SBK-L-13028, License Amendment Request 13-01 Application for Administrative Change and Corrections to the Technical Specifications

License Amendment Request 13-01 Application for Administrative Change and Corrections to the Technical Specifications

ML13070A010
Person / Time
Site:	Seabrook
Issue date:	03/01/2013
From:	Walsh K NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-13028, LAR 13-001
Download: ML13070A010 (24)

Text

NEXTera ENERGY.. yABROK March 1, 2013 10 CFR 50.90 SBK-L-13028 Docket No. 50-443 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Seabrook Station License Amendment Request 13-01 Application for Administrative Change and Corrections to the Technical Specifications In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), NextEra Energy Seabrook, LLC (NextEra) is submitting License Amendment Request (LAR) 13-01 to revise the Seabrook Station Technical Specification (TS).

The proposed changes delete the TS Index and make corrections to TS 3.4.8, Reactor Coolant System Specific Activity, and TS 6.8.1.6.a, Core Operating Limits Report.

The Enclosure to this letter provides NextEra's evaluation of the proposed changes. Attachment 1 to the enclosure provides markups of the TS that show the changes. As discussed in the evaluation, the proposed changes do not involve a significant hazards consideration pursuant to 10 CFR 50.92, and there are no significant environmental impacts associated with the change.

No new commitments are made as a result of this change.

The Station Operation Review Committee has reviewed this LAR. A copy of this LAR has been forwarded to the New Hampshire State Liaison Officer pursuant to 10 CFR 50.91 (b).

NextEra requests NRC review and approval of LAR 13-01 by March 1, 2014 and implementation within 60 days.

Should you have any questions regarding this letter, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.

NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L-13028 / Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on March 1, 2013.

Sincerely, Kevin T. Walsh Site Vice President NextEra Energy Seabrook, LLC Enclosure cc: NRC Region I Administrator NRC Project Manager NRC Senior Resident Inspector Perry E. Plummer, Acting Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

Enclosure NextEra Energy Seabrook's Evaluation of the Proposed Change

Subject:

Application for Administrative Change and Corrections to the Technical Specifications 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Significant Hazards Consideration 4.3 Conclusion

6.0 REFERENCES

A mk..................................... I.........

Attachment I - Markup of Technical Specifications 1

1.0

SUMMARY

DESCRIPTION License Amendment Request (LAR) 13-01 proposes to delete the index from the Seabrook Station Technical Specifications (TS) and make corrections to TS 3.4.8, Reactor Coolant System (RCS) Specific Activity, and TS 6.8.1.6.a, Core Operating Limits Report (COLR).

2.0 DETAILED DESCRIPTION

• TS index pages i through xii are deleted.

" TS Table 4.4-3, Reactor Coolant Specific Activity Sample and Analysis Program, currently specifies the following sample and analysis frequency for isotopic analysis for iodine:

Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds 1/pCi/gram DOSE EQUIVALENT 1-131 or I O0/E Ci/gram ofgross radioactivity This requirement is revised to provide the correct units of activity for 100/I- as indicated below.

Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds 1 pCi/gram DOSE EQUIVALENT 1-131 or 100/E microCi/gramofgross radioactivity

• TS 6.8, Core Operating Limits Report, item 2, requires operating limits for:

2. Cycle dependent maximum allowable combination of thermal power, pressurizer pressure and the highest operating loop average temperature (Tavg) for Specifications 2.1.1 and 2.1.2.

The reference to TS 2.1.2, Reactor Coolant System Pressure Safety Limit, is incorrect and is removed as shown below.

2. Cycle dependent maximum allowable combination of thermal power, pressurizer pressure and the highest operating loop average temperature (Tavg) for Specifications 2.1.1 andA-2.1.2.

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3.0 TECHNICAL EVALUATION

TS Index This LAR proposes to delete the index from the TS. Past submittals to the NRC have included the TS index to supplement requested TS changes, and a revised index has been included in issued amendments. This change will eliminate the need to include TS index pages as part of the license amendment request for future TS changes and to issue revised index pages with an amendment.

Although the index will not be considered part of the TS, NextEra will maintain and update the TS index. In addition, the distribution process currently in place will ensure that holders of controlled copies of the TS, including the NRC, will receive updated versions of the TS index similar to the way they currently receive changes to the TS and the TS Bases. When NextEra receives an approved change to the TS (including changes to the TS index) from the NRC or makes changes to the TS Bases, a transmittal form with the accompanying changes is sent to controlled copy holders, which includes offsite organizations that maintain controlled copies of the TS.

Changes to TS index pages would be distributed similarly following approval of this amendment request. Therefore, stakeholders will continue to be informed of any changes to the index after it is removed from the TS.

The TS index identifies the contents of the TS but does not contain any technical information. Therefore, this is an administrative change and no technical evaluation is required. Regulatory requirements related to this change are discussed in Section 4.0 of this LAR.

