Text

Enclosure to TSTF-507, Revision 0, Part 1 of 3
	ML081900432
Person / Time
Site:	Technical Specifications Task Force
Issue date:	07/03/2008
From:	; Technical Specifications Task Force
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML081900424	List: ML081900432; ML081900442; ML081900443; TSTF-08-10, Transmittal of TSTF-507, Revision 0, CLIIP Traveler to Increase Standardization;
References
	TSTF-507, Rev 0, TSTF-6-A, Rev 1, WOG-3.1, Rev 0
	Download: ML081900432 (295)
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{{#Wiki_filter:Enclosure to TSTF-507, Revision 0 Part 1 of 3

WOG-3.1, Rev. 0 TSTF-6-A, Rev. 1 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Add Exception for LCO 3.0.7 to LCO 3.0.1 NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 2) Consistency/Standardization Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Steve Wideman, (620) 364-4037, stwidem@wcnoc.com Add exception for LCO 3.0.7 to LCO 3.0.1 This change completes Revision 0 change NRC-03, C.5 which added LCO 3.0.7 to address test exception LCOs and was omitted by the original change. The CEOG, BWR-4 and BWR-6 ITS already contain this change. Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 15-Mar-95 Owners Group Comments: WOG-03, C.1 Owners Group Resolution: Approved Date: 11-Aug-95 TSTF Review Information TSTF Received Date: 05-Sep-95 Date Distributed for Review: 05-Sep-95 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 05-Sep-95 NRC Review Information NRC Received Date: 03-Oct-95 NRC Comments: 10/6/95 J. Leuhman review complete, accept change. 10/6/95 - to C. Grimes for review. 11/17/95 - C. Grimes approved change. Final Resolution: Superceded by Revision Final Resolution Date: 08-Jan-96 TSTF Revision 1 Revision Status: Active 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-3.1, Rev. 0 TSTF-6-A, Rev. 1 TSTF Revision 1 Revision Status: Active Revision Proposed by: TSTF Revision

Description:

Remarked the pages to use TSTF number instead of OG number. TSTF Review Information TSTF Received Date: 08-Jan-96 Date Distributed for Review: 08-Jan-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 08-Jan-96 NRC Review Information NRC Received Date: 08-Jan-96 NRC Comments: Date of NRC Letter: 27-Sep-96 1/30/96 - Changes processed and certified. 1/31/96 Control books, database, and TS+BBS updated. 2/1/96 TSTF-06 change approved in letter to NEI. Final Resolution: NRC Approves Final Resolution Date: 01-Feb-96 Affected Technical Specifications LCO 3.0.1 LCO Applicability 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

LCD Applicability 3.0 T,57F-6 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCD 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCD 3.0,~,_'_~~

~~

LCO 3.0.2 Upon discovery of a failure to meet an LCD, the Required Actions of the associated Conditions shall be met, except as provided in LCD 3.0.5 and LCD 3.0.6. If the LCD is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s} is not required, unless otherwise stated. LCD 3.0.3 When an LCD is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCD is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:

a. MODE 3 within 7 hours;

b. MODE 4 within 13 hours; and

c. MODE 5 within 37 hours.

Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCD 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. LCD 3.0.4 When an LCD is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This (continued) SWOG STS 3.0-1 Rev 1, 04/07/95

LCO Applicability 3.0

                                                                  -rSTP-6 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1         LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.~J ~§j)

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:

a. MODE 3 within 7 hours;

b. MODE 4 within 13 hours; and

c. MODE 5 within 37 hours.

Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This (continued) WOG STS 3.0-1 Rev 1, 04/07/95

WOG-6, Rev. 0 TSTF-19-A, Rev. 1 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Relocate the details of RTD and thermocouple calibration from the Channel Calibration definition NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 2) Consistency/Standardization Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Steve Wideman, (620) 364-4037, stwidem@wcnoc.com Relocate the details of RTD and thermocouple calibration from the Channel Calibration definition to the Bases associated with calibration of these components. The details associated with defining acceptable means by which a channel calibration can be accomplished for RTDs and thermocouples is proposed to be relocated to the Bases associated with the calibration of these components. The information contained in the definition is prescriptive in nature, better suited as Bases information consistent with other material relocated to the Bases. Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: Ginna Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 02-Nov-95 Owners Group Comments: (No Comments) Owners Group Resolution: Approved Date: 02-Nov-95 TSTF Review Information TSTF Received Date: 02-Nov-95 Date Distributed for Review: 02-Nov-95 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 14-Nov-95 NRC Review Information NRC Received Date: 16-Nov-95 NRC Comments: 11/19/95 - reviewer modified package. 12/7/95 - pkg to C. Grimes to review 6/11/96 - C. Grimes comment: TSTF-19 to be referred to a Tech Br. 9/18/96 - No change in status. 10/30/96 - Awaiting ICSB for support. 12/31/96 - NRC requested changes to TSTF-19. TSTF considering. 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-6, Rev. 0 TSTF-19-A, Rev. 1 OG Revision 0 Revision Status: Closed 1/31/97 - Revision sent to TSTF for review. 3/28/97 - Revision forwarded to NRC. Final Resolution: Superceded by Revision Final Resolution Date: 07-Apr-97 TSTF Revision 1 Revision Status: Active Revision Proposed by: NRC Revision

Description:

Letter from C. I. Grimes to James Davis dated 12/31/96 requested modifications to TSTF-19. 1) Remove the sentence, ",which may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel," and 2) add inserts similar to Insert 1 and 2 to BWOG SR 3.3.1.6, SR 3.3.17.2, SR 3.3.18.3, WOG SR 3.3.3.2, 3.3.4.3, and CEOG (analog and digital) 3.3.11.2 and 3.3.12.3. TSTF Review Information TSTF Received Date: 31-Dec-96 Date Distributed for Review: 03-Feb-97 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 21-Mar-97 NRC Review Information NRC Received Date: 07-Apr-97 NRC Comments: 4/10/97 - Forwarded to reviewer. 10/6/97 HICB recommended approval. 10/8/97 - to C. Schulten for TSB recommendation. 10/1/97 - NRC indicated that they would approve. 10/6/97 - HICB recommended approval. Final Resolution: NRC Approves Final Resolution Date: 01-Dec-97 Affected Technical Specifications 1.1 Definition of Channel Calibration SR 3.3.1.6 Bases RPS Instrumentation NUREG(s)- 1430 Only SR 3.3.17.2 Bases PAM Instrumentation NUREG(s)- 1430 Only SR 3.3.18.3 Bases Remote Shutdown System NUREG(s)- 1430 Only SR 3.3.1.12 Bases RTS Instrumentation NUREG(s)- 1431 Only SR 3.3.3.2 Bases PAM Instrumentation NUREG(s)- 1431 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-6, Rev. 0 TSTF-19-A, Rev. 1 SR 3.3.4.3 Bases Remote Shutdown System NUREG(s)- 1431 Only SR 3.3.11.2 Bases PAM Instrumentation (Analog) NUREG(s)- 1432 Only SR 3.3.11.2 Bases PAM Instrumentation (Digital) NUREG(s)- 1432 Only SR 3.3.12.3 Bases Remote Shutdown System (Analog) NUREG(s)- 1432 Only SR 3.3.12.3 Bases Remote Shutdown System (Digital) NUREG(s)- 1432 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-19,1?~ -1 INSERT 1 Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature detectors (RID) sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. INSERT 2 Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

Definitions 1.1 .-", TS rF/~ Rev. 1-1.0 USE AND APPLICATION 1.1 Definitions

     ---------------------~-------~-------NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Definition ACTIONS ACTIONS shall be that part of a Specification that

                                *prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ALLOWABLE THERMAL POWER ALLOWABLE THERMAL POWER shall be the maximum reactor core heat transfer rate to the reactor coolant permitted by consideration of the number and configuration *of reactor coolant pumps (RCPs) in operation. AXIAL POWER IMBALANCE AXIAL POWER IMBALANCE shall be the power in the top half of the core, expressed as a percentage of RATED THERMAL POWER (RTP), minus the power in the bottom half of the core, expressed as a percentage of RTP. AXIAL POWER SHAPING APSRs shall be control components used to control RODS (APSRs) the axial power distribution of the reactor core. The APSRs are positioned manually by the operator and are not trippable. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of ~~i~ adjustable devices in the channel.~: .~ (continued) BWOG STS 1.1-1 Rev 1, 04/07/95

Definitions 1.1 1.1 Definitions TSTF-/% ae.v,:L~

j CHANNEL CALIBRATION se.. !;II; element is '@f'llee&i; "he "8M" '8""~'8d (continued) CI tAHHEl ALI8IM1'ION shall ; "81 .... 8 I" iR,l ii8 811"S calibrath" .hal compar8s "hi ,.h81 88RI;",

818111ellt! w; th .hl reee""h ; 115 tall eci S8"3; Ii! by means' of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated. The CHANNEL CALIBRATION shall also include testing of safety related Reactor Protection System (RPS), Engineered Safety Feature Actuation System (ESFAS), and Emergency Feedwater Initiation and Control (EFIC) bypass functions for each channel affected by the bypass operation. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarms, interlocks, display, and trip functions. The ESFAS CHANNEL FUNCTIONAL TEST shall also include testing of ESFAS safety related bypass functions for each channel affected by bypass operation. CONTROL RODS CONTROL RODS shall be all full length safety and regulating rods that are used to shut down the reactor and control power level during maneuvering operations. CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactiVity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE (continued) BWOG STS 1.1-2 Rev 1. 04/07/95

RPS Instrumentation B 3.3.1 T[. TF-I<=j 'Rev, '1 I BASES SURVEILLANCE SR 3.3.1.6 (continued) REQUIREMENTS A CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift to ensure that the instrument channel remains operational between successive" tests. CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are within the assumptions of the unit specific setpoint analysis. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint analYSiS.) The Frequency is justified by the assumption of an [18] month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. ('-_ ~

                                                                      \;l;lSe~!:y SR 3.3.1. 7 This SR verifies individual channel actuation response times are less than or equal to the maximum values assumed in the accident analysis. Individual component response times are not modeled in the analyses. The analyses model the overall, or total, elapsed time from the point at which the parameter exceeds the analytical limit at the sensor to the point of rod insertion. Response time testing acceptance criteria for this unit are included in Reference 1.

A Note to the Surveillance indicates that neutron detectors are excluded from RPS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. Response time tests are conducted on an [18] month STAGGERED TEST BASIS. Testing of the final actuation devices, which make up the bulk of the response time, is included in the testing of each channel. Therefore, staggered testing results in response time verification of these devices every [18] months. The [18] month Frequency is based on unit operating experience, which shows that random failures of (continued) BWOG STS B 3.3-29 Rev 1, 04/07/95

PAM Instrumentation B 3.3.17 TS7f'-/9, Rwr1"* BASES SURVEILLANCE SR 3.3.17.2 REQUIREMENTS (continued) A CHANNEL CALIBRATION is performed every [18] months or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. This test verifies the channel responds to measured parameters within the necessary range and accuracy. A Note clarifies that the neutron detectors are not required to be tested as part of the CHANNEL CALIBRATION. There is no adjustment that can be made to the detectors. Furthermore, adjustment of the detectors is unnecessary because they are passive devices, with minimal drift. Slow changes in detector sensitivity are compensated for by performing the daily calorimetric calibration and the monthly axial channel calibration. For the Containment Area Radiation instrumentation, a CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr, and a one point calibration check of the detector below 10 R/hr with a gamma source. The Frequency is based on operating experience and consistency with the typical industry refueling cycle and is justified by the assumption of an [18] month calibration interval in the determination of the magnitude of equipment drift. REFERENCES 1. [Unit Specific Documents (e.g., FSAR, NRC Regulatory Guide 1.97 SER letter).]

2. Regulatory Guide 1.97.

3. NUREG-0737, 1979.

4. 32-1177256-00, "Technical Basis for Reactor Vessel Level Indication System (RVLIS) Action Statement,"

April 10, 1990. BWOG STS B 3.3-154 Rev 1, 04/07/95

Remote Shutdown System _B 3.3.18 BASES

                                                               -r:sTF- /7' Rt:.~.1.~

I ,. SURVEILLANCE SR 3.3.18.3 (continued) REQUIREMENTS because they are passive devices, with minimal drift. Slow changes in detector sensitivity are compensated for by performing the daily calorimetric calibration and the monthlY axial channel calibration.

  ~L9?rl-l                                             .

The Frequency is based on operating experlence and consistency with the typical industry refueling cycle and is justified by the assumption of an [18] month calibration interval in the determination of the magnitude of equipment drift. REFERENCES 1. 10 CFR 50, Appendix A, GOC 19. SWOG STS B 3.3-160 Rev 1, 04/07/95

Definitions 1.1 f S TF-/1 Re,/, 1. I 1.0 USE AND APPLICATION 1.1 Definitions

----------------------------~--------NOTE----~--------------------------------                .

The defined terms of. this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Definition ACTIONS ACTIONS shall be that part of a Specification that

                           . prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices. AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux (AFD) signals between the [top and bottom halves of a two section excore neutron detector]. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, display, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the

                                                , t le devi s 'n re~ ... ~ lie" GIIP.Ntl'b LIiAA+I8.1 Ihall ; R.llu.!  11ft itlpl aGe irarr i3l i grit.; 011 that Comp"!3 th! 0 ttfer SPAtiR! 8l8118Ate ~';i!h .he II~IIR*1) h"t1.lle.t
                                   . .                                              may e orme          eans of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

(continued) WOG STS 1.1-1 Rev 1, 04/07/95

RTS Instrumentation B 3.3.1 Ts( F/1 ReJ.1.

BASES SURVEILLANCE SR 3.3.1.11 (continued) REQUIREMENTS plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This Surveillance is not required for the HIS power range detectors for entry into MODE 2 or 1, and is not required for the NIS intermediate range detectors for entry into MODE 2, because the unit must be in at least MODE 2 to perform the test for the intermediate range detectors and MODE 1 for the power range detectors. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed on the [18] month Frequency. SR 3.3.1.12 SR 3.3.1.12 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10, every [18] months. This SR is modified by a Note stating that this test shall include

           / verification of the RCS reSi~ t~ure detector (RTD) bypass loop flow rate~~:..!.--J This test will verify the rate lag compensation for flow from the core to the RTDs.

The Frequency is justified by the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.1.13 SR 3.3.1.13 is the performance of a COT of RTS interlocks every [18] months.

The Frequency is based on the known reliability of the interlocks and the multichannel redundancy available, and has been shown to be acceptable through operating experience. (continued) WOG STS B 3.3-57 Rev 1, 04/07/95

PAM Instrumentation B 3.3.3 T5TF/~ ~ev.i BASES SURVEILLANCE SR 3.3.3.1 (continued) REQUIREMENTS should be compared to similar unit instruments located throughout the unit. Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized. The Frequency of 31 days is based on operating experience that demonstrates that channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels. SR 3.3.3.2 A CHANNEL CALIBRATION is performed every [18] months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. This SR is modified by a Note that excludes neutron detectors. The calibration method for neutron detectors is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." The Frequency is based on operating experience an consistency with the typical industry refueling cycle. REFERENCES 1. [Unit specific document (e.g., FSAR, NRC Regulatory Guide 1.97 SER letter).]

2. Regulatory Guide 1.97, [date].

3. NUREG-0737, Supplement-I, "TMI Action Items."

WOG STS B 3.3-137 Rev 1, 04/07/95

Remote Shutdown System B 3.3.4 Ts "TF-I'1, 'Rev.-1 BASES SURVEILLANCE SR 3.3.4.1 (continued) REQUIREMENTS within the criteria, it is an indication that the channels are OPERABLE. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized. The Frequency of 31 days is based upon operating experience which demonstrates that channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels. SR 3.3.4.2 SR 3.3.4.2 verifies each required Remote Shutdown System control circuit and transfer switch performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of the equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the unit can be placed and maintained in MODE 3 from the remote shutdown panel and the local control stations. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. (However, this Surveillance is not required to be performed only during a unit outage.) Operating experience demonstrates that remote shutdown control channels usually pass the Surveillance test when performed at the [18] month Frequency. SR 3.3.4.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. (continued) WOG STS B 3.3-142 Rev 1, 04/07/95

Definitions 1.1 7~TF/7. Re.J.1.. I /~ 1.1 Definitions lI CHANNEL CALIBRATION the entire channel, including the required sensor, (continued) alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining

                         . 'ustabl evic .         e channel. fJjl;1Wi_i\III~=---
                 )
                                               ~n~~ul~TION may e per ormed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or . status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog and bistable channels--the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarms, interlocks, display and trip functions;

b. Digital computer channels--the use of diagnostic programs to test digital computer hardware and the injection of simulated process data into the channel to verify OPERABILITY, including alarm and trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested. (continued) CEOG STS 1.1-2 Rev 1, 04/07/95

PAM Instrumentation (Analog) B 3.3.11 TS TF-t<f.I 'Rev.t. BASES SURVEILLANCE SR 3.3.11.1 (continued) REQUIREMENTS The Frequency of 31 days is based upon plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel during normal operational use of the displays associated with this LCD's required channels. SR 3.3.11. 2 A CHANNEL CALI BRATION is performed every [18] months or approximately every refueling. CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies the channel responds to the measured parameter within the necessary range and accuracy. A Note allows exclusion of neutron detectors from the CHANNEL CALIBRATION. r-At this unit, CHANNEL CALIBRATION shall find measurement ]

              ~rrors are within the following acceptance criteria:

For the Containment Area Radiation instrumentation, a CHANNEL CALIBRATION may consist of an electronic calibration of the r:r ~ l!::.r'I$e~ channel, not including the detector, for range decades above 10 R/hr, and a one point calibration check of the detector

              ~below 10 R/hr with a gamma source.

~~The Frequency is based upon operating experience and ~ consistency with the typical industry refueling cycle and is justified by an [18] month calibration interval for the determination of the magnitude of equipment drift. REFERENCES 1. Plant specific document (e.g., FSAR, NRC Regulatory Guide 1.97, SER letter).

2. Regulatory Guide 1.97.

3. NUREG-0737, Supplement 1.

4. NRC Safety Evaluation Report (SER).

CEOG STS B 3.3-151 Rev 1, 04/07/95

Remote Shutdown System (Analog) B 3.3.12 TS TF~ I~ Rev. i BASES SURVEILLANCE SR 3.3.12.3 (continued) REQUIREMENTS that the channel responds to the measured parameter within the necessary range and accuracy.~

 ~           The 18 month Frequency is based upon the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

The SR is modified by a Note, which excludes neutron detectors from the CHANNEL CALIBRATION. SR 3.3.12.4 SR 3.3.12.4 is the performance of a CHANNEL FUNCTIONAL TEST every 18 months. This Surveillance should verify the OPERABILITY of the reactor trip circuit breaker (RTCB) open/closed indication on the remote shutdown panels by actuating the RTCBs. The Frequency of 18 months was chosen because the RTCBs cannot be exercised while the unit is at power. Operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A, GDC 19,and Appendix R.

2. NRC Safety Evaluation Report (SER).

CEOG STS B 3.3-157 Rev 1, 04/07/95

PAM Instrumentation (Digital) B 3.3.11 rSTF~/'71 'Rev.,f BASES SURVEILLANCE SR 3.3.11.1 (continued) REQUIREMENTS which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel during normal operational use of the displays associated with this LCO's required channels. SR 3.3.11. 2 A CHANNEL CALIBRATION is performed every [18] months or approximately every refueling. CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies the channel responds to the measured parameter within the necessary range and accuracy. At this unit, CHANNEL CALIBRATION shall find measurement ] [ errors are within the following acceptance criteria: For the Containment Area Radiation instrumentation, a CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr, and a one point calibration check of the detector below 10 R/hr with a gamma source. The Frequency is based upon operating experience and consistency with the typical industry refueling cycle and is justified by the assumption of an [18] month calibration interval for the determination of the magnitude of equipment dri ft. REFERENCES 1. [Plant specific document (e.g., FSAR, NRC Regulatory Guide 1.97, SER letter).]

2. Regulatory Guide 1.97.

3. NUREG-0737, Supplement 1.

4. NRC Safety Evaluation Report (SER).

CEOG STS B 3.3-180 Rev 1, 04/07/95

Remote Shutdown System (Digital) B 3.3.12 7?:IF-I9, 'Rev.i. BASES SURVEI LLANCE SR 3.3.12.1 (continued) REQUIREMENTS CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized. The Frequency is based on plant operating experience that demonstrates channel failure is rare. SR 3.3.12.2 SR 3.3.12.2 verifies that each required Remote Shutdown System transfer switch and control circuit performs its intended function. This verification is performed from the reactor shutdown panel and locally, as appropriate. Operation of the equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the plant can be bro~ght to and maintained in MODE 3 from the reactor shutdown panel and the local control stations. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience demonstrates that Remote Shutdown System control channels seldom fail to pass the Surveillance when performed at a Frequency of once every [18] months. SR 3.3.12.3 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies (~'\ that the channel responds to the measured parameter within

~   ~~rf~     the necessary range and accuracy.~
  -           The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a (continued)

CEOG STS B 3.3-185 Rev 1, 04/07/95

BWOG-3, Rev. 0 TSTF-40-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Exempt RCP seal water injection or leakoff from the definition of Unidentified Leakage NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 1) Correct Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Paul Infanger, (352) 563-4796, paul.infanger@pgnmail.com Revise the definition of Unidentified Leakage from "All Leakage that is not identified leakage or controlled leakage." to "All leakage (except RCP seal water injection or leakoff) that is not identified leakage. The exception for controlled leakage from NUREG-0123 was revised in the definition of identified leakage to an exception for "RCP seal water injection or leakoff." However, the term was retained in the definition of Unidentified Leakage. This term, Controlled Leakage, should therefore be similarly revised and included as an exception in the definition of Unidentified Leakage since it is not considered to be Leakage, and the undefined term "Controlled Leakage" should not be used. This change is consistent with NUREG-1431. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: ANO-1 Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 24-Aug-95 Owners Group Comments: (No Comments) Owners Group Resolution: Approved Date: 08-Sep-95 TSTF Review Information TSTF Received Date: 22-Sep-95 Date Distributed for Review: 03-Oct-95 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: NUREG-1431 already has this change. N/A for BWRs. CEOG Accepts. TSTF Resolution: Approved Date: 25-Nov-95 NRC Review Information NRC Received Date: 03-Jan-96 NRC Comments: 6/11/96 - C. Grimes comment: decision on TSTF-40 to be made. 9/18/96 - no change in status 3/13/97 - Approved. 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWOG-3, Rev. 0 TSTF-40-A, Rev. 0 OG Revision 0 Revision Status: Active Final Resolution: NRC Approves Final Resolution Date: 13-Mar-97 Affected Technical Specifications 1.1 Definitions Change

Description:

Definition of Leakage 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

Defi nit ions 1.1 1.1 Definitions LEAKAGE 3. Reactor Coolant System (ReS) LEAKAGE (continued) through a steam generator (SG to th Secondary System; (exce + 'ReP S~Q I t....>e:r.ftv.

b. Unidentified LEAKAV lY1lt+/~1I'\ or JeakeH)

All LEAKAGE~hat is not identified LEAKAGE~

                           "'Msrrlll es tI!AlM8E;

c. Pressure Boundary LEAKAGE LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MODE A MODE shall correspond to anyone inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. NUCLEAR HEAT FLUX HOT FQ(Z) shall be the maximum local linear power CHANNEL FACTOR FQ(Z) density in the core divided by the core average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. NUCLEAR ENTHALPY RISE (FlH) shall be the ratio of the integral of HOT CHANNEL FACTOR (F~) linear power along the fuel rod on which minimum departure from nucleate boiling ratio occurs, to the average fuel rod power. OPERABLE--OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capab1eof performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). (continued) BWOG STS 1.1-5 Rev 1, 04/07/95

TsrF-4D Definitions 1.1 1.1 Definitions ENGINEERED SAFETY function (i.e., the valves travel to their FEATURE (ESF) RESPONSE required positions, pump discharge pressures reach TIME their required values, etc.). Times shall (continued) include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. The maximum allowable containment leakage rate, La' shall be [0.25]% of containment air weight per day at the calculated peak containment pressure (Pa)

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG to the Secondary System. /:

( Leta.!. '

                                                                + Rep S"eol JJak., ~'

b. Unidentified LEAKAGE 1o 1'I1.je.c.+- ", 0.,. leetl(o>,/f)

All LEAKAGE~hat is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

(continued) CEOG STS 1.1-4 Rev 1, 04/07/95

BWOG-5, Rev. 0 TSTF-42-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Revise the wording of SR 3.0.2 to be consistent with the other NUREGs NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 2) Consistency/Standardization Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Paul Infanger, (352) 563-4796, paul.infanger@pgnmail.com Delete the phrase "Required Action requires performance of a surveillance or its" so that SR 3.0.2 would read: "If a Completion Time requires periodic performance on a "once per..." basis, the above Frequency extension applies to each performance after the initial performance." The deleted phrase is unnecessary and potentially confusing. It is unnecessary since all "Once per..." basis requirements for Required Actions are in the Completion Time column. The deleted phrase is potentially confusing since it could be read to imply that the extension could only be applied if the periodic Required Action is a Surveillance. The proposed wording provides all the information that is necessary to correctly apply the extension. The proposed wording is consistent with NUREGs 1431, 1432, 1433, and 1434. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: ANO-1 Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 24-Aug-95 Owners Group Comments: (No Comments) Owners Group Resolution: Approved Date: 08-Sep-95 TSTF Review Information TSTF Received Date: 22-Sep-95 Date Distributed for Review: 03-Oct-95 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: N/A - WOG, CEOG, BWR/4, BWR/6 TSTF Resolution: Approved Date: 25-Nov-95 NRC Review Information NRC Received Date: 03-Jan-96 NRC Comments: Date of NRC Letter: 27-Sep-96 6/11/96 - C. Grimes comment: decision not made on TSTF-42. 9/18/96 - Approved. Final Resolution: NRC Approves Final Resolution Date: 18-Sep-96 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWOG-5, Rev. 0 TSTF-42-A, Rev. 0 OG Revision 0 Revision Status: Active Affected Technical Specifications SR 3.0.2 SR Applicability 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

T5TF--42 SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. Completion 1me requ1re perlO 1C per ormance on

                -;,'.-o-nc.....e-per . . ." bas is, the above Frequency extens i on applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications. SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance of the ' Surveillance. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be (continued) BWOG STS 3.0-4 Rev 1, 04/07/95

CEOG-1, Rev. 0 TSTF-47-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Eliminate "manipulation" from the definition of Core Alteration NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 2) Consistency/Standardization Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Patricia Furio, (410) 495-4374, patricia.s.furio@ccnppi.com The words, "or manipulation" were deleted from the definition of Core Alteration. When considering components in the Reactor Vessel such as Fuel , CEAs, or sources, it is improbable (by definition of manipulation) that fuel, CEAs or Sources could be manipulated without moving these components. Also this change would make the CE NUREG consistent with the other PWR NUREGs. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: Calvert Cliffs Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 14-Sep-95 Owners Group Comments: (No Comments) Owners Group Resolution: Approved Date: 02-Oct-95 TSTF Review Information TSTF Received Date: 27-Nov-95 Date Distributed for Review: 27-Nov-95 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: WOG - Not applicable to WOG TSTF Resolution: Approved Date: 21-Mar-96 NRC Review Information NRC Received Date: 25-Mar-96 NRC Comments: Date of NRC Letter: 27-Sep-96 4/22/96 R. Tjader recommended approval. Comment: Core Alts involve movement of fuel or components and manipulation is redundant. This change will make the consistent with the other PWRs. 6/11/96 - C. Grimes comment: TSTF-47 will be referred to a Tech Br. 9/18/96 - Approved. Final Resolution: NRC Approves Final Resolution Date: 18-Sep-96 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

CEOG-1, Rev. 0 TSTF-47-A, Rev. 0 Affected Technical Specifications 1.1 Definition of Core Alteration 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

Defi nit ions 1.1 T5TF-Lf7 1.1 Definitions (continued) CORE ALTERATION CORE ALTERATION shall be~t~e movement~

                           ~of               any fuel, sources, or reactivity
                           ~nents [excluding control element assemblies (CEAs) withdrawn into the upper guide structure], within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuriesjgram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [Table III of TID-14844, AEC, 1962, *Calculation of Distance Factors for Power and Test Reactor Sites,* or those listed in Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or ICRP 30, Supplement to Part 1, page 192-212, Table titled, *Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity"]. £-AVERAGE E shall be the average (weighted in proportion DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives> [15] minutes, making up at least 95% of the total noniodine activity in the coolant. ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval FEATURE (ESF) RESPONSE from when the monitored parameter exceeds its ESF TIME actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety (continued) CEOG STS 1~1-3 Rev 1, 04/07/95

CEOG-17, Rev. 0 TSTF-65-A, Rev. 1 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Use of generic titles for utility positions NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 3) Improve Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Patricia Furio, (410) 495-4374, patricia.s.furio@ccnppi.com On November 10, 1994, the NRC responded to a request by the BWR/6 owners to change the Section 2.0 "Safety Limits" and Section 5.0 "Administrative Controls" to allow the use of generic personnel titles as provided by ANSI/ANS 3.1 in lieu of plant-specific personnel titles [Letter from C. I. Grimes (NRC) to Lee Bush (WOG), Brian Mann (CEOG), Clinton Szabo (B&WOG) and Andrew Maron (BWROG), attached]. The NRC approved that request and provided example marked pages to the Rev. 0 Improved Technical Specifications. This change was not incorporated into Revision 1 of the ITS and is being proposed as a generic change for incorporation into the ITS. There is one difference between the NRC's proposed ITS changes and this change. The November 10, 1994 letter suggested using the bracketed phrase "a specified corporate executive position" to identify the utility corporate officer to be notified in case of a Safety Limit violation. This change proposes to use the phrase "the corporate officer with direct responsibility for the plant" (not bracketed) in lieu of inserting a plant-specific title. The attached letter from the NRC describes the justification for the change. Note that TSTF-05 deletes the portions of the Safety Limits affected by this change. Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: Calvert Cliffs Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 23-Jan-96 Owners Group Comments: (No Comments) Owners Group Resolution: Approved Date: 24-Jan-96 TSTF Review Information TSTF Received Date: 05-Mar-96 Date Distributed for Review: 07-Mar-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 16-Apr-96 NRC Review Information NRC Received Date: 12-Jun-96 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

CEOG-17, Rev. 0 TSTF-65-A, Rev. 1 OG Revision 0 Revision Status: Closed NRC Comments: Date of NRC Letter: 22-Sep-97 6/18/96 - recommend approval. Comment: Reviewers must be consistent in the use of brackets (Section 2.0). Why aren't brackets used in the Bases? 6/20/96 - pkg to C. Grimes to review. 9/18/96 - no change in status 3/18/97 - no change in status 4/17/97 - C. Grimes agreed to resolve 9/22/97 - Modify to be consistent in the use of brackets in the Bases, as well as in the Specifications. 10/2/97 - TSTF provided R.1. Final Resolution: Superceded by Revision Final Resolution Date: 02-Oct-97 TSTF Revision 1 Revision Status: Active Revision Proposed by: NRC Revision

Description:

Created Rev. 1 to address NRC comments regarding the use of brackets. TSTF Review Information TSTF Received Date: 01-Oct-97 Date Distributed for Review: 01-Oct-97 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 02-Oct-97 NRC Review Information NRC Received Date: 02-Oct-97 NRC Comments: 10/2/97 - Reviewer recommended approval and provided to W. Beckner for disposition. Final Resolution: NRC Approves Final Resolution Date: 02-Dec-97 Affected Technical Specifications 5.1.1 Administrative Controls - Responsibility 5.2.1 Administrative Controls - Onsite and Offsite Organizations 5.2.2.d Administrative Controls - Unit Staff 5.2.2.e Administrative Controls - Unit Staff 5.2.2.f Administrative Controls - Unit Staff 5.5.1 Administrative Controls - Programs and Manuals 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

CEOG-17, Rev. 0 TSTF-65-A, Rev. 1 5.7.1 Administrative Controls - High Radiation Area SL 2.2.6 Safety Limits NUREG(s)- 1430 Only SL 2.2.7 Safety Limits NUREG(s)- 1430 Only SL 2.2.7 Bases Safety Limits NUREG(s)- 1430 Only SL 2.2.4 Safety Limits NUREG(s)- 1431 Only SL 2.2.4 Bases Safety Limits NUREG(s)- 1431 Only SL 2.2.5 Safety Limits NUREG(s)- 1431 Only SL 2.2.5 Bases Safety Limits NUREG(s)- 1431 Only SL Vio 2.2.4 Safety Limits (Analog) NUREG(s)- 1432 Only SL Vio 2.2.4 Safety Limits (Digital) NUREG(s)- 1432 Only SL Vio 2.2.5 Safety Limits (Analog) NUREG(s)- 1432 Only SL Vio 2.2.5 Safety Limits (Digital) NUREG(s)- 1432 Only SL Vio 2.2.5 Bases RCS Pressure Safety Limits (Analog) NUREG(s)- 1432 Only SL Vio 2.2.5 Bases RCS Pressure Safety Limits (Digital) NUREG(s)- 1432 Only SL Vio 2.2.5 Bases Reactor Core Safety Limits (Analog) NUREG(s)- 1432 Only SL Vio 2.2.5 Bases Reactor Core Safety Limits (Digital) NUREG(s)- 1432 Only SL 2.2.3 Safety Limits NUREG(s)- 1433 1434 Only SL 2.2.3 Bases Safety Limits NUREG(s)- 1433 1434 Only SL 2.2.4 Safety Limits NUREG(s)- 1433 1434 Only SL 2.2.4 Bases Safety Limits NUREG(s)- 1433 1434 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

UNITED STATES NUCLEAR REGULATORY COMMISSION T5TF-6S" R~J WASHINGTON, D.C. 20555-0001 November 10, 1994 Mr. lee Bush Mr. Brian Mann Westinghouse Owners Group Combustion Engineering Owners Group c/o Commonwea1ln Edison Company c/o Baltimore Gas and Electric Company Zion Nuclear Power Station Calvert Cliffs Nuclear Power Plant 101 Shiloh Boulevard 1650 Calvert Cliffs Parkway Zion, Illinois 60099 lusby, Maryland 20657-4702 Mr. Clinton Szabo Mr. Andrew Maron Babcock &Wilcox Owners Group BWR Owners Group c/o Entergy Operations c/o Pennsylvania Power and light Arkansas Nuclear One Company Route 3, Box 137G Two North Ninth Street, A2-4 Russelvi11e, Arkansas 72801 Allentown, Pennsylvania 18101-1179 Gentlemen: The BWR/6 Owners recently proposed a change to Section 2.0 "Safety limits" and Section 5.0 "Administrative Controls" of the improved STS. The proposed change would allow the use of generic personnel tjt1es as provided by ANSI/ANS 3.1 in lieu of plant specific personnel titles. This change does not . eliminate any of the qualifications, responsibilities or requirements for ~j these positions, since the p1ant-spe~~!~~ personnel titles are currently identified in licensee controlled t h as the Fi t Anal sis Report (ESAR) or t e ~ ity Assurance (QA) Plan. In addition, the improved STS require that these positlons meet the~fications of Regulatory Guide (RG) 1.8 or an ANSI Standard acceptable to the NRC staff. The proposal provides a more direct link between the personnel qualifications as identified in the STS and STS-required responsibilities by utilizing the same ANSILANS St~dard position title. With the plant-specific personnel titles specified in the TS, a utility could utilize a person meeting the ANSI/ANS Standard qualifications to fulfill the TS qualification requirements while utilizing a separate person with the TS-identified title to perform the TS-required responsibilities. This is clearly not the intent of the TS requirements. The proposal will preclude this by utilizing the same generic position title for the responsibilities as contained in the qualifications requirements by reference to ANSI/ANS Standard or RG 1.8. The staff has reviewed the proposal and finds that lower case titles for all titles in the improved STS are acceptable. The titles selected by the 1icensee should agree with those titles in the ANSI Standard committed to in )~ improved STS Section 5.3. The relationship between the tjtles in the ANSI standard and the titles used b the license e de . 7 o e or an as appropr ate. Changes to titles in the FSAR/QA Plan shou1 be handled using normal FSAR/QA Plan change procedures (10 CFR

50. 54(a>>.

