12.3-7 6541 Lower Level GM 2.5 10-lI04 Non 1E 12.3-5 6543 Entrance GM >100 Non IE 12.3-5 6544 Entrance GM 2.5 Non 1E 12.3 -6 10'-l.04 6545, 6546 Charging Pump Room GM 110 3 Non 1E 12.3-5 6547

VIO - UFSAR Table 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE 6508- 1,2 PAB-HRAM Ion Chamber >100 10"2-104 r/hr 2 Non 1E 12.3 -5 6563-1,2 PAB-HRAM Ion Chamber >100 10-2-104 rlhr 2 Non 1E 12.3 -5 65 17-1,2 RHR - Pump Vault HRAM Ion Chamber >100 I10~-2-10 r/hr 2 Non 1E Fuel Storage Building 6549 Spent Fuel Pool Area GM 2.5 10.-104o 1 Non lE 12.3-15 6518 Spent Fuel - HIRAM Ion Chamber 2.5 10-2.104 r/hr 1 Non 1E 12.3-15 Control Room 6550 Main Control Board Area GM 0.5 10-10 1Non 1E 1.2-3 2 Waste Processing Building 6551 Waste Gas Processing Area GM 2.5 10".0 Non 1E 12.3-11 6552 Truck Loading Area GM 1.5 Non IE 12.3-10 6553 Radwaste Control Room GM 0.5 Non 1E 12.3-10 1

6554 Waste Management Control Panel GM 5 10-1.104 Non 1E 12.3-10 Area 6570 Extruder/Evaporator Manifold Area GM (Note 6) 10o.105 Non 1E 12.3-10 6571 Compacted Rad Waste Storage Area GM 2.5 10~-l.04 Non 1E 12.3-10

VIGO - UFSAR Table 12.3-14 Manipulator Crane and Spent Fuel Building Monitor Ranges (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE-- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Administration & Service Building 6555 Hot Chemistry Laboratory GM 0.5 10-1-104 1 Non 1E 12.3-16 6556 Decontamination Room GM 0.5 10"l104 1 Non IE 12.3-16 6557 RCA Shop (Note 2) GM 0.510I0 1 Non IE 12.3-16 6558 RCA Personnel Decontamination 0.5 10"1-104 1 GM Non IE 12.3-16 Area 6559 RCA Women's Locker Room GM 0.5 10-1-104 1 Non 1E 12.3-16 (Note 1) 6535-A and 6535-B will automatically terminate containment purge in the event of high radiation during fuel handling operations.

(Note 2) RCA - Radiologically Controlled Area.

(Note 3) Deleted.

(Note 4) GM - Geiger-Mueller.

(Note 5) Radiation monitoring setpoints are varied during operation to follow station conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the setpoints is found in the station operating procedures.

(Note 6) > 100

CL ~ ii~E!~

IE~ i2 !12! ~ 3 VllI- EP CALC 06-04 RVLIS Values for CG1/CS1I/CA1I 0

FPL- Energy Seabrook Station SEP# 2006021

SUBJECT:

Calculation of RVLIS Values for Emergency Action Levels FROM: D. L. Young DATE: May 8, 2006 TO" File Emergency Initiating Conditions CAl, CS1 and CG1 use Reactor Vessel Level Instrumentation System (RVLIS) readings in their associated Emergency Action Levels (EALs). These RVLIS values were calculated as follows.

Assumptions and Bases

1. These Initiating Conditions are evaluated in Mode 5. It was assumed that the accident was initiated in Mode 5, and that no prior accident had occurred in a higher Mode (i.e.,

there is no adverse containment).

2. Since containment conditions are normal, a RVLIS value without uncertainties was utilized.

3. Per Westinghouse Electric Corporation transmittal to Public Service Company of New Hampshire, Seabrook Station Units 1 and 2, Incore Thermocouples, NAH-3203 S.O. No.

NAH-4755/NAH-2.2.389, dated 01107/87 (film location, roll #715, frame #1081)

• From Attachment 2, a RVLIS full range indication of 66%, without uncertainties, is the mid:dle of the hot legs.

• From Attachment 2, a RVLIS full range indication of 55%, without uncertainties, is the top of the core.

• From Attachment 2, a RVLIS full range value of 35% equals 3.5 feet above the bottom of the core while a RVLIS full range value of 45% equals 8.5 feet above the bottom of the core. Therefore the relationship between RVLIS full range scaling and height in the reactor vessel is 2% per foot.

4. Per Procedure OS1000..12, Operation With RCS At Reduced Inventory/Midloop Conditions

• The radius of the hot legs (i.e., center to top/bottom) is 14.5".

• -71 inches equals the top of the hot leg nozzles.

• -85.5 inches equals the hot leg midplane

• -100 inches equals the bottom of the hot leg nozzles.

I II II II I IIII I III I II I II I I ill l

VII - EP CALC 06-04 RVLIS Values for CG1/CS1/CAI Calculation of RVLIS Values for Emergency Action Levels Page 2 Calculations

1. The RVLIS full range value at the bottom of th hot Jeaqs is 66% - (14.5"/12"ift)

• 2%/ft =

63.6%. This value is rounded to 64%.

2. The RVLIS full range value at 6" below t~he bottom of the hot leecis is 66% -((14.5" +

0 6"Y1 2"/ft)

• 2 0/ ft = 62.6%. This value is rounded to 63%.

3. As stated previously, the RVLUS full range value at the ton of the core is 55%.

The above calculation is EPCALC-06-04.

Prepared by:. _______________

David L Young Senior Nuclear Analyst Approved by: *J" Robert L Couture Principal Engineer - Nuclear DLY/dly cc:

IIIIIIII I I II II

Vll - EP CALC 06-04 RVLIS Values for CG1ICS1ICA1 RVLIS LEVEL VESSEL LEVEL

(%) (inches from vessel flange)

-108 119.8" 100 81.3" 90 31.8" 80 -17.7" 70 -67.2"

• 64 -96.9"

• 63 -101.9"

• 60 -116.7"

• 55 -141.5" RC-LI-9405, RC-LIT-9467, and

• 50 -166.2" the Tygon Tube do not indicate reactor vessel level when actual

• 40 -215.7" level is less than -95" due to the 30 -265.2" weir on the RCP discharge.