Similar amendments to remove the TS index were approved for Waterford Unit 3 in Amendment 200 in May 2005 [Reference 1], and for Arkansas Nuclear One, Unit 2 in Amendment 260 in June 2005 [Reference 2].

TS 3.4.8., Reactor Coolant System Specific Activity The limiting condition for operation (LCO) in TS 3.4.8 limits RCS specific activity to less than or equal to 100/P-microCuries per gram of gross radioactivity. If the LCO is not met, the Action requires:

With the specific activity of the reactor coolant greater than I microCurieper gram DOSE EQUIVALENT 1-131 or greater than 100/, microCuriesper gram, perform the samplingand analysis requirements of Item 4.a) of Table 4.4-3 until the specific activity of the reactorcoolant is restored to within its limits.

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However, item 4.a) of Table 4.4-3 referred to in the Action does not identify the correct threshold at which sampling and analysis is required. The value specified in the table, 100/I- Ci/gram of gross radioactivity, is much greater than the LCO limit and is inconsistent with the value of I 00/E microCuries specified in the Action. In the early version of the Seabrook Station TS (NUREG-1331 issued in May 1989),

Table 4.4-3 correctly specified a threshold value of I 00/P- microCuries per gram. In some subsequent revision of the TS, the prefix micro was dropped from the units of activity. When the error was introduced into the table is unknown.

The intent of the Action is to require increased sampling when RCS activity exceeds the limit of 100/E microCuries per gram established in the LCO. Therefore, the proposed change revises the units in TS Table 4.4-3, item 4.a) from 100/FB Curies per gram to 100/E- microCuries per gram. This change is of an administrative nature to correct an error that was inadvertently introduced into TS Table 4.4-3 at some time in the past.

TS 6.8, Core OperatingLimits Report (COLR)

TS 6.8 requires that core operating limits shall be established and documented in the COLR for certain parameters prior to each reload cycle. TS 6.8.1.6.a. 2 identifies one of the required parameters:

Cycle dependent maximum allowable combination of thermal power, pressurizer pressure and the highest operating loop average temperature (Tavg) for Specifications 2.1.1 and 2.1.2.

The reference to Specification 2.1.2 in the requirement above is incorrect.

Specification 2.1.2 (shown below) is related to the RCS pressure safety limit and is not a cycle dependent value.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained less than or equal to 2735 psig.

Specification 2.1.1, which requires the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR, is the only TS section 2.1 requirement applicable to this COLR requirement. Therefore, this proposed change removes from TS 6.8.1.6.a.2 the reference to Specification 2.1.2. This change is administrative in nature as it removes an incorrect reference to a TS that is not applicable.

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4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36 requires that each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications. The regulation also provides the categories of items that must be included in the TS:

1. Safety limits, limiting safety system settings, and limiting control settings

2. Limiting conditions for operation

3. Surveillance requirements

4. Design features

5. Administrative Controls The TS index, which this LAR proposes to remove from the TS, is not a required component of the TS in 10 CFR 50.36. Similar to the TS Bases, the index provides information about the TS but is not part of the TS. Because the index does not provide technical information required by 10 CFR 50.36, the proposed administrative change to remove the index from the TS is consistent with regulatory requirements.

4.2 Significant Hazards Consideration No Significant Hazards Consideration The proposed changes delete the index from the Seabrook Station Technical Specifications (TS) and make corrections to TS 3.4.8, Reactor Coolant System Specific Activity, and TS 6.8.1.6.a, Core Operating Limits Report.

In accordance with 10 CFR 50.92, NextEra has concluded that the proposed change does not involve a significant hazards consideration (SHC). The basis for the conclusion that the proposed change does not involve a SHC is as follows:

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1. The proposed change does not involve a significant increase in the probability or consequences of an accidentpreviously evaluated.

The proposed changes (1) remove the index from the TS, (2) correct an error in the units of activity for 100/F_ in TS 3.4.8, Reactor Coolant System Specific Activity, and (3) remove an incorrect, non-applicable reference in TS 6.8, Core Operating Limits Report. The proposed changes are all administrative in nature. The administrative changes are not initiators of any accident previously evaluated, and, consequently, the probability and consequences of an accident previously evaluated is not significantly increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed change does not create the possibility of a new or different kind of accidentfrom any previously evaluated The proposed changes are administrative in nature so no new or different accidents result from the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), a significant change in the method of plant operation, or new operator actions. The changes do not alter assumptions made in the safety analysis.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction in the margin of safety.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed administrative changes do not involve a change in the method of plant operation, do not affect any accident analyses, and do not relax any safety system settings.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, NextEra concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(b), and, accordingly, a finding of "no significant hazards consideration" is justified.

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4.3 Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment-will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

NextEra has evaluated the proposed amendment for environmental considerations.

The review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set for in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. NRC Letter "Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Re: Modification to Technical Specification (TS) 5.3.1, Fuel Assemblies, T.S. 5.6.1, Criticality, T.S. 6.9.1.11.1, Core Operating Limits Reports, and Deletion of TS Index (TAC No. MC3584)," May 9, 2005

2. NRC Letter "Arkansas Nuclear One, Unit 2 - Issuance of Amendments Re:

Deletion of Index Pages from the Technical Specification (TAC No.