L. Bush, et a1. November 10, 1994 Titles have been identified in improved STS Section 2.0 and 5.~. Appropriately de-capitalized and bracketed titles for the pos;+ions in improved STS Section 2.0 and 5.0 are included as Enclosures 1 through 5. In addition, a Reviewer's Note should be placed at the beginning of improved STS Chapter 5.0 detailing the use of generic titles and title changes, and Specification 5.2.1.a should be modified as shown to provide the link between the generic titles in the improved STS and the FSAR/OA Plan. In the letter dated October 25, 1993, from W. T. Russell to the chairpersons of the NSSS Owners Groups, the NRC staff provided its review of the Administrative Controls Chapter (5.0) of the improved STS. One of the comments provided changed 5.2.2.f to require that the Operations Manager or the Assistant Operations Manager hold an *active* senior reactor operator (SRO) license. Subsequently, the Owners Groups proposed in BWOG-09, change C.5 that this requirement be changed to read *active or inactive.* During meetings with the BWR/6 Owners, the basis for the original staff position, the acceptability of the Owners Groups proposal, and the use of various adjectives such as active, inactive, current, and valid as they relate to NRC-licensed operators and senior operators were raised as issues. Current ANSI Standard commitments require the Operations Manager to *ho1d at time of appointment ***

or .Obt~i nd hold ..*
an SRO license. An individual licensed pursuant to 10 CFR Par 55 is considered to hold a license. This license can be valid, current, . e or inactive.

When used to describe Part 55 licenses, valid and current are equivalent terms and simply mean that the license holder (1) has passed the NRC initial license examination, (2) is participating in and current in the facility licensee's licensed operator requa1ification program, and (3) has renewed or will renew the license in accordance with 10 CFR 55.57. An active license is a license that is not only current and valid, but one for which the individual has been standing the required proficiency watches of 10 CFR 55.53(e). An active license is only necessary if the individual will assume licensed operator watchstanding duties. Individuals not maintaining an active license in accordance with the requirements of 10 CFR 55.53(e) are considered to have an inactive license. An inactive license can still be current and valid. In addition, it is necessary for the Operations Manager or Assistant Operations Manager to hold a senior operator license in order to effectively (1) interface with the day-to-day operational aspects of control room activities and (2) communicate operational issues to higher levels of plant and utility management. In order to fulfill the job requirements of the Operations Manager or Assistant Operations Manager positions, it is not necessary for these individuals to maintain the watchstanding proficiency necessary to be considered the shift supervisor or to be considered the operator at the controls. In conclusion, stating in Section 5.2.2.f that the Operations Manager or Assistant Operations Manager holds an SRO license is sufficient. It is unnecessary to add to Section 5.2.2.f. the word *active* as proposed in the

crSTF~6S Rc::V , L. Bush, et al. November 10, 1994 version sent to the Owners Groups or the Owners Groups language of -active or inactive.- Further, requiring this level of operations management to hold a license is consistent with the requirements of most current TS. In some cases, current TS may specify a -valid- or -current- license of the individual in the senior operations management position. Finally, the use of -active- to describe the license status of individuals designated to assume the control room command function is correct. Enclosures 1 through 5 include proposed changes the improved STS to reflect this position. We would appreciate your written response to the above staff positions or, if you prefer, we can meet with you to discuss this matter. Sincerely, c;~~ Christopher I. Grimes, Chief Technical Specifications Branch Division of Project Support Office of Nuclear Reactor Regulation

Enclosures:

As stated (5) cc w/enclosure: J. Eaton, NEI D. Hoffman, EXCEL

TSTF-65 f(..." I INSERT A [ Reviewer's Note: Titles for members of the unit staff shall be specified by use of an overall statement referencing an ANSI Standard acceptable to the NRC staff from which the titles were obtained, or an alternative title may be designated for this position. Generally, the first method is preferable; however, the second methods is adaptable to those unit staffs requiring special titles because of unique organizational structures. The ANSI Standard shall be the same ANSI Standard referenced in Section 5.3, Unit Staff Qualifications. If alternative titles are used, all requirements of these Technical Specifications apply to the position with the alternative title as apply with the specified title. Unit staff titles shall be specified in the Final Safety Analysis Report or Quality Assurance Plan. Unit staff titles shall be maintained and revised using those procedures approved for modifying/revising the Final Safety Analysis Report or Quality Assurance Plan.] INSERTB including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications.

SLs 2.0 2.0 SLs 2.2 SL Violations .(continued) 2.2.3 In MODE 1 or 2, if SL 2.1.2 is not met, restore compliance within limits and be in MODE 3 within 1 hour. 2.2.4 In MODES 3, 4, and 5, if SL 2.1.2 is not met, restore RCS pressure to ~ [2750] psig within 5 minutes. 2.2.5 Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.6 2.2.7 Within 3 days, a Licensee Event Report pursua to 10 ~,e LER an the lant~, and

           ~~)o

r~

2.2.8 Operation of the p}ant shall not be resumed until authorized by the NRC.

                  -fAt.-    corfJc)r~* oH1t-e-r                tJl'fh  ol/*~~c.f (e .y.'a"Si 'b, 'II'~ ..fn.. .f-hL /..--k_",_-I_,  -~

BWOG STS 2.0-2 Rev 1, 04/07/95

                                                         . Reactor Core SLs B 2.1.1 75 7F-6s-BASES                                                                 f?~v I SAFETY LIMIT 2.2.7   (continued)

VIOLATIONS the nuclear lant, and the(utilit~~ ar er 10n. 7' /' L - Ir* '/t, - J/'Ie~f re~~~Hn*',*t'.j,~

                              ~~r~~~~ ~(c~r WI~            d           /           7 2.2.8             k  +~ ~Ia;,f.         -------~

If SL 2.1.1.1, SL 2.1.1.2, or SL 2.1.1.3 is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation. REFERENCES 1. 10 CFR 50, Appendix A, GOC 10.

2. FSAR, Section [ ].

3. 10 CFR 50.72.

4. 10 CFR 50.73.

BWOG STS B 2.0-5 Rev 1, 04/07/95

Responsibility 5.1 T:; 7F-6S-5.0 ADMINISTRATIVE CONTROLS I2ev , 5.1 Responsibility The lant peJ1~t&ri~ or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. 5.1.2 The [Shift Supervisor (55)] shall be responsible for the control room command function. During any absence of the [55] from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the [55] from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. SWOG STS 5.0-1 Rev 1, 04/07/95

..----~~-'- - Organization 5.2 TS T~-6S 5.0 ADMINISTRATIVE CONTROLS f(ev ) 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of* authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions or in equivalent forms of documentation. These

requlrements shall be documented in the [FSA~~/1 f1~ rn~~,.. ~-----;--:~

                 -- --   b. Th~~ant           iRten ~t shall be responsible for overall sa~-operation 0     t e plant and shall have control over those onsite activities necessary for safe operation and

c. :~::::i:;e:h:o:::::~~f~.i~i."~ll have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2* Unit Staff The unit staff organization shall include the following:

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator (continued)

SWOG STS 5.0-2 Rev 1, 04/07/95

Organization 5.2 I TS7F-6S

!i f  5.2 Organization                                                           II~II /

i Ir 5.2.2 Unit Staff (continued) t shall be assigned for each control room from which a reactor I i r is operating in MODES 1, 2, 3, or 4. I C TWO unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units. . ] I b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c. Shift crew composition may be less than the minimum requirement of 10 CFR SO.54(m)(2)(i) and 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

                          ~ ~tTeChniCian~all be on site when fuel is
            \--d-.--A---....l i~actor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, health physicists, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed: (continued) SWOG STS 5.0-3 Rev 1, 04/07/95

Organization 5.2 TSTF-6S' 5.2 Organization jt.~ t 5.2.2 Unit Staff (continued)

1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

                                                   .                       mqn~~r Any deviation fr~he a~uidelines shall be aut in advance by the"'~ant ~~                         hlS designee, in accordance with approved a mlnlstratlve procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that

  ~"-----L-~dfi~~eshall be reviewed monthly by th6l[4!lant nt or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

OR [ The amount of overtime worked by unit staff members performing safety related functions shall be limited and controlled in accordance with the NRC Policy Statement on wor~g hours (Generic letter 82-12).

f. The.....~erat ions tManager or tAss i stant ,dperations tManager shall hold an SRO license.

f

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shi ft.

SWOG STS 5.0-4 Rev 1, 04/07/95

Programs and Manuals 5.5 T5TF-6G 5.0 ADMINISTRATIVE CONTROLS f(etl ) 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained. 5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip' setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.2] and Specification [5.6.3].

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. Tnls documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; ~

Shall ~come ef tive after the approval of the ~ant

                ~~i)lij!ijtij~r, and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the (continued) BWOG STS 5.0-7 Rev 1, 04/07/95

[High Radiation Area] [5.7] TSrF-6s-5.0 ADMINISTRATIVE CONTROLS !<f-.J , [5.7 High Radiation Area] 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

            > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP *. Ind* iduals ualified in radia~ protection proce ures e.g                    ~echniciansJ1 or personnel continuously es rted by suc ndividuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates 5 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c. An individual Qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic

             ~~diation surveillance at the frequency specified by the
             ~ ~adiat;on~rotectionJ(anage~n the RWP.

5.7.2 In addition to the requirements of Specification 5.7.1, areas with radiation levels ~ 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel (continued) BWOG STS 5.0-24 Rev 1, 04/07/95

SLs 2.0 T5TF-6S"

                                                                      ~tI ,

2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1. 2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained

          ~ [2735] psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. 2.2.3 Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.5 Within 30 days a Licensee Event Repor~tER) shall be prepared pursuant to 10 CFR 50.73. The LER sh be submitted to the NRC,

         ~te review func&-;g~and the               lant~, and V. e  ~~~                     t-t@l.:a.-     ~

2.2.6 Operation of the unit shall not be resumed until authorized by the NRC. WOG STS 2.0-1 Rev 1, 04/07/95

Reactor Core Sls B 2.1.1 r:r:7F,6~ BASES f<e.tJ I APPLICABILITY 5, and 6, Applicability is not required since the reactor is (continued) not generating significant THERMAL POWER. SAFETY LIMIT The following Sl violation responses are applicable to the VIOLATIONS reactor core SLs. 2.2.1 If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this Sl is not applicable. The allowed 'Completion Time of 1 hour recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage. If SL 2.1.1 is violated, the NRC Operations Center must be notified within 1 hour, in accordance with 10 CFR 50.72 (Ref. 5). 2.2.5 If SL 2.1.1 is violated, a licensee Event Report shall be prepared and submitted within 30 days to the NRC in accordance with 10 CFR 50.73 (Ref. 6. A co of the report ~ shall also be provi ded to the6tl ant rriRtQffde nd the l~~Pzes~. ca--jJOra-k cxf{;ct"r u,¥/'" Jirecf n:~d-'>j"(1,.ft f'l -flu ;o(a"..1 (continued) WOG STS B 2.0-4 Rev 1, 04/07/95

Respons'i bil i ty 5.1 T5TF-Gs-5.0 ADMINISTRATIVE CONTROLS Re~ I 5.1 Responsibility The lant ~~ or his designee shall approve, prior to implementation, eac proposed test, experiment or modification to systems or equipment that affect nuclear safety. 5.1.2 The [Shift Supervisor (55)] shall be responsible for the control room comand function. During any absence of the [55] from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (5RO) license shall be designated to assume the control room command function. During any absence of the [55] from the control room while the unit is in HODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. WOG STS 5.0-1 Rev 1, 04/07/95

Organization 5.2 Ts TF-6S/-'~-\ 5.0 ADMINISTRATIVE CONTROLS

                                                                         !?e" I 5.2 Organization 5.2.1      Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and cOlllllunication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functiona-l descriptions of departmental responsibilities and relationships, and job descriptions for key personnel ositions, or in equivalent forms of documentation. These req~,ements shall be documented in the [FSAR]; ~

                          ~V                                 ~

b. The ant~r shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessar or safe operation and maintenance of the plant;

c. ~~specified corporate shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall include the following:

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator (continued)

WOG STS 5.0-2 Rev 1. 04/07/95

Organization 5.2 T5TF-6~ 5.2 Organization ~evl 5.2.2 Unit Staff (continued) shall be assigned for each control room from which a reactor is operating in MODES 1, 2, 3, or 4. C TWO unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units. ]

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew

               ~~~to within the min~ requirements.

d. A~~chnician( shall be on site when fuel is in the reactor. The~osition may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, health physicists, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time; (continued)

WOG STS 5.0-3 Rev 1, 04/07/95

Organization 5.2

                                                                      ,$,1F-' r

('~ 5.2 Organization ~'" J I ) 5.2.2 Unit Staff (continued)

2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for~h entire staff on a shift."JV)

I' lan~r OR [ The amount of overtime worked by unit staff members performing safety related functions shall be limited and controlled in accordance with the NRC Policy Statement 0rn wor~g hours (Generic letter 82-12).

f. The~erationsM(anager or ~sistant 'pperations~nager shall hold an SRO license.

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the Qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

WOG STS 5.0-4 Rev 1, 04/07/95

Programs and Manuals 5.5 TSTF-K 5.0 ADMINISTRATIVE CONTROLS ~II , 5.5 Programs and Manuals "rhe following programs shall be established, implemented, and maintained. 5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.2] and Specification [5.6.3].

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s} maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; r/

b. Shall become effective after the approval of the-~ant

~4iiJ- ~ perJn ; and

c. Shall be submi ted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the OOCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the (continued) WOG STS 5.0-7 Rev 1, 04/07/95

[High Radiation Areal [5.7] ISTF-6S 5.0 ADMINISTRATIVE CONTROLS J&..v I [5.7 High Radiation Area] 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

           > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controll d by requiring issuance of ~adiation Work Permit RWP. I iv'duals uali ied in radia n protection proce ures (e.g ,               , ~echnicians or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates ~ 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the

          .foll owing:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c, An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the ctivities within the area and shall perform periodic dj radiation surveillance at the frequency specified by the adiation~otection~anager~n the RWP. 5.7.2 In addition to the requirements of Specification 5.7.1, areas with radiation levels ~ 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in (continued) WaG STS 5.0-24 Rev 1, 04/07/95

SLs (Analog) 2.0 flSTP... 6s 2.0 SAFETY LIMITS (SLs) (Analog) ~f,tI I

2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, pressurizer pressure, and the highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figure 2.1.1-1.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1; 2, 3, 4, and 5, the RCS pressure shall be maintained

     -     S [2750] psia.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. 2.2.3 Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. ~~

                                                /Y'(::l "C? "..

2.2.4 Within 24 hours, notif~ the itJant ~ and~ C!~.'.!-N~.~~'li~~~ "---" 2.2.5 Within days, a Licensee Event Report (LER) shall be prepared pursua t 10 FR 50.73. The LE~all be submitted to the NRC and the ant ~ ande~~el'l!__~

                            ~~?J                                      ------

2.2.6 the unit shall not be resumed until authorized by the

Reactor Core SLs (Analog) B 2.1.1 c7_ST,F-65"' BASES f(~ J SAFm UMIT 2.2.1 VIOLATIONS (continued) If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE 1n which this SL is not applicable. The allowed Completion Time of 1 hour recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable and reduces the probability of fuel damage. 2.2.3 If SL 2.1.1 is violated, the NRC Operations Center must be notified within 1 hour, in accordance with 10 CFR 50.72 (Ref. 3). If SL 2.1.1 is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours. This 24 hour period provides time for the plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to senior management. . 2.2.5 2.2.6 If SL 2.1.1 is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and cOTfJor~1c o:Jfice, ?,),' J/~c-f ~~)(' ,~~

   ~..f~         /efii"f (continued)

CEOG STS B 2.0-4 Rev 1, 04/07/95

RCS Pressure SL (Analog) B 2.1.2 LTs("F-6~ BASES (continued) /(ttl , SAFETY LIMIT 2.2.2.2 (continued) VIOLATIONS compound the problem by adding thermal gradient stresses to the existing pressure stress. 2.2.3 If the RCS pressure SL is Violated, the NRC Operations Center must be notified within 1 hour, in accordance with 10 CFR 50.72 (Ref. 7). 2.2.4 If the RCS pressure SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours. This 24 hour period provides time for the plant operators and staff to take the appropriate immediate action and to assess the condition of the unit before reporting to senior management. 2.2.5 If the RCS pressure SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC in accordance with 10 CFR 50.73 (Ref. 8). A copy of the report shall also be provided~~

               ~nt" and the                 i -VAlc re' ... ~e 2.2.6 If the RCS pressure SL is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.

(continued) CEOG STS B 2.0-10 Rev 1, 04/07/95

SLs (Digital) 2.0 TSTF-bS-2.0 SAFETY LIMITS (SLs) (Digital) - '. (I.u I 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at ~ [1.19]. 2.1.1.2 In MODES 1 and 2, the peak linear heat rate (LHR) (adjusted for fuel rod dynamics) shall be maintained at

                   ~ [21.0] kW/ft.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at ~ [2750] psia. 2.2 SL Violations 2.2.1 If Sl 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. 2.2.3 Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.4 With~n ~: ~f;;~~!b~ p-,-::~~-----~~ ~ n ~ and~ 2.2.5 Within 30 days of the violati Event Report (LER) shall be prepared pursuant The lER ~~e submitted to the NRC d th an~ (continued) CEOG STS 2.0-1 Rev 1, 04/07/95

JAN-17-96 WED 11:52 EXCEL FAX NO, 3019847600 P. 06 Reactor Core SLs (Digital) 8 2.1.1 (TS TF-6~ BASES ~. ,R~"J SAFnY LIMIT 2.2.5 (cont;nued) VIOLATIONS

                   . report shall also be prov1~e~O~JMem~t
                    ~r plant, and the~~!e~~-ar)
                    ~                                                              -

If SL 2.1.1.1 or SL 2.1.1.2 is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that al' necessary reviews, analyses, and actions are completed before the unit begins

                      ;ts restart to normal operation.

REFERENCES 1. 10 eFR 50, Appendix At GDC 10, 1988.

2. FSAR, Section [ ].

3. 10 CFR 50.72.

4. 10 CFR 50.73.

corfXY~ It: cJh'U" cAt. cJ ir< ~.f reYJa-. P '6,'/,'} -he ~ r1ay,f CEOG STS B 2.0-5 Rev 1, 04/07/95

JAN-17-96 WED 11:55 EXCEL FAX NO, 3019841600 P. 10 Res Pressure SL (Digital) B 2.1.2 T.5TF- 6s:- BASES Re.J I SAFETY LIMIT 2.2.2.2 (continued) VIOLATIONS compound the problem by adding thermal gradient stresses to the existing pressure stress. If the ReS pressure SL is violated, the NRC Operations Center must be notified within 1 hour. in accordance with 10 CFR 50.72 (Ref. 7). If the ReS pressure Sl is violated, the appropriate senior management of the nuclear plant and the utility shall be notif1ed within 24 hours. This 24 hour period provides time for the plant operators and staff to take the appropriate immediate action and to assess the condition of the unit before reporting to the senior management. If the ReS pressure SL is violated, a licensee Event Report shall be prepared and submitted within 30 days to the NRC in accordance with 10 CFR 50.73 (Ref. S). A copy of the report shall also be prOVide~agement of the

                   ~ant.; and th u                'VV1~~'

Qf>~a.t',~ . 2.2.6 If the ReS pressure SL is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation. (continued) CEOG STS B 2.0-9 Rev 1, 04/07/95

Respons i bil ity 5.1 TSTF-6~ 5.0 ADMINISTRATIVE CONTROLS ~'1 5.1

          .The lan                    or his designee shall approve, prior to imple en ation, eac proposed test, experiment or modification to systems or equipment that affect nuclear safety.

5.1.2 The [Shift Supervisor (55)] shall be responsible for the control room command function. During any absence of the [SS] from the

     -     control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function.

During any absence of the [SS] from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. A CEOG STS 5.0-1 Rev 1, 04/07/95

Organization 5.2

                                                                 ~TsrF-6S 5.0 ADMINISTRATIVE CONTROLS                                           P.w ,

5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions or in equivalent forms of documentation. These req lr ent. s~_a.l1. b~,documented in the [FSA~~p.3J

b. The la shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and m~~ance of the plant; ~;~

c. ~ t./'specified corporate~~.:fp'gEitigR?;:l1 have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall include the following:

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator (continued)

CEOG STS 5.0-2 Rev 1, 04/07/95

Organization 5.2 T5(T-6~ 5.2 Organization /(e" J

  . 5.2.2      Unit Staff (continued) shall be assigned for each control room from which a reactor is operating in MODES 1, 2, 3, or 4.

D TWO unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units. ]

b. At least one 1icensed 'Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least

          -          one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c. Shift crew composition may be less than the minimum requirement of 10 CFR SO.S4(m)(2)(i) and S.2.2.a and S.2.2.g for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew co position to within the minim requirements.

~J A c n1 1 shall be on site when fuel is in the reactor. The pos; ion may be vacant for not more than 2 "hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, health physicists, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time; (continued)

CEOG STS 5.0-3 Rev 1, 04/07/95

Organization 5.2

                                                                        *T5TF-6s-5.2 Organization                                                               R~" J

. 5.2.2 Unit Staff (continued)

2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis

       -               and not for th entire staff on a shi t.

Any deviation fro ('llano. hall be authorized r in advance by the "or 1s eS1gnee, 1n accordance with ap oved a mlnlstrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. . ~ Controls shall~included in the procedures such th .,., individual ove me shall be reviewed monthly by the lant .~~~;; or his designee to ensure that excessive ours.have not been assigned. Routine deviation from the above guidelines is not authorized. [ The amount of overtime worked by unit staff members performing safety related functions shall be limited and cont~olled in accordance with the NRC Policy Statement on work ng hours (Generic Letter 82-12).

f. The perat ion..$. j(anager or 'ss i stant jPerati ons ~nager sha1 ra anSRcrrrcense. =--__ --=~ __---...::....---

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (55) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

CEOG STS 5.0-4 Rev 1, 04/07/95

Programs and Manuals 5.5 cTSTF-6~ 5.0 ADMINISTRATIVE CONTROLS te" , 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained. 5.5.1 Offsite Dose Calculation Manual eOOCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offs1te doses resulting from radioactive gaseous and liquid effluents, in the calculation*

of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

     '-    b. The aDCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.2.] and Specification [5.6.3].

Licensee initiated changes to the ODeM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),

2. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, ~int calculations;

~ b~~hall becom effective after the approval of the ant t~,;and

c. Shall be submi ted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the aDCM was made.

Each change shall be identified by markings in the margin of (continued) CEOG STS 5.0-7 Rev 1, 04/07/95

[High Radiation Area] [5.7] (~TSTP-6~ 5.0 ADMINISTRATIVE CONTROLS Rell I [5.7 High Radiation Area] 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as

          ,defined in 10 CFR 20, in which the intensity of radiation is
           > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of ~'ation Work Permit RWP         'viduals       . ied in radia'     protection proce ures {e. ,                 . techn i ci ans or personnel continuously escorte y such individuals may be exempt from the RWP issuance requirement during the performance of their
     -     assigned duties in high radiation areas with exposure rates S 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic adia . n rv' e at requency specified by the alation rotection anager in the RWP.

5.7.2 In addition to the requirements of Specification 5.7.1, areas with radiation levels ~ 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel (continued) CEOG STS 5.0-25 Rev 1, 04/07/95

Sls 2.0 T~1F-6S 2.0 SAFETY LIMITS (SLs) Ael/ } 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: THERMAL POWER shall be ~ 25% RTP. 2.1.1.2 With the reactor steam dome pressure ~ 785 psig and core flow ~ 10% rated core flow: MCPR shall be ~ [1.07] for two recirculation loop operation or ~ [1.08] for single recirculation loop operation. 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be ~ 1325 psig. 2.2 SL Violations With any SL violation, the following actions shall be completed: 2.2.1 Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.2 Within 2 hours: 2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 (continued) BWR/4 STS 2.0-1 Rev 1, 04/07/95

SLs 2.0 f!;Ti~-6S-2.0 SLs ~tJ 1 /-) 2.2 SL Violations (continued) 2.2.5 Operation of the unit shall not be resumed until authorized by the I NRC. . I 1

            -I-Juz. COff()'f&:<k officer WI:!£' d,'recf ~~hS;*t.,i~

I I ~ fot- .f-~ ;O/~;,-I: f t t II BWR/4 STS 2.0-2 Rev 1, 04/07/95

Reactor Core SLs B 2.1.1 T5TF-6S-BASES R.t." I SAFETY LIMIT VIOLATIONS (continued) If any SL is violated, pl ant and the s e ~u shall be noti le within 24 ours. e 4 our perlO provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the appropriate utility management. If any SL is violated, restart of the unit shall not commence until authorized by t~e NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation. REFERENCES 1. 10 CFR 50, Appendix A, .GOC 10.

2. NEOE-24011-P-A (latest approved revision).

3. XN-NF524(A), Revision 1, November 1983.

4. 10 CFR 50.72.

5. 10 CFR 100.

6. 10 CFR 50.73.

BWR/4 STS B 2.0-7 Rev 1, 04/07/95

Responsib'il ity 5.1

                                                                   ~TF-65" 5.0 ADMINISTRATIVE CONTROLS                                           f<erl I 5.1.2      The [Shift Supervisor (55)] shall be responsible for the control room command function. During any absence of the [55] from the control room while the unit is in HODE 1, 2, or 3, an individual with an active Senior Reactor Operator (SRO) license shall be .

designated to assume the control room command function. During any absence of the [55] from the control room while the unit is in MODE 4 or 5, an individual with an active 5RO license or Reactor Operator license shall be designated to assume the control room command function. BWR/4 STS 5.0-1 Rev 1, 04/07/95

Organization 5.2 Ts 7F-6s-5.0 ADMINISTRATIVE CONTROLS hvl 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional T

ii1 5erl- [3

          ~
             ~

descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These r_e_q~i~r e_n~ts~shall be docum nted in the [FSA~G?~ The lant ~~ shall be responsible for overall safe operation 0 the p ant and shall have control over those onsite activities necessary for safe operation and mai~ance of the plant; ~

c. ~~pecified corporat~89;'i8~~11 have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall include the following:

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator (continued)

BWR/4 STS 5.0-2 Rev 1, 04/07/95

Organization 5.2 TSTF-6s-5.2 Organization ~II , 5.2.2 Unit Staff (continued) shall be assigned for each control room from which a reactor is operating in MODES 1, 2, or 3. r r r.

,f                   D TWO unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units.                                                   ]

I r

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In f

addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. r c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.S4{m){2)(i) and S.2.2.a and S.2.2.g I for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew t co ition to within the min~requirements.

d. A aiechnician(shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, health physicists, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time; (continued)

BWR/4 STS 5.0-3 Rev 1, 04/07/95

Organization 5.2 TS TF-6S-5.2 Organization ~" 1 5.2.2 Unit Staff (continued)

2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any r

7 day period, all excluding shift turnover time;

3. A break of at least 8 hours should be allowed between work periods, including shift turnover time; I

f 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis r and not for ~he ntire staff on a shift. __---</y)c?h~r-t r Any deviation fro the a~{~l be au or zed in advance by the lant~or his designee, in f accordance with approved adminlstrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. ~ Controls shall~'ncluded in the procedures such th individ al

                       ,. ad e me shall be reviewed monthly by the or his designee to ensure that excessive ant ours ave not been assigned. Routine deviation from the above guidelines is not authorized.

OR [ The amount of overtime worked by unit staff members performing safety related functions shall be limited and wor . g hours (Generic Letter 82-12).

f. The erat ions "'nager or,iss i stant ~erations pranager shall hold an SRO license.

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

BWR/4 STS 5.0-4 Rev 1, 04/07/95

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented and maintained. 5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCH shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCH shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification [5.6.2] and Specification [5.6.3].

Licensee initiated changes to the ODCH:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effect* e after review and nsite review accePtan~

tion] and the approval of the~~~nt the __~r:.:-:p=}r1t , and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire DOCH as a part of or concurrent with the Radioactive Effluent Release RepQrt for the period of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page (continued) BWR/4 STS 5.0-7 Rev 1, 04/07/95

[High Radiation Area] [5.7] (sr~-6S 5.0 ADMINISTRATIVE CONTROLS

                                                                          ~I)l

[5.7 High Radiation Area] 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

             > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of~diation-work ermit RWP. I *viduals           i ie~ in radi on protection proce ures e.g.                 . Bechnicians) or personnel continuously escorte y suc ndividuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates S 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into s~ch areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is esponsible for providing positive control over the 5.7.2 i activities within the area anndall perform periodic radiation surveillance at the equency specified by the adiation;rrotectionJ(anager in the RWP.

In addition to the requirements of Specification 5.7.1, areas with radiation levels ~ 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work (continued) BWR/4 STS 5.0-23 Rev 1, 04/07/95

SLs 2.0 nTF-6~ 2.0 SAFETY LIMITS (Sls) ~I 2.1 SLs

2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: THERMAL POWER shall be ~ 25% RTP. 2.1.1.2 With the reactor steam dome pressure ~ 785 psig and core flow ~ 10% rated core flow: MCPR shall be ~ [1.07] for two recirculation loop operation or ~ [1.08] for single recirculation loop operation. 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be ~ 1325 psig. 2.2 SL Violations With any SL violation, the following actions shall be completed: 2.2.1 Within 1 hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.2 Within 2 hours: 2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 2.2.3 (continued) BWR/6 STS 2.0-1 Rev 1, 04/07/95 .~ r

2.0 Sls 2.2 Sl Violations (continued) 2.2.4 {BJ Operation of the unit shall not be resumed until authorized by the NRC. I BWR/6 STS 2.0-2 Rev 1, 04/07/95

Reactor Core SLs B 2.1.1 1 rsTF-6~ [ BASES /<..oJ , I f SAFETY LIMIT VIOLATIONS 2.2.3 (continued) If any SL is violated he pl ant and the u shall be notified within 24 hours. The 24 our period provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the appropriate utility management. i If any SL is violated, restart of the unit shall not J commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation. REFERENCES 1. 10 CFR SO, Appendix A, GOC 10.

2. NEOE-24011-P-A, (latest approved revision).

3. XN-NF524(A) , Revision 1, November 1983.

4. 10 CFR 50.72.

5. 10 CFR 100.

6. 10 CFR 50.73.

              /       CorPo'rc,ft oJj;(,,..~     LJ(:fJ, u"rec+ T¥>c)r.~,.~1,~    .;;.,. 1J..e..
             ~ f'l ,,+_,_

q BWR/6 STS B 2.0-7 Rev 1, 04/07/95

Responsibil ity 5.1 737F-t>S-5.0 ADMINISTRATIVE CONTROLS ~1 r I 5.1 I The ant r nd or his designee shall approve, prior to I implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. 5.1.2 The [Shift Supervisor (55)] shall be responsible for the control room command function. During any absence of the [55] from the control room while the unit is in MODE 1, 2, or 3, an individual with an active 5enior Reactor Operator (5RO) license shall be designated to assume the control room command function During any absence of the [55] from the control room while the unit is in MODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function. BWR/6 STS 5.0-1 Rev 1, 04/07/95

Organization 5.2 TSTF-6s-5.0 ADMINISTRATIVE CONTROLS Rel/ J 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affect'jng safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be defined and est~blished throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions' of departmental responsibilities and relationships, and job descriptions for key personnel posit*ons, or in equivalent f rms of documentation. These re u n shall be doc ted in the [FSA~];~4~~~

b. The f9ant S'p~i~tehde t shall be responsible for overall safe operation e plant and shall have control over those onsite activities necessary for safe operation and main!ence of the plant; ~~

                                               "r~CW'"         ~

c. ~ specified corporatesitilAf shall have corp rate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and'

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall include the following:

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator (continued)

BWR/6 STS 5.0-2 Rev 1, 04/07/95

Organization 5.2 7S7F-6s:- 5.2 Organization &\1' 5.2.2 Unit Staff (continued) shall be assigned for each control room from which a reactor is operating in MODES 1, 2, or 3. D TWO unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units. ]

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.S4(m)(2)(i) and S.2.2.a and S.2.2.g for a period of time not be exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew l __---------c;o~m4 sition to within the min~ requirements.

rC7d,~t;'tJ~ Yl/ T~e- ""'I'~II'\

d. A ~chnician(shall be on site when_fuel is f in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SROs, licensed ROs, health physicists, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour week, while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed: (continued) BWR/6 STS 5.0-3 Rev 1, 04/07/95

Organization 5.2 T5TF-6!J 5.2 Organization ~v J 5.2.2 Unit Staff (continued)

1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift. /YJan.q~

Any deviation fro~he above uidelines 1 be authorized in advance by the~an n or his designee, in accordance with approved admlnls rative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. ~ Controls shall*enCluded in the procedures such tha rJ~ individual ove e shall be reviewed monthly by the ~ant

               ~ or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

OR [ The amount of overtime worked by unit staff members performing safety related functions shall be limited and wor i g hours (Generic Letter 82-12).

f. The erations ~nager or JASsistant~erations ~nager shall hold an SRO license.