' 20 -314.7"

• 10 -364.2"

• 0 -413.7"

V12 - Manipulator Crane Monitor Ranges for CG1/CS1 TABLE 12.3-14 AREA RADIATION MONITORS (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE- DESCRIPTION TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE Containment Structure 6534 In-Core Instrument Seal Table GM 15 Non IE 12.3 -2 (Note 4) 10'-1.04 6535A, B Manipulator Crane GM 15 1E 12.3-3 2

(Note 1) (Note 4) 6536-1, 2 Personnel Hatch (Post-LOCA) Ion Chamber 2Non IE 12.3-3 Containment (Post-LOCA) Ion Chamber 25 100 -10s /hr 12.3-3 6576A, B 1E Primary Auxiliary Building 6537 Sampling Room GM 2.5 2 Non 1E 12.3 -6 6538, 6539 RHR Pump Area GM >100 Non 1E 12.3-4 6540 Volume Control Tank Area Ion Chamber 8X10 4 101.107 2 Non IE 12.3-7 6541 Lower Level GM 2.5 10-.0 1 Non 1E 12.3-5 4

6543 Entrance GM >100 10-1-10 Non 1E 12.3-5 1

6544 Entrance GM 2.5 10-1.104 Non IE 12.3-6 6545, 6546 Charging Pump Room GM 110 t 0-1.10 Non 1E 12.3-5 6547

V12 - Manipulator Crane Monitor Ranges for CG1/CS1 (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE..._- TYPE mr/hr mr/hr mr/hr OTY. CLASS REFERENCE DESCRIPTION 6508-1,2 PAB-HRAM Ion Chamber >100 10"z- 104 r/hr 2 Non 1E 12.3-5 6563-1,2 PAB-HRAM Ion Chamber >100 10"2-104 r/hr 2 Non IE 12.3-5 6517-1,2 RHR - Pump Vault HRAM Ion Chamber >100 10-2-104 r/hr 2 Non 1E Fuel Storage Building 6549 Spent Fuel Pool Area GM 2.5 1 Non 1E 12.3-15 6518 Spent Fuel - HRAM Ion Chamber 2.5 102-. 10 4 r/hr 1 Non 1E 12.3-15 Control Room 6550 Main Control Board Area GM 0.5 1--0 Non 1E 1.2-32 Waste Processing Building 6551 Waste Gas Processing Area GM 2.5 10-1.104 Non IE 12.3-Il 6552 Truck Loading Area GM 1.5 10-1.104 Non 1E 12.3-10 6553 Radwaste Control Room GM 0.5 10'-1.04 Non 1E 12.3-10 6554 Waste Management Control Panel GM 5 10'-l.04 Non 1E 12.3-10 Area 6570 Extruder/Evaporator Manifold Area GM (Note 6) 10°.105 Non 1E 12.3-10 6571 Compacted Rad Waste Storage Area GM 2.5 10'-l.04 Non 1E 12.3-10

V12 - Manipulator Crane Monitor Ranges for CG1/CS1 (Note 5)

DETECTOR ALARM INSTRUMENT BACK- RANGE SET UFSAR TAG NO. DETECTOR GRD. LOW-HIGH POINT DETECTOR IEEE FIGURE RE_..- DESCRIPTION TYPE mr/hr mr/hr mr/hr CLASS REFERENCE Administration & Service Building 6555 Hot Chemistry Laboratory GM 0.5 10-1.104 Non 1E 12.3-16 1

6556 Decontamination Room GM 0.5 10'-i.10 Non 1E 12.3-16 6557 RCA Shop (Note 2) GM 0.5 10'-.104 1 Non 1E 12.3-16 6558 RCA Personnel Decontamination GM 0.5 10'-l.04 Non 1E 12.3-16 Area 6559 RCA Women's Locker Room GM 0.5 Non 1E 12.3-16 (Note 1) 6535-A and 6535-B will automatically terminate containment purge in the event of high radiation during fuel handling operations.

(Note 2) RCA - Radiologically Controlled Area.

(Note 3) Deleted.

(Note 4) GM - Geiger-Mueller.

(Note 5) Radiation monitoring setpoints are varied during operation to follow station conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the setpoints is found in the station operating procedures.

(Note 6) > 100

V13 - UFSAR Section 6.2 Description of Containment Sumps for CG1/CS 1/CA1/ICU1 SEABROOK ENGINEERED SAFETY FEATURES Revision 15 STATION Containment Systems Section 6.2 UFSAR Page 51

g. Net Positive Suction Head Reqiuirements Adequate net positive suction head is assured by locating the RH-R and CBS pumps at the lowest level in the Auxiliary Building. The RHR and CBS pump available net positive suction heads from the containment sump were determined by assuming the limiting conditions in accordance with NRC Regulatory Guide 1.1 (pressure equal to atmospheric and temperature equal to 21 2°F). The CBS pump available and required net positive suction head is shown in Table 6.2-75.

The RHR pump available and required net positive suction head is shown in Table 6.3-1.

h. Heat Exchanger Surface Fouling The materials used for the CBS heat exchanger are listed in Table 6.1(B)-I. The shell side of the heat exchanger is cooled by the PCCW system which contains a corrosion-inhibiting agent and operates as a closed system. The tube side, which is in contact with the emergency core coolant, is corrosion resistant. The effect of corrosion fouling on heat exchanger surfaces will, therefore, be minimal; however, fouling factors were included in the detailed design of the units to assure the required heat removal capability through conservative design.

i. Heat Exchanger Performance The heat exchanger (residual heat removal, containment spray, primary component cooling water) temperatures have been selected based upon maximum service water (ultimate heat sink) temperatures and the amount of heat removal required. The flows, geometry, and surface area were studied in the evaluation of the containment pressure-temperature analysis. Those parameters selected and tabulated in Table 6.2-76 are those which meet the design basis requirement for containment cooling.

j. Containment Recirculation Sump and Strainer Design The containment recirculation sump collects and strains the water available for supplying the residual heat removal, containment spray, safety injection and high head charging pumps during the recirculation mode of operation following an accident. The sump is designed to meet the intent of Regulatory Guide 1.82.

Two completely independent sumps are located in the containment to maintain the "double train" concept as described in Subsection 6.2.2.2d.

One sump supplies water to Train A and the other sump supplies Train B. The arrangement of these sumps is shown in Figure 6.2-79. The minimum water level in containment during a loss-of-coolant accident is nominally Elevation (-)23.79 ft.

V13 - UFSAR Section 6.2 Description of Containment Sumps for CG1ICS1ICA 1/CU1I SEABROOK ENGINEERED SAFETY FEATURES IRevision 15 STATION Containment Systems Section 6.2 UFSAR jPage 52 A series of debris interceptors are provided on the containment floor within the recirculation flow paths. The debris interceptors reduce the quantity of debris transported to the sumps by trapping debris and allowing the remaining debris more time to settle prior to reaching the sumps.

Heavy particles are prevented from reaching the sumps by sloping the surrounding floor away from the sumps. This facilitates settling of debris on the floor prior to reaching the sump area.

A strainer is installed in each sump. Each strainer consists of rows of vertically oriented strainer panels, consisting of a framework sandwiched between two sets of wire cloth attached to perforated plates. The maximum hole size of the perforated plates and maximum width of a gap between bolted structures is 0.068 inches. The strainer would therefore prevent debris particles 0.068 inches or greater in diameter which may be generated following a large break LOCA from passing through or bypassing the strainer and entering the ECCS system. The minimum physical restriction in the ECCS flow path consists of 0.073 inches, which is the effective opening of the fuel assembly debris filter bottom nozzle in combination with the P-grid. Therefore, the strainer will prevent recirculation of debris particles of sufficient size to impede cooling flow to the core.

The strainer panels are mounted on a plenum structure within the sump. The plenum is sealed to the sump floor and at the sump wall adjacent to the ECCS pipe inlet to ensure that all water entering the sump passes through the strainer panels. Water is drawn through the strainer panels and plenum and into the lower portion of the sump.