MC4236)," June 22, 2005 7

Attachment 1 Markup of Proposed Technical Specifications Changes The attached markups reflect the currently issued version of the TS and Facility Operating License.

At the time of submittal, the Facility Operating License was revised through Amendment No. 132.

Listed below are the license amendment requests that are awaiting NRC approval and may impact the currently issued version of the Facility Operating License affected by this LAR.

LAR Title NextEra Energy Date Seabrook Letter Submitted 10-02 Application for Change to the Technical Specifications SBK-L-10074 05/14/2010 for the Containment Enclosure Emergency Air Cleanup System Changes to the Technical Specifications for New and 11-04 Spent Fuel Storage SBK-L-1 1245 01/30/2012 Application to Revise the Applicability of the Reactor 11-06 Coolant System Pressure -Temperature Limits and the SBK-L-11186 11/17/2011 Cold Overpressure Protection Setpoints Application for Technical Specification Improvement 12-06 to Extend the Inspection Interval for Reactor Coolant SBK-L-12235 12/20/2012 Pump Flywheels Using the Consolidated Line Item Improvement Process The following TS pages are included in the attached markup:

Technical Title Page Specification TS Index i through xii TS 3.4.8 Table 4.4-3, Reactor Coolant Specific Activity 3/4 4-21 Sample and Analysis Program TS 6.8.1.6.a.2 Core Operating Limits Report 6-16

INDE 1.0 DEFINITIONS SECTION PAGE 1.2 ACTUATION GIC TEST 1................1-1 1.3 ANALOG ANNEL OPERi ONAL TEST........................ . .......... ....... 1-1 1.4 A A..............."........ ............... ............. 1-1 1.5 C NEL CALIBRATION 1-1 1.6 HANNELCHECK. 1.................................

1.7 CONTAINMENT INTEGRITYi.....................1-2 11* CONTROLLED LEAKAGE ...- 2 1.9 CORE ALTERATION.............1-2 1.10 CORE OPERATING LIMIT..S------ 0--R..T............. ------ *.......... ..... 1-2 1.11 DIGITAL CHANNEL OPE IONAL TET............1-2 1.12 DOSE EQUIVALENT1- 1.................... 1-3 1.13 E - AVERAGE DI EGRiN ENERGY........................... 1-3 1.14 1.15 ENGINEEN FREQUE NOTATION.......

OTTON.... . iE.......................... 1-3 1-3 1.16 G 0I I 1-3 1.16 ~~~~~GASEO RADWASTE TREATMENT SYSTEM..........1-1.17 ID IFIED LEAKAGE1-3 1.18 ASTER RELAY TEST 1-4 1.19 MEMBER(S) OF THE PUBLIC........ 1-4 1.20 OFFSITE DOSE CALCULATION MANUAL........ .1-4 1.21 OPERABLE - OPERABILITY 1-4 1.22 OPERATIONAL MODE - MODE"............".1-. .

1.23 PHYSICS TESTS . . . . .. . . ... -4 1.24 PRESSURE BOU'NDARL GE.. 1-4 1.25 PROCESS CONTROL P RAM 1-5 1.26 1.6R PEP R IG ...

PURGE - PURGING.]..]I)) ................. -----------........ ........ 1- 1-5 1.27 QUADRANT POW TILT 1-5RAI 1.28 RATEDTHER LPOWER.1-5 1.29 REACT IP SYSTEM RESPONSE P IMEi-5,SE T . . . . 1-5 1.30 REPOR ABLE EVENT.................. 1-5 1.31 CONTAINMENT ENCLOSURI'IE'BU'ILD*ING ET 1-5 1.32 SHUTDOWN MARGIN .................................... 1-6 1.33 SITE BOUNDARY .......................................... 1-6 1.34 SLAVE RELAY TEST ........................................ 1-6 1.35 (NO T US ED )............ .................................... .............................. ................ 1-6 1.36 SOURCE CHEC ........................... ................ 1-6 1.37 STAGGERE EST BASIS..............1-6 1.38 THERMA OWER ....................................... 1-6 1.39 TRIP UATING DEVIC.E. OPERATIONANL" T"1.............................. 1-6 1.40 UNI NTIFIED LEAKAGE 1.41 RESTRICTED AREA 1-7 1.42 (NOT USED) ...... D.A. ......................... . ..... ................................ 1-1.43 VENTING TABLE 1.1 FREQUENCY NOTA N. ................................ 1-8 TABLE 1.2 OPERATIONAL MODES 1-8 SEABROOK - UNIT 1 i Amendment No. 8, 66

INDEX 20SAFETY LIMITS AND UV*q6ýNG/SAFETY SYSTEM SETTINGS SECTIO.__N

• PAGE

.1 AC ETY T C S SL)I.......... ......... ......................... ......... 2-1 2.1 3A F TY LIM T

• I

• I NS--------------.........................