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

BWR/6 STS 5.0-4 Rev 1, 04/07/95

Programs and Manuals 5.5 \ TSTP-65 5.0 ADMINISTRATIVE CONTROLS /<.11 f 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained. 5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.2] and Specification [5.6.3].

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; ~

b. Shall becom.e effective after review and acceptan by the onsite review function] and the approval of the lant

                  ,--.;:~r~'J;..)n tUft! ~and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of, or concurrent with, the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

(continued) BWR/6 STS 5.0-7 Rev 1, 04/07/95

[High Radiation Area] [5.7]

                                                                     /.5rF-6s-5.0 ADMINISTRATIVE CONTROLS
                                                                        ~11  }

[5.7 High Radiation Area] 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

           > 100 mremjhr but < 1000 mrem/hr, shall be barricaded and conspicuously post d as a high radiation area~and entrance thereto shall be contro e by requiring issuance of          aiation Work Permit RWP. n iv' 1              . ied in radia on protection proce ures (e.g., .             . ~chnicians or personnel continuously escorted y such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates ~ 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is recei ved. Entry into such .areas with this monitoring device may be made after the dose rate .

levels in the area have been established and personnel are aware of them.

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is esponSible for providing positive control over the 5.7.2 i radiation surveillance at the~~Uency activities within the area an all perform periodic adiation~otectionpinagerl'i~1 the RWP.

specified by the In addition to the requirements of Specification 5.7.1, areas with radiation levels ~ 1000 mremjhr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel (continued) BWR/6 STS 5.0-24 Rev 1, 04/07/95

WOG-36, Rev. 0 TSTF-104-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Relocates discussion of exceptions from LCO 3.0.4 to the Bases NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 1) Correct Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Steve Wideman, (620) 364-4037, stwidem@wcnoc.com This change removes the additional discussion provided in LCO 3.0.4 with respect to the use of exceptions and provides the necessary discussion in the Bases. This change provides consistency with LCO 3.0.3 by moving the discussion of exceptions from the LCO to the Bases. In addition, this change reduces the potential for confusion by revising the discussion to eliminate the repeated use of the phrase "Modes or other specified conditions in the Applicability" to increase clarity. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: Ginna Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 30-Oct-95 Owners Group Comments: Ginna 3 Owners Group Resolution: Approved Date: 09-Nov-96 TSTF Review Information TSTF Received Date: 27-Nov-95 Date Distributed for Review: 27-Nov-95 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: CEOG accepts BWOG accepts BWROG accepts TSTF Resolution: Approved Date: 30-Apr-96 NRC Review Information NRC Received Date: 17-Jul-96 NRC Comments: 9/18/96 - Review pending. 3/14/97 - NRC approves. Final Resolution: NRC Approves Final Resolution Date: 14-Mar-97 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-36, Rev. 0 TSTF-104-A, Rev. 0 OG Revision 0 Revision Status: Active Affected Technical Specifications LCO 3.0.4 LCO Applicability LCO 3.0.4 Bases LCO Applicability 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

LCO Applicability 3.0 T~T~-IO'l 3.0 LCO APPLICABILITY LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. SR 3.0.4 is only applicable for entry into a Mode or other specified condition in the Applicability in Modes 1, 2, 3 and 4. Reviewer's Note: LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, 3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. Before this version of LCO 3.0.4 can be implemented on a plant-specific basis, the licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be included in individual LCOs to justify this change; such an evaluation should be summarized in a matrix of all existing LCOs to facilitate NRC staff review of a conversion to the STS. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. (continued) SWOG STS 3.0-2 Rev 1, 04/07/95

LCO Applicability B 3.0 TS TF-IOr-t BASES LCO 3.0.4 proVlslons of LCO 3.0.4 shall not prevent changes in MODES (continued) or other specified conditions in the Applicability that result from any unit shutdown. Exceptions to LCO 3.0.4 are stated in the individual S ecifications. Exceptions may apply to all the ACTIONS or to a specl lC Required Action of a Specification. LCO 3.0.4 is only applicable when entering MODE 4 from MODE 5, MODE 3 from MODE 4, MODE 2 from MODE 3, or Mode 1 from Mode 2. Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODES 1, 2, 3, or 4. The requirements of LCO 3.0.4 do not apply in MODES 5 and 6, or in other specified conditions of the Applicability (unless in MODES 1, 2, 3, or 4) because the ACTIONS or individual specifications sufficiently define the remedial measures to be taken. [In some cases (e.g., ** ) these ACTIONS provide a Note that states "While this LCO is not met, entry into a MODE or other specified condition in the Applicability is not permitted, unless required to comply with ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.] Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LCD 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been* removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with (continued) BWOG STS B 3.0-6 Rev 1, 04/07/95

LCO Applicability 3.0 TSTF-/o'/ ~~, 3.0 LCO APPLICABILITY LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO 3.0.4 is only applicable for entry into a MODE or others specified condition in the Applicability in MODES 1, 2, 3, and 4. Reviewers's Note: LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, 3, and 4: The MODE change restrictions in LCO 3.0.4 were previously appl icabl e in all MODES. , Before thi s version of LCO 3.0.4 can be implemented ona plant-specific basis, the licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be included in individual LCOs to justify this change; such an evaluation should be summarized in a matrix of all existing LCOs to facilitate NRC staff review of a conversion to the STS. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. (continued) WaG STS 3.0-2 Rev 1, 04/07/95

LCO Applicability B 3.0 7STF-IOi BASES LCO 3.0.4 that are required to comply with ACTIONS. In addition, the (continued) provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. Exceptions to LCO 3.0.4 are stated in the individual Specifications w Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification. LCO 3.0.4 is only applicable when entering MODE 4 from MODE 5, MODE 3 from MODE 4, MODE 2 from MODE 3, or MODE 1 from MODE 2. Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODES 1, 2, 3, or 4. The requirements of LCO 3.0.4 do not apply in MODES 5 and 6, or in other specified conditions of the Applicability (unless in MODES 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. [In some cases (e.g., *. ) these ACTIONS provide a Note that states "While this LCO is not met, entry into a MODE or other specified condition in the Applicability is not permitted, unless required to comply with ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.] Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to (continued) WOG STS B 3.0-6 Rev 1, 04/07/95

LCO Applicability 3.0 (~rF-lo'l 3.0 LCO APPLICABILITY LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, 3, and 4. Reviewers's Note: LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, 3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. Before this version of LCO 3.0.4 can be implemented on a plant-specific basis, the licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be included in individual LCOs to justify this change; such an evaluation should be summarized in a matrix of all existing LCOs to facilitate NRC staff review of a conversion to the STS. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. (continued) CEOG STS 3.0-2 Rev 1, 04/07/95

LCO Applicability B 3.0 7STt:-I01 BASES LCO 3.0.4 The provisions of LCO 3.0.4 shall not prevent changes in (continued) MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. Exceptions to LCO 3.0.4 are stated in the individual

'(-!:..~~~L-_ _isj5P..:..ecrisf;pi~Cattr1i10cnlsRE. Exceptions may apply to all the ACTIONS or
~                  0   a speCl lC Required Action of a Specification.

LCO 3.0.4 is only applicable when entering MODE 4 from MODE 5, MODE 3 from MODE 4, MODE 2 from MODE 3, or MODE 1 from MODE 2. Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODES 1, 2, 3, or 4. The requirements of LCO 3.0.4 do not apply in MODES 5 and 6, or in other specified conditions of the Appl icabil ity (unless in MODES 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. [In some cases (e.g., *. ) these ACTIONS provide a Note that states "While this LCO is not met, entry into a MODE or other specified condition in the Applicability is not permitted, unless required to comply with ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.] Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with (continued) CEOG STS B 3.0-6 Rev 1, 04/07/95

LCO App1i cab-j 1i ty 3.0 T5TF-IOi 3.0 LCD APPLICABILITY LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3. Reviewers's Note: LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and

3. The MODE change restrictions in lCO 3.0.4 were previously applicable in all MODES. Before this version of LCO 3.0.4 can be implemented on a plant-specific basis, the licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be included in individual LCOs to justify this change; such an evaluation should be summarized in a matrix of all existing LCOs to facilitate NRC staff review of a conversion to the STS.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. (continued) BWR/6 STS 3.0-2 Rev 1, 04/07/95

LCO Applicability B 3.0 T5TF--/o,/ BASES LCO 3.0.4 interpreted as endorsing the failure to exercise the good (continued) practice of restoring systems or components to OPERABLE . status before entering an associated MODE or other specified condition in the Applicability. The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. Exceptions to LCO 3.0.4 are stated in the individual Specifications. Exceptions may apply to all the ACTIONS or o a speci lC Required Action of a Specification. LCO 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2. Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3. The requirements of LCO 3.0.4 do not apply in MODES 4 and 5, or in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. [In some cases (e.g., .. ) these ACTIONS provide a Note that states "Wh"ile this LCO is not met, entry into a MODE or other specified condition in the Applicability is not permitted, unless required to comply with ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.] Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4, or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. (continued) BWR/6 STS B 3.0-6 Rev 1, 04/07/95