The strainer will also act as a vortex preventor to further preclude air intrusion into the ECCS piping.

The strainers have been designed to accommodate the debris generated and transported to the sump during the recirculation phase of a LOCA. The head loss due to debris on the strainer is less than the available NPSH margins for operating ECCS pumps, thereby ensuring that cavitation of the ECCS pumps will not occur.

Therefore, the design meets the intent of Regulatory Guide 1.82.

The potential for clogging of the sump strainers by equipment and piping insulation or loose insulation in the containment is minimized by the type of insulation used.

V13 - UFSAR Section 6.2 Description of Containment Sumps for CG1/CS 1/CA 1ICU1 SEABROOK ENGINEERED SAFETY FEATURES Revision 15 STATION Containment Systems Section 6.2 UFSAR Page 53 The thermal insulation inside the containment for piping and equipment except the reactor pressure vessel is fiberglass blanket insulation of the type commercially known as Nukon, manufactured by Owen's-Coming Fiberglass. The outside surface of the insulation blankets is covered with a stainless-steel jacket or is encapsulated in stainless steel wire mesh. Nukon is consistent with the recommendations of Regulatory Guide 1.36. The reactor pressure vessel is insulated with stainless-steel reflective insulation or fiberglass blanket.

Clogging of the strainers by nonsafety-related equipment is unlikely due to the remote location of the sumps relative to the NNS equipment and physical barriers separating the sumps from other areas in the containment. The supplementary neutron shielding around the reactor vessel which could be displaced by blowdown forces during an accident is designed to remain anchored and intact; hence, it is not a potential source of strainer blockage during an accident.

The design of the sump suction piping ensures that adequate flow and net positive suction head are available to all pumps under the most limiting containment conditions, as required by Regulatory Guide 1.1. The two sumps and the pumps they service are designed so that any single active or passive failure will not cause the loss of both A and B Train components.

The sumps are visually inspected on a periodic basis to assure that they are clean, free of debris and that all strainers are intact and in position. The containment sump line isolation valves are exercised periodically to assure operability within Technical Specification requirements.

k. Periodic Testing The provisions for periodic testing and inspection of the containment spray system are discussed in Subsection 6.2.2.4.

Applicable Codes, Standards and Guides The codes, standards and guides applicable to the containment spray system are summarized in Table 6.2-79.

m. Remote Manual Operation of the Containment Building Spray System The CBS system is designed to function completely automatically under accident conditions, hence there are no operations which must be performed manually by the operator from the main control board to initiate the proper function of the system during an accident. After the suctions of the CBS pumps are automatically switched over from the RWST to the containment recirculation sumps, the isolation valves (CBS-V2, CBS-V5) in the discharge line from the RWST will be closed by the operator.

V14 - SAMG Calculation Aid CA-3 for Hydrogen Flammability for CG1.2clContainment Potential Loss Number Title Revision CA-3 HYDROGEN FLAMMABILITY IN CONTAINMENT Rev. 0 Purpose The purpose of CA-3 is:

• Define whether the hydrogen in the containment atmosphere is flammable, and

• Estimate the hydrogen concentration in the containment atmosphere based on an estimated oxidation percentage.

Assumptions

• This aid is valid for scenarios where the containment fans are on or off.

• Hydrogen released into containment is created with the reactor core through zirconium oxidation.

• 6% hydrogen concentration by volume is defined as the minimum concentration that will sustain a global burn.

• The hydrogen burn is based on a simplified adiabatic isochoric complete combustion for hydrogen.

• No hydrogen ignitors are considered and no previous bumns are assumed to have occurred.

• All energy released from the bumn is absorbed by the containment atmosphere as increased internal energy.

• The aid assumes the air, steam, and hydrogen within the containment atmosphere are thoroughly mixed at 100% humidity and all behave as ideal gases.

• The air volume is based on initial conditions, plus 10% to account for instrument air leakage, and remains constant throughout each scenario.

• During the core/concrete scenario, the core is assumed to be ex-vessel and the containment atmosphere is superheated.

• During the core/concrete scenario, no additional noncondensable, noncombustible gases are released while additional hydrogen is released.

• For the venting scenarios, the air, steam, and hydrogen are released in the same ratio as they exist in containment when venting takes place.

• When combustion occurs, all hydrogen is assumed to be consumed.

• Expected containment failure has been defined at the pressure at which there is a 5%

probability of containment failure, minus 10 psi.

Additional Data The figures displayed in this aid are for a wet hydrogen measurement only. Wet measurement is chosen because the post accident sampling is performed with steam in the sample. For sampling cases above 300 °F, or cases where heat tracing is lost, condensation in the analyzer will occur and conditions will approach that of a dry measurement where a higher concentration is read resulting in the appearance of a more limiting condition.

Page 2 of 8

V15 - Tech Spec Table 1.2 Cold Shutdown Temp Limit for CU3/CA3.1I TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

SFCP In accordance with the Surveillance Frequency Control Program TABLE 1.2 OPERATIONAL MODES REACTIVITY  % RATED AVERAGE COOLANT MODE CONDITION~keff THERMAL POWER* TEMPERATURE

1. POWER OPERATION > 0.99 > 5% > 350°F

2. STARTUP > 0.99 _<5% > 350°F

3. HOT STANDBY < 0.99 0 > 35 0oF

4. HOT SHUTDOWN < 0.99 0 3500 F > Tavg >2000°F

5. COLD SHUTDOWN < 0.99 0 < 2000 F

6. REFUELING** < 0.95 0 < 140°F

• Excluding decay heat.

• • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

SEABROOK - UNIT 11-Amnet14 1-8 Amendment 141

Vlti - i-roceciure U:57000.CMJ De~im~tIon 0? Kecluced Inventory tar CA3 Figure 3: Shutdown and Refueling Related Definitions (Sheet 1 of 2)

1. CORE ALTERATIONS A CORE ALTERATION is defined in Technical Specifications as definition 1.9.

Conservation decision making during refueling equipment problems must be used to ensure Technical Specification compliance is maintained.

OE34388 - Non-conservative Implementation of a CORE ALTERATION During core offload activities, a fuel handling equipment failure resulted in a fuel assembly being suspended, motionless in the core region for several days time. The station initially concluded a CORE ALTERATION was not in progress based on no assembly movement and secured the SRO position providing oversight of fuel handling activities. Subsequently, the station determined the subject condition should have been treated as a CORE ALTERATION since the fuel assembly remained in the core region and the subject move sheet had been started but not completed. In this case, SRO oversight of the bundle should have been maintained per 10CFR50.54.

2. RCS Loops Filled Technical Specifications T.S. 3.4.1.4.1 and T.S. 3.4.1.4.2 contain different restrictions in MODE 5 depending on RCS loops filled or RCS loops not filled. See Technical Specification Bases 3/4.4.1, Reactor Coolant Loops and Coolant Circulation for all conditions needed to declare the RCS "Loops filled."

3. Reactivity Additions Transferring water to the RCS, refueling cavity, refueling canal, transfer canal, or spent fuel pool that is lower in boron concentration is acceptable provided that the boron concentration is greater than the refueling boron concentration requirement. Likewise, transferring water to the RCS, refueling cavity, refueling canal, transfer canal, or spent fuel pool that is lower in temperature (down to 50°F) than the water contained in those volumes is also acceptable.