......................... "......2-2.1.2 REACTOR COOLANT TEM PRESSURE SL .............. 2-1 2.1.3 SAFETY LIMIT VIO IONS..................2-1 FIGURE 2.1-1 (TH IGURE IS NOT USED) ............................ . 2-2 2..* REACTOR TRIP SYSTEM INSTRUMENT N SETPOINTS........... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM TRUMENTATION TRIP SETPOINTS 2-4 BASES 2.1 SAFETY LIMI.T's (SWs 2.1.1 RE ORCORE SLs.................. .. ..................... .. B2-1 2.1.2 (1EfACTOR COOLANT SYSTEM PRESS URE SL..................... B 2-2a 2.1.3 SAFETY LIMIT VIOLATIONS .................................................... B-2-2b 2.2 LIMITING SAFETY SYSTEM SETTIN6 2.2.1 REACTOR TRIP SYSTEM I RUMEN TATION SETPOINTS ........... B 2-3 3.0/4.0 LIMITING CONDITI I'll FOR OPERATI( ON AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLIC IT.. .... .................................. o3/40-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - Tavg Greater Th "20F F............................. 3/41-1 Shutdown Margin - Tavg Less T Equal to 2001F........-'or ......... 3/4 1-3 Moderator Temperature Co cient nt..........

....... ....................... * .......... . . . . 3/4 1-Minimum Temperature Criticality ............. ................... 1-6 SEABROOK - UNIT 1 ii Amendment No. 96

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEIU.ANG-E REQUIREMENTS SECTION PAGE 3/4.1.2 BORATION SYSTE Isol n of Unborated Water Sources - Shutdown 3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height .... ................. .... ...................................... 3/4 1-15 TABLE 3.1-1 ACCIDENT ANA ES REQUIRING REEVALUATION IN THE EVENT 6F INOPERABLE FULL-LENGTH ROD............ 3/4 1-17 Position ication Systems - Operating ..................................... ......... 3/4 1-18

- Shutdown .......... ..............

Pood Drop Time...... System . . .. . ........

on Indication 3/4 1-19 ShudDowropT i .............

nse rto ..................................... ..... .................................. 3/4 1-20 Shutdown Rod Insertion Limit 3/4 1-21 Control Rod Insertion Limits 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFEREN 3/4 2-1 3/4.2.2 HEAT FLUX HOT C NEL FACTOR F (Z.............)...................

-... ...... 3/4 2-4 3/4.2.3 NUCLEAR ENT PY RISE HOT CHANNEL FACTOR 3/4: 2-8 3/4.2.4 QUADRANT WER TILT RATIO 3/4 2-9 3/4.2.5 DNB P METERS...................................... 3/4 2-10 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRU TATION.... ............... 3/4. 3-1 TABLE 3.3-1 REACTOR TRIP S EM INSTRUMENTATION 3/4 3-2 SEABROOK- UNIT 1 iii endment No. 8, 93

ING CONDITIONS FON .. ....... SURVEILLANCE REQUIREMENTS .... 3ND 3Y SECTIOABLN AGYE TABLE 3.-5 (This table number is not used)

TA B--*.3-1 REACTOR TRIP SYSTEM INSTRU TATION SURVEILLANCE

~REQUIREMENTS ....... * ................ 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FE RES ACTUATION SYSTEM RaIoN MENTATIOn For.P.a.Operations ......................... 3/4 3-16 TABLE 3.3-6 INSTRUMENTATION RGINAIOD SAFETY FEATURES ACTUATION SYSM 3UVELLN3EQIRMET INST RUMENTATION S....... FORP 3/4P3-2N 3/4.E3.3- MOITRNGINEDSTRUENTYFATURSATUTOSSE TAB 4.3-2 ENGINEERED SAFETY FEATURES ACT kAION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ........... 3/4 3-31 3/4.3.3 MONITORING INSTRUMENTATIO Radiation Monitoring For NUMBperations IS...N.......

........................ ....... 3/4 3-36 TABLE 3.3-6 RADIATION 3NITORING INSTRUMENTATION FOR P OPERATIONS NUMBER.IS.O3/4T TABLE 4.3-3 (AB NUMBER ISTNO US T TABE.3.TIONS NURVEILLANCEREQUIREMENTS...... ..... 3/4 3-39 (THIS SPECIFICATION NUMBER IS NOTUUSD.............. ............... 3/4 3-40 TABLE 3.3-7 (THIS TABLE NUMBER IS NOT US D ).....................

....................... 3/4 3-42 TABLE 4.3.4 (THIS TABLE NUMBER IS N T ED) .................................................. 3/4 3-43 (THIS SPECIFICATION MBER IS NOT USED) ................................ 3/4 3-44 TABLE 3.3-8 (THIS TABLE :fvBER IS NOT USED) .................................................. 3/43-45 Remo hutdown System ........................................ "-3/4 3-46 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM..............3/43-47 Accident Monitoring Instrumentation........................ 3/4 3-49 TABLE 3.3-10 ACCIDENT MONITORING IN MENTATION . ......... 3/4 3-50 TABLE 3.3-1 1 (This table number is n ed) .................................. 3/4 3-53 Radioactive Liquid Effluent Monitoring Instrumentation ..........................