LCO Applicability B 3.0 Tsr;:--/o,-/ BASES LCO 3.0.4 that are required to comply with ACTIONS. In addition, the (continued) provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. Exceptions to LCD 3.0.4 are stated in the individual

 ~~~~         ~S~e~c~if~i~c~a~tl~*o~n~s~. Exceptions may apply to all the ACTIONS or o a speciflc Required Action of a Specification.

LCO 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2. Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3. The requirements of LCO 3.0.4 do not apply in MODES 4 and 5, or in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual specifications sufficiently define the remedial measures to be taken. [In some cases (e.g., .. ) these ACTIONS provide a Note that states "While this LCO is not met, entry into a MODE or other specified condition in the Applicability is not permitted, unless required to comply with ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.] Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with LCO 3.0.5 the applicable Required Action(s>> to allow the performance (continued) BWR/4 STS B 3.0-6 Rev 1, 04/07/95

LCO Applicability 3.0 Ts TF-IOLj .~. 3.0 LCO APPLICABILITY LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCD 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3. Reviewer's Note: LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCD 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. Before this version of LCD 3.0.4 can be implemented on a plant-specific basis, the licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be included in individual LCOs to justify this change; such an evaluation should be summarized in a matrix of all existing LCOs to facilitate NRC staff review of a conversion to the STS. LCD 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. (continued) BWR/4 STS 3.0-2 Rev 1, 04/07/95

TSTF-I04 Insert The exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time.

WOG-41, Rev. 0 TSTF-106-A, Rev. 1 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Change to Diesel Fuel Oil Testing Program NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 3) Improve Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Steve Wideman, (620) 364-4037, stwidem@wcnoc.com The proposed change will make it clear that the TS requirement is only applicable to the new fuel, must be done within 31 days following addition to the storage tanks, and is only required to be done one time. As worded, paragraph b of the Diesel Fuel Oil Testing Program can be, and has been, misinterpreted. Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: Ginna Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 30-Oct-95 Owners Group Comments: Ginna 16 Owners Group Resolution: Approved Date: 09-Nov-95 TSTF Review Information TSTF Received Date: 29-Nov-95 Date Distributed for Review: 30-Nov-95 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: CEOG Accepts BWOG Accepts BWROG Accepts TSTF Resolution: Approved Date: 30-Apr-96 NRC Review Information NRC Received Date: 17-Jul-96 NRC Comments: Date of NRC Letter: 27-Sep-96 9/18/96 - Approved. NOTE: BWOG frequency was not changed to 31 days. TSTF to correct on final change. 10/30/96 - NRC requests revision 1 to change BWOG frequency from 30 to 31 days. TSTF to provide. Final Resolution: Superceded by Revision Final Resolution Date: 23-Jan-97 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-41, Rev. 0 TSTF-106-A, Rev. 1 TSTF Revision 1 Revision Status: Active Revision Proposed by: NRC Revision

Description:

Changed BWOG frequency from 30 to 31 days. TSTF Review Information TSTF Received Date: 30-Oct-96 Date Distributed for Review: 20-Nov-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 30-Oct-96 NRC Review Information NRC Received Date: 23-Jan-97 Final Resolution: NRC Approves Final Resolution Date: 23-Jan-97 Affected Technical Specifications 5.5.13 Diesel Fuel Oil Testing Program NUREG(s)- 1430 1431 1432 Only 5.5.10 Diesel Fuel Oil Testing Program NUREG(s)- 1433 1434 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-I06 Insert verify that the properties of the new fuel oil, other than those addressed in a., above, . are within limits for ASTM 2D fuel oil.

Programs and Manuals 5.5 T~TF-I06 5.5 Programs and Manuals (continued) 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. 'rhe purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific graVity within 1 imits,

2. A flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and 5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the foll owing:

A change in the TS incorporated in the license; or A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59. (continued) CEOG STS 5.0-16 Rev 1, 04/07/95

Programs and Manuals 5.5 Ji5TF-/Ot 5.5 Programs and Manuals 5.5.13 Diesel Fuel Qil Testing Program (continued)

2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and b.

c. 5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not invQlve either of the following:

1. A change in the*TS incorporated in the license; or

2. A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of 5.5.14b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71Ce).

(continued) BWQG STS 5.0-16 Rev 1, 04/07/95

Programs and Manuals 5.5 i5TP-!Ob 5.5 Programs and Manuals 5.5.10 Diesel Fuel Oil Testing Program (continued)

a. Acceptability of new fuel oil for use priOr to addition to storage tanks by determining that the fuel 'oi1 has:

1. an API gravity or an absolute specific gravity within 1imits,

2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil,

3. a clear and bright appearance with proper color;

b. ar it' i t d a lt~torage

                                                             .~ne~~/£D

c. Total particulate concentration of the fuel oil is ~ 10 mg/l when tested every 31 days in accordance with ASTM 0-2276, Method A-2 or A-3.

5.5.11 Technical Specifications (TS) Bases Control Program This. program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Basc~ of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees ~ay make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of 5.5.IIb above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without (continued)

BWR/6 STS 5.0-15 Rev 1, 04/07/95

Programs and Manuals 5.5 T5T~-ID6 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program (continued)

2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. a clear and bright appearance with proper color; b.

c. 5.5.14 Technical Specifications (TS) Bases Control Program This' program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

(continued) WOG STS 5.0-16 Rev 1, 04/07/95

Programs and Manuals 5.5 T5TF-/06 5.5 Programs and Manuals (continued) 5.5.10 piesel Fuel Qi1 Testing Program A diesel fuel oil. testing program to implement" required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits, .

2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and

3. a clear and bright appearance with proper color; 5.5.11 Technical Specifications (IS) Bases Control Program Ihis program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the IS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the IS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

(continued) BWR/4 SIS 5.0-15 Rev 1, 04/07/95

Programs and Manuals 5.5 i5T;:-/ob 5.5 Programs and Manuals 5.5.10 Diesel Fuel Oil Testing Program (continued)

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel 'oi1 has:

1. an API gravity or an absolute specific gravity within limits,

2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil,

3. a clear and bright appearance with proper color;

b. ~ s j;f)S~!-gDJ~'\~a~it~J.n.-4ilJrl't :c...

3'~~followfng pm~i ~ad~it~to storage tanks; and C~-f -/1 ---=-f~/--'/i1

                                                               . (1:/    ,,~ ne...J "jt(e   0' c .. Total particulate concentration of the fuel'oil is ~ 10 mg/1 when tested every 31 days in accordance with ASTM 0-2276, Method A-2 or A-3.

5.5.11 Technical Specifications (TS) Bases Control Program This. program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Basc~ of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees ~ay make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of 5.5.IIb above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without (continued)

BWR/6 STS 5.0-15 Rev 1, 04/07/95

BWOG-10, Rev. 0 TSTF-118-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Administrative Controls Program Exceptions NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 3) Improve Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Paul Infanger, (352) 563-4796, paul.infanger@pgnmail.com Revise Administrative Controls 5.5.9 to add the following sentence, "The provisions of SR 3.0.2 are applicable to the Steam Generator Tube Surveillance Program test frequencies," and revise 5.5.13 to add the following sentence, "The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program testing frequencies." These sentences provide consistency with the current application of these requirements as provided in 5.5.6, "Pre-Stressed Concrete Containment Tendon Surveillance Program," and in 5.5.11, "Ventilation Filter Testing program." SR 3.0.2 and SR 3.0.3 are already applicable to the surveillances which reference these programs, and, therefore, the lack of an applicability statement in the Programs introduces confusion. Further, the applicability of SR 3.0.2 and SR 3.0.3 to the program surveillances is consistent with current licensing basis and the old STS. SR 3.0.3 is not addressed for the SGTSP since these tests cannot be performed during operation. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: ANO-1 Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 15-Dec-95 Owners Group Comments: 1/15/96 - Approved by TE Owners Group Resolution: Approved Date: 01-Feb-96 TSTF Review Information TSTF Received Date: 01-Jul-96 Date Distributed for Review: 31-Jul-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: CEOG - Applicable, CEOG accepts WOG - Applicable and accepts BWROG - applicable and accepts TSTF Resolution: Approved Date: 10-Oct-96 NRC Review Information NRC Received Date: 22-Jan-97 NRC Comments: 3/13/97 - NRC approves. 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWOG-10, Rev. 0 TSTF-118-A, Rev. 0 OG Revision 0 Revision Status: Active Final Resolution: NRC Approves Final Resolution Date: 13-Mar-97 Affected Technical Specifications 5.5.9 Steam Generator Tube Surveillance Program NUREG(s)- 1430 1431 1432 Only 5.5.13 Diesel Fuel Oil Testing Program NUREG(s)- 1430 1431 1432 Only 5.5.10 Diesel Fuel Oil Testing Program NUREG(s)- 1433 1434 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

Programs and Manuals 5.5 TS7F-IIB 5.5 Programs and Manuals (continued) 5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as foll ows:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities . testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

5.5.9 Steam Generator (SG) Tube Surveillance Program j ReViewer's Note: The Licensees current licensing basis steam ] generato.r tube surveillance requirements shall be relocated from the LCO and included here. An appropriate administrative controls program format should be used. ~.. . "-TfJ~ P~c/JI:;:;;:;~7SR 1~cF2?ire-eyp/"c~ lie fo -Ih~ S-(; leA'=.)

      '----... 5"c-{ r\~"c' '1/ r ,.,'0' ?r.-.'q,C(;y1  +e:;-r- j,/C''~e'J~ ~ o.          . . - - - -..- -

(continued) BWOG STS 5.0-11 Rev 1, 04/07/95

Programs and Manuals 5.5

                                                                               -'-ST~-IIB 5.5 Programs and Manuals 5.5.13      Diesel Fuel Oil Testing Program (continued)

2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and

3. a clear and bright appearance with proper color;

b. Other properties for ASTM 20 fuel oil are within limits within 30 days'following sampling and addition to storage tanks; and

c. Total particulate concentration of the fuel oil is ~ 10 mg/1 when tested every 31 days in accordance with ASTM 0-2276, Method A-2 or A-3.

5.5.14 Technical Specifications (IS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. A change in the TS incorporated in the license; or

2. A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of 5.5.14b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

TAe-pr~;;;'~-;J --:s-,;5.;;~;-~:d-S;;'-30:Tcue :';I'/ca~

1. Ih. /).

TO -r e VI r-:;<? I .{:<-u ( 0:( "kd';'v]q

../

f,.<:)c;(CU'V7 V

                                                                    ~'<9
                                                                          ./

f:,e'7I/1~"'("/~J, //

(continued) BWOG STS 5.0-16 Rev 1, 04/07/95

Programs and Manuals 5.5 TSTF-lt8 5.5 Programs and Manuals (continued) 5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

5.5.9 Steam Generator (SG) Tube Surveillance Program Reviewer's Note: The Licensee's current licensing basis steam generator tube surveillance requirements shall be relocated from the LCO and included here. An appropriate administrative controls

      ~~program format should be used.

The r:ruv.jioll'l.r f!:>-f S!! ].0. L cV cy~/"ca6Je -It:} -fh~ 56 Tub<

            .5"CA...riJ'eI'/hVlCC Prvqrc.(l'Y",  le.sf ~eq{;lt*,c"ec (continued)

WaG STS 5.0-11 Rev 1, 04/07/95

Programs and Manuals 5.5 TS TF-118 ~- 5.5 Programs and Manuals { 5.5.13 Oiese1 Fuel Oil Testing Program (continued)

2. a flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and

3. a clear and bright appearance with proper color;

b. Other properties for ASTM 20 fuel oil are within limits within 31 days following sampling and addition to storage tanks; and

c. Total particulate concentration of the fuel oil is S 10 mg/l when tested every 31 days in accordance with ASTM 0-2276, Method A-2 or A-3.

5.5.14 Technical Specifications (TS) Bases Control Program Thfs program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

   -rhe     ?r{)v"S/O~H of SR 3,0. '2. a~d SI? 7.0. J a.r~

4rF 1,ca bk. -tv .fh" O'es;~1 F"e I 0,1 "Ted,,,; Proy>r-Q

  +t"'-~*I       I-:r(.'~:'.
  --       {fcr; (continued)

WOG STS 5.0-16 Rev 1, 04/07/95

Programs and Manuals 5.5 {3TF-II13 5.5 Programs and Manuals (continued) 5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semi annua 11 y or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

5.5.9 Steam Generator (SG) Tube Surveillance Program Reviewer's Note: The Licensees current licensing basis steam generator tube surveillance requirements shall be relocated from the LCO and included here. An appropriate administrative controls ______... program format should be used. J h e- frc/v,'s.,'o~~ 0'; $1( ],u,2. cz-,e ar~/"ct:ttle -fo fh.e S(~ IZtIoe S--'-'<.-"'XI'jia~ce Pr~,':(',Y1. (continued) CEOG STS 5.0-11 Rev 1, 04/07/95

Programs and Manuals 5.5

                                                                          ,STF-118         /---.

5.5 Programs and Manuals (continued) 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 20 fuel oil, and

3. A clear and bright appearance with proper color;

b. Other properties for ASTM 20 fuel oil are within limits within 31 days following sampling and addition to storage tanks; and

c. Total particulate concentration of the fuel oil is ~ 10 mg/l when tested every 31 days in accordance with ASTM 0-2276, Method A-2 or A-3.

5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall 'be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

A change in the TS incorporated in the license; or A change to the updated FSAR or Bases that involves an h eviewed safety question as defined in 10 CFR 50.59.

  ~ e.     /?rc.Jvt's/on!-  0+   S(p  J,(),2. C('7c! 51? ].0, 1  Cl-ye   el~/,~
  """0  +he D,'e-sel    h,a ( Dlf    ks-!/,,'1  PrUCjrahl  ks.-f,";) -f>~?Ae"c.,e.r, .,.,/

_____-- ---- - ./ ., (continued) CEOG STS 5.0-16 Rev 1, 04/07/95

r Programs and Manuals f l ! 5.5 l f I {STF-/18 i 5.5 Programs and Manuals 5.5.10 Diesel Fuel Oil Testing Program (continued)

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within 1imits,

2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil,

3. a clear and bright appearance with proper color;

b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and c; Total particulate concentration of the fuel oil is 5 10 mg/1 when tested every 31 days in accordance with ASTM D-2276, ethod A-2 or A-3. . ~J ~

This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Basc~ of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees ~ay make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of 5.5.IIb above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without (continued)

BWRj6 STS 5.0-15 Rev 1, 04/07/95

Programs and Manuals 5.5 TSTF-tIB 5.5 Programs and Manuals (continued) 5.5.10 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits,

2. a flash point and kinematic viscosity within limits for ASTH 20 fuel oil, and

3. a clear and bright appearance with proper color;

b. Other properties for ASTM 20 fuel oil are within limits within 31 days following sampling and addition to storage tanks; and 5.5.11 This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

(continued) BWR/4 STS 5.0-15 Rev 1, 04/07/95

BWOG-16, Rev. 0 TSTF-124-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Delete Specific Reference to Bypasses in CFT and Calibration Definitions NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 2) Consistency/Standardization Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Paul Infanger, (352) 563-4796, paul.infanger@pgnmail.com Revise the definition of CHANNEL CALIBRATION to delete the following sentence: "The CHANNEL CALIBRATION shall also include testing of safety related Reactor Protection System (RPS), Engineered Safety Feature Actuation System (ESFAS), and Emergency Feedwater Initiation and Control (EFIC) bypass functions for each channel affected by the bypass operation." and revise the definition of CHANNEL FUNCTIONAL TEST to delete the last sentence, "The ESFAS CHANNEL FUNCTIONAL TEST shall also include testing of ESFAS safety related bypass functions for each channel affected by bypass operation." Separate delineation of the inclusion of bypasses in the CFT and CAL are not necessary. The current industry standards clearly indicate the operating bypasses are a part of the system function (IEEE-603, which replaced IEEE-279, and is endorsed by RG 1.153) and that CFT should include the testing of such bypasses (IEEE 338, Section 6.3.2). Further , such application of including the required channel bypasses is consistent with the application provided in the ISTS NUREGs for all other vendors. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: ANO-1 Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 15-Dec-95 Owners Group Comments: 1/15/96 - TE Comment - Were bypass functions removed from the Instrument TSs because of the definition? 2/7/96 - Approved by FPC Owners Group Resolution: Approved Date: 01-Feb-96 TSTF Review Information TSTF Received Date: 01-Jul-96 Date Distributed for Review: 31-Jul-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: CEOG - Not applicable WOG - NA BWROG - NA TSTF Resolution: Approved Date: 10-Oct-96 NRC Review Information NRC Received Date: 22-Jan-97 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWOG-16, Rev. 0 TSTF-124-A, Rev. 0 OG Revision 0 Revision Status: Active NRC Comments: 4/17/97 - This change will likely be superseded by the Traveler to revise the Channel Calibration and Channel Functional Test definitions. Final Resolution: NRC Approves Final Resolution Date: 01-Oct-97 Affected Technical Specifications 1.1 Definition of Channel Calibration 1.1 Definition of Channel Functional Test 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

Definitions 1.1 1.1 Definitions CHANNEL CALIBRATION sensing element is replaced, the next required (continued) CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST CONTROL RODS CONTROL RODS shall be all full length safety and regulating rods that are used to shut down the reactor and control power level during maneuvering operations. CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE (continued) BWOG STS 1.1-2 Rev 1, 04/07/95

BWOG-17, Rev. 0 TSTF-125-A, Rev. 1 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Delete EFPD Definition NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 2) Consistency/Standardization Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Paul Infanger, (352) 563-4796, paul.infanger@pgnmail.com Delete the definition of EFFECTIVE FULL POWER DAYS. This definition is unnecessary, and inconsistent with the other ISTS NUREGs. EFPD is a very well understood, common term which does not require definition. This term is consistently used throughout the nuclear industry, and has been consistently applied for many years without a TS definition. Additionally, the term is not used as a defined term in the Bases. A search of the ISTS NUREGs reveals that the term "EFPD" is used in 37 Specifications or Bases in the PWR NUREGs. It is only defined in one occurrence (NUREG-1430, LCO 3.2.2 Bases). EFPD is not a defined term in the other ISTS NUREGs. Therefore, eliminating the defined term from NUREG-1430 and eliminating the one occurrence of it's definition from the NUREG-1430 Bases makes NUREG-1430 internally consistent in its usage and with the other NUREGs (both definitions, Specifications, and Bases). Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: ANO-1 Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 15-Dec-95 Owners Group Comments: 1/15/96 - Approved by TE 1/16/96 - Approved by FPC Owners Group Resolution: Approved Date: 01-Feb-96 TSTF Review Information TSTF Received Date: 01-Jul-96 Date Distributed for Review: 31-Jul-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: CEOG - Not applicable WOG - NA BWROG - NA TSTF Resolution: Approved Date: 10-Oct-96 NRC Review Information NRC Received Date: 22-Jan-97 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWOG-17, Rev. 0 TSTF-125-A, Rev. 1 OG Revision 0 Revision Status: Closed NRC Comments: 2/7/97 - Reviewer recommended approval. 2/13/97 - To C. Grimes for disposition. 3/18/97 C. Grimes modified pkg. The proposed changes are intended to establish consistent terminology in the B&W STS. However, it is not clear from the justification how the changes achieve consistency both within the B&W STS, and between B&W and the other STS, relative to the proper terminology and use of defined terms. 4/17/97 NRC report states that TSTF preparing revision package. Final Resolution: Superceded by Revision Final Resolution Date: 17-Apr-97 TSTF Revision 1 Revision Status: Active Revision Proposed by: TSTF Revision

Description:

NRC requested additional information on how the proposed change made the B&W NUREG more consistent with the other NUREGs and internally. A search of the ISTS NUREGs reveals that the term "EFPD" is used in 37 Specifications or Bases in the PWR NUREGs. It is only defined in one occurrence (NUREG-1430, LCO 3.2.2 Bases). EFPD is not a defined term in the other ISTS NUREGs. Therefore, eliminating the defined term from NUREG-1430 and eliminating the one occurrence of it's definition from the NUREG-1430 Bases makes NUREG-1430 internally consistent in its usage and with the other NUREGs (both definitions, Specifications, and Bases). TSTF Review Information TSTF Received Date: 11-Jan-97 Date Distributed for Review: 15-Jan-98 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 05-Feb-98 NRC Review Information NRC Received Date: 20-Feb-98 Final Resolution: NRC Approves Final Resolution Date: 21-Apr-98 Affected Technical Specifications 1.1 Definitions - EFPD Change

Description:

Definition Deleted LCO 3.2.2 Bases APSR Insertion Limits 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

                                                             --I  .s (  F . 12 S, f~ ,

Definitions 1.1 1.1 Definitions CORE ALTERATION ALTERATIONS shall not preclude completion of (continued) movement of a component to a safe position. CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or those listed in Table E-7 of Regulatory Guide 1.109, Rev. I, NRC, 1977, or ICRP 30, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit ActiVity"]. 'E-AVERAGE 'E shall be the average (weighted in proportion DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives> [15] minutes, making up at least 95% of the total noniodine activity in the coolant. EMERGENCY FEEDWATER The EFIC RESPONSE TIME shall be that time INITIATION AND CONTROL interval from when the monitored parameter (EFle) RESPONSE TIME exceeds its EFIC actuation setpoint at the channel sensor until the emergency feedwater equipment is (continued) SWOG STS 1.1-3 Rev 1, 04/07/95

APSR Insertion Limits B 3.2.2 BASES T~ TF,. /2~ I?e~j I LCO controlling the power distribution within an acceptable (continued) range. The fuel cycle desi n assumes_~SR withdrawal at the c r EFpqvburnup window specified in t e OLR. Prior to t is window, the APSRs cannot be maintained fully withdrawn in steady state operation. After this window, the APSRs are not allowed to be reinserted for the remainder of the fuel cycle. Error adjusted maximum allowable setpoints for APSR insertion are provided in the COLR. The setpoints are derived by adjustment of the measurement system independent limits to allow for THERMAL POWER level uncertainty and rod position errors. Actual alarm setpoints implemented in the unit may be more restrictive than the maximum allowable setpoint values to allow for additional conservatism between the actual alarm setpoints and the measurement system independent limits. APPLICABILITY The APSR physical insertion limits shall be maintained with the reactor in MODES 1 and 2. These limits maintain the power distribution within the range assumed in the accident analyses. In MODE 1, the limits on APSR insertion specified by this LCO maintain the axial fuel burnup design conditions assumed in the reload safety evaluation analysis. In MODE 2, applicability is required because ~ff 0.99. Applicability in MODES 3, 4, and 5 is not required, because the power distribution assumptions in the accident analyses would not be exceeded in these MODES. ACTIONS For steady state power operation, a normal position for APSR insertion is specified in the station operating procedures. The APSRs may be positioned as necessary for transient AXIAL POWER IMBALANCE control until the fuel cycle design requires them to be fully withdrawn. (Not all fuel cycles may incorporate APSR withdrawal.) APSR position limits are not imposed for gray APSRs, with two exceptions. If the fuel cycle design incorporates an APSR withdrawal (usually near end of cycle (EOC)), the APSRs may not be maintained in the fully withdrawn position prior to the fuel cycle burnup for (continued) SWOG STS B 3.2-13 Rev 1. 04/07/95

WOG-65, Rev. 0 TSTF-152-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Revise Reporting Requirements to be Consistent with 10 CFR 20 NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 3) Improve Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Steve Wideman, (620) 364-4037, stwidem@wcnoc.com This change revises 5.6.1, Occupational Radiation Exposure Report, and 5.6.3, Radioactive Effluent Release Report, to be consistent with NRC letter dated 7/28/95 by C. I. Grimes on changes to Tech Specs resulting from 10 CFR 20 and 50.36a changes. NRC provided guidance on how to change the ITS in a proposed Generic Letter that was never issued. These changes reflect the rule changes. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: Diablo Canyon Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 16-Aug-96 Owners Group Comments: (No Comments) Owners Group Resolution: Approved Date: 16-Aug-96 TSTF Review Information TSTF Received Date: 27-Sep-96 Date Distributed for Review: 27-Sep-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: CEOG - Applicable, accepts BWROG - Applicable, accepts WOG - Applicable, accepts TSTF Resolution: Approved Date: 21-Oct-96 NRC Review Information NRC Received Date: 23-Jan-97 NRC Comments: 3/13/97 - NRC approves. Final Resolution: NRC Approves Final Resolution Date: 13-Mar-97 04-Aug-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-65, Rev. 0 TSTF-152-A, Rev. 0 Affected Technical Specifications 5.6.1 Occupational Radiation Exposure Report 5.6.3 Radioactive Effluent Release Report 04-Aug-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

Reporting Requirements 5.6 TS//-/~2 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4. 5.6.1 Occupational Radiation Exposure Report

             -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should' combine sections common to all units at the station. A tabulati on an annual ba' of the number of tation, utility, and oth personnel (inclu 'ng contractors) r. eiving exposures

             > 100 rem/yr and their sociated man re xposure according to wor and job function (e.g., reactor 0 rations and surveill e, i ervice inspecti ,routine mainte ce, special mainten ce describe mainte nce], waste proc sing, and refuelin . This tabulation su ements the requ' ements of 10 CFR 20 06. The dose assign nts to various d y functions may b stimated based on pocket osimeter, therm minescent dosimet        (TLD), or film badge asurements. Sma exposures totall'         < 20% of the indi aual total dose ed not be account for. In the a egate, at least % of the total w e body dose recei ternal sources ould be assigned       specific major wo functions. Th eport shall be s mitted by April 30          each year. [The' itial report shal be submitted by A 11 30 of the year follo ng the initial cr,' icality.]

5.6.2 Annual Radiological Environmental Operating Report

            -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (continued) SWOG STS 5.0-18 Rev 1, 04/07/95

Reporting Requirements 5.6 TITF/5'2.. 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results* of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. [The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.] In the event that some individual results are not av~ilable for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon ~s possible. 5.6.3 Radioactive Effluent Release Re ort

                            .      iJ,al
                  ------------------ ------------NOTE-------------------------------

A single subm'ttal may be made for a multiple unit station. The submittal ad combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. The Radioactive Effluent Release Report covering the operation of the unit shall be sub' ted in accordance with 10 CFR 50.36a. The report s a inc ude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the pr/cr1o objectives outlined in the aDCM and Process Control Program and in conformance with 10 CFR SO.36a and 10 C@FR50'APpendixI' fY'a r:J 1. of Section IV.B.l. D

                                                              )cuf;

-eat: t, :J~t( r (continued) BWOG STS 5.0-19 Rev 1, 04/07/95

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS I -:s /F- /Sl. 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4. 5.6.1 Occupational Radiation Exposure Report

             -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. A tabulatio n an annual basis of ~ number of statio, utility, and other ersonnel (includin ntractors) receivi exposures

            > 100 em/yr and their as lated man rem expos e according to war nd job functions        .g., reactor operate s and surveillance, i ervice inspection outine maintenance pecial maintenance describe maintena e], waste processi ,and refueling).           s tabulation supp ents the requirem s of 10 CFR 20.220. The dose assignme s to various duty nctions may be est' ated based on pocket      imeter, thermolum' scent dosimeter ( ), or film badge me urements. Small e osures totalling          0% of the indivi al total dose nee ot be accounted f . In the aggr- gate, at least 80% f the total whole dy dose received e ernal sources sho        be assigned to s cific major work unctions. The reg t shall be submit d by April 30 of ach year. [The initi      report shall be bmitted by April 0 of the year following e initial critica ty.]

5.6.2 Annual Radiological Environmental Operating Report

            -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (continued) WOG STS 5.0-18 Rev 1, 04/07/95

Reporting Requirements 5.6

                                                                    /3T,c-/S.2 5.6 Reporting Requirements 5.6.2       Annual Radiological Environmental Operating Report   (continued)

(ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. [The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.] In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.3 Radioactive Effluent Release Re ort s~~11

           ----------------- -----------NOTE-------------------------------

A single sUbmitta may be made for a multiple unit station. The submittal combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. The report sha include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR,50, Appendix I, Secti on IV. B.l. (f;j; (continued) WOG STS 5.0-19 Rev 1, 04/07/95

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS N TP-/~2 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4. 5.6.1 Occupational Radiation Exposure Report

              -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. A tabulation n an annual basi of the number of station, utility, and other rsonnel (includ' g contractors) recei 'ng exposures

              > 100     m/yr and their sociated man rem exp re according to work nd job functions e.g., reactor opera" ns and surveill ce, Q

Insert ~ i rvice inspectio , routine maintenanc special mainten e describe mainte ce], waste processi ,and refueling tabulation sup ements the requirem s of 10 CFR 20. 06. The This dose assignm ts to various duty nctions may be imated based on pocket simeter, thermolum" escent dosimeter LD), or film badge m surements. Small osures totallin ~ 20% of the indiv' ual total dose nee not be accounted or. In the agg gate, at least 80% f the total whol ody dose recei e ernal sources sho be assigned to ecific major wo unctions. The re t shall be submi ed by April 30 each year. [The init* report shall be submitted by Ap* 30 of the year following *nitial criticality.] 5.6.2 Annual Radiological Environmental Operating Report

             -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (continued) CEOG STS 5.0-19 Rev 1, 04/07/95

Reporting Requirements 5.6 T~T/'-IS-l ".-- 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. [The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.] In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.3 Radioactive Effluent Release Report Shetll

           ----------------        ----------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. The Radioactive Effluent Release Report covering the operation of the unit'shall be submitted in accordance with 10 CFR 50.36a. The eport shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CF~ Appendix I, Section IV.B.1. c0 Monthly Operating Reports Routine reports of operating statistics and shutdown experience[, including documentation of all challenges to the pressurizer (continued) CEOG STS 5.0-20 Rev 1. 04/07/95

Reporting Requirements 5.6 TSTF-/s2.. r'" i . 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4. 5.6.1 Occupational Radiation Exposure Report

             -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. tabulation on a annual basis of the number of station, utility, and other pers el (including con r-a'Ctors) receiving exposures

            > 100 mrem/ and their associa                man rem exposure acc ding to work ~nd' functions (e.g. eactor operations and rveillance, inservi       inspection, rout' e maintenance, specia aintenance

[de~ lbe maintenance], ste processing, and r ueling). This ta ation supplement he requirements of 1 FR 20.2206. The se assignments to arious duty function ay be estimated b ed on pocket dosimet ,thermoluminescent simeter (TLD), or 1m badge measureme s. Small exposures talling < 20% of e individual to 1 dose need not be ounted for. In e aggregate, least 80% of the t al whole body do received from external ources should be ass' ned to specific jor wor~ functi s. The report shall e submitted by ril 30 of eac year [The initial report hall be submitt by April 30 the yea* fo 11 owi ng in it i a1 c~ t i ca 1ity. ] 5.6.2 Annual Radiological Environmental Operating Report

            -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (cont oj nued) BWR/6 STS 5.0-18 Rev 1, 04/07/95

Reporting Requirements 5.6 5.6 Reporting Requi~ements Ts 7F-!S-:< 5.6.2 Annual Radiological Environmental Operating Report (continued) (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. [The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the ;xposure period associated with each result.] In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.3 Radioactive Effluent Release Report A-;i~~l;-;~b;itt~-b;-;;d;-~~~E~-;~lti;l;-~~it-;t;ti~~~--Th;-- submittal~combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. The Radioactive Effluent Release Report covering the operation of the unit shall be submitte in accordance with 10 CFR 50.36a. The repor shall inclu e a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the pr/v.r +0 unit. The material provided shall be consistent with the {VlaJ.1 J- objectives outlined in the ODCM and Process Control Program and in conformance with 10 eFR 50.36a and 10 e~o~ Appendix I, faGh fe~r- Section IV.B.1. arf 5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience[, including documentation of all challenges to the safety/relief (continued) BWR/6 STS 5.0-19 Rev 1, 04/07/95

Reporting Requirements 5.6 TS TP-ISl 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C~ The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979J. [The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.J In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.3 Radioactive Effluent Release Report A-~i~~l;-~~b;it~-b~-;~d~-~~~E~-;~lti;l;-~~it-~t~ti~~~--Th;-- sUbmittal~combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. du,ri,,", fhe prev/()Ys fjea.- The Radioactive Effluent Release Report covering the operation of t e unit shall be submitted in accordance with 10 CFR 50.36a. The epor s a lnclu e summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the pnOr -fo unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in rna::; 1 o-f conformance with 10 CFR 50.36a and 10 C(i£FR50, Appendix I, Section IV.B.1. eaG"'. :Je~r ev- f 5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience[, including documentation of all challenges to the safety/relief (cont"j nued) BWR/4 STS 5.0-19 Rev 1, 04/07/95

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS

                                                                 -rs IF -/S-2 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Occupational Radiation Exposure Report

             -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. om

            -----.'----~-~---_.--:------~

5.6.2 Annual Radiological Environmental Operating Report

            -------------------------------NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station. The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (continued) BWR/4 STS 5.0-18 Rev 1, 04/07/95

TSTF-/5~ INSERT A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent> 100 mrems and the associated collective deep dose equivalent (reported in person - rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CPR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescence dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling < 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year. [The initial report shall be submitted by April 30 of the year following the initial criticality.]

WOG-78, Rev. 0 TSTF-166-A, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Correct Inconsistency Between LCO 3.0.6 and the SFDP Regarding Performance of an Evaluation NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 1) Correct Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Steve Wideman, (620) 364-4037, stwidem@wcnoc.com Revise LCO 3.0.6 to explicitly require an evaluation per the Safety Function Determination Program. Delete statement "additional . . . limitations may be required" from LCO 3.0.6. There is an inconsistency between LCO 3.0.6, the Safety Function Determination Program (SFDP), and the LCO 3.0.6 Bases. As currently written, LCO 3.0.6 does not explicitly require an evaluation in accordance with the SFDP, rather it states that additional evaluations may be required. Both the SFDP and the LCO 3.0.6 Bases state that upon entry into LCO 3.0.6, an evaluation shall be made to determine if a loss of safety function exists. In addition, because LCO 3.0.6 states that the evaluation be done in accordance with the SFDP and the SFDP states that other appropriate actions may be taken, there is no need for the statement "additional . . . limitations may be required" in LCO 3.0.6. Revision History OG Revision 0 Revision Status: Active Revision Proposed by: Prairie Island Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 10-Oct-96 Owners Group Comments: (No Comments) Owners Group Resolution: Approved Date: 10-Oct-96 TSTF Review Information TSTF Received Date: 11-Oct-96 Date Distributed for Review: 29-Oct-96 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: CEOG - Applicable, accepts BWROG - Applicable, accepts BWOG - Applicable, accepts TSTF Resolution: Approved Date: 03-Dec-96 NRC Review Information NRC Received Date: 27-Mar-97 NRC Comments: 4/16/97 - Reviewer recommends approval. 4/16/97 - Forwarded to C. Grimes for disposition. 04-Aug-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-78, Rev. 0 TSTF-166-A, Rev. 0 OG Revision 0 Revision Status: Active 5/2/97 C. Grimes approved changes. Final Resolution: NRC Approves Final Resolution Date: 02-May-97 Affected Technical Specifications LCO 3.0.6 LCO Applicability 04-Aug-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

LCD Applicability 3.0 TSrf-!t' 3.0 LCO APPLICABILITY (continued) LCD 3.0.6 When a supported system LCD is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to

                 ~~for the supp rted s ste             n this event,
     ~~= eva1uatio*                                         be.--re u    in l~          accordance with Specification 5.5.15, "Safety Function Determination Program. If a loss of safety function is II determined to exist by this program, the appropriate Conditions and Required Actions of the LCD in which the loss of safety function exists are required to be entered.

When a support systemls Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. LCO 3.0.7 Test Exception LCOs [3.1.9, 3.1.10, 3.1.11 and 3.4.19] allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. SWOG STS 3.0-3 Rev 1, 04/07/95

LCD Applicability 3.0 IS. rF-/66 3.0 LCD APPLICABILITY (continued) LCD 3.0.6 When a supported system LCD is not met solely due to a support system LCD not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCD ACTIONS are required to be entered. This is an exception to LCD 3.0.2 for the supeorte tern. In this ev t a . . a evaluation~ a i* atio rna e r i d in accor ance with Speciffca 10n . . , a e y unc 10n Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCD in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCD 3.0.2. LCD 3.0.7 Test Exception LCOs [3.1.9, 3.1.10, 3.1.11, and 3.4.19] allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCD is desired to be met but is not met, the ACTIONS of the Test Exception LCD shall be met. When a Test Exception LCD is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. WOG STS 3.0-3 Rev 1, 04/07/95

LCD Applicability 3.0 TSTF-16t 3.0 LCO APPLICABILITY (continued) LCD 3.0.6 When a supported system LCD is not met solely due to a support system LCD not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCD ACTIONS are required to be entered. This is an exception to LCD ~2 for the su~po ted system. In this ev

      ~ ~@d'i=Oi17evaluatio
      ~ - accordance with Speci ication 5.5.15, IIS afety Function Determination Program (SFDP).II If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCD in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCD 3.0.2. LCD 3.0.7 Special test exception (STE) LCOs [in each applicable LCD section] allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with STE LCOs is optional. When an STE LCD is desired to be met but is not met, the ACTIONS of the STE LCO shall be met. When an STE LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications. CEOG STS 3.0-3 Rev 1, 04/07/95

LCO Applicability 3.0 TS TF-/66 3.0 LCO APPLICABILITY (continued) LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the suppo ted system. In this event,

         ~ k:~~evaluatio                                                 in
         ~      accor ance with Specification 5.5.12, "Safety Function Determination Program (SFDP) .11 If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support systemls Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. LCO 3.0.7 Special Operations LCOs in Section 3.10 allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Special Operations LCOs is optional. When a Special Operations LCO is desired to be met but is not met, the ACTIONS of the Special Operations LCO shall be met. When a Special Operations LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications. BWR/4 STS 3.0-3 Rev 1, 04/07/95

lCO Applicability 3.0 TSTF-/~b 3.0 LCO APPLICABILITY (continued) lCO 3.0.6 When a supported system lCO is not met solely due to a support system lCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to lCO 3.0.2 for the suppor ed s stem. In this event

           ~;,~DiOevaluatio~                   "                      . in
           ~      accordance with Specification 5.5.12, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the lCO in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and ReqUired Actions shall be entered in accordance with lCO 3.0.2. f [ lCO 3.0.7 Special Operations leOs in Section 3.10 allow specified Technical Specifications (TS) requirements to be changed to I permit performance of special tests and operations. Unless otherwise specified, all other IS requirements remain lCO 3.0.7 unchanged. Compliance with Special Operations lCOs is I optional. When a Special Operations LCO is desired to be met but is not met, the ACTIONS of the Special Operations lCO shall be met. When a Special Operations LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications. BWR/6 STS 3.0-3 Rev 1, 04/07/95

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Revision of Channel Calibration, Channel Functional Test, and Related Definitions NUREGs Affected: 1430 1431 1432 1433 1434 Classification: 1) Correct Specifications Recommended for CLIIP?: (Unassigned) Correction or Improvement: (Unassigned) Industry

Contact:

Tom Silko, (802) 258-4146, tsilko@entergy.com In a meeting between the NRC and the NEI TSTF on April 17, 1997, the NRC described problems that had been found with the ISTS definitions of Channel Calibration, Channel Functional Test, and related definitions. The NRC proposed revised definitions for these terms. Based on the NRC's suggestions, revised definitions of these terms have been developed and Bases statements have been added. 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 The revised definitions eliminate a current ambiguity and possible misinterpretation of Channel Calibration, Channel Functional Test, Actuation Logic Test, Channel Operational Test, Trip Actuating Device Operational Test, and Logic System Functional Test. The current definitions use phrases similar to "required sensor, alarm, interlock, display and trip functions," and "required relays and contacts, trip units, solid state logic elements, etc." There is ambiguity in the application of the word "required" and whether the list is inclusive or representative. Therefore, this list has been replaced with phrases similar to, "all devices in the channel required for channel OPERABILITY." This clarifies the use of the word "required" and makes clear that the components that are required to be tested or calibrated are only those that are necessary for the channel to perform its safety function. The list of components is eliminated from the definition. These changes will clarify the requirements and allow for consistent application of the definitions, tests, and calibrations. In addition, in the statement at the end of each definition addressing the allowance to have the test "... performed by means of any series of sequential, overlapping or total channel steps. . ." the statement "... so that the entire channel/relay is tested/calibrated. . ." is deleted. This deletion is purely to remove the conflict between the verbatim reading of the definition where it is stated "... of all devices in the channel required for channel OPERABILITY.. . ." and the flexibility of testing permitting a "... successful test to be the verification of the change of state of a single contact of the relay. . ." as stated in the Bases. The revised CHANNEL FUNCTIONAL TEST definition does not address the method of the testing of all of the required channel devices. As discussed in the Peach Bottom NRC Inspection Report addressing channel testing methods, a successful test of a channel relay and associated required contacts may be the verification of a single contact and that all contacts of the required device need not be tested provided the required channel contact is otherwise tested. The Bases of applicable Surveillances are modified to include this clarification of the acceptable methods of testing. This clarification is applied to all Channel Functional Tests, Channel Operational Tests, and TADOTs. In the Bases of the SRs a statement is added to indicate that, "A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable [CHANNEL FUNCTIONAL TEST / CHANNEL OPERATIONAL TEST / TADOT] of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests." This statement is modified for BWRs only to address the BWR specifics of the scram contactors as they pertain to the test of the Manual Scram Function. This statement is necessary to clarify what verification is required to support a successful test. This Bases statement to address the specifics of how the tests as defined may be performed is appropriate and acceptable because: 1) the entire scope of the required test is still being performed - only the acceptance criteria is modified to require verification of a certain portion of the instrument functions to have a successful test, and 2) all portions of the scope of the required test required for OPERABILITY are being tested, and 3) provision for the acceptance of the verification of change of the state of a single contact of the relay as desired by the NRC. Other changes are made for consistency of the definitions between the ISTS NUREGs. The NUREG-1430 Channel Functional Test and NUREG-1431 Channel Operational Test definitions are modified to include the sentence, "The CHANNEL FUNCTIONAL (OPERATIONAL for NUREG-1431) TEST may be performed by means of any series of sequential, overlapping, or total channel steps." This allowance currently exists in the CEOG, BWR/4 and BWR/6 definitions of Channel Functional Test and is understood to apply to the BWOG and WOG definitions, although not stated. Editorial changes to the WOG Channel Calibration definition are made for consistency with the other ITS NUREGs. The changes proposed increase the consistency of the five NUREGs and are not intended to change the meaning or intent of the affected definitions. Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: Grand Gulf Revision