4. RCS Reduced Inventory Condition As discussed in NRC Generic Letter 88-17, an RCS reduced inventory condition exists whenever the RCS water level is lower than negative 36 inches (as measured from the reactor vessel flange).

OS1000.09 Rev. 29 Page 39 of 48

V17 - Site Specific EAL Design Basis for 25 PSIG RCS Pressure Range for CA3.2 Per the EAL design Basis, The wide-range RCS pressure transmitters have *arange of 0 to 3,000 psig. The main control boards have two post-accident monitoring qualified meters, one for each wide-range RCS pressure transmitter. These meters have major divisions at 100 psig intervals and minor divisions at 50 psig intervals. Since it is possible to read the approximate mid-point between minor divisions, the value is set to 25 psig.

V18 - UFSAR Section 8.3.2 - DCV 105 Limit for MGS/MS8/CU4 ISTATION. .5;*~{:**:i{:.. - {* i' On{:i:

site Power Systems .:;:;i<>{-g"*;{ .*!:- ::: ISection 8.3 2:,

UF-A .:::*}i<::t * .:ii.. -'... -**:;:::!!i.t*-::j" Page 68#!{}.::.*!;:

8.3.2 DC Power System 8.3.2.1 Description The station DC power system is comprised of the battery chargers, station batteries and 125V Distribution System. It provides the sources of power for direct current load groups, vital control and instrumentation systems, and control and operation of Class 1E and non-Class I1E electrical equipment. It is a two-wire ungrounded system.

The battery chargers (rectifiers) provide the normal steady-state DC power; the station batteries provide for normal transient loads and also act as the reserve source upon failure of the rectifier or the AC supply to it. Figure 8.3-2, Figure 8.3-37 and Figure 8.3-3 8 present the one-line diagrams for the station DC electric power system, and show the connections to the AC Vital Instrumentation and Control Power System.

The safety-related portion of the station DC power system shown on Figure 8.3-37 consists of four 125-volt batteries, chargers and DC buses. The loads supplied from the buses include inverters for redundant vital instrument buses, distribution panels for power to the Class lE direct current loads, power for control and operation of the Class lB systems for Engineered Safety Features, and power for selected non-Class lB loads.

Each DC bus consists of metal-enclosed 125-volt DC switchgear consisting of vertical sections housing buses, circuit breakers, instruments and accessory equipment. The breakers are low voltage manual power circuit breakers. Figure 8.3-37 shows that the safety-related DC system incorporates mechanically interlocked manual circuit breakers which will permit the connection of two DC supply buses within the same train to a single battery, but prevents paralleling the two batteries in the train.

The nonsafety-related portion of the station DC systems shown on Figure 8.3-38 consists of two 125-volt batteries, chargers and DC buses. The loads supplied fr'om the buses include inverters for the computer and auxiliary power panels feeding nonvital equipment requiring constant supply and control power feeders to nonvital equipment (13 .8-ky switchgear, turbine generator emergency oil pump, etc.).

a. System Separation, Ventilation and Redundancy Four safety-related 125-volt batteries are supplied. Each battery is housed in an individual room in the seismic Category I Control Building. Separate ventilating systems are provided for the battery rooms of each train (see Subsection 9.4.10).

The batteries are seismically qualified and are mounted on seismic Category I racks. The safety-related battery chargers and DC buses are also seismically qualified.

V18 - UFSAR Section 8.3.2 -. DCV 105 Limit for MG8/MS8/CU4

S"I:'AB:<**doK<*: FL':.CTR**IC POWER

.<:"..J~":"( . ".-:!~* !;i§Revision 16.:i;.:.i.)

Each battery has its own charger and DC bus. The battery chargers and DC buses of each train are located in an area adjacent to their associated battery rooms and are physically separated from the char gers and buses associated with the redundant train (see Figure 8.3-27 and Figuie 8.3-36).

Four DC supplies are provided for the four NSSS inverteis for vital instrument buses and the power and control requirements of the two .engineered safety features trains (see Subsection 8.3,.1~d). Equipment is located and cables are routed in a manner to assure continued independence and separation so that the loss of DC supply to either train does not prevent the minimum safety function of the other train from being performed.

One nonsafety-related inverter for the station computer is powered from the Train A DC system through a Class lE breaker on Bus lilC. One nonsafety-related DC power panel is powered from the Train B DC system through a subfeed from safety-related DC power panel PP-i111B. All remaining nonsafety-related loads (DC motors, other nonsafety-related inverters, nonvital control panels) are connected to the nonsafety-related batteries (Figure 8.3:.38).

b. Station Battery Capacity The safety-related station batteries are lead-calcium, power station type. Each battery consists of 59 cells, and has a nominal 8-hour rating of 2280-ampere hours.

Each safety-related battery is sized to supply its safety-related and nonsafety-related loads for the durations indicated in Table 8.3-5. Battery B-1C is capable of providing power to the nonvital computer inverter, 1-2A, for 15 minutes while supplying its safety-related loads; the inverter load is automatically disconnected from the DC system after the 15-minute period. This disconnection is accomplished by a safety-related trip circuit on the Class lE breaker feeding inverter I-2A. This cir'cuit, which monitors the time the inverter draws power from the battery, is testable.

Tn addition, each safety-related battery is sized to have sufficient capacity to serve as the source, for the duration indicated in Table 8.3-5, for two load groups of the same train during the period when one battery is out of service (see Figure 8.3-37). Figure 8.3-51 shows the separate and combined load profiles for the safety-related batteries.

The safety-related station batteries also have sufficient capacity for the four-hou" Station Blackout coping duration. The Station Blackout battery sizing evaluation includes the one battery/two bus configuration (see Section 8.4.4.2).

V18 - UFSAR Section 8.3.2 - DCV 105 Limit far MG8IMSBICU4

• STATION Onsite. Power Systems Section 8.3it There aic two nonsafety-related batteries (13-2A and B-2B) provided in the Turbine Building. The nonsafety-related station baffezies are lead calcium, power station type, consisting of 59 cells. Battery B-2A supplies various DC motors for the turbine auxiliaries, various control panels and the Turbine Building DC lighting. Battery B-2B supplies the computer inverter T-2B, the nonvital instrument inverter 1-4, Control Building DC lighting and various control panels.

Each Class lE battery was sized in accordance with the recommended practices in IE:EE Standard 485-1978. These practices were applied as follows:

1. The system maximum voltage (140 volts) and maximum equalizing cell voltage (2.33V per cell) were selected. This resulted in a selection of 59 cells which include margin between the equalizing voltage (137.5 V),

and the system's maximum voltage.

2. A duty cycle diagram was developed, based upon the combined known and anticipated loads for both DC buses of the same train (see Figure 8.3-5 1).

3. The battery capacity data were selected from the manufacturer's data, based upon the minimum cell voltage (1.78V per cell permitted by the system minimum voltage of 105V).