TABLE 3.3-12 (THIS TABLE NUMBER IS NOT USED) ..................... 3/4 3-56 SSEABROOK - UNIT 1 iv Amendment No. 59, 66

~INDEX CONDITIONS FOR OPERATIONP97-ý§-UREILLANCE REQUIREMENTS SECTION, T

• PAGE_

TABLE 4.3-5 (THIS LE NUMBER IS NOT USED) ......................... 3/4 3-58 xplosive Gas Monitoring Instrumentation .................... 3/43-60 TABLE . -13 EXPLOSIVE GAS MONITORING INS MENTATION .......... 3/4 3-61 T 4.3-6 EXPLOSIVE GAS MONITORING TRUMENTATION SURVEILLANCE REQUIRE TS........................ 3/43-64 3/4*3.4 (THIS SPECIFICATION N ER IS NOT USED) ............................................ 3/4 3-67 3/4A4 REACTOR COO i SYSTEM 3/4,4.1 REAC"T OOLANT LOOPS AND COOLANT CIRCULA

ý--*S artup and Power Operation ............................... ...... .................................. 3/4 4-1 Hot Standby ..................................................................................... 3/4 4-2 Hot Shutdown 3/44-4 Cold Shutdown - Loops Filled ..... .......................................... 3/4 4-6 Cold Shutdown - Loops Not F d ....................................................................... 3/4 4-7 3/4.4.2 SAFETY VALVES S tdown ..................................... 3/44-8 perati ................................................................................................ 3/4 4-9 3/4.4.3 PRESSURIZER 3/44-10 3/4.4.4 RELIEF VALVES. 3/44-11 3/4.4.5 STEAM GENERATORS (SG) TUBE EGRITY. .......... ........ 3/4 4-13 3/4.4.6 REACTOR COOLANT SYS LEAKAGE Leakage Detection tems .................................... 4 4-14 Operational Le ge ......................................................................... 3/4 4-15 I

3/4.4.7 (THIS SPECIFICATION NUMBER IS NOT USED)........ ......................... 3/4 4-18 I 3/4.4.8 SPECIFIC ACTIVITY. .................................... 3/4 4-19 II SEABROOK- UNIT 1 v A ndment No. 5,Q-66,-93, 115 Corrected By Letter Dated May 17, 2007

,INDEX SLIMITING CONDITIONS FOR OPERATION AND SSURVýNCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE E ALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY VERSUS PERCENT OF RATED THERMAL POWER WIT REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/gram-E EQUIVALENT 1-131.................................... 3/44-20 ABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVIT MPLE AND ANALYSIS PROGRAM ..................................... 3/4 4-21 3/4.4.9 PRESSURE/TEMPERATURE I- S General ............. 34 3/4...............................

4-22 FIGURE 3.4-2 RE R COOLANT SYSTEM HEATUP LIMITATIONS -

APPL LE UP TO 11. 1 EFPY................3/44-23 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -

APPLICABLE UP TO 11.1 EFPY.3/4 4-24 Pressurizer.

Overpressure75Protection . .... Systems--------------------------iiiiiiill3/4

.... ....................... ................................ 3/4 4-25 4-26 FIGURE 3.4-4 RCS COLD OVERPRES E PROTECTION SETPOINTS 3/44-29 3/4.4.10 DELETED ............................................... 3/4 4-30 3/4.4.11 REACTOR CO T SYSTEM VENTS ......................... 3/4 4-31 3/4.5 ENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Hot Standby, Startup, and Power Operation ...... . ..................... 3/4 5-1 Shutdow n........................................... . ...... .......................................... 3 4 5-3 3/4.5.2 ECCS SUBSYSTEMS - Tavg G TER THAN OR EQUAL TO 350F ...... 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS EO S S T* -

_T0oT ESS THAN 350°F sT/5-10 ECCS SYBSYSE - Tavg Equal To or Less Tha F........ .

3/4.5.4 REFUELI ER STORAGE TANK .................... ...... 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity ............................. .. ....................................... 3/4 6-1

\ C nta nm e t Leakage Containment L aka e...................

........... ........ 3/46-2 SEABROOK - UNIT 1 vi Amendm o. 70, 89,115,-126 Corrected By Lette Dated May 17, 2007

INDEX LIITING CONDITIONS FOR OPERATIO _N4 USUR ELLANCE REQUIREMENTS PAGE SECTION 3/4 6-7 Coj: ment Air Locks ..........

nternal Pressure ----------------------------------------- 3/4 6-9 Air Ter.............................

Tem perature .................................................. 3/4 6-10 Containment Vessel Structural Integrity ...... . ..................... .....................