Description:

Original Issue 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 OG Revision 0 Revision Status: Closed Owners Group Review Information Date Originated by OG: 19-May-97 Owners Group Comments: Approved by the BWROG on 5/19/97 Owners Group Resolution: Approved Date: 19-May-97 TSTF Review Information TSTF Received Date: 07-Jul-97 Date Distributed for Review: 07-Jul-97 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: Preliminary review and comment by the TSTF on 7/9/97. Changes made and Traveler approved for wider review distribution for review. Distributed for comment to all plants involved in Owner's Group ITS activities on 7/14/97. Discussed with NRC at October TSTF/NRC meeting. Originally approved by the TSTF on 10/13/97. 4/21/98 - TSTF agreed to provide to NRC with Bases changes by end of May, 1998. 7/2/98 - Revision 1 created. TSTF Resolution: Superceeded Date: 02-Jul-98 TSTF Revision 1 Revision Status: Closed Revision Proposed by: TSTF Revision

Description:

The Rev 1 is provided to explain the proposed changes in the definitions and to provide for the Industry position of allowing the Bases to further explain the specifics of how the testing is performed. TSTF Review Information TSTF Received Date: 02-Jul-98 Date Distributed for Review: 02-Jul-98 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 10-Jul-98 NRC Review Information NRC Received Date: 01-Sep-98 NRC Comments: 10/12/98 - NRC directs application of the insert to all CFTs and COTs. TSTF to modify. Final Resolution: Superceded by Revision Final Resolution Date: 12-Oct-98 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 TSTF Revision 2 Revision Status: Closed Revision Proposed by: NRC Revision

Description:

TSTF-205 is modified to apply the Insert to the Bases of all Channel Functional Tests, COTs, TADOTs, and LSFTs per NRC comment on 10/12/98. TSTF Review Information TSTF Received Date: 29-Oct-98 Date Distributed for Review: 29-Oct-98 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 02-Nov-98 NRC Review Information NRC Received Date: 03-Nov-98 NRC Comments: 11/12/98 - NRC will approve or provide status by 11/20/98. Final Resolution: Superceded by Revision Final Resolution Date: 22-Dec-98 TSTF Revision 3 Revision Status: Active Revision Proposed by: NRC Revision

Description:

This revision addresses NRC comments. The Traveler is revised to incorporate NRC requested changes to NUREG-1431 definitions of Master Relay Test and Slave Relay Test to make the wording of these definitions more consistent with similar definitiions in other NUREGs. Also incorporated is the the NRC request to add the statement, "at least once per refueling interval with applicable extensions" to the Bases inserts to clarify that the Frequency of the testing of all of the other required contacts of the relay is performed at least once per refueling interval (18 or 24 months, as appropriate) and that the SR 3.0.2 extension, or other applicable extensions, apply. TSTF Review Information TSTF Received Date: 17-Dec-98 Date Distributed for Review: 17-Dec-98 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments: (No Comments) TSTF Resolution: Approved Date: 17-Dec-98 NRC Review Information NRC Received Date: 24-Dec-98 NRC Comments: Date of NRC Letter: 13-Jan-99 1/11/98 - Reviewer recommends approval. 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 TSTF Revision 3 Revision Status: Active Final Resolution: NRC Approves Final Resolution Date: 13-Jan-99 Affected Technical Specifications 1.1 Definitions - Channel Calibration 1.1 Definitions - Channel Functional Test NUREG(s)- 1430 1432 1433 1434 Only SR 3.3.1.4 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1430 Only SR 3.3.2.1 Bases Reactor Protection System (RPS) Manual Reactor Trip NUREG(s)- 1430 Only SR 3.3.3.1 Bases Reactor Protection System (RPS)Reactor Trip Module (RTM) NUREG(s)- 1430 Only SR 3.3.4.1 Bases CONTROL ROD Drive (CRD) Trip Devices NUREG(s)- 1430 Only SR 3.3.5.2 Bases Engineered Safety Feature Actuation System (ESFAS) NUREG(s)- 1430 Only Instrumentation SR 3.3.6.1 Bases Engineered Safety Feature Actuation System (ESFAS) Manual NUREG(s)- 1430 Only Initiation SR 3.3.7.1 Bases Engineered Safety Feature Actuation System (ESFAS) NUREG(s)- 1430 Only Automatic Actuation Logic SR 3.3.8.2 Bases Emergency Diesel Generator (EDG) Loss of Power Start NUREG(s)- 1430 Only (LOPS) SR 3.3.11.2 Bases Emergency Feedwater Initiation and Control (EFIC) System NUREG(s)- 1430 Only Instrumentation SR 3.3.12.1 Bases Emergency Feedwater Initiation and Control (EFIC) Manual NUREG(s)- 1430 Only Initiation SR 3.3.13.1 Bases Emergency Feedwater Initiation and Control (EFIC) Logic NUREG(s)- 1430 Only SR 3.3.14.1 Bases Emergency Feedwater Initiation and Control (EFIC)-Emergency NUREG(s)- 1430 Only Feedwater (EFW)Vector Valve Logic SR 3.3.15.2 Bases Reactor Building (RB) Purge IsolationHigh Radiation NUREG(s)- 1430 Only SR 3.3.16.2 Bases Control Room IsolationHigh Radiation NUREG(s)- 1430 Only SR 3.4.12.7 Bases Low Temperature Overpressure Protection (LTOP) System NUREG(s)- 1430 Only SR 3.4.15.2 Bases RCS Leakage Detection Instrumentation NUREG(s)- 1430 Only SR 3.7.5.6 Bases Emergency Feedwater (EFW) System NUREG(s)- 1430 Only 1.1 Definitions - Actuation Logic Test NUREG(s)- 1431 Only 1.1 Definitions - Channel Operational Test NUREG(s)- 1431 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 1.1 Definitions - Master Relay Test NUREG(s)- 1431 Only 1.1 Definitions - Slave Relay Test NUREG(s)- 1431 Only 1.1 Definitions - Trip Actuating Device Operational Test NUREG(s)- 1431 Only SR 3.3.1.4 Bases Reactor Trip System (RTS) Instrumentation NUREG(s)- 1431 Only SR 3.3.1.7 Bases Reactor Trip System (RTS) Instrumentation NUREG(s)- 1431 Only SR 3.3.1.8 Bases Reactor Trip System (RTS) Instrumentation NUREG(s)- 1431 Only SR 3.3.1.9 Bases Reactor Trip System (RTS) Instrumentation NUREG(s)- 1431 Only SR 3.3.1.13 Bases Reactor Trip System (RTS) Instrumentation NUREG(s)- 1431 Only SR 3.3.1.14 Bases Reactor Trip System (RTS) Instrumentation NUREG(s)- 1431 Only SR 3.3.1.15 Bases Reactor Trip System (RTS) Instrumentation NUREG(s)- 1431 Only SR 3.3.2.5 Bases Engineered Safety Feature Actuation System (ESFAS) NUREG(s)- 1431 Only Instrumentation SR 3.3.2.7 Bases Engineered Safety Feature Actuation System (ESFAS) NUREG(s)- 1431 Only Instrumentation SR 3.3.2.8 Bases Engineered Safety Feature Actuation System (ESFAS) NUREG(s)- 1431 Only Instrumentation SR 3.3.2.11 Bases Engineered Safety Feature Actuation System (ESFAS) NUREG(s)- 1431 Only Instrumentation SR 3.3.4.4 Bases Remote Shutdown System NUREG(s)- 1431 Only SR 3.3.5.2 Bases Loss of Power (LOP) Diesel Generator (DG) Start NUREG(s)- 1431 Only Instrumentation SR 3.3.6.4 Bases Containment Purge and Exhaust Isolation Instrumentation NUREG(s)- 1431 Only SR 3.3.6.6 Bases Containment Purge and Exhaust Isolation Instrumentation NUREG(s)- 1431 Only SR 3.3.7.2 Bases Control Room Emergency Filtration System (CREFS) Actuation NUREG(s)- 1431 Only Instrumentation SR 3.3.7.6 Bases Control Room Emergency Filtration System (CREFS) Actuation NUREG(s)- 1431 Only Instrumentation SR 3.3.8.2 Bases Fuel Building Air Cleanup System (FBACS) Actuation NUREG(s)- 1431 Only Instrumentation SR 3.3.8.4 Bases Fuel Building Air Cleanup System (FBACS) Actuation NUREG(s)- 1431 Only Instrumentation SR 3.3.9.1 Bases Boron Dilution Protection System (BDPS) NUREG(s)- 1431 Only SR 3.4.12.8 Bases Low Temperature Overpressure Protection (LTOP) System NUREG(s)- 1431 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 SR 3.4.15.2 Bases RCS Leakage Detection Instrumentation NUREG(s)- 1431 Only SR 3.4.19.2 Bases RCS Loops-Test Exceptions NUREG(s)- 1431 Only SR 3.1.5.4 Bases Control Element Assembly (CEA) Alignment (Digital) NUREG(s)- 1432 Only SR 3.1.5.6 Bases Control Element Assembly (CEA) Alignment (Analog) NUREG(s)- 1432 Only SR 3.3.1.4 Bases Reactor Protective System (RPS) Instrumentation-Operating NUREG(s)- 1432 Only (Analog) SR 3.3.1.6 Bases Reactor Protective System (RPS) Instrumentation-Operating NUREG(s)- 1432 Only (Analog) SR 3.3.1.7 Bases Reactor Protective System (RPS) Instrumentation-Operating NUREG(s)- 1432 Only (Analog) SR 3.3.1.7 Bases Reactor Protective System (RPS) Instrumentation-Operating NUREG(s)- 1432 Only (Digital) SR 3.3.1.9 Bases Reactor Protective System (RPS) Instrumentation-Operating NUREG(s)- 1432 Only (Digital) SR 3.3.1.11 Bases Reactor Protective System (RPS) Instrumentation-Operating NUREG(s)- 1432 Only (Digital) SR 3.3.1.13 Bases Reactor Protective System (RPS) Instrumentation-Operating NUREG(s)- 1432 Only (Digital) SR 3.3.2.2 Bases Reactor Protective System (RPS) Instrumentation-Shutdown NUREG(s)- 1432 Only (Analog) SR 3.3.2.2 Bases Reactor Protective System (RPS) Instrumentation-Shutdown NUREG(s)- 1432 Only (Digital) SR 3.3.2.3 Bases Reactor Protective System (RPS) Instrumentation-Shutdown NUREG(s)- 1432 Only (Analog) SR 3.3.2.3 Bases Reactor Protective System (RPS) Instrumentation-Shutdown NUREG(s)- 1432 Only (Digital) SR 3.3.3.1 Bases Reactor Protective System (RPS) Logic and Trip Initiation NUREG(s)- 1432 Only (Analog) SR 3.3.3.2 Bases Reactor Protective System (RPS) Logic and Trip Initiation NUREG(s)- 1432 Only (Analog) SR 3.3.3.3 Bases Control Element Assembly Calculators (CEACs) (Digital) NUREG(s)- 1432 Only SR 3.3.3.3 Bases Reactor Protective System (RPS) Logic and Trip Initiation NUREG(s)- 1432 Only (Analog) SR 3.3.3.5 Bases Control Element Assembly Calculators (CEACs) (Digital) NUREG(s)- 1432 Only SR 3.3.4.1 Bases Reactor Protective System (RPS) Logic and Trip Initiation NUREG(s)- 1432 Only (Digital) 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 SR 3.3.4.2 Bases Engineered Safety Features Actuation System (ESFAS) NUREG(s)- 1432 Only Instrumentation (Analog) SR 3.3.4.2 Bases Reactor Protective System (RPS) Logic and Trip Initiation NUREG(s)- 1432 Only (Digital) SR 3.3.4.3 Bases Engineered Safety Features Actuation System (ESFAS) NUREG(s)- 1432 Only Instrumentation (Analog) SR 3.3.4.3 Bases Reactor Protective System (RPS) Logic and Trip Initiation NUREG(s)- 1432 Only (Digital) SR 3.3.5.1 Bases Engineered Safety Features Actuation System (ESFAS) Logic NUREG(s)- 1432 Only and Manual Trip (Analog) SR 3.3.5.2 Bases Engineered Safety Features Actuation System (ESFAS) NUREG(s)- 1432 Only Instrumentation (Digital) SR 3.3.5.2 Bases Engineered Safety Features Actuation System (ESFAS) Logic NUREG(s)- 1432 Only and Manual Trip (Analog) SR 3.3.5.5 Bases Engineered Safety Features Actuation System (ESFAS) NUREG(s)- 1432 Only Instrumentation (Digital) SR 3.3.6.1 Bases Engineered Safety Features Actuation System (ESFAS) Logic NUREG(s)- 1432 Only and Manual Trip (Digital) SR 3.3.6.2 Bases Diesel Generator (DG)-Loss of Voltage Start (LOVS) (Analog) NUREG(s)- 1432 Only SR 3.3.6.3 Bases Engineered Safety Features Actuation System (ESFAS) Logic NUREG(s)- 1432 Only and Manual Trip (Digital) SR 3.3.7.2 Bases Containment Purge Isolation Signal (CPIS) (Analog) NUREG(s)- 1432 Only SR 3.3.7.2 Bases Diesel Generator (DG)-Loss of Voltage Start (LOVS) (Digital) NUREG(s)- 1432 Only SR 3.3.7.3 Bases Containment Purge Isolation Signal (CPIS) (Analog) NUREG(s)- 1432 Only SR 3.3.7.5 Bases Containment Purge Isolation Signal (CPIS) (Analog) NUREG(s)- 1432 Only SR 3.3.8.2 Bases Control Room Isolation Signal (CRIS) (Analog) NUREG(s)- 1432 Only SR 3.3.8.3 Bases Containment Purge Isolation Signal (CPIS) (Digital) NUREG(s)- 1432 Only SR 3.3.8.3 Bases Control Room Isolation Signal (CRIS) (Analog) NUREG(s)- 1432 Only SR 3.3.8.4 Bases Containment Purge Isolation Signal (CPIS) (Digital) NUREG(s)- 1432 Only SR 3.3.8.5 Bases Containment Purge Isolation Signal (CPIS) (Digital) NUREG(s)- 1432 Only SR 3.3.8.5 Bases Control Room Isolation Signal (CRIS) (Analog) NUREG(s)- 1432 Only SR 3.3.8.8 Bases Containment Purge Isolation Signal (CPIS) (Digital) NUREG(s)- 1432 Only SR 3.3.9.2 Bases Chemical and Volume Control System (CVCS) Isolation Signal NUREG(s)- 1432 Only (Analog) 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 SR 3.3.9.2 Bases Control Room Isolation Signal (CRIS) (Digital) NUREG(s)- 1432 Only SR 3.3.9.3 Bases Control Room Isolation Signal (CRIS) (Digital) NUREG(s)- 1432 Only SR 3.3.9.5 Bases Control Room Isolation Signal (CRIS) (Digital) NUREG(s)- 1432 Only SR 3.3.10.1 Bases Shield Building Filtration Actuation Signal (SBFAS) (Analog) NUREG(s)- 1432 Only SR 3.3.10.2 Bases Fuel Handling Isolation Signal (FHIS) (Digital) NUREG(s)- 1432 Only SR 3.3.10.2 Bases Shield Building Filtration Actuation Signal (SBFAS) (Analog) NUREG(s)- 1432 Only SR 3.3.10.3 Bases Fuel Handling Isolation Signal (FHIS) (Digital) NUREG(s)- 1432 Only SR 3.3.10.4 Bases Fuel Handling Isolation Signal (FHIS) (Digital) NUREG(s)- 1432 Only SR 3.3.12.4 Bases Remote Shutdown System (Analog) NUREG(s)- 1432 Only SR 3.3.12.4 Bases Remote Shutdown System (Digital) NUREG(s)- 1432 Only SR 3.3.13.2 Bases [Logarithmic] Power Monitoring Channels (Analog) NUREG(s)- 1432 Only SR 3.3.13.2 Bases [Logarithmic] Power Monitoring Channels (Digital) NUREG(s)- 1432 Only SR 3.4.12.6 Bases Low Temperature Overpressure Protection (LTOP) System NUREG(s)- 1432 Only SR 3.4.15.2 Bases RCS Leakage Detection Instrumentation NUREG(s)- 1432 Only SR 3.4.17.2 Bases Special Test Exceptions (STE) RCS Loops NUREG(s)- 1432 Only 1.1 Definitions - Logic System Functional Test NUREG(s)- 1433 1434 Only SR 3.3.1.1.4 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1433 Only SR 3.3.1.1.5 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1433 Only SR 3.3.1.1.9 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1433 Only SR 3.3.1.1.12 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1433 Only SR 3.3.1.2.5 Bases Source Range Monitor (SRM) Instrumentation NUREG(s)- 1433 Only SR 3.3.1.2.6 Bases Source Range Monitor (SRM) Instrumentation NUREG(s)- 1433 Only SR 3.3.2.1.1 Bases Control Rod Block Instrumentation NUREG(s)- 1433 Only SR 3.3.2.1.2 Bases Control Rod Block Instrumentation NUREG(s)- 1433 Only SR 3.3.2.1.3 Bases Control Rod Block Instrumentation NUREG(s)- 1433 Only SR 3.3.2.1.6 Bases Control Rod Block Instrumentation NUREG(s)- 1433 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 SR 3.3.2.2.2 Bases Feedwater and Main Turbine High Water Level Trip NUREG(s)- 1433 Only Instrumentation SR 3.3.4.1.1 Bases End of Cycle Recirculation Pump Trip (EOC-RPT) NUREG(s)- 1433 Only Instrumentation SR 3.3.4.2.2 Bases Anticipated Transient Without Scram Recirculation Pump Trip NUREG(s)- 1433 Only (ATWS-RPT) Instrumentation SR 3.3.5.1.2 Bases Emergency Core Cooling System (ECCS) Instrumentation NUREG(s)- 1433 Only SR 3.3.5.2.2 Bases Reactor Core Isolation Cooling (RCIC) System Instrumentation NUREG(s)- 1433 Only SR 3.3.6.1.2 Bases Primary Containment Isolation Instrumentation NUREG(s)- 1433 Only SR 3.3.6.1.5 Bases Primary Containment Isolation Instrumentation NUREG(s)- 1433 Only SR 3.3.6.2.2 Bases Secondary Containment Isolation Instrumentation NUREG(s)- 1433 Only SR 3.3.6.3.2 Bases Low-Low Set (LLS) Instrumentation NUREG(s)- 1433 Only SR 3.3.6.3.3 Bases Low-Low Set (LLS) Instrumentation NUREG(s)- 1433 Only SR 3.3.6.3.4 Bases Low-Low Set (LLS) Instrumentation NUREG(s)- 1433 Only SR 3.3.7.1.2 Bases [Main Control Room Environmental Control (MCREC)] System NUREG(s)- 1433 Only Instrumentation SR 3.3.8.1.2 Bases Loss of Power (LOP) Instrumentation NUREG(s)- 1433 Only SR 3.3.8.2.1 Bases Reactor Protection System (RPS) Electric Power Monitoring NUREG(s)- 1433 Only SR 3.4.6.2 Bases RCS Leakage Detection Instrumentation NUREG(s)- 1433 Only SR 3.9.2.2 Bases Refuel Position One-Rod-Out Interlock NUREG(s)- 1433 Only SR 3.3.1.1.4 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1434 Only SR 3.3.1.1.5 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1434 Only SR 3.3.1.1.9 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1434 Only SR 3.3.1.1.12 Bases Reactor Protection System (RPS) Instrumentation NUREG(s)- 1434 Only SR 3.3.1.2.5 Bases Source Range Monitor (SRM) Instrumentation NUREG(s)- 1434 Only SR 3.3.1.2.6 Bases Source Range Monitor (SRM) Instrumentation NUREG(s)- 1434 Only SR 3.3.2.1.1 Bases Control Rod Block Instrumentation NUREG(s)- 1434 Only SR 3.3.2.1.2 Bases Control Rod Block Instrumentation NUREG(s)- 1434 Only SR 3.3.2.1.3 Bases Control Rod Block Instrumentation NUREG(s)- 1434 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-38, Rev. 0 TSTF-205-A, Rev. 3 SR 3.3.2.1.4 Bases Control Rod Block Instrumentation NUREG(s)- 1434 Only SR 3.3.2.1.8 Bases Control Rod Block Instrumentation NUREG(s)- 1434 Only SR 3.3.4.1.1 Bases End of Cycle Recirculation Pump Trip (EOC-RPT) NUREG(s)- 1434 Only Instrumentation SR 3.3.4.2.2 Bases Anticipated Transient Without Scram Recirculation Pump Trip NUREG(s)- 1434 Only (ATWS-RPT) Instrumentation SR 3.3.5.1.2 Bases Emergency Core Cooling System (ECCS) Instrumentation NUREG(s)- 1434 Only SR 3.3.5.2.2 Bases Reactor Core Isolation Cooling (RCIC) System Instrumentation NUREG(s)- 1434 Only SR 3.3.6.1.2 Bases Primary Containment Isolation Instrumentation NUREG(s)- 1434 Only SR 3.3.6.2.2 Bases Secondary Containment Isolation Instrumentation NUREG(s)- 1434 Only SR 3.3.6.3.2 Bases Residual Heat Removal (RHR) Containment Spray System NUREG(s)- 1434 Only Instrumentation SR 3.3.6.4.2 Bases Suppression Pool Makeup (SPMU) System Instrumentation NUREG(s)- 1434 Only SR 3.3.6.5.1 Bases Relief and Low-Low Set (LLS) Instrumentation NUREG(s)- 1434 Only SR 3.3.7.1.2 Bases [Control Room Fresh Air (CRFA)] System Instrumentation NUREG(s)- 1434 Only SR 3.3.8.1.2 Bases Loss of Power (LOP) Instrumentation NUREG(s)- 1434 Only SR 3.3.8.2.1 Bases Reactor Protection System (RPS) Electric Power Monitoring NUREG(s)- 1434 Only SR 3.4.7.2 Bases RCS Leakage Detection Instrumentation NUREG(s)- 1434 Only SR 3.9.1.1 Bases Refueling Equipment Interlocks NUREG(s)- 1434 Only SR 3.9.2.2 Bases Refuel Position One-Rod-Out Interlock NUREG(s)- 1434 Only 31-Jul-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-205, Rev 3 BASES INSERTS INSERT A A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST ofa relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. INSERT B A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST ofa relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. In accordance with Reference 9, the scram contactors must be tested as part of the Manual Scram Function. INSERT C A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL OPERATIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. INSERTD A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable TADOT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

T5TF-2os Definitions r(g.v S 1.1 /""""'- - 1.0 USE AND APPLICATION 1.1 Definitions

          ---------------------~---------------NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times. ALLOWABLE THERMAL POWER ALLOWABLE THERMAL POWER shall be the maximum reactor core heat transfer rate to the reactor coolant permitted by consideration of the number and configuration -of reactor coolant pumps (RCPs) in operation. AXIAL POWER IMBALANCE AXIAL POWER IMBALANCE shall be the power in the top half of the core, expressed as a percentage of RATED THERMAL POWER (RTP), minus the power in the bottom half of the core, expressed as a percentage of RTP. AXIAL POWER SHAPING APSRs shall be control components used to control RODS (APSRs) the axial power distribution of the reactor core. The APSRs are positioned manually by the operator and are not trippable. CHANNEL CALIBRATION Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a (continued) BWOG STS 1.1-1 Rev 1, 04/07/95

T..5Tf:'-::;.OS Definitions ?fAN 3 1.1 1.1 Definitions CHANNEL CALIBRATION sensing element is replaced, the next required (continued) CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of ~Y/se.:~;s~f ~equential overla . *

                                   ~kt~ne12Tr.ztt.!h~"fYr                                               et71'1 The CHANNEL CALIBRATION shall also include testing of safety related Reactor Protection System (RPS),

Engineered Safety Feature Actuation Syste~ (ESFAS), and Emergency Feedwater Initiation and Control (EFIC) bypass functions for each channel affected by the bypass operation. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ._--.----...

              -r1 ~ c,H4:::C-z..-~~-~:;'~:-;1. ~~ ~T ~;;.;- -~~~/*;;;~/11.u2
                                       ~ ,..,.,         oS 4.'" i t2.r O"f"" ~":i ~

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                  '1     I

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                                                                                              " .:~ ...... .....

(continued)

                                                                                                          ."     _--.--~-.,......, .

SWOG STS 1.1-2 Rev 1, 04/07/95

                                                                      /'STPeRtJS p<tW3 RPS Instrumentation B 3.3.1 BASES SURVEILLANCE    SR 3.3.1.2    (continued)

REQUIREMENTS a small fraction of [2J% in any 24 hour period. Furthermore, the control room operators monitor redundant indications and alarms to detect deviations in channel outputs. SR 3.3.1. 3 A comparison of power range nuclear instrumentation channels against incore detectors shall be performed at a 31 day Frequency when reactor power is > 15% RTP. A Note clarifies that 24 hours is allowed for performing the first Surveillance afte, reaching 15% RTP. If the absolute

               *difference between the power range and incore measurements is ~ [2J% RTP, the power range channel is not inoperable, but a CHANNEL CALIBRATION that adjusts the measured imbalance to agree with the incore measurements is necessary. If the power range channel cannot be properly recalibrated, the channel is declared inoperable. The calculation of the Allowable Value envelope assumes a difference in out of core to incore measurements of 2.5%.

Additional inaccuracies beyond those that are measured are also included in the setpoint envelope calculation. The 31 day Frequency is adequate, considering that long term drift of the excore linear amplifiers is small and burnup of the detectors is slow. Also, the excore readings are a strong function of the power produced in the peripheral fuel bundles, and do not represent an integrated reading across the core. The slow changes in neutron flux during the fuel cycle can also be detected at this interval. SR 3.3.1. 4 A CHANNEL FUNCTIONAL TEST is performed on each required RPS channel to ensure that the entire channel will perform the intended function. SetDoints must be found within the Allowab e a ues specified in Table 3.3.1-1. Any setpoint adjustment shall be consistent with the assumptions of the current unit specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the (continued) SWOG SiS S :.:-27 R,::\,' 1

                                                                     ... , 0'""",'10-/
                                                                                    / 0-'~1:

                                                                             .: ;~:"}~
                                                                                    ~ .
                                                                              .  ~, ~'~"'"

RPS Manual Reactor Trip B 3.3.2 TSTF-20~t'~.3 BASES ACTIONS C.l (continued) or 5, the unit must be placed in a MODE in which manual trip is not required. To achieve this status, all CRD trip breakers must be opened. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to open all CRD trip breakers without challenging unit systems. SURVEILLANCE SR 3.3.2.1 REQUIREMENTS This SR requires the performance of a CHANNEL FUNCTIONAL TEST of the Manual Reactor Trip Function. This test verifies the OPERABILITY of the Manual Reactor Tri b actuation of the CRD trip breakers. The Frequen~y shall be once prior to each reactor startup if not performed within the preceding 7 days to ensure the OPERABILITY of the Manual Reactor Trip Function prior to achieving criticality. The Frequency was developed in consideration that these Surveillances are only performed during a unit outage. REFERENCES 1. FSAR, Chapter [7].

SWOG STS B 3.3-33 Rev 1, 04/07/95

RPS-RTM B 3.3.3 TS(F~20~, ~L.I.g BASES SURVEILLANCE SR 3.3.3.1 (continued) REQUIREMENTS this Frequency, the licensee must justify the Frequency as ] [ required by the NRC Staff SER for the topical report. The SRs include performance of a CHANNEL FUNCTIONAL TEST every [45] days on a STAGGERED TEST BASIS. This test shall verify the OPERABILITY of the RTM and its ability to receive and properly respond to channel trip and reactor trip si nals. Calculations have shown that the Frequency (45 days) maintains a high level of reliability of the Reactor Trip System in BAW-10I67 (Ref. 2). REFERENCES 1. FSAR, Chapter [7].

2. BAW-IOI67, May 1986.

SWOG STS B 3.3-38 Rev 1, 04/07/95

CRD Trip Devices B 3.3.4 T~TF-20~ ~w.~s BASES ACTIONS C.1 and C.2 (continued) The 1 hour Completion Time is sufficient to perform the Required Action. D. 1, D. 2. 1, and D. 2. 2 If the Required Actions of Condition A, B, or C are not met within the required Completion Time in MODE 1, 2, or 3, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3, with all CRD trip breakers open or with all power to the CRD System removed within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. E.1 and E.2 If the Required Actions of Condition A, B, or C are not met within the required Completion Time in MODE 4 or 5, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, all CRD trip breakers must be opened or all power to the CRD System removed within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to open all CRD trip breakers or remove all power to the CRD System without challenging unit systems. SURVEILLANCE SR 3.3.4.1 REQUIREMENTS SR 3.3.4.1 is to perform a CHANNEL FUNCTIONAL TEST every 31 days. This test verifies the OPERABILITY of the trip devices by actuation of the end devices. Also, this test independently verifies the undervoltage and shunt trip mechanisms of the AC breakers.j'The Frequency of 31 days is based on operating experience, which has demonstrated that failure of more than one channel of a given function in any 31 day interval is a rare event. REFERENCES 1. FSAR, Chapter [7]. SWOG STS B 3.3-44 Rev 1, 04/07/95

                                                                   -rs -rr::: t::f 0 S-1frJ-v 3 ESFAS Instrumentation B 3.3.5 BASES SURVEILlANCE    SR 3.3.5.2   (continued)

REQUIREMENTS as inoperable, although during this time period it cannot initiate ESFAS. This allowance is based on the inability to perform the Surveillance in the time permitted by the Required Actions. Eight hours is the average time required to perform the Surveillance. It is not acceptable to routinely remove channels from service for more than 8 hours to perform required Surveillance testing. A CHANNEL FUNCTIONAL TEST is performed on each required ESFAS channel to ensure the entire channel will perform the intended functions. Any setpoint adjustment shall be conslsten Wlt e assumptions of the current unit specific setpoint analysis. The Frequency of 31 days is based on unit operating experience, with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given function in any 31 day interval is a rare event. SR 3.3.5.3 CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift to ensure that the instrument channel remains operational between successive tests. CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are within the assumptions of the unit specific setpoint analysis. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint analysis. This Frequency is justified by the assumption of an [18] month calibration interval to determine the magnitude of equipment drift in the setpoint analysis. SR 3.3.5.4 SR 3.3.5.4 ensures that the ESFAS actuation channel response times are less than or equal to the maximum times assumed in the accident analysis. The response time values are the (continued) B1;JOG STS B 3.3-61 Rev 1. Od/07/95

ESFAS Manual Initiation B 3.3.6

                                                          ;s, {F-2o~tJw.*S BASES APPLICABILITY   to respond by manually operating the ESF components, if (continued)   required.

ACTIONS A Note has been added to the ACTIONS indicating separate Condition entry is allowed for each ESFAS manual initiation Function. Condition A applies when one manual initiation channel of one or more ESFAS Functions becomes inoperable. Required Action A.I must be taken to restore the channel to OPERABLE status within the next 72 hours. The Completion Time of 72 hours is based on unit operating experience and administrative controls, which provide alternative means of ESFAS Function initiation via individual component controls. The 72 hour Completion Time is consistent with the allowed outage time for the safety systems actuated by ESFAS. B.I and B. 2 Required Action B.1 and Required Action B.2 apply if Required Action A.1 cannot be met within the required Completion Time. If Required Action A.I cannot be met within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on-operating experience, to reach the required MODES from full power conditions in an orderly manner and without challenging unit systems. SURVEILLANCE SR 3.3.6.1 REQUIREMENTS This SR requires the performance of a CHANNEL FUNCTIONAL TEST of the ESFAS manual initiation. This test verifies that the initiating circuitry is OPERABLE and will actuate the end device (i.e., pump, valves, etc.). e [18] month Frequency is based on the need to perform this Surveillance (continued) BWOG STS B 3.3-65 Rev 1, 04/07/95

ESFAS Automatic Actuation Logic B 3.3.7

                                                           -rS  TF~2D5j 7Pevl3 C".

BASES ACTIONS A.l and A.2 (continued) equivalent to the automatic actuation logic performing its safety function ahead of time. In some cases, placing the component in its engineered safeguard configuration would violate unit safety or operational considerations. In these cases, the component status should not be changed, but the supported system component must be declared inoperable. Conditions which would preclude the placing of a component in its engineered safeguard configuration include, but are not limited to, violation of system separation, a~tivation of fluid systems that could lead to thermal shock, or isolation of fluid systems that are normally functioning. The Completion Time of 1 hour is based on operating experience and reflects the urgency associated with the inoperability of a safety system component. Required Action A.2 requires entry into the Required Actions of the affected supported systems, since the true effect of automatic actuation logic failure is inoperability of the supported sys~em. The Completion Time of 1 hour is based on operating experience and reflects the urgency associated with the inoperability of a safety system component. SURVEILLANCE SR 3.3.7.1 REQUIREMENTS SR 3.3.7.1 is the performance of a CHANNEL FUNCTIONAL TEST on a 31 day STAGGERED TEST BASIS. The test demonstrates that every automatic actuation logic associated with one of the two safety system trains successfully performs the two-out-of-three logic combinations every 31 days. All automatic actuation logics are thus retested every 62 days. The test simulates the required one-out-of-three inputs to the logic circuit and verifies the successful operation of the automatic actuation 10 ic. The Frequency is based on operating experience that demonstrates the rarity of more than one channel failing within the same 31 day interval. Automatic actuation logic response time testing is incorporated into the response time testing required by LCO 3.3.5. (continued) BWOG STS B 3.3-70 Rev 1, 04/07/95

EDG LOPS B 3.3.8 T3TF;2~7Pev,3-' BASES SURVEILLANCE SR 3.3.8.1 (continued) REQUIREMENTS the probability of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with this LCO's required channels. SR 3.3.8.2 The Note allows channel bypass for testing without defining it as inoperable although during this time period it cannot actuate a diesel start. This allowance is based on the assumption that 4 hours is the average time required to perform channel Surveillance. The 4 hour testing allowance does not significantly reduce the probability that the EOG will start trip when necessary. It is not acceptable to routinely remove channels from service for more than 4 hours to perform required Surveillance testing. A CHANNEL FUNCTIONAL TEST is performed on eacn required EOG LOPS channel to ensure the entire channel will perform the intended funct' n Any setpoint adjustments shall be consistent with the assumptions of the current unit specific setpoint analysis. The Frequency of 31 days is considered reasonable based on the reliability of the components and on operating experience that demonstrates channel failure is rare. SR 3.3.8.3 A CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The setpoints and the response to a loss of voltage and a degraded voltage test shall include a single point verification that the trip occurs within the required delay time, as shown in Reference 1. CHANNEL CALIBRATION shall find that measurement setpoint errors are within the assumptions of the unit specific setpoint analysis. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint analysis in Reference 4. (continued) BWOG STS B 3.3-78 Rev 1, 04/07/95

                                                                     -13""~~~S
                                                                            ;fQ.N" ~

EFIC Instrumentation B 3.3.1l BASES SURVEILLANCE SR 3.3.11.1 (continued) REQUIREMENTS criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Off scale low current loop channels are verified to be reading at the bottom of the range and not failed downscaleA The Frequency, about once every shift, is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK

                 ~upp1ements less formal, but more frequent, checks of channel operability during normal operational use of the
              . displays associated with the LCO required channels.

SR 3.3.11. 2 A CHANNEL FUNCTIONAL TEST verifies the function of the required trip, interlock, and alarm functions of the chaone1* Setpoi nts for both tri p and bypass removal unctl0n~ must be found within the Allowable Value specified in the LCD. (Note that the Allowable Values for the bypass removal functions :re specified in the Applicable MODES or Other Specified Condition column of Table 3.3.11-1 as limits on applicability for the trip Functions.) Any setpoint adjustment shall be consistent with the assumptions of the current unit specific setpoint analysis. The Frequency of 31 days is based on unit operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given function in any 31day interval is a rare event. SR 3.3.11.3 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The test verifies the channel responds to a measured parameter within the necessary range (c'Jntinued) BWOG SiS B 3.3-109 Rev 1, 04/07/95

                                                                -;S/Fd?()S
                                                                       ~Q..v3.

EFIC Manual Initiation B 3.3.12 BASES (continued) SURVEILlANCE SR 3.3.12.1 REQUIREMENTS This SR requires the performance of a CHANNEL FUNCTIONAL TEST to en~ure that tne channels can perform their intended functions. i!Ujevet)~~r MFW and Main Steam Line Isolation, the tes need not include actuation of the end device. This is due to the risk of a unit transient caused by the closure of valves associated with MFW and Main Steam Line Isolation or actuating EFW during testing at power. The Frequency of 31 days is based on operating experience that demonstrates the rarity of more than one channel failing within the same 31 day interval. REFERENCES 1. IEE£-279-1971, April 1972. BWOG STS B 3.3-115 Rev 1, 04/07/95

EFIC Logic B 3.3.13 T~TF-2os-;7P~~.3 BASES ACTIONS A.I (continued) Therefore, the failed channel(s) must be restored to OPERABLE status to re-establish the system's single-failure tolerance. Condition A can be thought of as equivalent to failure of a single train of a two train safety system (e.g., the safety function can be accomplished, but a single failure cannot be taken). Thus, the Completion Time of 72 hours has been chosen to be consistent with Completion Times for restoring one inoperable ESF System train.

The EFIC System has not been analyzed for failure of one train of one Function and the opposite train of the same Function. In this condition, the potential for system interactions that disable heat removal capability on EFW has not been evaluated. Consequently, any combination of failures in both channels A and B is not covered by Condition A and must be addressed by entry into LCO 3.0.3.

B.1 and B.2 If Required Action A.I cannot be met within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODES from full power conditions in an orderly manner and without challenging unit systems. SURVEILLANCE SR 3.3.13.1 REQUIREMENTS This SR requires the performance of a CHANNEL FUNCTIONAL TEST to ensure that the channels can perform their intended functions. This test verifies MFW and Main Steam Line Isolation and EFW initiation automatic actuation logics are function~ This test simulates the required inputs to the logic circuit and verifies successful operation of the automatic actuation logic. The test need not include actuation of the end device. This is due to the risk of a unit transient caused by the closure of valves associated (continued) BWOG STS B 3.3-120 Rev 1, 04/07/95

EFIC--EFW--Vector Valve Logic 8 3.3.14 TSTF.. .205"11P~* ? BASES ACTIONS A.l (continued) These conflicting requirements result in the necessity for two valves in series, in parallel with two valves in series, and a four channel valve command system. With one channel inoperable, the system cannot meet the single-failure criterion and still meet the dual functional criteria previously described. Therefore, when one vector valve logic channel is inoperable, the channel must be restored to OPERABLE status within 72 hours. This is analogous to having one EFW train inoperable; wherein a 72 hour Completion Time is provided by the Required Actions of LCD 3.7.4, "EFW System." As such, the Completion Time of 72 hours is based on engineering judgement. 8.1 and 8.2 If Required Action A.l cannot be met within the required Completion Time, the unit must be brought to a MODE in which the LCD does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. SURVEILLANCE SR 3.3.14.1 REQUIREMENTS SR 3.3.14.1 is the performance of a CHANNEL FUNCTIONAL TEST every 31 days. This test demonstrates that the EFIC--EFW--vector valve logic performs its function as desired.? The Frequency is based on operating experience that demonstrates the rarity of more than one channel failing within the same 31 day interval. REFERENCES None. SWOG STS B 3.3-125 Rev 1, 04/07/95

RB Purge Isolation--High Radiation B 3.3.15 nTF-;2.o:r; f'e-VI 3 BASES SURVEILLANCE SR 3.3.15.1 {continued} REQUIREMENTS including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. The Frequency, about once every shift, is based on operating experience that demonstrates channel failure is rare. [At this unit, the following administrative controls and design features {e.g., downscale alarms} immedi~tely alert operators to loss of function.] SR 3.3.15.2 This SR requires the performance of a CHANNEL FUNCTIONAL TEST once every 92 days to ensure that the channels can perform their intended functions. This test verifies the capability of the instrumentation to provide the RB isolation." Any setpoint adjustment shall be consistent with t e assumptions of the current unit specific setpoint analysis. In MODES 1, 2, 3, and 4, the test does not include the actuation of the purge valves, as these valves are normally closed. The justification of a 92 day Frequency, in view of the fact that there is only one channel, is Draft NUREG-1366 {Ref. 4}. SR 3.3.15.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the unit specific setpoint analysis. {continued} BWOG STS B 3.3-131 Rev 1, 04/07/95

Control Room Isolation--High Radiation B 3.3.16 i s TF-;LO~ 1f'ev, ~~ . BASES SURVEILLANCE SR 3.3.16.2 (continued) REQUIREMENTS average time required to perform channel surveillance. It is not acceptable to routinely remove channels from service for more than 3 hours to perform required surveillance testing. SR 3.3.16.2 is the performance of a CHANNEL FUNCTIONAL TEST once every 92 days to ensure that the channels can perform their intended functions. This test verifies the capability of the instrumentation to provide the automatic Control Room Isolation. Any setpoint adjustment shall be consistent with e assumptions of the current unit specific setpoint analysis. The justification of a 92 day Frequency, in view of the fact that there is only one channel, is Draft NUREG-l:66 (Ref. 3). SR 3.3.16.3 This SR requires the performance of a CHANNEL CALIBRATION with a setpoint Allowable Value of ~ [25] mR/hr to ensure that the instrument channel remains operational with the correct setpoint. This test is a complete check of the instrument loop and the transmitter. CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the unit specific setpoint analysis. The Frequency is based on the assumption of an [18J month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis and is consistent with the typical refueling cycle. (continued) SWOG STS B 3.3-138 Rev 1, 04/07/95

LTOP System B 3.4.12 {..5 TF-2oS;~,3 BASES SURVEILLANCE REQUIREMENTS SR 3.4.12.7 (cont;nued'.t::!!l1se"f AJ using the PORV for LTOP. PORV ~ctuation is not needed, as it could depressurize the RCS. The [12] hour Frequency considers the unlikelihood of a low temperature overpressure event during the time. The 31 day Frequency is based on industry accepted practice and is acceptable by experience with equipment reliability. SR 3.4.12.8 The performance of a CHANNEL CALIBRATION is required every [18] months. The CHANNEL CALIBRATION for the LTOP setpoint ensures that the PORV will be actuated at the appropriate

              ~CS pressure by verifying the accuracy of the instrument string. The calibration can only be performed in shutdown.

The Frequency considers a typical refueling cycle and industry accepted practice. REFERENCES 1. 10 CFR 50, Appendix G.

2. Generi c Letter 88-11.

3. FSAR, Sect ion 15.

4. 10 CFR 50.46.

5. 10 CFR 50, Appendix K.

SWOG STS B 3.4-67 Rev 1, 04/07/95

RCS Leakage Detection Instrumentation B 3.4.15 T~ IF2 os, ;fe-v* .? BASES ACTIONS C.1 and C.2 (continued) experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. With both required monitors inoperable, no automatic means of monitoring leakage are available, and immediat~ plant shutdown in accordance with LCO 3.0.3 is required. SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that each channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.15.2 SR 3.4.15.2 requires the performance of a CHANNEL FUNCTIONAL TEST of the required containment atmosphere radioactivity monitor~The test ensures that the monitor can perform its function 1n the desired manner. The test verifies the alarm setpoint and relative accuracy of the instrument~tring. The Frequency of 92 days considers instrument reliability, and operating experience has shown it proper for detecting degradation. SR 3.4.15.3 and SR 3.4.15.4 These SRs require the performance of a CHANNEL CALIBRATION for each of the required RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of [18] months is a typical refueling cycle and considers channel reliability. (continued) ) BWOG STS B 3.4-86 Rev 1, 04/07/95

EFW System B 3.7.5 T..5rF-2o~~.3 BASES SURVEILLANCE SR 3.7.5.5 REQUIREMENTS (continued) This SR ensures that the EFW System is properly aligned by verifying the flow paths to each steam generator prior to entering MODE 2 after more than 30 days in MODE 5 or 6. OPERABILITY of EFW flow paths must be demonstrated before sufficient core heat is generated that would require the operation of the EFW System during a subsequent shutdown. The Frequency is reasonable, based on engineering judgment, in view of other administrative controls to ensure that the flow paths are OPERABLE. To further ensure EFW System alignment, flow path OPERABILITY is verified, following extended outages to determine no misalignment of valves has occurred. This SR ensures that the flow path from the CST to the steam generator is properly aligned. (This SR is not required by those units that use EFW for normal startup and shutdown.) SR 3.7.5.6 and SR 3.7.5.7 For this facility, the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION for the EFW pump suction pressure interlocks are a~follows: REFERENCES 1. FSAR, Section [9.2.7].

2. FSAR, Section [9.2.8].

3. ASME, Boiler and Pressure Vessel Code. Section XI.

BWOG STS B 3.7-31 Rev 1, 04/07/95

Ts. TF-';2oS-Definitions 7(e:w :3 1.1 1.0 USE AND APPLICATION 1.1 Defi nit ions

-------------------------------------NOTE----~--------------------------------

The defined terms of. this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken uQder designated Conditions within specified Completion Times. ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic stat and the verification of the required logic ou pu. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices. AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux (AFD) signals between the [top and bottom halves of a two section excore neutron detector]. CHANNEL CALI B ()(/-ff'v+" Valu J2. ~ o.f.-f'h.a.. fc;.re..I"J2.-f4/" *~+

 ...f1...4. C. ""~"" ~ I 1>1 0  fit /'+  0 f'~,

(continued) WOG STS 1.1-1 Rev 1, 04/07/95

TSTF-20~ Definitions ~tMJ ~ 1.1

                                                                                                                     .. ~.-

1.1 Definitions (continued) CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL OPERATIONAL TEST COT CORE ALTERATIOtf CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [Table III of T1D-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or those listed in Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or ICRP 30, Supplement to Part 1, page

       --,.;., ~ CO"- I'?"t:f.j Iu.. ~ 4--r1" ~ ~                                         .
                              -f r::J.-.#'\ s: ~;-~r 0 ~ ,S" ~?'I; ~:A-z-(                 J
        .,.",,.Q.. e:.,.."r 0 .            '1    '""In -re..I c h  A 1'1 rr¢.' ~ .,J.(2. ~ r.

r-'

olla.r-/~""I'I/1'~ J ~ / (continued) WOG STS 1.1-2 Rev 1, 04/07/95

                                       ~ M /.I {7"'ZO( t<1!l.A- 'f TIE~r                                  / 5 TF--)oS"
                                        /'f'1 ~ I.u.. ~~,..,~ h ~ "., ~e-1.r                            Definitions RI:tlI~3 or ~            sa.,j"Q../'  o.,c ~~,,~~~II 1.1
                                       ~ ..... Ie..,;,,;,,~/ IT"'"  -n ~I  r-/-Q.~rl-' - ~--

1.1 Definitions

                                                                                         -_ .. _-- .. -   -_ ... _,..-- ~-._""". -"

LEAKAGE systems or not to be pressure boundary (continued) LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System;

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified L~KAGE; Pressure Boundary LEAKAGE a JI "" ",sr~" "'Il./""y.s c.

I~ 1Ni. c..h.e:.II't'" ~ J LEAKAGE (except SG LEAKAGE) through a I'~";"~A ~ nonisolable fault in an RCS component body, pipe wall, or vessel wall. aft AM'Ifl,./* OI'LI<Jt1(JI'-IT"/ A S RELAY T ST shall consist of energizing eiSAMa&tQF pelay and verifying the OPERABILITY of eac relay. The MASTER RELAY TEST shall include a continuity check of each associa ave relay~~