4. The calculated minimum required cell size was increased by 25 percent for end-of-life compensation.

5. Temperature correction factors were applied to the calculated minimum required cell size, to allow for operation at the minimum design temperature (65°F for batteries B-lA and B-1C, and 60°F for batteries B-lB and B3-1D). These temperatures also apply to Station Blackout.

6. Sizing calculations were performed using methods similar to Figure 3 of IEEE 485-1978 to determine the minimum required cell size.

7. A minimum design mar'gin of 15 percent was included in the original battery purchase specification calculated cell size to allow capacity for Tuture loads.

V19 - NUHOMS HSM Technical Specification Dose Rates for EAL EU1 HSM-H Dose Rate Evaluation Program 5.4 5.4 HSM-H Dose Rate Evaluation Program This program provides a means to help ensure that the cask (DSC) is loaded properly and that the facility will meet the off-site dose requirements of 72.104(a).

1. As part of its evaluation pursuant to 10 CFR 72.212, the licensee shall perform an analysis to confirm that the limits of 10 CFR Part 20 and 10 CER 72.104 will be satisfied under the actual site conditions and configurations considering the planned number of HSMs to be used and the planned fuel loading conditions.

2. On the basis of the analysis in TS 5.4.1, the licensee shall establish a set of HSM-H dose rate limits which are to be applied to 32PTH DSCs used at the site. Limits shall establish peak dose rates for:

a. HSM-H front surface,

b. HSM-H door centerline, and

c. End shield wall exterior.

3. Notwithstanding the limits established in TS 5.4.2, the dose rate limits may not exceed the following values as calculated for a content of design basis fuel as follows:

a. 800 mremlhr at the front bird screen,

b. 2 mremlhr at the door centerline, and

c. 2 mrelr at the end shield wall exterior.

4. If the measured dose rates do not meet the limits of TS 5.4.2 or TS 5.4.3, whichever are lower, the licensee shall take the following actions:

a. Notify the U.S. Nuclear Regulatory Commission (Director of the Office of Nuclear Material Safety and Safeguards) within 30 days,

b. Administratively verify that the correct fuel was loaded,

c. Ensure proper installation of the HSM-H door,

d. Ensure that the DSC is properly positioned on the support rails, and

e. Perform an analysis to determine that placement of the as-loaded DSC at the ISFSI will not cause the ISFSI to exceed the radiation exposure limits of 10 CER Part 20 and 72 and/or provide additional shielding to assure exposure limits are not exceeded.

NUHOMS HD System Technical Specifications5- 5-9

V20 - Site Specific CSFST Care Coaling far MG1/MS5/Fuel Clad and CantainmentPatentialLass Number Title Revision F-0.2 CORE COOLING (C) 20 (RED) GO TO FR-C.1 GO TO FR-C.1 GO TO FR-C.2 GO TO FR-C.2 GO TO FR-C.3 GO TO FR-C.2

• • .*

• GO TO FR-C.3

C)CSF SAT G:\Word\lmagesP\lmagesP.OP\F02.ds4 1 of 1

INumber Title F-0.4 V21 - Site Specific CSFST Integrity INTEGRITY (P)

IRevision 21 (FIGURE 1)

W 4In 1W o

T1 T2 COLD LEG TEMPERATURE GO TO FR-P.1

-mmI-------- (ORANGE) GO T GOFR.P.1T N (* GO TO (YELLOW)@ w@*

• FR-P.2

~'ES I ,

IDECREASE IN NO (GREEN) L

~ )SAT CSF ENTER D-. ALL COLD LEGS TEMPERATURE LESS THAN 1000 F GOTO IN THE LAST YE SIXTY MINUTES (ORANGE) E1-l -ol FR-P.1 ATURES R THAN YE

• GO TO (YELLOW)0 0@* FR-P.2 FL (GREEN)

O SAT CSF (GREEN)

G:\Word~lmagesP~mages_.P.OP\FO4.ds4 FL O SAT CSF 1 of 3

V21 - Site Specific CSFST Integrity F-0.4 INTEGRITY (P) 21 FIGURE 1 2500 2800 2560 2500 226°F/

2050PS 2000 (5

U) 03.

uJ 1500 rDr w) 1000 n) 500 RED ORANGE -- GREEN

-J 0

4 - 1 100 194°/ 200 300 400 T1 T2 RCS COLD LEG TEMPERATURE (OF)

F04-1 2 of 3

V21 - Site Specific CSFST Integrity Number Title Revision F-0.4 INTEGRITY (P) 21 FIGURE 2 LTOP PRESSURE LIMITS Reactor Coolant System Pressure vs Temperature Curve 2800 Max. PORV i Setpoint 2385 PSIG ':

2600 I 1600 PSID 2400 (wiRtC SP-zSRG Liub itl j* "*/

2200 2000 1800 1 II LTOP Pr~Q'~I irn I I I I!1 I I 1I 1I 1 1 1rl-* l IJ l I t I I I I I /] I I ,I I l I II I i51600 Li**t711-iJ i ii i i i*lq i ii i iJ"i ii i i ii i ii:i i i i*"i i iii Yi ii ;ii S1400

. .. . .. . .. . [...i-... . .. . .. .. . .. . ... .... ... -

2000F Cs~1200 For RCP Operation

• L 100°F 1000 (Admin. 325 PS*

800 LTOP Min.

Setpoint 426 PSIG Lsub ooio0 50OF Subcooling 6O00 RCP Required NPSH 400 200 0

0 50 100 150 200 250 300 350 400 450 500 550 600 650 RCS Temperature (F) 3 of 3

V22 - Site Specific CSFST Heat Sink for MS5IRCS PotentialLoss SNumber Title Revision F-O.3 HEAT SINK (H) 21 GO TO FR-H.1 0

(YELLOW) (1h KY***** GO TO FR-H.2 ENTER--I (YELLOW) (

S GO TO

• ******' FR-H.3 (YELLOW) ,*GO TO 0 ** FR-H.4 GO TO FR-H.5 G:\Word'lmagesP\lmagesP.O P\F03.ds4 K__*Q (GREEN)

SAT CSF 1 of 1

V23 - EPCALC- 06-01 Rad Values for Fission ProductBarrierLoss and PotentialLoss FPL Energy Seabrook Station SEP# 2006012

SUBJECT:

Calculation of Containment Dose Rate Values Used in FPB EALs FROM: D. L. Young DATE: March 28, 2006 TO: File Attached are the calculations forthe containment dose rate values used in the Fission Product Barrier (FPB) Emergency Action Levels (EALs). The calculation number is EPCALC-06.01.

Senior Nuw DLY/dly cc: W. Cash

• ' I I II II " .... ... I IIII I I 'I I' I I III

Calculation of Fission Product Barrier Matrix EPCALC-06-O1 EAL Values for Containment Dose Rates Containment BarrierPotential Loss - Containment Ra*diation-Monitor Reading 'm ..

ra,,

NEIE 99-01. Rev. 4, Basis: Unless there is a (site-specific) analysis justifying a higher value, it is re*:ommended "

that a radiation monitor reading corresponding to 20% fuel clad damage be specified here. '-

Calculation Basis: SBC-1023, Update of METPAC Data Affected by Power Uprate and AST Implementation, O The calculation was perforated using an assumed value of 20% fuel clad failure.