........................................................ 3/4 6-1 6-11 Contaiment ystem ..... ....

• entiltion System----------------------3/46-12 Containment Ventilation 3/4.6.2 DEPRESSURIZATION A OLING SYSTEMS Containm ent ay System................................................................................... 3/4 6-14 S pray itive System........................................................................................... 3/4 6-15 3/4.6. CONTAINMENT ISOLATION VALVES 3/4.6.4 COMBUSTIBLE GAS CONTROL (THIS SPECIFICATION NUMBER IS NOT t E D) ........................................... 3/4 6-1 9 (THIS SPECIFICATION NUMBER IS NON Hydrogen Mixing System .............. ...... 3/4 6-20 3/4.6.5 CONTAINMENT ENCLO E BUILDING Containment E sure Emergency Air CleE 314 6-25 Containme Siyse ................................

nclosure Building Integrity ..... anup.......................

Cont .................. 3I4. 6-25 ent Enclosure Building Structural I 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves ................................. 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLO LE POWER R ANGE NEUTRON FLUX HIGH SETPOINT WITH I ERABLE STEAM LINE SAFETY VALVES DURIN FOUR LOOP RATIONS .......... LOOP ii.iiiiiiil 3/43/7-2 7-2 TABLE 3.7-2 S MLINE SAFETY VALVESSPE Auxiliary Feedwater System ........................... 3/4 7 -7 Condensate Storage Tank ..............................

• Specific Activity ........................

ý ED )......................................... ..... k -

TABLE 4.7-1 (THIS TABLE NUMBE NOT USE 7-9 L . ....... ........ 3/4 Main Steam Line Isolation Valves Atmospheric Relief Valves Amendment No.-1.3/4 7-10 JAmendment No.-14, -2, 99 SEABROOK - UNIT 1 vii

INDEX LIMTIG ONDITIONS FOR OEAXTION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.2/-'S'TEAM GIENEERATOR PRESSURE/TEMPERATU IY*IMITATION ....... 3/4 7-11 3/47 PRIMARY COMPONENT COOLING WATE STEM 3/47-12

'/4.7.4 SERVICE WATER SYSTEM / ULTIMAT EAT SINK ................ 3/4 7-13 3/4.7.5 (THIS SPECIFICATION NUMBER OT USED) .............................. 3/4 7-14 3/4.7.6 CONTROL ROOM SUBSYSTT. .. .......................... 3/47-16 Emergency Makeup Air Filtration............... 3/47-16 Air Conditioning............................................. 4 7-18a 3/4.7.7 SNUBBERS.. ............................................ 3/4 7-19 3/4.7.8 SEALED RCE CONTAMINATION..............3/4 7-20 3/4.7.9 T* PECIFICATION NUMBER IS 3/47-22 3/4.7.1 HIS SPECIFICATION NUMBER IS NOT USED3/47-23 TAB E 3.7-3 (THIS TABLE NUMBER IS NOTU ....................................... 3/47-24 3/4.8 ELECTRICAL POWER SYSTEMS*

Operating ......... . ............................... ................................................ 3/4 8-1 TABLE 4.8-1 (TH ABLE NUMBER IS NOT USED) ....... ............. 3/48-10 Shutdown .......................................

3/4.8.2 D.C. SOURCES O perating .............................................. .............................................. 3/4 8-12 TABLE 4.8-2 BATTERY SURVEIL CE REQUIREMENTS 3/48-14 Shutdown ........................................

3/4.8.3 ONSITE POW ISTRIBUTION Operatin................................................... .............................. 3/4 8-16 Shut n "..---. 3/48-18 Trip Circuit for inverterIi-2A..2 ...................... ..- 3/4 8-19 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTI EVICES A.C. Circuits Inside Primary Co ment ........................

Motor-Operated Valves T al Overload Protection..... ... ....... 3/4 8-24 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 3/49-1 3/4.....................

3/4 9.2 INSTRUMENTATION ............................... 3/4 9-2 3/49.3 DECAY TIME 3/49-3 SEABROOK - UNIT 1 viii Amend ntNo. 5, 59,638, 125

INDEX*

LIIIGCONDITIONS FOR OPERATION A NQ-:ýELACE REQUIREMENTS 3/4.9.4 3/4.9.5 CONTAINM...

(THIS ,PCIFICATION .. NUMBER BUILDING PENETRATIONS IS NOT UIE * .