                                                                                     .,r MODE                                         A MODE shall correspond to anyone c USlve combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE--OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s). PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are: (continued) WOG STS 1.1-4 Rev 1, 04/07/95

TS TF-";;' os-7J, .t<.. oS", A /I ~ R'=LA '1 -rc:.s r Definitions ~Y g M~~ J:u:... ;?~.,4,..,.,,~ ,b., "'Q..~~ 1.1 "f a...'1 r~,.:tU" a ~ ~u.a...,L;A 1.1

                                            ~ JlQ~/~IJ;'"        Qrf"n TA I ~-,'a.1" r.

1-----, ~ ..

1. I Defi ni ti ons SHUTDOWN MARGIN (SDM) a. All rod cluster control assemblies (RCCAs) are (continued) fylly inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable

                                      ,           of being fully inserted, the reactivity worth of the RCCA must be accounted for in the
    ~    1/ S   /d..vc. I"'~r I" determination of SDM; and
    '1')...L "c.A a,., "e I .,..~ fJ rr.c.A
  .~ ~hd...""""L'                            b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the [nominal zero ap ICRI'f014-' .,."                           power design level].                             .

SLAVE RELAY TES~ ~~ ...~~~;::~LA~Y~T~ES~T shall consist of energizing

                                                           "         nd verifying the OPERABILITY of eac slave relal. The SLAVE RELAY TEST shall include~; a ;+ftjM~ a continuity check of associatea estable actuation devices.~---------~

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function. THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE OPERATIONAL TEST {TADOT}

                                                                   /h.L "P1oa -r /t'Ie-;
                                                                   ~   7?'1/ULn.s'  CJ -f ~ '1 Ss:1JIJ~'" ~/.1   0 'rI.Q.r I~~~~",~ J
                                                                 -f~'-"a.1 C-J,...~"LI ~~;t'~.

WaG STS 1.1-6 Rev 1, 04/07/95

RTS Instrumentation B 3.3.1 j5TF -2o~ ;f~,3 BASES SURVEILLANCE SR 3.3.1.4 REQUIREMENTS (continued) SR 3.3.1.4 is the performance of a TADOT every 31 days on a STAGGERED TEST BASIS. This test shall verify OPERABILITY by

 .-'-         'D    actuation of the end devices.
  --1_ Y1~er'1   l--------~--~

The RTB test shall include separate verification of the undervoltage and shunt trip mechanisms. Independent verification of RTB undervoltage and shunt trip Function is not required for the bypass breakers. No capability is provided for performing such a test at power. The independent test for bypass* breakers is included in ~R 3.3.1.14. The bypass breaker test shall include a local shunt trip. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service. The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry operating experience, considering instrument reliability and operating history data. SR 3.3.1. 5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested every 31 days on a STAGGERED TEST BASIS, using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function. The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry operating experience, considering instrument reliability and operating history data. SR 3.3.1.6 SR 3.3.1.6 is a calibration of the excore channels to the incore channels. If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify the f(~I) input to the overtemperature ~T Function. (continued) WOG STS B 3.3-53 Rev 1, 04/07/95

RTS Instrumentation B 3.3.1 T3 TF - :2 os: ,f'ev, 3 BASES SURVEI LLANCE SR 3.3.1.6 (continued) REQUIREMENTS A Note modifies SR 3.3.1.6. The Note states that this Surveillance is required only if reactor power is > 50% RTP and that [24] hours is allowed for performing the first surveillance after reaching 50% RTP. The Frequency of 92 EFPD is adequate. It is based on industry operating experience considering instrument t reliability and operating history data for instrument drift. SR 3.3.1. 7 SR 3.3.1.7 is the performance of a COT every [92J days. A COT is performed on each required channel to ens~h~ ,-, entire channel will perform the intended Function.~ ~erf~ Setpoints must be within the Allowable Values specified in Table 3.3.1-1. The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology. The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology. The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 7. SR 3.3.1.7 is modified by a Note that provides a 4 hour delay in the requirement to perform this Surveillance for source range instrumentation when entering MODE 3 from MODE

2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the RTBs are open and SR 3.3.1.7 is no longer required to be performed. If the unit is to be in MODE 3 with the RTBs closed for> 4 hours this Surveillance must be performed prior to 4 hours after entry into MODE 3.

The Frequency of [92J days is justified in Reference 7. (continued) WOG STS B 3.3-54 Rev 1, 04/07/95

RTS Instrumentation B 3.3.1 T.5TF~ ).. OS; t{Je-v,.*s. BASES SURVEIllANCE SR 3.3.1.8 REQUIREMENTS (continued) SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6 and P-10 interlocks are in their required state for the existing unit condition. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed within [92] days of the Frequencies prior to reactor startup and four hours after reducing power below P-10 and P-6. The Frequency of "prior to startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "4 hours after reducing power below P-10~ (applicable to intermediate and power range low channels) and "4 hours after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. The Frequency of every 92 days thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 or < P-6 for more than 4 hours, then the testing required by this surveillance must be performed prior to the expiration of the 4 hour limit. Four hours is a reasonable time to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and after reducing power into the applicable MODE << P-10 or < P-6) for periods

               > 4 hours.

SR 3.3.1.9 SR 3.3.1.9 is the performance of a TADOT and is performed every [92] days, as justified in Reference 7. ~ I DJ,;. .- -----.-1 (continued) WOG STS B 3.3-55 Rev I, 04/07/95

RTS Instrumentation B 3.3.1 BASES Ts TF -,2 OS; rrev.3 SURVEILLANCE SR 3.3.1.11 (continued) REQUIREMENTS plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This Surveillance is not required for the NIS power range detectors for entry into MODE 2 or 1, and is not required for the NIS intermediate range detectors for entry into MODE 2, because the unit must be in at least MODE 2 to perform the test for the intermediate range detectors and MODE 1 for the power range detectors. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed on the [18] month Frequency. SR 3.3.1.12 SR 3.3.1.12 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10, every [18] months. This SR is modified by a Note stating that this test shall include verification of the RCS resistance temperature detector (RTD) bypass loop flow rate. This test will verify the rate lag compensation for flow from the core to the RTDs. The Frequency is justified by the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.1.13 ern SR 3.3.1.13 is the performance of a COT of RTS interlocks

     ~~ r J. C ;i';'-_ _e_ve_r~y-,,-[1_8.. :;..]_m_o_n_th_s.....1' The Frequency is based on the known reliability of the interlocks and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

(continued) WOG STS B 3.3-57 Rev 1, 04/07/95

RTS Instrumentation B 3.3.1 TSrF-20~Reu,~ BASES SURVEILLANCE SR 3.3.1.14 REQUIREMENTS (continued) SR 3.3.1.14 is the performance of a TADOT of the Manual Reactor Trip, RCP Breaker Position, and the SI Input from ESF S. This TADOT is performed every [18] months. The test shall independently verify the OPERABILITY of the . undervoltage and shunt trip mechanisms for the Manual Reactor Trip Function for the Reactor Trip Breakers and Reactor Trip Bypass Breakers. The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip. The Frequency is based on the known reliability of*the Functions and the multichannel redundancy available, and has been shown to be acceptable through operating experience. The SR is modified by a Note that excludes verification of setpoints from the TADOT. The Functions affected have no setpoints associated with them. SR 3.3.1.15 SR 3.3.1.15 is the performance of a TADOT of Turbine Trip Functions. This TADOT is as described in SR 3.3.1.4, except that this test is performed prior to reactor startup. A Note states that this Surveillance is not required if it has been performed within the previous 31 days. Verification of the Trip Setpoint does not have to be performed for this Surveil 1ance. Performance of th is test will ensure that the turbine trip Function is OPERABLE prior to taking the reactor critical. This test cannot be performed with the reactor at power and must therefore be performed prior to reactor startup. SR 3.3.1.16 SR 3.3.1.16 verifies that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in Technical Requirements Manual, Section 15 (Ref. 8). Individual component response times are not modeled in the analyses. (continued) WaG STS B 3.3-58 Rev 1, 04/07/95

ESFAS Instrumentation B 3.3.2 T.5 TF-.2 oS; 1f'-ev..3 _ BASES SURVEILLANCE SR 3.3.2.3 (continued) REQUIREMENTS tester is not used and the continuity check does not have to be performed, as explained in the Note. This SR is applied to the balance of plant actuation logic and relays that do not have the SSPS test circuits installed to utilize the semiautomatic tester or perform the continuity check. This test is also performed every 31 days on a STAGGERED TEST BASIS. The Frequency is adequate based on industry operating experience, considering instrument reliability and operating history data. SR 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. This test is performed every 31 days on a STAGGERED TEST BASIS. The time allowed for the testing (4 hours) and the surveillance interval are justified in Reference 8. SR 3.3.2.5 SR 3.3.2.5 is the performance of a COT. A COT is performed on each reqUired channel to ensure the entire channel will perform the intended Function. Setpoints must be found within the Allowable Values s ecified in Table 3.3.1-1. The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology. The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology. The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of the (continued) WaG STS 8 3.3-116 Rev 1, 04/07/95

ESFAS Instrumentation B 3.3.2 T5TF-.2.o~~~". g BASES SURVEILLANCE SR 3.3.2.5 (continued) REQUIREMENTS surveillance interval extension analysis (Ref. 8) when applicable. The Frequency of 92 days is justified in Reference 8. SR 3.3.2.6 SR 3.3.2.6 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment. Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay. This test is performed every [92] days. The Frequency is adequate, based on industry operating experience, considering instrument reliability and operating history data. SR 3.3.2.7 SR 3.3.2.7 is the performance of a TADOT every [92] days. This test is a check of the Loss of Offsite Power, Undervoltage RCP, and AFW Pump Suction Transfer on Suction Pressure- Low Functions. Each Function is tested up to, and including, the master transfer relay coils. The test also includes trip devices that provide actuation signals directly to the SSPS. The SR is modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION. The Frequency is adequate. It is based on industry operating experience, considering instrument reliability and operating history data. (continued) WOG STS B 3.3-117 Rev 1, 04/07/95

ESFAS Instrumentation B 3.3.2 i s TF- 205"1 iPe-u,:g BASES SURVEILLANCE SR 3.3.2.8 REQUIREMENTS (continued) SR 3.3.2.8 is the performance of a TADOT. This test is a check of the Manual Actuation Functions and AFW pump start on trip of all MFW pumps. It is performed every [18] months. Each Manual Actuation Function is tested up to, and includin the master rela coi In some lns ances, the test lnc u es actuation of the end device (i.e., pump starts, valve cycles, etc.). The Frequency is adequate, based on industry operating experience and is consistent with the typical refueling cycle. The SR is modified by a Note that excludes verification of setpoints during the TADOT for manual initiation Functions. The manual initiation Functions have no associated setpoints. SR 3.3.2.9 SR 3.3.2.9 is the performance of a CHANNEL CALIBRATION. A CHANNEL CALIBRATION is performed every [18] months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology. The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology. The Frequency of [18] months is based on the assumption of an [18] month calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology. This SR is modified by a Note stating that this test should include verification that the time constants are adjusted to the prescribed values where applicable. SR 3.3.2.10 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the (continued) WOG STS B 3.3-118 Rev 1, 04/07/95

ESFAS Instrumentation B 3.3.2

                                                            ---r3.TF~;Lo!";  ~v, .?

BASES SURVEILLANCE REQU 1REMENTS . SR 3.3.2.11 (continued) (Tns eA -D~ Trip Interlock, and the Frequency is once per RTB cycle. This Frequency is based on operating experience demonstrating that undetected failure of the P-4 interlock sometimes occurs when the RTB is cycled. The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no associated setpoint. REFERENCES l. FSAR, Chapter [6].

2. FSAR, Chapter [7].

3. FSAR, Chapter [15].

4. lEEE-279-1971.

5. 10 CFR 50.49.

6. RTS/ESFAS Setpoint Methodology Study.

7. NUREG-1218, April 1988.

8. WCAP-I0271-P-A, Supplement 2, Rev. 1, June 1990.

9. Technical Requirements Manual, Section 15. "Response Times."

WOG STS B 3.3-120 Rev 1, 04/07/95

Remote Shutdown System B 3.3.4 T~ TF-?-OS; R~/3_ BASES SURVEILLANCE SR 3.3.4.3 (continued) REQUIREMENTS The Frequency of [18] months is based upon operating experience and consistency with the typical industry refueling cycle. SR 3.3.4.4 SR 3.3.4.4 is the performance of a TADOT every 18 months. This test should verify the OPERABILITY of the reactor trip breakers (RTBs) open and closed indication on the remote shutdown panel, by actuating the RTBs The Frequency is based upon operating experience and consistency with the typical industry refueling outage. REFERENCES 1. 10 CFR 50, Appendix A, GDC 19. WOG STS B 3.3-143 Rev 1, 04/07/95

LOP DG Start Instrumentation B 3.3.5 T5 TF-;;"os, ~e,.;.;g BASES SURVEILLANCE SR 3.3.5.2 ( TV1~e~-I- OJ REQUIREMENTS (continued) SR 3.3.5.2 is the performance of a TADOT. t This test is performed every [31 days]. The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment. For these tests, the relay Trip Setpoints are verified and adjusted as necessary. The Frequency is based on the known reliability of the relays and controls and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

SR 3.3.5.3 SR 3.3.5.3 is the performance of a CHANNEL CALIBRATION.

The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay, as shown in Reference 1. A CHANNEL CALIBRATION is performed every [18] months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. The Frequency of [18] months is based on operating experience and consistency with the typical industry refueling cycle and is justified by the assumption of an [18] month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. REFERENCES 1. FSAR, Section [8.3].

2. FSAR, Chapter [15].

3. Unit Specific RTS/ESFAS Setpoint Methodology Study.

WOG STS B 3.3-149 Rev 1, 04/07/95

Containment Purge and Exhaust Isolation Instrumentation B 3.3.6

                                                     -/-£ T F - ;;2.0 S, ,(>~I $

BASES SURVEILLANCE SR 3.3.6.1 (continued) REQUIREMENTS channels during normal operational use of the displays associated with the LCO required channels. SR 3.3.6.2 SR 3.3.6.2 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. Through t-he semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function. In addition, the master relay coil is pulse tested for continuity. This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. This test is performed every 31 days on a STAGGERED TEST BASIS. The Surveillance interval is acceptable based on instrument reliability and industry operating experience. SR 3.3.6.3 SR 3.3.6.3 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. This test is performed every 31 days on a STAGGERED TEST BASIS. The Surveillance interval- is acceptable based on instrument reliability and industry operating experience. SR 3.3.6.4 A COT is performed every 92 days on each required channel to ensure the entire channel will perform the intended Function.yThe Frequency is based on the staff recommenoation for increasing the availability of radiation monitors according to NUREG-1366 (Ref. 2). This test verifies the capability of the instrumentation to provide the containment purge and exhaust system isolation. The (continued) WOG STS B 3.3-156 Rev 1, 04/07/95

Containment Purge and Exhaust Isolation Instrumentation B 3.3.6 I! T-S TF-;2o~/jJw,.5 BASES SURVEILLANCE SR 3.3.6.4 (continued) REQUIREMENTS setpoint shall be left consistent with the current unit specific calibration procedure tolerance. SR 3.3.6.5 SR 3.3.6.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways .. Actuation equipment that may be operated in the design mitigation mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment. Actuation equipment that may not be operated in the design mitigation mode is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay. This test is performed every [92] days. The Frequency is acceptable based on instrument reliability and industry operating experience. SR 3.3.6.6 SR 3.3.6.6 is the performance of a TADDT. This test is a check of the Manual Actuation Functions and is performed every [18] months. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.). The test also includes trip devices that provide actuation signals directly to the SSPS, bypassing the analog process control equipment. The SR is modified by a Note that excludes verification of setpoints during the TADDT. The Functions tested have no setpoints associated with them. The Frequency is based on the known reliability of the Function and the redundancy available, and has been shown to be acceptable through operating experience. (continued) WOG STS B 3.3-157 Rev I, 04/07/95

CREFS Actuation Instrumentation B 3.3.7 TS-TF-:; o~ I'e.,...$ BASES SURVEILLANCE SR 3.3.7.1 (continued) REQUIREMENTS including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the dis~lays associated with the LCO required channels. SR 3.3.7.2 A COT is performed once every 92 days on each required channel to ensure the entire channel' will perform the intended function. This test verifies the capability of the__----, instrumentation to provide the CREFS actuation. The ~~f~Y setpoints shall be left consistent with the unit specific ~ calibration procedure tolerance. The Frequency is based on-------- the known reliability of the monitoring equipment and has been shown to be accep~ab1e through operating experience. SR 3.3.7.3 SR 3.3.7.3 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. Through the semiautomati~ tester, all possible logic combin~tions, with and without applicable permissives, are tested for each protection function. In addition, the master relay coil is pulse tested for continuity. This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. This test is performed every 31 days on a STAGGERED TEST BASIS. The Frequency is justified in WCAP-I0271-P-A, Supplement 2, Rev. 1 (Ref. 1). SR 3.3.7.4 SR 3.3.7.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity (continued) WOG STS B 3.3-165 Rev 1, 04/07/95

CREFS Actuation Instrumentation B 3.3.7 IS. TF- ;loS; ~,-.g BASES SURVEILLANCE SR 3.3.7.4 (continued) REQUIREMENTS check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. This test is performed every 31 days on a STAGGERED TEST BASIS. The Frequency is acceptable based on instrument reliability and industry operating experience. SR 3.3.7.5 SR 3.3.7.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment. Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay. This test is performed every [92] days. The Frequency is acceptable based on instrument reliability and industry operating experience. SR 3.3.7.6 SR 3.3.7.6 is the performance of a TADOT. This test is a check of the Manual Actuation Functions and is performed every [18] months. Each Manual Actuation Function is tested up to, and including, the master relay coils. ~In~s~om~e~~ ~ instances, the test includes actuation of the end device -r- ~ (Le., pump starts, valve cycles, etc.). _Y1~e.,.. The test also includes trip devices that provide actuation signals directly to the Solid State Protection System, bypassing the analog process control equipment. The Frequency is based on the known reliability of the Function and the redundancy available, and has been shown to be acceptable through operating experience. The SR is modified by a Note that excludes verification of setpoints during the (continued) WOG STS B 3.3-166 Rev 1, 04/07/95

FBACS Actuation Instrumentation B 3.3.8 T.s TF -;2 0 S; tPev,$ BASES SURVEILLANCE SR 3.3.8.1 (continued) REQUIREMENTS channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels. SR 3.3.8.2 A COT is performed once every 92 days on each required channel to ensure the entire channel will perform the intended function. This test verifies the capability of the instrumentation to provide the FBACS actuation. The setpoints shall be left consistent with the unit specific calibration procedure tolerance. The Frequency af 92 days is based on the known reliability of the monitoring equipment and has been shown to be acceptable through operating experience. SR 3.3.8.3 SR 3.3.8.3 is the performance of an ACTUATION LOGIC TEST. The actuation logic is tested every 31 days on a STAGGERED TEST BASIS. All possible logic combinations, with and without applicable permissives, are tested for each protection function. The Frequency is based on the known ~ (continued) WOG STS B 3.3-173 Rev 1, 04/07/95

FBACS Actuation Instrumentation B 3.3.8

                                                        -r~r,c-;20S: ~~,S' BASES SURVEILLANCE   SR 3.3.8.3   (continued)

REQUIREMENTS [ reliability of the relays and controls and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

SR 3.3.8.4 SR 3.3.8.4 is the performance of a TADOT. This test is a check of the manual actuation functions and is performed every [18] months. Each manual actuation function is teste~d up to, and including, the master relay coils.~In some instances, the test includes actuation of the end device

                                                                               ~~~r
                                                                                    =J (e.g., pump starts, valve cycles, etc.). The Frequency is ~l)~ _

based on operating experience and is consistent with the typical industry refueling cycle. The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them. SR 3.3.8.5 A CHANNEL CALIBRATION is performed every [18] months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. The Frequency is based on operating experience and is consistent with the typical industry refueling cycle. REFERENCES 1. 10 CFR 100.11.

2. Unit Specific Setpoint Calibration Procedure.

',lOG STS B 3.3-174 Rev 1, 04/07/95

BOPS B 3.3.9 T£ TF-;2or: tf4l:~ BASES ACTIONS B.l. B.2.1. B.2.2.1. and B.2.2.2 (continued) once per 12 hours thereafter. This backup action is intended to confirm that no unintended boron dilution has occurred while the BOPS was inoperable, and that the required SDM has been maintained. The specified Completion Time takes into consideration sufficient time for the initial determination of SOM and other information available in the control room related to SDM. SURVEILLANCE The BDPS trains are subject to a COT and a CHANNEL REQUIREMENTS CALIBRATION. SR 3.3.9.1 SR 3.3.9.1 requires the performance of a COT every [92] days, to ensure that each train of the BOPS and associated trip setpoints are fully operational. This test shall include verification that the boron dilution alarm setpoint is equal to or less than an increase of twice the count rate within a 10 minute period. The Frequency of [92] day~ is consistent with the requirements for source range channels in WCAP-I0271-P-A (Ref. 2). SR 3.3.9.2 SR 3.3.9.2 is the performance of a CHANNEL CALIBRATION every [18] months. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. For the BOPS, the CHANNEL CALIBRATION shall include verification that on a simulated or actual boron dilution flux doubling signal the centrifugal charging pump suction valves from the RWST open, and the normal CVCS volume control tank discharge valves close in the required closure time of ~ 20 seconds. The Frequency is based on operating experience and consistency with the typical industry refueling cycle. (continued) WOG STS B 3.3-178 Rev 1, 04/07/95

LTOP System B 3.4.12 TSTF~;20S';Rev:S BASES SURVEILLANCE SR 3.4.12.6 (continued) REQUIREMENTS The 72 hour Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify that the PORV block valve remains open. SR 3.4.12.7 Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction valve and RHR suction isolation valve are open and by testing it in accordance with the Inservice Testing Program. (Refer to SR 3.4.12.4 for the RHR suction valve Surveillance and for a description of the requirements of the Inservice Testing Program.) This Surveillance is only performed if the RHR suction relief valve is being used to satisfy this LCD. Every 31 days the RHR suction isolation valve is verified locked open, with power to the valve operator removed, to ensure that accidental closure will not occur. The "locked open" valve must be locally verified in its open position with the manual actuator locked in its inactive position. The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve position. SR 3.4.12.8 Performance of a COT is required within 12 hours after decreasing RCS temperature to 5 [275]OF and every 31 days on each required PORV to verify and, as necessary, adjust its lift set oint The COT will verify the setpoint is within the PTLR allowed maximum limits in the PTLR. PORV actuation could depressurize the RCS and is not required. The 12 hour Frequency considers the unlikelihood of a low temperature overpressure event during this time. A Note has been added indicating that this SR is required to be met 12 hours after decreasing RCS cold leg temperature to 5 [275]OF. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP (continued) WOG STS B 3.4-71 Rev 1, 04/07/95

RCS Leakage Detection Instrumentation B 3.4.15 BASES T:; TF-;2.0S; lP~v, =? ACTIONS E.l and E.2 (continued) required plant conditions from full power conditions in an orderly manner and without challenging plant systems. With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required .. SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner.~ The test verifies the alarm setpoint and relatlve accuracy of the instrument string. The Frequency of 92 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation. SR 3.4.15.3. [SR 3.4.15.4. and SR 3.4.15.51 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of [18] months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this Frequency is acceptable. (continued) WOG STS B 3.4-91 Rev I. 04/07/95

RCS Loops--Test Exceptions B 3.4.19

                                                           / 5 TF--2 oS; ~evi.5' BASES (continued)

ACTIONS . A.l When THERMAL POWER is ~ the P-7 interlock setpoint 10%, the only acceptable action is to ensure the reactor trip breakers (RTBs) are opened immediately in accordance with Required Action A.l to prevent operation of the fuel beyond its design limits. Opening the RTBs will shut down the reactor and prevent operation of the fuel outside of its design limits. SURVEILLANCE SR 3.4.19.1 REQUIREMENTS Verification that the power level is < the P-7 interlock setpoint (10%) will ensure that the fuel design criteria are not violated during the performance of the PHYSICS TESTS. The Frequency of once per hour is adequate to ensure that the power level does not exceed the limit. Plant operations are conducted slowly during the performance of PHYSICS TESTS and monitoring the power level once per hour is sufficient to ensure that the power level does not exceed the limit. SR 3.4.19.2 The power range and intermediate range neutron detectors and the P-7 interlock setpoint must be verified to be OPERABLE and adjusted to the proper value. A COT is performed within 12 hours prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the

i required degree of core protection during the performance of

c y')s~1""+

C-the PHYSICS STS."1'The time limit of 12 hours is sufficient to ensure that t e instrumentation is OPERABLE shortly before initiating PHYSICS TESTS. REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50, Appendix A, GDC 1, 1988.

WOG STS B 3.4-108 Rev 1, 04/07/95

YS TI="-20S-Definitions ~~.;': 1.1 I\Jl .

                                                    ;::=;-tz. ::zIV  I=tn't. M  ~   ,.J       ....

1.0 USE AND APPLICATION 0,-.)1- '1 1.1 Definitions

NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken uijder designated Conditions within specified Completion Times. AXIAL SHAPE INDEX (ASI) ASI shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core. ASI = lower - upper lower + upper AZIMUTHAL POWER TILT AZIMUTHAL POWER TILT shall be the power asymmetry (Tq)-Digital between azimuthally symmetric fuel assemblies. AZIMUTHAL POWER TILT AZIMUTHAL POWER TILT shall be the maximum of the (Tq) -Analog difference between the power generated in any core quadrant (upper or lower) (Pqu~d) and the average. power of all quadrants (P avg ) 1n that half (upper or lower) of the core, divlded by the average power of all quadrants in that half (upper or lower) of the core. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass (continued) CEOG STS 1.1-1 Rev 1, 04/07/95

TSTF-;(os Definitions ~4.".3' 1.1 1.1 Definitions CHANNEL CALIBRATION tle en i s c n~l ~ s~ulridlasJiJl"l07' (continued) ~ ahd1~a(l r ' ._~~~f~e CHANNEL FUNCTIONAL TEST. Ca 1 ra lon (;)./1 t:::hz..v I ~ Q.S I'" \ ument channels with res i stance temperature

                                           \. detector (RTD) or thermocouple sensors may cons i st
    -1"AtJ.. C. A~" ~I                        of an i np 1ace qual itat i ve assessment of sensor
     -rf2-4v" r~ ~                          '\ behavior and normal cal ibration of the remaining
          ~                                      adjustable devices in the channel. Whenever a cJ,...~,.. g,.J.                           . sensing element is repl aced, the next required p.~I:.~I"1~/I"ITY C4."~(/ CHANNEL CALIBRATION shall include an inplace cross
   ~
                                         - ~

calibration that compares the other sensiog elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential in , o tot~lrchannel steps. ~t r t'h CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog and bistable channels--the injection of a simulated or actual signal into the channel as close to the sensor as practicable to .

CJ.p "'"' 1/ d JZ v i ""'S I ~,~~9*~?J~m¥tJ71:~(J:l!f!Ji1Xf) I;' ~ CAa"f1J2.I/ b. -r~/,;; r~~ ~ ! c.h"-;1I1~/ 0;; £tR"" J l.. ,-r Y The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested. (continued) CEGG STS 1.1-2 Rev 1, 04/07/95

RPS Instrumentation--Operating (Analog) B 3.3.1

                                                                    --rS~      :({} .~,~,

BASES ~~.3 SURVEILLANCE SR 3.3.1.4 (continued) REQUIREMENTS Change, every [92] days to ensure the entire channel will perfol'lll its intended function when needed. I: In addition to power supply tests, The RPS CHANNEL

                                                                        -f:?  Pr FUNCTIONAL TEST consists of three overlapping tests as described in Reference 7. These tests verify that the RPS is capable of performing its intended function, from bistable input through the RTCBs. They include: '

Bistable Tests The bistable setpoint must be found to trip within the Allowable Values specified in the LCO and left set consistent with the assumptions of the plant specific setpoint analysis (Ref. 5). As found and as left values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis. The requirements for this review are outlined in Reference 6. A test signal is superimposed on the input in one channel at a time to verify that the bistable trips within the specified tolerance around the setpoint. This is done with the affected RPS channel trip channel bypassed. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the Frequency extension analysis. The requirements for this review are outlined in Reference [9]. Matrix Logic Tests Matrix Logic tests are addressed in lCO 3.3.3. This test is performed one matrix at a time. It verifies that a coincidence in the two input channels for each Function removes power from the matrix relays. During testing, power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. This test will detect any short circuits around the bistable contacts in the coincidence logic, such as may be caused by faulty bistable relay or trip channel bypass contacts.

                                                                 ~(continued)

CEOG STS B 3.3-31 Rev 1, 04/07/95

CEA Alignment (Analog) B 3.1.5 TS T F ;2-0 s-" ;r.v:g BASES

\.

SURVEILLANCE SR 3.1.5.3 (continued) REQU IREHENTS can be detected, and protection can be provided by the CEA deviation circuits. SR 3.1.5.4 Demonstrating the CEA deviation circuit is OPERABLE verifies the circuit is functional. The 31 day Frequency takes into account other information continuously available*to the operator in the control room, so that during CEA movement, deviations can be detected, and protection can be provided by the CEA motion inhibit. SR 3.1. 5.5 Verifying each CEA is trippable would require that each CEA be tripped. In MODES 1 and 2, tripping each CEA would result in radial or axial power tilts, or oscillations. Therefore, individual CEAs are exercised every 92 days to provide increased confidence that all CEAs continue to be trippable, even if they are not regularly tripped. A l movement of [5 inches] is adequate to demonstrate motion"/ without exceeding the alignment limit when only one CEA is being moved. The 92 day Frequency takes into consideration other information available to the operator in the control room and other surveillances being performed more frequently, which add to the determination of OPERABILITY of the CEAs. Between required performances of SR 3.1.5.5, if a CEA(s)is discovered to be immovable, but remains trippable and aligned, the CEA is considered to be OPERABLE. At any time, if a CEA(s) is immovable, a determination of the trippability (OPERABILITY) of the CEA(s) must be made, and appropriate action taken. SR 3.1. 5.6 Performance of a CHANNEL FUNCTIONAL TEST of each reed switch position transmitter channel ensures the channel is OPERABLE and capable of indicating CEA position over the entire len th of the CEA's travel. Since this Surveillance must be per orme w en the reactor is shut down, an 18 month Frequency to be coincident with refueling outage was (continued) CEOG STS B 3.1-32 Rev 1, 04/07/95

I CEA Alignment (Digital) B 3.1.5 T.5 TF- ;2.CS//(ev,s BASES SURVEILLANCE SR 3.1.5.3 (continued) REQUIREMENTS which add to the determination of OPERABILITY of the CEAs (Ref. 7). Between required performances of SR 3.1.5.3, if a CEA(s) is discovered to be immovable but remains trippable and aligned, the CEA is considered to be OPERABLE. At anytime, if a CEA(s) is immovable, a determination of the trippability (OPERABILITY) of that CEA(s) must be made, and appropriate action taken. SR 3.1. 5.4 Performance of a CHANNEL FUNCTIONAL TEST of each reed switch position transmitter channel ensures the channel is OPERABLE and capable of indicating CEA position over the entire 1 h of the CEA's travel. Since this test must be performed w en e reac or is shut down, an 18 month Frequency to be coincident with refueling outage was selected. Operating experience has shown that these components usually pass this Surveillance when performed at a Frequency of once every 18 months. Furthermore, the Frequency takes into account other surveillances being performed at shorter Frequencies, which determine the OPERABILITY of the CEA Reed Switch Indication System. SR 3.1. 5.5 Verification of full length CEA drop times determines that the maximum CEA drop time permitted is consistent with the assumed drop time used in the safety analysis (Ref. 7). Measuring drop times prior to reactor criticality, after reactor vessel head removal, ensures the reactor internals and CEDM will not interfere with CEA motion or drop time, and that no degradation in these systems has occurred that would adversely affect CEA motion or drop time. Individual CEAs whose drop times are greater than safety analysis assumptions are not OPERABLE. This SR is performed prior to criticality due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power. (continued) CEOG STS B 3.1-32 Rev 1, 04/07/95

RPS Instrumentation--Operating (Analog) B 3.3.1 7S'~~(Jr

     .!  BASES                                                                                 A'eu$
         -SU-R-VE~I-LLA-NC-E---S-R- 3.-3-.1-.-6---------/~:::T:=::t1>=;:=::~==Pr==)::---

REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST on the Lo s of Load, Power Rate of Change, and Manual Trip channels is performed prior to a reactor startup to ensure the entir channel will perform its intended function if required. The Loss of Load pressure ,sensor cannot be tested during reactor operation without closing the high pressure TSV, which would result in a turbine trip or reactor trip. The Power Rate of Change--High trip Function is required during startup operation and is bypassed when shut down or > 15% RTP. The Manual Trip Function can either be tested at power or shutdown; however, the simplicity of this circuitry and the absence of drift concern makes this Frequency adequate. Additionally, operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once per 7 days prior to each reactor startup. SR 3.3.1.7 is a CHANNEL FUNCTIONAL TEST similar to

 -".                         SR 3.3.1.4, except SR 3.3.1.7 is applicable only to bypass

,~) Functions and is performed once within 92 days prior to each s ar up Proper operation of bypass permissives is critical during plant startup because the bypasses must be in place to allow startup operation and must be removed at the appropriate points during power ascent to enable certain reactor trips. Consequently, the appropriate time to verify bypass removal function OPERABILITY is just prior to startup. The allowance to conduct this test within 92 days of startup is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 9). Once the operating bypasses are removed, the bypasses must not fail in such a way that the associated trip Function gets inadvertently bypassed. This feature is verified by the trip Function CHANNEL FUNCTIONAL TEST, SR 3.3.1.4. Therefore, further testing of the bypass function after startup is unnecessary. SR 3.3.1.8 SR 3.3.1.8 is the performance of a CHANNEL CALIBRATION every [18] months. (continued) CEOG STS B 3.3-33 Rev 1, 04/07/95

RPS Instrumentation--Shutdown (Analog) B 3.3.2 J3~.:(O.C BASES ~1JA.J3 ( 1 SURVEILLANCE SR 3.3.2.2 (continued) REQUIREMENTS required. The Power Rate of Change-High trip Function is required during startup operation and is bypassed when shut down or > 15% RTP. Additionally, operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once every 92 days prior to each reactor startup. . SR 3.3.2.3 SR 3.3.2.3 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.2.2, except SR 3.3.2.3 is applicable only to bypass Functions and is performed once within 92 days prior to each startup. Proper operation of bypass permissives is critical during plant startup because the bypasses must be in place to allow startup operation and must be removed at the appropriate points during power ascent to enable certain reactor trips. Consequently, the appropriate time to verify bypass removal function OPERABILITY is just prior to startup. The allowance to conduct this Surveillance within 92 days of startup is based on the reliability analysis presented in topical report CEN-327, "RPSjESFAS Extended Test Interval Evaluation" (Ref. 5). Once the operating bypasses are removed, the bypasses must not fail in such a way that the associated trip Function gets inadvertently bypassed. This feature is verified by the trip Function CHANNEL FUNCTIONAL TEST, SR 3.3.2.2. Therefore, further testing of the bypass function after startup is unnecessary. SR 3.3.2.4 SR 3.3.2.4 is the performance of a CHANNEL CALIBRATION every [18J months. CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains (cont i nued ) ,'" \ CEGG STS B 3.3-46 Rev 1, 04/07/95

RPS Logic and Trip Initiation (Analog) B 3.3.3

                                                              --'--~TF-;2.CJS, tfeu.;:S BASES (continued)

SURVEILLANCE Reviewer's Note: In order for a pl ant to take credit f o r ] REQUIREMENTS topical reports as the basis for justifying Frequencies, topical reports must be supported by an NRC staff Safety [ Evaluation Report that establishes the acceptability of each topical report for that unit (Ref. 4). SR 3.3.3.1 A CHANNEL FUNCTIONAL TEST on each RPS Logic channel and RTCB channel is performed every [92] days to ensure tne entire channel will perform its intended function when needed In addition to power supply tests, the RPS CHANNEL FUNCTIONAL TEST consists of three overlapping tests as described in Reference 3. These tests verify that the RPS is capable of performing its intended function, from bistable input through the RTCBs. The first test, the bistable test, is addressed by SR 3.3.1.4 in LCO 3.3.1. This SR addresses the two tests associated with the RPS Logic: Matrix Logic and Trip Path. Matrix Logic Tests These tests are performed one matrix at a time. They verify that a coincidence in the two input channels for each . Function removes power from the matrix relays. During testing, power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. The Matrix Logic tests will detect any short circuits around the bistable contacts in the coincidence logic such as may be caused by faulty bistable relay or trip channel bypass contacts. Trip Path Tests These tests are similar to the Matrix Logic tests, except that test power is withheld from one matrix relay at a time, allowing the initiation circuit to de-energize, opening the affected set of RTCBs. The RTCBs must then be closed prior to testing the other three initiation circuits, or a reactor tri p may result. (continued) CEOG STS B 3.3-59 Rev 1, 04/07/95

RPS Logic and Trip Initiation (Analog) B 3.3.3 BASES

                                                          /!;fF--JoSi 1?'V;.s\

SURVEILLANCE Trip Path Tests (continued) REQUIREMENTS The Frequency of [92] days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). SR 3.3.3.2 A CHANNEL FUNCTIONAL TEST on the Manual Trip channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required. The Manual Trip Function can be tested either at power or shutdown. However, the simplicity of this circuitry and the absence of drift concern makes this Frequency adequate. Additionally, operating experience has shown that these components usually pass the Surveillance when performed once within 7 days prior to each reactor startup. SR 3.3.3.3 Each RTCa is actuated by an undervo1tage coil and a shunt , trip coil. The system is designed so that either ,,~ de-energizing the undervoltage coil or energizing the shunt trip coil will cause the circuit breaker to open. When an RTCB is opened, either during an automatic reactor trip or by using the manual push buttons in the control room, the undervoltage coil is de-energized and the shunt trip coil is energized. This makes it impossible to determine if one of the coils or associated circuitry is defective. Therefore, once every 18 months, a CHANNEL FUNCTIONAL TEST is performed that individually tests all four sets of undervoltage coils and all four sets of shunt trip coils. During undervoltage coil testing, the shunt trip coils shall remain de-energized, preventing their operation. Conversely, during shunt trip coil testing, the undervo1tage coils shall remain energized, preventing their operation. This Surveillance ensures that every undervoltage coil and every shunt trip coil is capable of performing its intended function and that no single active failure of any RTCB component will prevent a reactor trip. e 18 month Frequency is based on the need to perform his Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient of the Surveillance CEOG STS B 3.3-60 (continued) Rev 1, 04/07/95

ESFAS Instrumentation (Analog) B 3.3.4

                                                                 /Sr;t=,;(6/

BASES ~a-.v $ SURVEILLANCE SR 3.3.4.1 (continued) REQUIREMENTS the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of CHANNEL OPERABILITY during normal operational use of displays associated with the LCO required channels. SR 3.3.4.2 A CHANNEL FUNCTIONAL TEST is performed every [92] days to ensure the entire channel will perform its intended function when needed. The CHANNEL FUNCTIONAL TEST tests the individual sensor subsystems using an analog test input to each bistable. A test signal is superimposed on the input in one channel at a time to verify that the bistable trips within the specified tolerance around the setpoint. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [8]. SR 3.3.4.3 SR 3.3.4.3 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.4.2, except 3.3.4.3 is performed within 92 days prior to startup and is only applicable to bypass Functions. These include the Pressurizer Pressure--Low bypass and the

             , ~team Generator Pressure-- Low bypass       .J The CHANNEL FUNCTIONAL TEST for proper operation of the bypass removal Functions is critical during plant heatups because the bypasses may be in place prior to entering MODE 3 but must be removed at the appropriate points during plant startup to enable the ESFAS Function. Consequently, just prior to startup is the appropriate time to verify (continued)

CEGG STS B 3.3-83 Rev 1, 04/07/95

ESFAS Logic and Manual Trip (Analog) B3.3.5 BASES n TF- ;lor;;', ~w. S ( ACTIONS 0.1 and 0.2 (continued) Condition 0 is entered when the Required Action and associated Completion Time of Condition C are not met. If Required Action C.1 cannot be met within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.3.5.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed every 92 days to ensure the entire channel will perform its intended function when needed. Sensor subsystem tests are addressed in Tn.se.A- A r-_---:L=C=O.....;3:..::.~3.:...4..:. .;.:-. -,T. :. .;h. :. . :i. : ,s-.. :s:.:.R:. . : ad::d:.:.r.:es::s:.:e:.s.:...A:.:c,:tu::a:.:::t~io::n:-.:Lo::g:.:i.=c_t:e:s.:ts:.: Actuation Logic Tests Actuation subsystem testing includes injecting one trip l, orO 0 signal into each two-out-of-four logic subsystem in each ESFAS Function and using a bistable trip input to satisfy the trip logic. Initiation relays associated with the affected channel will then actuate the individual ESFAS components. Since each ESFAS Function employs subchannels of Actuation Logic, it is possible to actuate individual components without actuating an entire ESFAS Function. Note 1 requires that Actuation Logic tests include operation of initiation relays. Note 2 allows deferred at power testing of certain relays to allow for the fact that operating certain relays during power operation could cause plant transients or equipment damage. Those initiation relays that cannot be tested at power must be tested in accordance with Note 2. These include [SIAS No.5, SIAS No. 10, CIAS No.5, and MSIS No.1.] These relays actuate the following components, which cannot be tested at power:

RCP seal bleedoff isolation valves; (continued)

CEOG STS B 3.3-98 Rev 1, 04/07/95

ESFAS Logic and Manual Trip (Analog) B 3.3.5 IS TF-.,2os* I(J~. S ( BASES SURVEILLANCE Actuation Logic Tests (continued) REQUIREMENTS

Service water isolation valves;
VCT discharge valves;
Letdown stop valves;
CCW to and from the RCPs;
MSIVs and feedwater isolation valves; and
Instrument air containment isolation valves.

The reasons that each of the above cannot be fully tested at power are stated in Reference 1. . These tests verify that the ESFAS is capable of performing its intended function, from bistable input through the actuated components. The Frequency of [92] days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 2). SR 3.3.5.2 \. ./ A CHANNEL FUNCTIONAL TEST is performed on the manual ESFAS actuation circuitry, de-energizing relays and providing I VI ser-t A~ M_a_n_ua_l_T_r_iP_O_f_t_h_e_Fu_n_c_t1_'o_n . This Surveillance verifies that the trip push buttons are capable of opening contacts in the Actuation Logic as designed, de-energizing the initiation relays and providing Manual Trip of the Function. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every [18J months. REFERENCES 1. FSAR, Section [7.3].

2. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989.

CEOG STS B 3.3-99 Rev 1, 04/07/95

DG--LOVS (Analog) BASES B 3.3.6 7-S-n-: ,::;~ J