Time NG Core NG NG Cone uCi-hr Men tinventory (assumed) I .

(Note 1) (Note 2) (Note 3) (Note 4) (Note 5) "fuel clad damager for EAL l CD

.... o....................................... ......... ..... ........................

2 3,622+08 1.002-02 4,732+01 4.712-02 "1,003 -

4 5 3,27E+08 1.00E-02 4.262+01 8,93E-02 615 Free volume of conta~nment 2,702÷06 ft per UJFSAR Sec. 15B,2,11,A 10 2.982+08 1,00OE-02 3.882+01 1 ,05B.-01 389 Free volum~e of containment 7.066E+10 ml [2.7E+06 *28315.846592 ml/ft3j 20 2.53E+08 1.002-02 3.3'tE+0t 1,552-01 212 F~ree volume of containment 7,602+10 cc per UFSAR Sec. iSB.2.Ul.A

• Free volume of containment 7,665+04 mu per UFSAR Sec. 15B.2,ll.A5 Note I -Values taken from SBC-1023, Table 3,10"1 Note 2-5%

• 2O% 1%

Note 3 - [(Core NG

• 12+6 uCllCi) *.01*]/7.662+10 cc .

Nlote 4.. Values taken from SBCo1 023, Tables 4.5 and 4.6 Nlote 5 - Cont NG ConcI'J(t) '

0)3 m'-

oD

-I

-o

_ _ a Hey. 0 Page1 3/2412006 IIII I I II I II I I III II I I ,ill i ,i ,, , ,,,,,,,,

Calculation of Fission Product Barrier Matrix EPCALC-06-01 EAL Values for Containment Dose Rates CA~

iRCe:Barrioi Loss - Contalnment Radiation Monitor Reading -o NEt 99-01, Rev. 4, Basis: The reading should be calculated assuming the instantaneous release and dispersal 0 of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e.. I-wvithin T/S) into the containment atmosphere.

Oatculatlon Basis: SBC-t023, Update of METPAC Vata Affected by Power Uprate and AST Implementation. 9 The calculation was performed using RCS actlvity valufes at Technical Specification limits.

0 Decay RCS ROB Cnmt J(t) Cnmt Volume of water inRiCS 1.152+04 UFSAR Table 5.1-i*

Time NG Cone NG NG Cons uCI-hi Man Gallons of watter In RCS 8,622+04 gal [12100 ft3

• 7.48 gal/ft3]

(his) (uCilgm) (uCI) uCflco cc-R Rlhr lbs of waler In RCS 7,t95+05 lbs 190505 gal

• 8.345 lbs/gal]

(Note 1) (Note 1) (Note 2) (Note 3) (Note 4) (Note 5) gins or cc of waler In RCS 3,275+08 gin [755269 lbs *454 gnislIb] 0) 0.5 1.51tE÷02 4.95E+t0 l8o47E-01 3.305.032 20 CD 1 1.465+02 4,788+10 6,245-01 3,DIE-02 16 Free volume of conlalnrnnett 2.70E+00 ftJ per UFSAR Sac, 15B.2,11,A C,)

2 1.395+02 4.585+10 5,94E-0I 4,715.02 13 Froe volume of contulnmenl 7.665÷10 ml [2,75+05

• 283'16,846592 ml/ftS] 0 5 1.265+02 4,';3E+l0 5,395-01 8,935-02 8 Free volume of containinont 7,1305+10 cc 10 1,152÷02 3,70E÷10 4,915-01 1.05E-01 5 Free volunio of containment 7.600E04 per UFSAR Sec. 15B.2.I.,A C,)

20 1,035+02 3.385+10 4.42E-01 1.5135-01 3 C,)

Note I - Values ftom 850-1 023, Table 3.2 and 3.9, 0 Also see UFSAR Table 150..2 Note 2. ((uCi/gm)*3.272+8 gin) -a Note 3 - (u~i)/7.60E+10 cc 0 Nlote4- Values taken from SBC-1023, Tables 4,5 and 4.6 Nlote 5- Cent NG Conc/J(t) w 0)

CD I-.

0 Co C,)

0) 0 e.g.

CD 0) 0 C,)

C,)

R*. 0 Page2 Rev, 0Page 23/24/2001 O

I i il Iml i i Ill I i III I IIII II I

11111 I Calculation of Fission PrOduct Barrier Matrix EPCALC-tJ6-01 EAL Values for Containment Dose Rates 3~l Fuel Clad Barrior: Loss - Containment Radiation Monitor Reading M21 99-01, Rev. 4, Basis: The reading should be caicu~ated assuming the instantaneous release and dispersal of

'he reactor coolant noble gas and iodine inventory associated with a concentration of 300 uCi/gm dose equivalent 1-131 into the containment atmosphere.

3alouiatron Basis: $BC-1023, Update of METPAC Data Affected by Power Uprate and AST Implementation. The atculatlon was performed using ROS activity' values associated wilh a concentration of 300 uCi/gm dose 6 7~4

• quivalenl 1-131.

Total Ci of DE1-131 in ROS=

f300 uCi/cc *3.27E+8 gml~l E+6 uCi/CI = 9.81 E+4 Cl 0.

Decay DEI..131I NG DEl131t NG In NG cone J~t) Cnmt Time in RCS in Core in Core Cant in Conlt uGi.hr Posl:.LOOA CD (hrs) (Cl) (CI) (Cl) (Ci) (uCl/coo) cc.R ... (Rlhr) CD (Note I) (Note 2) (Note 3) (Note 4) (Note 5) (Note 5) (Note?7) CA 0.5 9,81E+04 4,33E+08 "1.32E+08 3,21E+05 4.19E+D0 3,301*.02 127 0 I 9.81B+04 3,90P=÷8 '].32Eq÷08 2.912+08 3.79E+-00 3.912.02 97 2 9,81E+04 3,22E÷08 1.302+05 2.73E+05 3.662+00 4.712-02 78 5 9,81E+04 3,27E+a8 1,25E.08 2.5132+05 3.342400 6..932.02 46 (A CA 10 9.812+04 2,96g+08 '1.IBE+08 2,4,52+05 3.202÷00 1.062-01 31 20 9.812+04 2.532+08 1.08E+08 2.31E+05 3.012+00 1,562-01 19 Note 1- See caiculation above Notoe5- (NG In Cent Ci/7.652+1O cc) 1E+6 uCi/Ci -I, Note 5- Values taken from SBC-1 023, Tables 4.6 and 4.6 0 Note 2- Values taken from SB C-I023, Note 3.. Values taken from SBC-1023, Table 5.5 Table 3.1 0 Note 7 - NO cone in ContlJ~t} 0 Nate 4 -9.81 E+4" (NG in core/DEl-131 in core) p.,.

Free volume of containment Volume of RCS 7.66E+i0 cc 3.27E+0& gn, w

CD

• '1

~*5 CD I-.