........ []II((. . 3/4 3/4 9-4 9-5 349.96 (THIS SPECIFICATION NUMBER IS NOT USED) ................... 3/4 9-1 3/4,9.10 WATERIS LEIFIRATIONUR VESSEL.....~~......................3/4 9-3/4.9.8 RESIDUAL HEAT REMOVAL EM ANG C E CYIRCLNLTION High W ater Low Water LevelLevel

• (( ...... ...................................... 3/4 9-8 3/49-9 3/4 9.9 (THIS SPECIFICNL S BL ISTOT. ......................... 3/4 9-10 3/49.10 WATER LEVELN-R TOR VESSEL STORAGE ................... 3/49-11 3/49.11 WATER LEVELGTORAGE POOL ........................... _4 9-12 3/4.9.12 FUEL ST , INEIN, ANDRPOWEY STR LLEANIN0 34U.. 3. C T STS....................... . .......... ......... ........ 3/4" -3 3/49.10.4 ( SPE S CIFICATIN P NTORAGEU...................... .UMBER. ........... 3/49-10 3/4.9.14 ICATIONSY.T.-HUTD EW FUEL ASSEMBLNDE ......... ....... 3/49-18 3/4.10 SPECIAL TEST EX N TI I3NS 3/4.10.1 SHUTDOWNAGIN ........................................ 3/410-1 3(4.10.2 3/4.Q.31,0. GRUSP GHT, ....................................

I_. 5 CS TESTS ISNPSCTIO N, AUND POWE. ... UM.ITS.... STRIBUTIN LN

...................................................... 3/410-2 3/4 10-3 3/4.10.4 (THIS SPECIFICATION NUMBER ISOT USED) . ............................ 3/410-46 3/4 .10.5 P O S ITIO N INDICAT IO N S Y S ~ g-S HUT DO W N . ................... 3/4 10-5 3/4.11 RADIOACTIVE EFFL ,'TS 3/4.11.1 LIQUID EFFLUENTIO M-(THIS SPECIFTION NUMBER IS NOT USED) ................... 3/411-1 (THIS SPE *ICATION NUMBER IS NOT USED) ................................

• 3/411-2 (THIS SPP IFICATION NUMBER IS NOT US ......................... 3/4 11-3 Liquid/H oldup Tanks ..................................... ..................................................... 3/4 11I-4 3/4.11.2 GAy ,OUS EFFLUENTS (TFIS SPECIFICATION NUMBE*A NOT USED) . ............ ................. 3/411-5 rHIS SPECIFICATION NU R SNO SE )......................... 3/4 11-6

'(THIS SPECIFICATIO MBER ISNOT USED)......................3/1 7 (THIS SPE F N N MBE I NO US D)... ........................... 3/4 11-8 Explo " s Mixture - System ............................. . ................................ 3/4 11-9 3/4.11.3 (THIS SPECIFICATION NUMBER IS NOT Us ............................................ 3/411-10 3/4.1.4 (THIS SPECIFICATION NUMBER IS NOT U I D) ........................................... 3/411-12 3/412 RADIOLOGICAL ENVIRONMENTAL ONITORING 3/4.12.1 (THIS SPECIFICATION NUMB RI O S D ..................................... .... / 12-1 SEABROOK - UNIT 1 Ix Amendment No. 6, 34, 66,:72,--95, 91

FOR

,INDEX OPERATON..lq`SURVE,.LANOE REQUIREM LIMI.TING CONDITIONS SECTION 3/4.12.2 (THIS SPECIFICATI NUMBER IS NOT USED) ...............................

PAGE 3/4 12-3

\

3/4.12.3 (THIS SPECIFIC ION NUMBER IS NOT USED)............ ....... 3/4 12-5 3.R/4.o BASeci I

~~.Lyll,u II3Ir-AI umr(Lo 5.1 SITE 5.1.1 EXCLUSIONAREA ................................................................ .......... 5-1 5.1.2 LOW POPULATI ZONE ......................... 5-1 5.1.3 MAPS DEFI G UNRESTRICTED AREAS AND SI BOUNDARY R RAD ACTIVE GASEOUS AND LIQUID EF UENTS 1 FIGURE 5.1- SITE AND EXCLUSION AREA BO DARY ----------------- 5-3 FIGURE 5.1-2 LOW POPULATION ZONE -5 FIGURE 5.1-3 LIQUID EFFLUENT DISCHA E LOCATION...5-7 5.2 CONTAINMENT 5.2.1 CONFIGURATION .................................. 5-1 5.2.2 DESIGN PRES RE TEMPERATURE5-5.3 REACTOR 5.3.1 FUEL ASSEMBLIES ........................................ 5-9 5.3.2 CONTROL ROD ASSEMBLIES ................................ 5-9 5.4 REACTOR COOLANT SYSTE 5.4.1 DESIGN PRESSURE A TEMPERATURE 5-9 5.4.2 VOLUME ............................................... 5-9 5.5 (OTHIS P NUMBER IS NOT USED) .................... 5-9 5.6 FUEL STORAGE 5.6.1 CRITICALITY 5-10 5.6.2 DRAINAGE ...................... 5-10 5.6.3 CAPACITY .............................................. 5-10 SEABROOK- UNIT 1 x Amendment No. 50,66,74, 89, 93,115, 116

5.0 DESIGN FEATURES CNO SECTION PG 5.7 COMPON *CYCLICT OR TRANSIENT LIMIT .. 5-10 TABLE 5.7-1 OR I ONENT........................ 5-11

,6.0/ ADMINISTRATIVE CONTROLS TABLE 6.2-1 IMUM SHIFT CREW COMPOSITION 63....................