~~~

SURVEI LLANCE SR 3.3.6.1 (continued) REQUIREMENTS The Frequency, about once every shift, is based upon operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. SR 3.3.6.2 A CHANNEL FUNCTIONAL TEST is performed every [92] days to ensure that the entire channel will perform its intended function when needed. The Frequency of [92] days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of

     \
      ,                a given function in any [92] day Frequency is a rare event.

--..-) Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [6J. SR 3.3.6.3 SR 3.3.6.3 is the performance of a CHANNEL CALIBRATION every 18 months. The CHANNEL CALIBRATION verifies the accuracy of each component within the instrument channel. This includes calibration of the undervoltage relays and demonstrates that the equipment falls within the specified operating characteristics defined by the manufacturer. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to (continued) CEGG STS B 3.3-107 Rev 1. 04/07/95

CPIS (Analog) B 3.3.7 7'!>7P~6r BASES ~~:2 SURVEILLANCE SR 3.3.7.1 (continued) REQUIREMENTS The Frequency, about once every shift, is based on 'operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels ..

         ,    SR 3.3.7.2 A CHANNEL FUNCTIONAL TEST is perfor ed on each containment radiation monitoring channel to ens re the entire channel
         \   will perform its intended function. Any setpoint adjustment shall be consistent with the assumptions of the current I    plant specific setpoint analysis.

The Frequency of [92] days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any [92] day interval is a rare event. SR 3.3.7.3 Proper operation of the initiation relays is verified by de-energizing these relays during the CHANNEL FUNCTIONAL TEST of the Actuation Logic every [31] days. This will actuate the Function, operating all associated equipment. Proper operation of the equipment actuated by each train is thus verifie. A Note indicates this Surveillance includes erlfication of operation for each initiation relay.,,\ The Frequency of [31] days is based on plant operating experience with regard to channel OPERABILITY, which demonstrates that failure of more than one channel of a given Function in any [31] day interval is a rare event. (continued) CEOG STS B 3.3-114 Rev 1, 04/07/95

CPIS (Analog) B 3.3.7

                                                                 '- / ~ TF- .20S: A:'ev,:3 BASES SURVEILLANCE    SR 3.3.7.4 REQUIREMENT (continued)   CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis.                                  .

The Frequency is based upon the assumption of an [18] month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.7.5 Every [18] months, a CHANNEL FUNCTIONAL TEST is performed on the manual CPIS actuation circuitry. TVlser+-A J,------------- This Surveillance verifies that the trip push buttons are capable of opening contacts in the Actuation Logic as designed, de-energizing the initiation relays and providing Manual Trip of the Function. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every 18 months. SR 3.3.7.6 This Surveillance ensures that the train actuation response times are less than or equal to the maximum times assumed in the analyses. The 18 month Frequency is based upon plant operating experience, which shows random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. Testing of the final actuating devices, which make up the bulk of the response time, is included. Testing (continued) CEOG STS B 3.3-115 Rev 1, 04/07/95

CRIS (Analog) B 3.3.8 j S 7t:= ,d;llJ r BASES I?~::;' (~) SURVEILLANCE SR 3.3.8.1 (continued) REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. At this unit, the following administrative controls and design features (e.g., downscale alarms) immediately alert operations to loss of function in the nonredundant channels. ("0) At this unit, verification of sample system alignment and J'~ [ operation for gaseous, particulate, and iodine monitors is required as follows: SR 3.3.8.2 A CHANNEL FUNCTIONAL TEST is performed on the requi ed control room radiation monitoring channel to ensure the entire channel will perform its intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis. The requirements for this review are outlined in Reference [4]. The Frequency of [92] days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any [92] day interval is a rare event. (continued) CEOG STS B 3.3-122 Rev 1, 04/07/95

CRIS (Analog) B 3.3.8

                                                                /5 TF-..2os,R~.:.g>

BASES SURVEILLANCE SR 3.3.8.3 REQUIREMENTS (continued) Proper operation of the individual initiation relays is verified by de-energizing these relays during the CHANNEL FUNCTIONAL TEST of the Actuation Logic every [31] days. This will actuate the Function, operating all associated

 ~
             +A--   equipment. Proper operation of the equipment actuated by
   --j v)5~(

J each train is thus verified.

                  ~-----------

The Frequency of [31] days is based on plant operating experience with regard to channel OPERABILITY, which demonstrates that failure of more than one channel of a given Function in any [31] days interval is a rare event. Note 1 indicates this Surveillance includes verification of operation for each initiation relay. Note 2 indicates that relays that cannot be tested at power are excepted from the Surveillance Requirement while at power. These relays must, however, be tested during each entry into MODE 5 exceeding 24 hours unless they have been , tested within the previous 6 months. SR 3.3.8.4 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [4]. The Frequency is based upon the assumption of an [18] month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis. (continued) CEOG STS B 3.3-123 Rev 1, 04/07/95

CRrs (Analog) B 3.3.8 BASES T~ TF-;105; f t;? SURVEILLANCE SR 3.3.8.5 REQUIREMENTS (continued) Every [18] months, a CHANNEL FUNCTIONAL TEST is performed on the manual eRIS actuation circuitry. rrl.$~r+ 14 ~----------- This test verifies that the trip push buttons are capable of opening contacts in the Actuation Logic as designed, de-energizing the initiation relays and providing Manual Trip of the function. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every [18] months. SR 3.3.8.6 This Surveillance ensures that the train actuation response times are less than the maximum times assumed in the analyses. The [18] month Frequency is based upon plant ( operating experience, which shows that random failures of instrumentation components causing serious response time~J' degradation, but not channel failure, are infrequent occurrences. Testing of the final actuating devices, which make up the bulk of the response time, is included in the Surveillance testing. REFERENCES 1. FSAR, Chapter [15].

2. "Plant Protection System Selection of Trip Setpoint Values. 1I

3. 10 CFR 50, Appendix A, GDC 19.

4. [] .

CEOG STS B 3.3-124 Rev 1, 04/07/95

CVCS Isolation Signal (Analog) B 3.3.9

                                                        -rSn=:tos BASES                                                           A'a.w- '3 SURVEILLANCE   SR 3.3.9.1    (continued)

REQUIREMENTS minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. SR 3.3.9.2 A CHANNEL FUNCTIONAL TEST is performed on each channel to ensure he entire channel will perform its intended unc ion. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [3]. The Frequency of 31 days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. Proper operation of the individual subgroup relays is verified by de-energizing these relays during the CHANNEL FUNCTIONAL TEST of the Actuation Logic every 31 days. This will actuate the Function, operating all associated equipment. Proper operation of the equipment actuated by each train is thus verified. Note 1 indicates this test includes verification of operation for each initiation relay. [At this unit, the verification is conducted as foll ows:] Note 2 indicates that relays that cannot be tested at power are excepted from the SR while at power. These relays must, however, be tested during each entry into MODE 5 exceeding 24 hours unless they have been tested within the previous 6 months. [At this unit, the basis for this test exception is as foll ows:] (continued) CEOG STS B 3.3-131 Rev 1, 04/07/95

SBFAS (Analog) B 3.3.10 TSTF-.20~Ir'~ .$I BASES (continued) ACTIONS When the number of inoperable channels in a trip Function exceeds those specified in the Conditions associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered if applicable in the current MODE of operation. A.l Condition A applies to the failure of one SBFAS Manual Trip channel or of one Actuation Logic associated with the Chemical and Volume Control System Isolation Signal or SBFAS. Required Action A.1 requires restoration of the inoperable channel to restore redundancy of the affected Function. The Completion Time of 48 hours is consistent with the Completion Time of other ESFAS Functions employing similar logic and should be adequate for most repairs while minimizing the risk of operating with an inoperable channel for a manually actuated Function. B.1 and B.2 Condition B specifies the shutdown track to be followed if the Required Action and associated Completion Time of Condition A are not met. If Required Action A.I cannot be met within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The Completion Times are reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.3.10.1 REQUIREMENTS The SBFAS can be initiated either on a Safety Injection Actuation Signal (SIAS) or manually. This Surveillance is a restatement of SR 3.3.5.1 on the SIAS Function. Performing SR 3.3.5.1 satisfies thiS.~~~~~~llance.~The Frequency is the same as that for SR 3~ Cl'--r-v1s-el'.;...-.C2)--"AJ - J- - - - ' (continued) CEOG STS B 3.3-134 Rev 1, 04/07/95

SBFAS (Analog) B 3.3.10 IS TF- ;;LoS; {J~.g BASES SURVEILLANCE SR 3.3.10.2 REQUIREMENTS (continued) Every [18] months, a CHANNEL FUNCTIONAL TEST is performed on the manual SBFAS actuation circuitry. ~.rVl~e~+ IiJ This Surveillance verifies that the trip push buttons are capable of opening contacts in the Actuation Logic as designed, de-energizing the initiation relays and providing Manual Trip of the Function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every [18] months. REFERENCES 1. FSAR, Chapter [15]. CEOG STS B 3.3-135 Rev 1, 04/07/95

Remote Shutdown System (Analog) B 3.3.12 7~/t=~cr BASES A'~g SURVEILLANCE SR 3.3.12.3 (continued) REQUIREMENTS that the channel responds to the measured parameter within the necessary range and accuracy. The 18 month Frequency is based upon the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power .

             . The SR is modified by a Note, which excludes neutron detectors from the CHANNEL CALIBRATION.

SR 3.3.12.4 SR 3.3.12.4 ;s th performance of a CHANNEL FUNCTIONAL TEST every 18 months. This Surveillance should verify the OPERABILITY of the reactor trip circuit breaker (RTCS) open/closed indication on the remote shutdown panels by actuating the RTCSs. The Frequency of 18 months was chosen because the RTCSs cannot be exercised while the unit is at power. Operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A, GOC 19,and Appendix R.

2. NRC Safety Evaluation Report (SER).

CEOG STS B 3.3-157 Rev 1, 04/07/95

[Logarithmic] Power Monitoring Channels (Analog)

                                                                   , B 3.3.13
                                                               ~~20J'-

BASES Rt2AJS SURVEILLANCE SR 3.3.13.1 (continued) REQUIREMENTS Agreement criteria are determined by the plant staff and should be based on a combination of the channel instrument uncertainties including control isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. If the channels are within the criteria, it is an indication that the channels are OPERABLE. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of displays associated with the LCO required channels. A CHANNEL FUNCTIONAL TEST is performed every [92] days to ensure that the entire ch nnel is capable of properly indicating neutron flux. Internal test circuitry is used to feed preadjusted test signals into the preamplifier to verify channel alignment. It is not necessary to test the detector, because generating a meaningful test signal is difficult; the detectors are of simple construction, and any failures in the detectors will be apparent as change in channel output. This Frequency is the same as that employed for the same channels in the other applicable MODES. [At this unit, the channel trip Functions tested by the CHANNEL FUNCTIONAL TEST are as follows:] SR 3.3.13.3 SR 3.3.13.3 is the performance of a CHANNEL CALIBRATION. A CHANNEL CALIBRATION is performed every [18] months. The Surveillance is a complete check and readjustment of the [logarithmic] power channel from the preamplifier input through to the remote indicators. The Surveillance verifies (continued) CEOG STS B 3.3-161 Rev 1, 04/07/95

RPS Instrumentation--Operating (Digital) B 3.3.1

                                                                      -rS rn= ~ {) r--

BASES ~~:s SURVEILLANCE Matrix Logic Tests REQUIREMENTS (continued) Matrix logic tests are addressed in lCO 3.3.4. This test is performed one matrix at a time. It verifies that a coincidence in the two input channels for each Function removes power from the matrix relays. During testing, power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. This test will detect any short circuits around the bistable contacts in the coincidence logic, such as may be caused by faulty bistable relay or trip channel bypass contacts. Trip Path Tests Trip path (Initiation Logic) tests are addressed in LCD 3.3.4. These tests are similar to the Matrix Logic tests, except that test power is withheld from one matrix relay at a time, allowing the initiation circuit to de-energize, thereby opening the affected set of RTCSs. The RTCBs must then be closed prior to testing the other three ", initiation circuits, or a reactor trip may result. The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 9). The CPC and CEAC channels and excore nuclear instrumentation channels are tested separately. The excore channels use preassigned test signals to verify proper channel alignment. The excore logarithmic channel test signal is inserted into the preamplifier input, so as to test the first active element downstream of the detector. The power range excore test signal is inserted at the drawer input, since there is no preamplifier. The quarterly CPC CHANNEL FUNCTIONAL TEST is performed using software. This software includes preassigned addressable constant values that may differ from the current values. Provisions are made to store the addressable constant values on a computer disk prior to testing and to reload them after testing. A Note is added to the Surveillance Requirements to verify that the CPC CHANNEL FUNCTIONAL TEST includes the correct values of ~ddressable constants. (continued) CEOG STS B 3.3-35 Rev 1, 04/07/95

TSr;::~U~ RPS Instrumentation--Operating (Digital) B 3.3.1 R.vvs-BASES SURVEILLANCE SR 3.3.1.8 REQUIREMENTS (continued) A Note indicates that neutron detectors are excluded from CHANNEL CALIBRATION. A CHANNEL CALIBRATION of the power range neutron flux channels every 92 days ensures that the channels are reading accurately and within tolerance (Ref. 9). The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the interval between surveillance interval extension analysis. The requirements for this review are outlined in Reference [9]. Operating experience has shown this Frequency to be satisfactory. The detectors are excluded from CHANNEL CALIBRATION because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in detector sensitivity are compensated for by performing the daily calorimetric calibration (SR 3.3.1.4) and the monthly linear subchannel gain check (SR 3.3.1.6). In addition, the associated control room indications are monitored by the operators. SR 3.3.1. 9 The characteristics and Bases for this Surveillance are as described for SR 3.3.1.7. This Surveillance differs from SR 3.3.1.7 only in that the CHANNEL FUNCTIONAL TEST on the Loss of Load functional unit is only required above 55% RTP. When above 55% and the trip is in effect, the CHANNEL FUNCTIONAL TEST will ensure the channel will perform its /flSeL .,- equipment protective function if ne ded. The Note allowing ours a er reac 1ng % R P is necessary for Surveillance A- performance. This Surveillance cannot be performed below 55% RTP, since the trip is bypassed. (continued) CEOG STS B 3.3-36 Rev 1, 04/07/95

RPS Instrumentation--Operating (Digital) B 3.3.1

                                                                               -r5*~~4r BASES                                                                 ~~:S SURVEILLANCE    SR 3.3.1.10 REQUIREMENTS (continued)  SR 3.3.1.10 is the performance of a CHANNEL CALIBRATION every [18] months.

CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [9]. The Frequency is based upon the assumption of an [18] month

.. ~.- ""~.                calibration interval for the determination of the magnitude
          \                of equipment drift in the setpoint analysis as well as

.... -... operating experience and consistency with the typical [18] month fuel cycle. The Surveillance is modified by a Note to indicate that the neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in detector sensitivity are compensated for by performing the daily calorimetric calibration (SR 3.3.1.4) and the monthly linear subchannel gain check (SR 3.3.1.6). SR 3.3.1.11 Every [18] months, a CHANNEL FUNCTIONAL TEST is performed on the CPCs. The CHANNEL FUNCTIONAL TEST shall include the injection of a signal as close to the sensors as practicable to verify OPERABILITY including alarm and trip Functions~ The basis for the [18] month Frequency is that the CPCs perform a continuous self monitoring function that eliminates the need for frequent CHANNEL FUNCTIONAL TESTS. This CHANNEL FUNCTIONAL TEST essentially validates the self (continued) CEOG STS B 3.3-37 Rev 1, 04/07/95

RPS Instrumentation--Operating (Digital) B 3.3.1

                                                                   -rS- n= e:?() r. , .
                                                                        ~a-v3' BASES SURVEILLANCE      SR 3.3.1.11   (continued)

REQUIREMENTS monitoring function and checks for a small set of failure' modes that are undetectable by the self monitoring function. Operating experience has shown that undetected CPC or CEAC failures do not occur in any given [18J month interval. SR 3.3.1.12 The three excore detectors used by each CPC channe~ for axial flux distribution information are far enough from the core to be exposed to flux from all heights in the core, although it is desired that they only read their particular level. The CPCs adjust for this flux overlap by using the predetermined shape annealing matrix elements in the epe software. After refueling, it is necessary to re-establish the shape annealing matrix elements for the excore detectors based on more accurate incore detector readings. This is necessary because refueling could possibly produce a significant change in the shape annealing matrix coefficients. Incore detectors are inaccurate at low power levels. THERMAL POWER should be significant but < 70% to perform an accurate axial shape calculation used to derive the shape annealing matrix elements. By restricting power to ~ 70% until shape annealing matrix elements are verified, excessive local power peaks within the fuel are avoided. Operating experience has shown this Frequency to be acceptable. SR 3.3.1.13 SR 3.3.1.13 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.1.7, except SR 3.3.1.13 is applicable only to bypass functions and is performed once within 92 days prior to each 8 X su+Pr _ startup.J' Proper operation of bypass permissives is critical

        ~       aurlng plant startup because the bypasses must be in place to allow startup operation and must be removed at the appropriate points during power ascent to enable certain reactor trips. Consequently, the appropriate time to verify bypass removal function OPERABILITY is just prior to (continued)

CEOG STS B 3.3-38 Rev 1, 04/07/95

RPS Instrumentation--Shutdown (Digital) B 3.3.2

                                                               -rS" ,co?DJ BASES                                                              A'-"~.~-.s SURVEILLANCE   SR 3.3.2.1    (continued)

REQUIREMENTS instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limits. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels.

                                                                               ,~'-~

SR 3.3.2.2 \ . ".:;.... ' A CHANNEL FUNCTIONAL TEST on each channel, except Loss of Load and power range neutron flux, is performed every 92 days to ensure the entire channel will perform its intended function when needed. This SR is identical to SR 3.3.1.7. Only the Applicability d ~ In addition to power supply tests, the RPS CHANNEL FUNCTIONAL TEST consists of three overlapping tests as described in the FSAR, Section [7.2] (Ref. 3). These tests verify that the RPS is capable of performing its intended function, from bistable input through the RTCSs. They include: Bistable Tests A test signal is superimposed on the input in one channel at a time to verify that the bistable trips within the specified tolerance around the setpoint. This is done with the affected RPS channel trip channel bypassed. Any CEGG STS B 3.3-50 Rev 1, 04/07/95

RPS Instrumentation--Shutdown (Digital) B 3.3.2

                                                                  ~~~br
                                                                   / ... ~ .$ .

BASES ,,~ . SURVEILLANCE SR 3.3.2.3 (continued) REQUIREMENTS functions and is performed once within 92 days prior to each startup. This SR is identical to SR 3.3.1.13. Only the Applicability differs. roper operation of bypass permissives is critical during plant startup because the bypasses must be in place to allow startup operation and must be removed at the appropriate points during power ascent to enable certain reactor trips. Consequently, the appropriate time to verify bypass. removal function OPERABILITY is just prior to startup. The allowance to conduct this Surveillance within 92 days of startup is based on the reliability analysis presented in topical report CEN-327, wRPS/ESFAS Extended Test Interval Evaluation* (Ref. 6). Once the operating bypasses are removed, the bypasses must not fail in such a way that the associated trip Function gets inadvertently bypassed. This feature is verified by the trip Function CHANNEL FUNCTIONAL TEST, SR 3.3.2.2. Therefore, further testing of the bypass functi on after startup is' unnecessary. . SR 3.3.2.4 SR 3.3.2.4 is the performance of a CHANNEL CALIBRATION every 18 months. This SR is identical to SR 3.3.1.10. Only the Applicability differs. CHANNEL CALIBRATION is a complete check of the instrument channel excluding the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [3]. The Frequency is based upon the assumption of an [18] month calibration interval for the determination of th magnitude (continued) CEOG STS B 3.3-52 Rev 1, 04/07/95

CEACs (Digital)

                                                                              . B 3.3.3
       ~                                                                        )~~:I,r BASES                                                                 A'~:;

SURVEILLANCE SR 3.3.3.2 (continued) REQUIREMENTS more autorestarts of a nonbypassed CPC occur within a 12 hour period, the CPC may not be completely reliable. Therefore, the Required Action of Condition D must be performed. The Frequency is based on operating experience that demonstrates the rarity of more than one channel failing within the same 12 hour interval. SR 3.3.3.3 A CHANNEL FUNCTIONAL TEST on each CEAC channel is performed every 92 days to ensure the entire channel will perform its intended function when needed. The quarterly CHANNEL FUNCTIONAL TEST is performed using test software. The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended

                      ~~Test I~rval Evaluation" (Ref. 5~
               ~lnS2(" I A SR 3.3.3.4 SR 3.3.3.4 is the performance of a CHANNEL CALIBRATION every

[18] months. CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measur-ed parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [5]. The Frequency is based upon the assumption of an [18] month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis and includes operating experience and consistency with the typical [18] month fuel cycle. (continued) CEGG STS B 3.3-63 Rev 1, 04/07/95

CEACs (Digital) B 3.3.3

                                                                   -rS.,.,=~or*.\

BASES If~.s:* .. )

~------------------------------=-..:....--

SURVEILLANCE SR 3.3.3.5 REQUIREMENTS (continued) Every [18] months, a CHANNEL FUNCTIONAL TEST is performed on the CEACs. The CHANNEL FUNCTIONAL TEST shall include the injection of a signal as close to the sensors as practicable to verify OPERABILITY, including alarm and trip Functions. The basis for the [18] month Frequency is that the CEACs perform a continuous self monitoring function that eliminates the need for frequent CHANNEL FUNCTIONAL TESTS. This CHANNEL FUNCTIONAL TEST essentially validates the self monitoring function and checks for a small set of failure modes that are undetectable by the self monitoring function. Operating experience has shown that undetected CPC or CEAC failures do not occur in any given [18] month interval. SR 3.3.3.6 The isolation characteristics of each CEAC CEA position isolation amplifier and each optical isolator for CEAC to CPC data transfer are verified once per refueling to ensure that a fault in a CEAC or a CPC channel will not render t'~

                                                                             !.~  .... :.. .

another CEAC or CPC channel inoperable. The CEAC CEA \.:~:

                                                                                 .~.-
                                                                                             ~

position isolation amplifiers, mounted in CPC cabinets A and 0, prevent a CEAC fault from propagating back to CPC A or D. The optical isolators for CPC to CEAC data transfer prevent a fault originating in any CPC channel from propagating back to any CEAC through this data link. The Frequency is based on plant operating experience with regard to channel OPERABILITY, which demonstrates the failure of a channel in any [IS] month interval i~rare. REFERENCES 1. 10 CFR 50.

2. 10 CFR 100.

3. FSAR, Section [7.2].

4. NRC Safety Evaluation Report, [Date].

5. CEN-327, June 2, 19S6, including Supplement 1, March 3, 19S9.

CEOG STS B 3.3-64 Rev 1, 04/07/95

RPS Logic and Trip Initiation (Digital) B 3.3.4

                                                       -r~ TF-;2 O~ 1Pev. ~

BASES f ACTIONS 0.1 (continued) If the affected RTCB cannot be opened, Required Action E is entered. This would only occur if there is a failure in the Manual Trip circuitry or the RTCB(s). E.1 and E.2 Condition E is entered if Required Actions associated with Condition A, B, or 0 are not met within the required Completion Time or, if for one or more Functions, more than one Manual Trip, Matrix Logic, Initiation Logic, or RTCB channel is inoperable for reasons other than Condition A or D. If the RTCBs associated with the inoperable channel cannot be opened, the reactor must be shut down within 6 hours and all the RTCBs opened. A Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required plant conditions from full power conditions in an orderly manner and without challenging plant systems and for opening RTCSs. All RTCBs should then be opened, placing the I plant in a MODE where the LCO does not apply and ensuring no ~~J CEA withdrawal occurs. SURVEILLANCE In order for a unit to take credit for topical reports REQUIREMENTS as the basis for justifying Frequencies, topical reports mlJst be supported by an NRC staff Safety Evaluation Report that establishes the acceptability of each topical report for that unit (Ref. 4). SR 3.3.4.1 A CHANNEL FUNCTIONAL TEST on each RPS Logic channel and RTca channel is performed every [92] days to ensure the entire channel will perform its intended function when needed. In addition to power supply tests, the RPS CHANNEL FUNCTIONAL TEST consists of three overlapping tests as described in Reference 3. These tests verify that the RPS is capable of performing its intended function, from (continued) CEOG STS B 3.3-76 Rev 1, 04/07/95

ESFAS Instrumentation (Digital) B 3.3.5

                                                                   -rs:,.-r=~or     .... 0, BASES   (continued)                                                    /fa-v:5            !

SURVEILLANCE SR 3.3.5.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are (:.c~~ OPERABLE. \$ The Frequency, about once every shift, is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent,' checks of channel OPERABILITY during normal operational use of displays associated with the LCO required channels. SR 3.3.5.2 A CHANNEL FUNCTIONAL TEST is performed every 92 days to ensure the entire channel will perform its intended function when needed. The CHANNEL FUNCTIONAL TEST is part of an overlapping test sequence similar to that employed in the RPS. This sequence, consisting of SR 3.3.5.2, SR 3.3.6.1, and CEGG STS B 3.3-104 Rev 1. 04/07/95

ESFAS Instrumentation (Digital) B 3.3.5

                                                                 -; "5;'"7"(=- ~4 r BASES                                                               /<Q-V 5        ()
                                                                                    ,~ :." ....-

SURVEI LLANCE SR 3.3.5.3 (continued) REQUIREMENTS The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. SR 3.3.5.4 This Surveillance ensures that the train actuation response times are within the maximum values assumed in the safety analyses. Response time testing acceptance criteria are included in Reference 10. ESF RESPONSE TIME tests are conducted on a STAGGERED TEST BASIS of once every [18] months. The [18] month Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. (~

SR 3.3.5.5 SR 3.3.5.5 is a CHANNEL FUNCTIONAL TEST sim"ilar to SR 3.3.5.2, except SR 3.3.5.5 is performed within 92 days prior to startup and is only applicable to bypass functions. Since the Pressurizer Pressure--Low bypass is identical for both the RPS and ESFAS, this is the same Surveillance performed for the RPS in SR 3.3.1.13. The CHANNEL FUNCTIONAL TEST for proper operation of the bypass permissives is critical during plant heatups because the bypasses may be in place prior to entering MODE 3 but must be removed at the appropriate points during plant startup to enable the ESFAS Function. Consequently, just prior to startup is the appropriate time to verify bypass function OPERABILITY. Once the bypasses are removed, the bypasses must not fail in such a way that the associated ESFAS Function is inappropriately bypassed. This feature is verified by SR 3.3.5.2. (continued) ,~~~~ CEOG STS B 3.3-106 Rev 1, 04/07/95

ESFAS Logic and Manual Trip (Digital) B 3.3.6 T5 T ~-.2 OS: Reu. (i5'.:::; BASES (continued) '.

                                                                                         . ."~."