0 CA CA 0

CD CD 0

CR CR Roy. 0 Rev.O~~Page 331406 312412006 II I III II II I IIII I I I . al II

IIII . J "

Calculation of Fission Product Barrier Matrix EPCALC-06-01 EAL Values for Containment Dose Rates Prepared by; Reviewed by: ________

Li~L Bill Gash Radlallon Proteotion Manager Dete "Co 0o Approved by: 3J&O(o Date

-o 0=

Rev. 0 Page 4 Rtv.IIIPanIIIIII24II00

V24 - Site Specific CSFST Containmentfor Containment PotentialLoss I Number Title Revision F-0.5 CONTAINMENT (Z) 20 (RED) GO TO ENTER FR-Z.1 (RED) GO TO FR-Z.1

' u (ORANGE) (* GO TO J* GO TO

---

• FR-Z.2 (YELLOW) (h GO TO ALL CONTAINMENT N PHSPENETRATIONS YS (YELLOW) GO TO ISOLATED * *

• FR-Z.1 (YELLOW)GOT I (YELOW* GO TO 0* FR-Z.3 ENO YSS I/Hr YE G:\Word~lmagesP'lmages_P.OP\F05.ds4 liQ CSF SAT 1 of 1

V25 - Tech Spec Containment Spray Setpoint for MU7lContainment Potential Loss TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) _z s TRIP SETPOINT ALLOWABLE VALUE

1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generator, Phase "A" Isolation, Containment Ventilation Isolation, and Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency Fan/Filter Actuation, and Latching Relay).

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

c. Containment Pressure-Hi-I 4.2 0.71 1.67 <*4.3 psig < 5.3 psig

d. Pressurizer Pressure-Low N.A. N.A. N.A. _>1800 psig _>1786 psig

e. Steam Line Pressure-Low 13.1 10.71 1.63 _>585 psig >_568 psig*

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. N.A, N.A,

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Containment Pressure--H i-3 0.71 1.67 3.0 _ 18.0 psig < 18.7 psig SEABROOK - UNIT 13/324AedntN.3 3/4 3-24 Amendment No. 33

V26 - UFSAR Section 8.4.2 SBO Coping Basis for MG1 SEABROOK STATION ELECTRIC POWER Compliance with 10 CFR 50.63, Loss of All Alternating IRevision 14 Section 8.4 UFSAR Current Power (Station Blackout) Page 1 8.4 COMPLIANCE WITH 10 CFR 50.63. LOSS OF ALL ALTERNATING CURRENT POWER (STATION BLACKOUT) 8.4.1 Basic Requirements This section describes Seabrook's compliance with 10 CFR 50.63 which requires that each light-water-cooled nuclear power plant be able to withstand and recover from a loss of all alternating current power or station blackout (loss of both offsite power and onsite emergency power). Regulatory Guide 1.155 (RG 1.155), "Station Blackout," provided a method for complying with 10 CFR 50.63. RG 1.155 stated that NUMARC 87-00, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," also provided acceptable guidance for meeting the requirements of 10 CFR 50.63.

Seabrook followed NUMARC 87-00 except where the Regulatory Guide took precedence.

Station Blackout is considered a non-design basis accident. No other active or passive failures or design basis events are required to be considered during Station Blackout. Safe shutdown for Station Blackout (see 10 CFR 50.2 and NUMARC 87-00) means bringing the plant to a hot shutdown or hot standby condition. The Seabrook analysis and procedures proceed with plant cooldown until secondary side pressure is reduced to about 250 psig and the plant is in a hot standby condition.

Seabrook responds to Station Blackout as an AC Independent plant relying only on the station batteries as a source of electrical power for the coping duration specified in Section 8.4.2. When the Station Blackout analysis was initially performed there were no alternate AC power sources to support response as an Alternate AC (AAC) plant. The Supplemental Emergency Power System (SEPS) was subsequently installed and may be used as a source of power during a station blackout event but will not be credited as an alternate AC power source.

8.4.2 Station Blackout Duration Seabrook Station's Blackout coping duration is four hours. This is based on evaluation of the offsite power design characteristics, emergency AC power system configuration and emergency diesel generator (EDG) reliability. The offsite power design characteristics included the expected frequency of grid-related loss of offsite power, the estimated frequency of loss of offsite power from severe and extremely severe weather, the number of switchyards and the type of bus transfers. Site-specific weather data were used to evaluate the reliability of offsite power relative to weather-caused outages. One out of two emergency diesel generators is required to operate safe shutdown equipment following a loss of offsite power. A target EDG reliability of 0.975 will be maintained by implementation of the Maintenance Rule EDG performance criteria.

8.4.3 Procedures Station procedures address the action necessary to cope with a Station Blackout (loss of all AC power) including actions such as opening cabinet doors. Also, as required by RG 1.155, procedures address AC power restoration and severe weather conditions.

V27- EC282 184 Seismic Monitoring System Upgrade Description of OBE Lights for HU2 signal proportional to the magnitude of the motion to the three recorders located in SM-CP-5 8.AThe recorders monitor the motion signals and if a level exceeds preset trigger - - -- -

values, then the recorders trigger to record the event, a red indicating light is illuminated on SM-CP-58, and a main plant computer alarm is actuated indicating that a seismic event is in progress. After the event recording ends, the central controller is notified and the data is retrieved, analyzed, and reports are generated. This process typically takes 2-3 minutes. Jf the magnitude on the free field accelerometer exceeds OBE levels, then a - - --

yellow indicating light is illuminated on SM-CP-58. The Operators then take the actions specified in the applicable procedures.

EN-AA-205-1 100-F01, Revision 4 ECFORM- 110

V28 - Verification of Fire Alarms for IC HU4 Verification of Fire Alarms for NiEJ 99-0 1 Rev. 6 IC I{U4 Per FP62296, Seabrook Station Pre-Fire Strategies Manual Table 112 Location Control Room Alarm Location Control Room Alarm Condensate Storage Tank Fuel StorageBuilding FP-.CP-454 Enclosure*

Containment FP-CP*-376 Primary Auxiliary Building FP-CP-446/FP-CP-4531FP-CP-559 Control Building FP-CP-558 Service Water Pump House FP-CP-3801FP-CP-474 Cooling Tower FP-CP-381 Steam and Feedwater Pipe FP-CP-45 1 EAST

__________Chases FP-CP-452 WEST Diesel Generator Buildings FP-CP-561(A) & FP- North Tank Farm*

___________________ CP-562(B) ____________ __________

Emergency Feed water FP-CP-560 Startup Feedwater Pump Pump House Area*___________

RHIR Equipment Vaults ..... FP-CP-475 __________

• Fire Protection Design Basis Document (DBD-FP-02), Rev. 0, documents NRC approved deviations from 10 CFR 50, Appendix R, fire protection guidelines for location of fire detection systems in the indicated locations.