6.2.3 IS SPECIFICATION NUMBER IS NOT ED).................................. ........... 6-4 6.2.4 SHIFT TECHNICALADVISOR.. ..... ........................ 6-4 6.3 (THIS SPECIFICATION NUM IS NOT USED) 6.4 THIS SPEC IFICATION NUMBER IS NOT USED) ........................ 6-4 6.5 (THIS SPECIFICATION NUMBER IS NOT USED ........................... 6-4 6.6 (THIS SP IFICATION NUMBER IS NOT USED) 6-4 6.7 P CXEDURES AND PROGRAMS 6 6.8 *St Jru R. uesport ZREPORTING REQUIREMENTS ...................

............ ........................ 6-14 6.8.1 StnartUp ROUTINE Report REPORTS....................6-14 A nnual R eports ----------------------------- *. ... .""......... ..- 6 1 Annual Radiological Environmental Operating R ort............................6-15 Annual Radioactive Effluent Release Repo ..... ....................................... 6-15 CORE OPERATING LIMITS REPORT........................ 6-16 Steam Generator Tube Inspection R ort....... ........... ................ 6-21 SEABROOK - UNIT 1 A Arendment No. , 63, 89, 116

lr*ý INDEX 6.0 ADMINISTRATIVE CONTROLS SECTION PAGE 6.8.2 SPECIAL REPORTS .............. .......................... 6-21 6.9 (TH ISSPECT, 6.9 THS SPEC ONNU .MER

....... N. OT .... ..ii ATION NUMBER IS NOT USED i.

i.

iii..

.. 62

.......6..-21 6.10 RA AT~iON PROTECTION PROGRAM..............6-22

6. HIGH RADIATION AREA 62 6.12 PROCESS CONTROL PROG (PCP) .......................... 6-23 6.13 OFFSITE DOSE CALC, "*TION MANUAL (ODCM) .................. 6-23 6.14 MAJOR CHA 'S TO LIQUID, GASEOUS, AND SOLID RADWAST TREATMENT SYSTEMS . .--. 6-24 6.15 (eKTAINMENT LEAKAGE RATE TESTING PROG . 6-25 SEABROOK - UNIT 1 xii A~ecr~nt No. 6,66ý, 85, 91,116

a -' ,

TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED

1. Gross Radioactivity At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. 1,2,3,4 Determination

2. Isotopic Analysis for DOSE 1 per 14 days. I EQUIVALENT 1-131 Concentration

3. Radiochemical for E Determination* 1 per 6 months** I

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the 1#, 2#, 3#, 4#, 5#

Including 1-131, 1-133, and 1-135 specific activity exceeds 1 pCi/gram DOSE EQUIVALENT 1-131 or 100/* Ci/gram of gross radioactivity, and b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 1, 2, 3 following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period.

• A radiochemical analysis for F shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in the reactor coolant.

The specific activities for these individual radionuclides shall be used in the determination of F for the reactor coolant sample. Determination of the contributors to F shall be based upon those energy peaks identifiable with a 95% confidence level.

• • Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

1. Until the specific activity of the Reactor Coolant System is restored within its limits.

SEABROOK - UNIT 1 3/4 4-21 Amendment No. 4.6-

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.8.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

1. Cycle dependent Overpower AT and Overtemperature.AT trip setpoint parameters and function modifiers for operation with skewed axial power profiles for Table 2.2-1 of Specification 2.2.1.

2. Cycle dependent maximum allowable combination of thermal power, pressurizer pressure and the highest operating loop average temperature (Tavg) for Specification, 2.1.1

3. SHUTDOWN MARGIN and minimum boron concentration limits for MODES 1, 2, 3, and 4 for Specification 3.1.1.1.

4. SHUTDOWN MARGIN and minimum boron concentration limits for MODE 5 for Specification 3.1.1.2.

5. Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3.1.1.3.

6. The minimum boron concentration for Modes 4, 5, and 6 for Specification 3.1.2.7.

7. Shutdown Rod Insertion limit for Specification 3.1.3.5.

8. Control Rod Bank Insertion limits for Specification 3.1.3.6.

9. AXIAL FLUX DIFFERENCE limits for Specification 3.2.1

10. Heat Flux Hot Channel Factor, FRTP -Q and K(Z) for Specification 3.2.2.

11. Nuclear Enthalpy Rise Hot Channel Factor, and FRP for Specification 3.2.3.

12. Cycle dependent DNB-related parameters for reactor coolant system average temperature (Tavg), and pressurizer pressure for Specification 3.2.5.

13. The boron concentration limits for MODES 1, 2 and 3 for Specification 3.5.1.1.

14. The boron concentration limits for MODES 1, 2, 3 and 4 for Specification 3.5.4.

15. The boron concentration limits for MODE 6 for Specification 3.9.1.

SEABROOK - UNIT 1 6-16 Amendment No. 66,104,441-*-