V29 - Site Specific CSFST Subcriticalityfor MS5IMA5IMU5 Number Title IRevision F-0.1 SUBCRITICALITY (S) 20 (RED) GO TO FR-S.I (ORANE)EE E GO TO mn m un n u FR-S.I ENTER, (YELLOW)

GO TO FR-S.2 CSF:

SAT (GREEN)

GO TO (YELLOW) (*

OOOOO FR-S.2 CSF NOTE: USE GAMMA-METRICS SAT (GREEN)

G:\Word~lmagesP\lmagesP.OP\FO01.ds4 1 of 1

V/30 - EPCALC-06-01 Letdown Monitor Value for MU3 0 FPL Energy Seabrook Station SEP# 2006020

SUBJECT:

Calculation of Value for Initiating Condition SU4, EAL #1 FROM: 0. L. Young DATE: April 17, 2006 TO: File Initiating provides Condition indication SU4, of fuelEAL clad#1degradation addresses greater a letdown radiation than monitor Technical reading that Specification allowable limits. This value was calculated as follows.

• The equipment tag identifier for the Letdown Monitor is 1-RM-6520-1

• ROS noble gas activity at 10OIEb, = 1.887E+2 uCi/grn (per SBC-1023, Table 3.2)

• Noble gas activity will govern the monitor reading, not iodine

• Letdown Monitor Conversion Factor = 14.148 mR/hr/uCi/ml (per ER 4.3, Figure 12) therefore, The 1-RM-6520-1 reading at the Technical Specification allowable limit =

1.887E+2 *iig

• 14.148 mR/hr/u*fm = 2,669.73 mR/hr or 2,670 mR/hr The above calculation is EPCALC-06-03.

Prepared by:.

David SeniorL. Young Analyst Nuclear Approved by:

William T. Cash Radiation Protection Manager DLY/dly CC:

V31 - Tech Spec 3/4.4.8 Specific Activity for MU3 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microCurie per gram DOSE EQUIVALENT 1-131, and

b. Less than or equal to 100/E microCuries per gram of gross radioactivity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

• ---------NOTE...------..............

LCO 3.0.4.c is applicable to DOSE EQUIVALENT 1-131.

MODES 1, 2, and 3*:

a. With the specific activity of the reactor coolant greater than 1 microCurie per gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 5000 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and

b. With the specific activity of the reactor coolant greater than 100/IE microCuries per gram, be in at least HOT STANDBY with Tavg less than 5000 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than 1 microCurie per gram DOSE EQUIVALENT 1-131 or greater than 100/I: microCuries per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-3 until the specific activity of the reactor coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-3.

• With Tavg greater than or equal to 5000 F.

SEABROOK- UNIT 1 3/4 4-19 SEABOOK-NIT 3/44-19Amendment No. 4-44, 115

V32 - UFSAR Table 11.5 Monitor Values for RU1 TABLE 11.5-1 PROCESS AND EFFLUENT RADIATION MONITORS (Note 5)

Det Back Range Alarm Instrument Tag Detector Grd Low-High Set Point Reference Detector Safety Energy* Loop Diag. P&ID No. Re- Deescription Typ mr/hr (ucilcc) Isotooe_ Qtv Class Level 1-NH-Y I-NHY 58 13 1 137 6454 St*term Drains Gamma 0.5 10-6 10.2 Co ,! ,CS Non lE Note 2 506765 SD-20404 Scint 6502 WTaste Gas Inlet to Gamma 15.0 10-2 10+2 Xe 133 Non 1E Note 1 506897 20772 Caarbon Delay Beds Scint 6503 [aste Gas Gamma 15.0 10-3 10k' Krs 5 Note 1 506898 20770 ompressor Inlet Scint 6504 H 2Gas Compressor Gamma 15.0 10-3 10+1 Note 1 506899 20773 isch. Scint 133 6505 C'ondenser Air Evac Beta 0.5 Note 3 Xe Note 1 506055 20774 Scint 6515,6516 Primary Component Gamma 2.5 10-7 10-3 co 5 s,11'3 1,C5 137 2 Note 2 506190, 20211, Cooling Water Scint 506194 20205 6509 Liquid Waste Test Tk 10-6 10"2 C58,I131 ,S137 Gamma 2.5 Note 2 506927 20831 Disch to CWS Scint c58,I131,C 137 6510, 6511, Steam Gen B lowdown Gamma 2.5 10-6 10-2 4 Note 2 506815 20521 6512, 6513 Sample Loops 1,2,3,4 Scint 6519 Steam Gen Blowdown Gamma 2.5 10-7 10-3 Co 58 , 1131,CS 137 Note 2 506734 20626 Flash Tank Drain Scint 6520 Reactor Coolant Gross GM 15 10-1 10+4 mr/hr Co 58 ,I 13 1,CS 137 Note 2 506269 20722 Activity Monitor See Table 11.5-1 (Sheet 3) for notes.

V32 - UFSAR Table 11.5 Monitor Values for RU1 (Note 5)

Det Back Range Alarm Detector Grd Low-High Set Point Reference Detector Safety Energy* Loop Diag. P&ID Instrument Tag No. Re- Tye mr/hr (iclc (ui/c Isotone Oty Class Level 1-NHY 1-NHY Description 33 6481-1, 6482-1 Main Steam Line Gamma 2.5 10-° 10+Smrihr Xe' - 506551 20580 6481-2, 6482-2 Monitor Scint -2, -3, -4 20581 6490 Aux Steam Cond Gamma 0.5 10- 7 10"3 Co5 s,i'3 , CS' 37 1 Nonl1E Note 2 507165 20908 Scint 6521 Turb. Bldg. Sump Liq. Gamma 2.5 10-6 10-2 COss, 1131, CS' 3 7 20195 Non lE Note 2 506713, Monitor Scint 506716 6527A1, A2 COP Monitors GM 2.5 10' -l0 6 cpm Xe 133 4 1E Note 1 506211 20504 B1, B2 33 6560 Resin Sluice Line GM 100 Note 4 Xe1 1 Non 1E 506694 20252 33 6561 Resin Transfer Line GM 100 Note 4 Xe' 1 Non lE 506694 20735 33 6564 Sluice Pump Line GM 100 Note 4 Xe1 1 Non lE 586692 20252 58 37 6473 Water Treatment Gamma 2.5 10 Co , i131. CS' 1 Non lE Note 2 506976 20040 Liquid Effluent Scint Radiation Monitor

1~

V32 - UFSAR Table 11.5 Monitor Values for RU1 SEABROOK RADIOACTIVE WASTE MANAGEMENT Revision: 13 STATION UFSAR TABLE 11.5-1 Sheet: 3 of 3 Max. Beta Energy Predominant Gamma Energy (Mev)

Note Isotopes 1 Xe' 33 0.346 0.08 1 Xe' 3 5 0.92 0.249 Kr85 0.67 0.5 14 KrsSm 0.82 0.150 2 ~ 131 0.606 0.364 1133 1.27 0.53 CS'3 4 0.662 0.604 Cs' 3 7 0.514 0.662 Co58 0.474 0.81 Co6 ° 0.3 14 1.17, 1.33 3 Condenser Air Evacuation Monitor to have output in counts per min (cpm) (10' to 106).

4 Monitors 6560, 6561, 6564 have output in mr/hr (100 to 105).

5 Radiation monitoring setpoints are varied during operation to follow station operating conditions. Setpoints are maintained within the bounds established in the Technical Specifications. The methodology for establishing the setpoints is found in the Station Offsite Dose Calculation Manual (ODCM) and/or station operating procedures.

A