SURVEILLANCE SR 3.3.6.1 r,; .'. REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed every [92] days to ensure the entire channel will perform its intended function when needed:.1' The CHANNEL FUNCTIONAL TEST is part of an overlapping test sequence similar to that employed in the RPS. This sequence, consisting of SR 3.3.5.2, SR 3.3.6.1, and SR 3.3.6.2, tests the entire ESFAS from the bistable input through the actuation of the individual subgroup relays. These overlapping tests are described in Reference 1. SR 3.3.5.2 and SR 3.3.6.1 are normally performed together and in conjunction with ESFAS testing. SR 3.3.6.2 verifies that the subgroup relays are capable of actuating their respective ESF components when de-energized. These tests verify that the ESFAS is capable of performing its intended function, from bistable input through the actuated components. SR 3.3.5.2 is addressed in LCD 3.3.5. SR 3.3.6.1 includes Matrix Logic tests and trip path (Initiation Logic) tests. Matrix Logic Tests, These tests are performed one matrix at a time. They verify that a coincidence in the two input channels for each function removes power to the matrix relays. During testing, power is applied to the matrix relay test coils, preventing the matrix relay contacts from assuming their energized state. The Matrix Logic tests will detect any short circuits around the bistable contacts in the coincidence logic, such as may be caused by faulty bistable relay or trip channel bypass contacts. Trip Path (Initiation Logic) Tests These tests are similar to the Matrix Logic tests, except that test power is withheld from one matrix relay at a time, allowing the initiation circuit to de-energize, opening one contact in each Actuation Logic channel. The initiation circuit lockout relay must be reset (except for EFAS, which lacks initiation circuit lockout relays) prior to testing the other three initiation circuits, or an ESFAS actuation may result. (continued) CEOG STS B 3.3-126 Rev 1, 04/07/95

ESFAS Logic and Manual Trip (Digital) B 3.3.6 To! TF~oS; IP~.:;' BASES { SURVEILLANCE SR 3.3.6.3 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on the manual ESFAS actuation circuitry, de-energizing relays and providin9L manual actuation of the function. < "'-{Ir1se,..../-ii] This test verifies that the trip push buttons are capable of opening contacts in the Actuation Logic as designed. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every [18J months. REFERENCES 1. FSAR, Section [7.3].

2. CEN-327, May 1986, including Supplement 1, March 1989.

3. CEN-403.

CEOG STS B 3.3-128 Rev 1, 04/07/95

DG--LOVS (Digital) B 3.3.7

                                                              -r:s .,-p. IJ rb BASES                                                            ~Q..v~

SURVEILLANCE SR 3.3.7.1 (continued) REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of channel instrument uncertainties, including indication and readability. If the channels are within the criteria, it is an indication that the channels are OPERABLE. The Frequency, about once every shift, is based upon operating experience that demonstrates channel failure is rare. Since the probability of two random failures*in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. SR 3.3.7.2 A CHANNEL FUNCTIONAL TEST is performed every [92] days to ensure that the entire channel will perform its i ntended /~~ function when needed. \~ The Frequency of [92] days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any [92] day Frequency is a rare event. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [6]. SR 3.3.7.3 SR 3.3.7.3 is the performance of a CHANNEL CALIBRATION every [18] months. The CHANNEL CALIBRATION verifies the accuracy of each component within the instrument channel. This inc 1udes can brat ion of the undervoltage relays and demonstrates that the equipment falls within the specified operating characteristics defined by the manufacturer. The (cont i nued) ""...~,:--: CEGG STS B 3.3-136 Rev 1. 04/07/95

CPIS (Digital) B 3.3.8

                                                                     ~-r1=':?lJr BASES                                                             ;(~5 SURVEILLANCE  SR 3.3.8.3    (continued)

REQUIREMENTS (Ref. 4). The Frequency of 92 days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 92 day Frequency is a rare event. . A Note to the SR indicates this Surveillance is applicable in MODES 1, 2, 3, and 4 only. SR 3.3.8.4 A CHANNEL FUNCTIONAL TEST is performed on the required containment radiation monitoring channel to ensure the entire channel will erform its intended function.~ e p01nts must be oun wit 1n t e owab e a ues specified inSR 3.3.8.4 and left consistent with the assumptions of the plant specific setpoint methodology (Ref. 4). The Frequency of 92 days is based on plant operating experience with regard to channel OPERABILITY and

 ~
   \               drift, which demonstrates that failure of more than one

.~/ channel of a given Function in any 92 day interval is a rare event. A Note to the SR indicates that this test is only applicable during CORE ALTERATIONS or during movement of irradiated fuel assemblies within containment. SR 3.3.8.5 Proper operation of the individual initiation relays is verified by actuating these relays during the CHANNEL FUNCTIONAL TEST of the Actuation Logic every [18] months.* This will actuate the Function, operating all associated equipment. Proper operation of the equipment actuated by each train is thus verified. The Frequency of [18] months is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function during any [18] month interval is a rare event. A Note to the SR indicates that this Surveillance includes verification of operation for each initiation relay. (continued) CEOG STS B 3.3-145 Rev 1, 04/07/95

CPIS (Digital) B 3.3.8 TS TF-;2. os-; ~. ..? BASES SURVEILLANCE SR 3.3.8.6 REQUIREMENTS (continued) CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [5]. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. SR 3.3.8.7 This Surveillance ensures that the train actuation response times are less than or equal to the maximum times assumed in the analyses. The [18] month Frequency is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. Testing of the final actuating devices, which make up the bulk of the response time, is included in the Surveillance. SR 3.3.8.8 Every [18] months, a CHANNEL FUNCTIONAL TEST is performed on the CPIS Manual Trip channel*f----CTVl~e~ J[) This test verifies that the trip push buttons are capable of opening contacts in the Actuation Logic as designed, de-energizing the initiation relays and providing manual actuation of the Function. The [18] month Frequency is based on the need to perform this Surveillance under the (continued) CEOG STS B 3.3-146 Rev 1, 04/07/95

CRIS (Digital) B 3.3.9

                                                                 -r s -;;;:= e:?fJ .r BASES                                                                 ~~.'.f SURVEILLANCE    SR 3.3.9.1    (continued)

REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. At this unit, the following administrative controls and design features (e.g., downscale alarms) immediately alert operations to loss of function in the nonredundant channels. [ At this unit, verification of sample system alignment and operation for gaseous, particulate, and iodine monitors is required as follows: ] SR 3.3.9.2 A CHANNEL FUNCTIONAL TEST is performed on the required control room radiation monitoring channel to ensure the entire channel wi 11 rform its intended function. ny se pOln adjustment shall e consistent with the assumptions of the current plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis. The requirements for this review are outlined in Reference [4]. The Frequency of [92] days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any [92] day interval is a rare event~ (continued) CEGG STS B 3.3-153 Rev 1, 04/07/95

CRIS (Digital) B 3.3.9 T.5rF-;2o~R~;; BASES SURVEILLANCE SR 3.3.9.3 REQUIREMENTS (continued) Proper operation of the individual initiation relays is verified by de-energizing these relays during the CHANNEL FUNCTIONAL TEST of the Actuation Logic every [18] months. This will actuate the Function, operating all associated equipment. Proper operation of the equipment actuated by each train is thus verified. The Frequency of [18] months is based on plant operating experience with regard to channel OPERABILITY,.which demonstrates that failure of more than one channel of a given Function in any [18] month interval is a rare event. Note 1 indicates this Surveillance includes verification of operation for each initiation relay. Note 2 indicates that relays that cannot be tested at power are excepted from the Surveillance Requirement while at power. These relays must, however, be tested during each entry into MODE 5 exceeding 24 hours unless they have been tested within the previous 6 months. SR 3.3.9.4 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances.* CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the surveillance interval extension analysis. The requirements for this review are outlined in Reference [4]. The Frequency is based upon the assumption or an [18] month calibration interval for the determination or the magnitude of equipment drift in the setpoint analysis. (continued) CEOG STS B 3.3-154 Rev 1, 04/07/95

CRrs (Digital) B 3.3.9 T5TF-20~A1w.~ BASES

 .SURVEILLANCE     SR '3.3.9.5 REQUIREMENTS (cont in.ued) Every [18] months, a CHANNEL FUNCTIONAL TEST is performed on the manual CRIS actuation circuitry. ~ IY15erl- ~ J This test verifies that the trip push buttons are capable of opening contacts in the Actuation Logic as designed, de-energizing the initiation relays and providing Manual Trip of the function. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every [18] months.

SR 3.3.9.6 This Surveillance ensures that the train actuation response times are less than the maximum times assumed in the analyses. The [18] month Frequency is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. Testing of the final actuating devices, which make up the bulk of the response time, is included in the Surveillance testing. REFERENCES 1. FSAR, Chapter [15].

2. "Plant Protection System Selection of Trip Setpoint Values."

3. 10 C~ 50, Appendix A, GDC 19.

4. [ ]

CEOG STS B 3.3-155 Rev 1, 04/07/95

FHIS (Digital) B 3.3.10

                                                                           ~rt=      ;;6r BASES   (continued)                                                     1<a.v3 SURVEILLANCE        SR 3.3.10 .1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between the two instrument cbannels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the.criteria, it may be an indication that the transmitter or the signal processing equipment has drifted

     \                     outside its limit.

The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. [ For this plant, the CHANNEL CHECK verification of sample system alignment and operation for gaseous, particulate, iodine, and gamma monitors is as follows: ] SR 3.3.10.2 A CHANNEL FUNCTIONAL TEST is performed on the required fuel building radiation monitoring channel to ensure the entire channel will perform its intended functi~Any setpoint .;.:i (continued) CEOG STS B 3.3-161 Rev 1, 04/07/95

FHIS (Digital) B 3.3.10 BASES T5 TF-;2DS; Rt!AJ .? 1 SURVEILLANCE SR 3.3.10.2 (continued) REQU IREMENTS

             ,adjustment shall be consistent with the assumptions of the current plant specific setpoint analysis.

The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis. The requirements for this review are outlined in Reference [4]. The Frequency of 92 days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 92 day Frequency is a rare event. SR 3.3.10.3 Proper operation of the individual initiation relays is verified by actuating these relays during the CHANNEL FUNCTIONAL TEST of the Actuation Logic every [18] months. This will actuate the Function, operating all associated equipment. Proper operation of the equipment actuated by t each train is thus verified. The Frequency of [18] months ~ is base on p ant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function during any [18] month Frequency is a rare event. A Note to the SR indicates that this Surveillance includes verification of operation for each initiation relay. [At this unit, the verification is conducted as-follows:] SR 3.3.10.4 Every 18 months, a CHANNEL FUNCTIONAL TEST is performed on the FHIS Manual Trip channel. ~ II'/~~"f A J This Surveillance verifies that the trip push buttons are capable of opening contacts in the Actuation logic as designed, de-energizing the initiation relays and providing Manual Trip of the Function. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the (continued) CEOG STS B 3.3-162 Rev 1, 04/07/95

Remote Shutdown System (Digital) B 3.3.12

                                                                 /S/t=;';() f . "

BASES "cz...., LJ. -3' ,/', r\ .'i

-------------------.:....-_-------..:.....--::=..

SURVEILLANCE SR 3.3.12.3 (continued) REQUIREMENT plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. SR 3.3.12.4 SR 3.3.12.4 is the performance of a CHANNEL FUNCTIONAL TEST every 18 months. This Surveillance should verify the OPERABILITY of the reactor trip circuit breaker (RT£B) open/closed indication on the remote shutdown panels by actuating the RTCSs. The Frequency of 18 months was chosen becau s cannot be exercised while the unit is at power. Operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A, GDC 19.

2. 10 CFR 50, Appendix R.

3. NRC Safety Evaluation Report (SER).

CEOG STS B 3.3-186 Rev 1, 04/07/95

[Logarithmic] Power Monitoring Channels (Digital) B 3.3.13 BASES

                                                                   --rSRlL.vg
                                                                        -n=. ~ 6 ~ '..

OJ SURVEILLANCE SR 3.3.13.1 (continued)

REQUIREMENTS continues to operate properly between each CHANNEL CALI BRATI ON. Agreement criteria are determined by the plant staff and should be based on a combination of the channel instrument uncertainties including control isolation, indication, and readability. If a channel is outside of the criteria, it may be an indication that the transmitter or the signal p~ocessing equipment has drifted outside of its limits. If the channels are within the criteria, it is an indication that the channels are OPERABLE. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of displays associated with the LCO required channels. SR 3.3.13.2 A CHANNEL FUNCTIONAL TEST is performed every [92] days to ensure that the entire channel is capable of properly indicating neutron flux. Internal test circuitry is used to feed preadjusted test signals into the preamplifier to verify channel alignment. It is not necessary to test the

                         ,                 ing a meaningful test signal is difficult; the detectors are of simple construction, and any failures in the detectors will be apparent as change in channel output. This Frequency is the same as that employed for the same channels in the other applicable MODES.

At this unit, the channel trip Functions tested by the ] [ CHANNEL FUNCTIONAL TEST are as follows: SR 3.3.13.3 SR 3.3.13.3 is the performance of a CHANNEL CALIBRATION. A CHANNEL CALIBRATION is performed every [18] months. The (continued) CEOG STS B 3.3-190 Rev 1, 04/07/95

LTOP System B 3.4.12 T5TF-2o~ ;P~.$ BASES SURVEILLANCE SR 3.4.12.5 (continued) REQUIREMENTS locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure event. The 72 hour Frequency considers operating experience with accidental movement of valves having remote control .and position indication capabilities available where easily monitored. These considerations include the administrative controls over main control room access and equipment control. SR 3.4.12.6 Performance of a CHANNEL FUNCTIONAL TEST is required every 31 days to verify and, as necessary, adjust the PORV open setpoints. The CHANNEL FUNCTIONAL TEST will verify on a monthly basis that the PORV lift setpoints are within the LCO limit.p PORV actuation could depressurize the RCS and is not required. The 31 day Frequency considers experience with equipment reliability. , ........ A Note has been added indicating this SR is required to be performed [12] hours after decreasing RCS cold leg temperature to S [285]OF. The test cannot be performed until the RCS is in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within 12 hours after entering the LTOP MODES. SR 3.4.12.7 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required every [18] months to adjust the whole channel so that it responds and the valve.opens within the required LTOP range and with accuracy to known input. The [18] month Frequency considers operating experience with equipment reliability and matches the typical refueling outage schedule.

(continued) CEOG STS B 3.4-66 Rev 1, 04/07/95

RCS Leakage Detection Instrumentation B 3.4.15 r,pry:::. 0/ 0 J BASES I?~g ACTIONS E.1 and E.2 (continued) power conditions in an orderly manner and without challenging plant systems. If all required monitors are inoperable, no automatic means of monitoring leakage are available and immediate plant

             *shutdown in accordance with LCO 3.0.3 is required ..

SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitors. The check gives reasonable confidence the channel is operating properly. The Frequency of [12] hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.15.2 SR 3.4.15.2 requires the performance of a CHANNEL FUNCTIONAL TEST of the required containment atmosphere radioactivity monitors. The test ensures that the monitor can perform its function in the desired manner. The test verifies the alarm setpoint and relative accuracy of the instrument string.~ The Frequency of 92 days considers instrument reliabilit~~ ) and operating experience has shown it proper fordetecti~

   ;:..~     degradation.        __- - - - - - - - - - - - - - - -
   ~----------

SR 3.4.15.3. SR 3.4.15.4, [and SR 3.4.15.51 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of [18] months is a typical refueling cycle and considers channel reliability. Operating experience has shown this Frequency is acceptable. (continued) CEOG STS B 3.4-86 Rev 1, 04/07/95

RCS Loops--Test Exceptions B 3.4.17

                                                                  ~rPtR6r BASES (continued)                                                      ~~.3 SURVEILLANCE      SR 3.4.17.1 REQUIREMENTS THERMAL POWER must be verified to be within limits once per hour to ensure that the fuel design criteria are not violated during the performance of the PHYSICS TESTS. The hourly Frequency has been shown by operating practice to be sufficient to regularly assess conditions for potential degradation and verify operation is within the LCO limits.

Plant operations are conducted slowly during the performance of PHYSICS TESTS, and monitoring the power level once per hour is sufficient to ensure that the power level does not exceed the limit. SR 3.4.17.2 Within 12 hours of initiating startup or PHYSICS TESTS, a CHANNEL FUNCTIONAL TEST must be performed on each logarithmic power level and linear power level neutron flux monitoring channel to verify OPERABILITY and adjust setpoints to proper values. This will ensure that the Reactor Protection System is properly aligned to provide the required degree of core protection during startup or the erformance of the PHYSICS TESTS. The interval is adequate o ensure t at e appropriate equipment is OPERABLE prior to the tests to aid the monitoring and protection of the plant during these tests. REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50, Appendix A, GDC 1, 1988.

CEOG STS B 3.4-96 Rev 1, 04/07/95

TS TF-:J,oS-Definitions ~"v~S 1.1 1.0 USE AND APPLICATION 1.1 Definitions

    -------------------------------------NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Deffnition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times. AVERAGE PLANAR LINEAR The APLHG~ shall be applicable to a specific HEAT GENERATION RATE planar height and is equal to the sum of the (APLHGR) [LHGRs] [heat generation rate per unit length of fuel rod] for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle [at the height]. CHANNEL CALIBRATION 0.11 d fl.1J ,.t:,.Q"S' I'"", ",..,... a.. c 1'It::.." ".1 Q..I I"~i ~~ ..f:.rr c""..,., 1"'1 r.z-I ()PI::.-r<~B Ji-J -ry ~ CHANNEL CHECK (continued) BWR/4 STS 1.1-1 Rev 1, 04/07/95

1.1 Definitions CHANNEL CHECK status derived from independent instrument (continued) channels measuring the same parameter. CHANNEL FUNCTIONAL TEST o-t c./J df2."II'G~S I r,.., ~~ e.-ht::4,,,, <<./ r/2.j 6J~ 1".q..Q. W C/::::v,,, 2- J

  ",o£.~", 8Jt.ITV.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe positi on. CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) prOVides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuriesjgram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose (continued) BWRj4 STS 1.1-2 Rev 1, 04/07/95

1.1 Definitions (continued) LOGIC SYSTEM FUNCTIONAL TEST r AofLOGIC SYSTEM FUNCTIONAL TEST s all be a test a~1~Y4peiDlogic component1.*{i:e~! :11 . -_< re'f'llrQa..j:i~a;i;'iRi se.. !.e!!, '1"1' Y"li8- !ol,d

         ""~"r~+J.,t state lo9l8 elellle"t:3, e!e., of a logic circuit,

( __-4t.. from as .close to the sensor as practicable up to,

     ....pPiEJ'tJ'l81I..Jr'Y       but not including, the actuated device, to verify
       ---------------       ~j    OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MAXIMUM FRACTION The MFLPD shall be the largest value of the OF LIMITING fraction of limiting power density in the core. POWER DENSITY (MFLPD) The fraction of limiting power density shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type. MINIMUM CRITICAL- POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core [for each class of fuel]. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. MODE A MODE shall correspond to anyone inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s} and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s} are also capable of performing their related support function(s}. PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. (continued) BWRj4 STS 1.1-5 Rev 1, 04/07/95

BASES

                                                                                     .**:*::*~.~1~"~:~,,
                                                                                                         -i; SURVEILLANCE       SR 3.3.1.1.3     (continued)                             ~:
                                                                               . ".'                f REQUIREMENTS accurately reflects the required setpoint as a function of .                    '   .....

flow. Each flow signal from the respective flow unit must* be ~ 105% of the calibrated flow signal. If the flow unit signal is not within the limit, one required APRM that receives an input from the inoperable flow unit must be declared inoperable. The Frequency of 7 days is based on engineering jUdqment, operating experience, and the reliability of this instrumentation. SR 3 . 3 . 1. 1. 4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended functi~ Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM and APRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be . performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9). SR 3.3. 1. 1. 5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the

                .       d function A Frequency of 7 days provides an acceptable level of system average availability over the (continued)

BWR/4 STS B 3.3-27 Rev 1, 04/07/95

RPS Instrumentation B 3.3.1.1 T~ 7F-;;20S;~. 3 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued) REQUIREMENTS A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flu~ profile for appropriate representative input to the APRM System. The 1000 MWD/T Frequency is based on operating experience with LPRM sensitivity changes. SR 3.3.1.1.9 and SR 3.3.1.1.12 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be conslstent Wl the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1. 10 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be (continued) BWR/4 STS B 3.3-29 Rev 1, 04/07/95

SRM Instrumentation B 3.3.1.2

                                                             ;5 T~- ;2o~ I'&.J.3'-

BASES SURVEILLANCE SR 3.3.1.2.4 REQUIREMENTS {continued} This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient "source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate .. To accomplish this, the SR is modified by a Note that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical. The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours to 24 hours. SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is required in MODE 5, and the 7 day Frequency ensures that the channels are OPERABLE while core reactivity changes could be in progress. This Frequency is reasonable, based on operating experience and on other Surveillances {such as a CHANNEL CHECK}, that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4. Since core reactivity changes do not normally take place, the Frequency has been extended from 7 days to 31 days. The 31 day Frequency is based on operating experience and on other Surveillances {such as {continued} BWR/4 STS B 3.3-41 Rev 1, 04/07/95

Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

                                                                  -n T r -;Los-, iP~I.*;

SURVEILLANCE Reviewer's Note: Certain Frequencies are based on approved ] REQUIREMENTS topical reports. In order for a licensee to use these Frequencies, the licensee must justify the Frequencies as [ required by the staff SER for the topical report. As noted at the beginning of the SRs, the SRs for each Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1. The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary. SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control Multiplexing System inpu~ Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analyses (Ref. 8). SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with (continued) BWRj4 STS B 3.3-51 Rev 1, 04/07/95

Control Rod Block Instrumentation B 3.3.2.1 DTF-2o~ ~..3 BASES SURVEILLANCE SR 3.3.2.1.5 (continued) REQUIREMENTS setpoint must be verified periodically to be ~ [10]% ~rp. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Frequency is based on the trip setpoint methodology utilized for the low power setpoint channel. SR 3.3.2. 1. 6 A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch--Shutdown Position Function to ensure that the entire channel will perform the intended function. The CrJnl'I'~~ FUNCTIONAL TEST for the Reactor Mode Switch--Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs. As noted in the SR, the Surveillance is not required to be performed until 1 hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 18 month Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3 . 2. 1. 7 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel (continued) BWRj4 STS B 3.3-53 Rev 1, 04/07/95

J: l;.:'

                                                                                              . :~:/~;i;** :,'

Feedwater and Main Turbine High lIater Level Trip Instrumentation B3 3 2 2

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BASES .

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SURVEILLANCE SR 3.3.2.2.1 (continued) REQUIREMENTS indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to -verifying.the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are detenlined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limits. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays. associated with the ch~nnels required by the LCO. SR 3.3.2.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the

                . nded function. Any setpoint adjustment shall be consistent with the assumptions of the current plant speci fj c setpo i nt methodo logy . ~

The Frequency of 92 days is based on reliability analysis (Ref. 2). SR 3.3.2.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the chann~l adjusted to account for instrument drifts between succeSSlve (continued) BWRj4 SiS 8 3.3-61 Rev 1, 04/07/95

EOC-RPT Instrumentation B 3.3.4.1 TE. T F- ;l Os-, 7tJ4,"p BASES SURVEILLANCE analysis demonstrated that the 6 hour testing allowance REQUIREMENTS does not significantly reduce the probability that the . (continued) recirculation pumps will trip when necessary. SR 3 .3 .4. 1. 1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the

                .               tion. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The Frequency of 92 days is based on reliability analysis of Reference 5. SR 3 .3 .4. 1. 2 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in SR 3.3.4.1.3. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on assumptions of the reliability analysis (Ref. 5) and on the methodology included in the determination of the trip setpaint. SR 3.3.4.1. 3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. (continued) BWRj4 STS B 3.3-87 Rev 1, 04/07/95

                                                                                                ;:~;t;~:*
                                                                                          . *.:;.. A%~£*

ATWS-RPT .Instrumentation '. "::;'i~'~;: B 3.3.4.2. ,:.:~';:;~j;~. BASES 7"""~'~-7"'~-="i~:~:::~C>~~~1.~ 1'\ ~ .., ";J": ..* ;'Z'!l.~':'I"...."",:,

-SU-R-V-EI-L-LAN-C-E---S-R-3-.-3-.4 -.-2-.I-(-C-o-nt-i-n-Ue-d-)------------*- -~;~~

REQUIREMENTS something even more serious. A CHANNEL CHECK will detect ..*~::;1S;~:, gross channel fa il ure; thus, ; tis key to veri fyi ng the ..

                      ~~~~~~eg~~U~T~g~~inues to operate properly between each                 . *..)P.:;f.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument 'uncertainties, . ."' .~.~ including indication and readability. If a channel ~is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCD. SR 3.3.4.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 2. SR 3.3.4.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in SR 3.3.4.2.4. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the (continued) BWR/4 STS B 3.3-98 Rev 1, 04/07/95

                                                                                                  -.j.
                                                                                                   ""V Eees Instrumentation       -.:":.~1*~~:
                                                                                          "~~~~"~
                                                                 -rsB 3.3. 5 *1  "" "

ri= J:Jbr" "~:'~'::if~.~ _M_S_~~~~~~~~~~~~~~~~~~~~~~~~~~~_~~cc*'*:~ . SURVEILLANCE SR 3.3.5."1.1 ..' ':f:gi;t;:::' REQUIREMENTS , :':'l:?;:~:'~' (continued) . Performance of the CHANNEL CHECK once every 12 hours ensures -.' : . :~..:>..;

                                                                                     ; ... :.:~: ~:l_:

that a gross failure of instrumentation has not occurred. A. CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read '- approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited to 12 hours; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3 . 5. 1. 2 A CHANNEL FUNCTIONAL TEST is performed an each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analyses of Reference 5. (continued) BWR/4 SiS B 3.3-136 Rev 1, 04/07/95

                                                                                               ~    '.    '  .: .

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RCIC System Instrumentation ._,"':;:)::.<~~~

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I BASES SURVEILLANCE SR 3.3.5.2.1 (continued) REQUIREMENTS someth i ng even more sed ous

A CHANNEL CHECK wi 11 detect. .-'. ':\~f~~~

gross channel failure; thus, it is key to verifying the . :.f"!-.:/:S~ instrumentation continues to operate properly between each - ~~'~:-::j' CHANNEL CALIBRATION. . '. Agreement criteria are determined by the plant staff based on. a combination of the channel instrument uncertainties, including indication and readability. If" a channel ,is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodo1ogy ~ The Frequency of 92 days is based on the reliability analysis of Reference 1. SR 3.3.5.2.3 The calibration of trip units provides a check of the actual trip setpoints; The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.2-1. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of t~e plant safety analysis. Under these conditions, the setpolnt (continued) BI;lR/4 SiS B 3.3-149 Rev 1, 04/07/95

                                                                                       .._e.:.... *
                                                                                       ~
                                                                                      '.         ~

Primary Containment Isolation SURVEILLANCE SR 3.3.6.1.1 (continued) REQU I REMENTS CHANNEL CHECK is normally a comparison of the parameter inaicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of .'..

*. ' :.':,~

excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to ope~ate properly between each . CHANNEL CALIBRATION. . Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCD. SR 3.3.6.1.2 and SR 3.3.6,1.5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function, Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodo logy" The 92 day Frequency of SR 3.3.6.1.2 is based on the reliability analysis described in References 6 and 7. The 184 day Frequency of SR 3.3.6.1.5 is based on engineering judgment and the reliability of the components (time delay relays exhibit minimal drift). (continued) BWRj4 SiS B 3.3-181 Rev 1,' 04/07/95

                                                                                                        '."~
                                                                                             *,~I Secondary Containment Isolation Instrumentati on';' >'*~*;<~~1*

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-SU-R-V-EI-L-LAN-C-E---SR--3-.3-.-6-.2-.-1-(-C-on-t-i-nu-e-d-)--.,---------------.:-~~

REQUIREMENTS channe1s. It is based on th~ assumpt i on that i nstrument**:\*;:~:S{ channels monitoring the same parameter should read approximately the same value. Significant deviations

                   . between the instrument channels could be an indication of
                   . excess i ve i nstrul11ent dri ft in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to~verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCD. SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of References 5 and 6. SR 3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inope~able if the trip setting is discovered to be less conservatlve than the Allowable Value specified in Table 3.3.6.2-1. If the trip setting is discovered to be less conservative than (continued) BWRj4 SiS B 3.3-194 Rev 1, 04/07/95

BASES SURVEILLANCE SR 3.3.6.3.2. SR 3.3.6.3.3, and SR 3.3.6.3.4 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent Wl the assumptions of the current plant specific setpoint methodology. , The 92 day Frequency is based on the. reli~pility analysis of Reference 3. A portion of the SjRV tailpipe pressure switch instrument channels are located inside the primary containment. The Note for SR 3.3.6.3.3, "Only required to be performed prior to entering MODE 2 during each scheduled outage> 72 hours when entry is made into primary containment," is based on the location of these instruments, ALARA considerations, and compatibility with the Completion Time of the associated Required Action (Required Action B.1). SR 3.3.6.3.5 The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint methodology. The Frequency of every 92 days for SR 3.3.6.3.5 is based on the reliability analysis of Reference 3. SR 3.3.6.3.6 CHANNEL CALIBRATION is a complete check of the instrument loop and sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive (continued) BWR/4 SiS B 3.3-205 Rev 1, 04/07/95

BASES SURVEILLANCE SR 3.3.7.1.1 (continued) REQUIREMENTS outside the criteria, it may be an indication that the. instrument has drifted outside its limit. ;.' .

                                                                             .>. :~~.~~.~

The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent,checks of channel status during normal operational use of the displays associated with channels required by the LCO. , ". SR 3 . 3 . 7 . 1. 2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodoloqy. The Frequency of 92 days is based on the reliability analyses of References 5 and 6. SR 3.3.7.1.3 The calibration of trip units provides a check of the actual trip setpoints. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.7.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analyses of References 5 and 6. (continued) 8WR/4 SiS 83.3-217 Rev 1, 04/07/95

BASES

 .........                                                                 .- '~". ~; -'

Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of.xne parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with channels required by the LCD. SR 3 .3 .8. 1. 2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent Wl the assumptions of the current plant specific setpoint methodology .. The Frequency of 31 days is based on operating experience with reaard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. (continued) BWR/4 SiS Rev 1, 04/07/95

RPS Electric Power Monitoring B 3.3.8.2 TSTF-;2o~ tPe<J l :! BASES ACTIONS 0.1. 0.2.1. and 0.2.2 (continued) In addition, action must be immediately initiated to either restore one electric power monitoring assembly to OPERABLE status for the inservice power source supplying the required instrumentation powered from the RPS bus (Required Action D.2.1) or to isolate the RHR Shutdown Cooling System (Required Action 0.2.2). Required Action D.2.1 is provided because the RHR Shutdown Cooling System may be needed to provide core cooling. All actions must continue until the applicable Required Actions are completed. SURVEILLANCE SR 3.3.8.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the entire channel will perform the intended function. ny setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is only required to be performed while the plant is in a condition in which the loss of the RPS bus will not jeopardize steady state power operation (the design of the system is such that the power source must be removed from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance. The 184 day Frequency and the Note in the Surveillance are based on guidance provided in Generic Letter 91-09 (Ref. 2). SR 3.3.8.2.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. (continued) BWR/4 STS B 3.3-232 Rev 1, 04/07/95

RCS Leakage Detection Instrumentation B 3.4.6 T5 T F - ).. 0 S; ri'eu. .;' BASES (continued) SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR is for the performance of a CHANNEL CHECK of the required primary containment atmospheric monitoring system. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.6.2 This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their function in the desired manner. The test also verifies the alarm. setpoint and relative accuracy of the instrument string. The Frequency of 31 days considers instrument reliability, and operating experience has shown it proper for detecting degradation. SR 3.4.6.3 This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of [18] months is a typical refueling cycle and considers channel reliability. Operating experience has proven this Frequency is acceptable. REFERENCES 1. 10 CFR 50, Appendix A, GDC 30.

2. Regulatory Guide 1.45, May 1973.

3. FSAR, Section [5.2.7.2.1].

4. GEAP-5620, April 1968.

5. NUREG-75/067, October 1975.

6. FSAR, Section [5.2.7.5.2].

BWR/4 STS B 3.4-32 Rev 1, 04/07/95

Refuel Position One-Rod-Out Interlock B 3.9.2 TSTF- ;2OSi Rev,s BASES ACTIONS A.l and A.2 (continued) fuel assemblies. Action must continue until all such control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted. SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation. The Frequency of 12 hours is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation. SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into

 )- +A

,t- ~SP~ the logic. The CHANNEL FUNCTIONAL TEST may be performed by J ----:l'1rfV"<:s'Oe~ri"ilei(s;-'of sequential, overlapping, or total channel steps so that the entire channel is tested. The 7 day Frequency is considered adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert the operator to control rods not fully inserted. To perform the required testing, the applicable condition must be entered (i.e., a control rod (continued) BWR/4 STS 8 3.9-7 Rev 1, 04/07/95

TSTF'-:<o~ Definitions ;?w S 1.1 1.0 USE AND APPLICATION 1.1 Definitions

  --~----------------------------------NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable througho'lt these Technical Specifications and Bases. Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

 ~VERAGE  PLANAR LINEAR        The APLHGR shall be applicable to a specific HEAT GENERATION RATE          planar height and is equal to the sum of the (APLHGR)         .            [LHGRs] [heat generation rate per unit length of fuel rod] for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle [at the height].

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds withln the necessary range and accuracy to known values of the parameter that the channel 0.11 d~lca.r /:, -11." moni EL CALIBRATION shall encom ass cAc..nl')~1 ~lJ~r.A a rPftJ;, n 1~. 1n ~ IJ nr?

                                 ~~~rtfie CHANNEL FUNCTIONAL TEST. Calibration
~ C Itt:.."" a.. I           ~o~l~n~s~trument channels with resistance temperature detector (RTD) or thermocouple sensors may consist OPIi7e~QI'-Jry ~JJ.            of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total
                              ~/ll2l1l11TYf/l!l CHANNEL CHECK                 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or (continued)

BWR/6 STS 1.1-1 Rev 1, 04/07/95

TSTF-;}oS Definitions -r(~3 1.1 1.1 Definitions CHANNEL CHECK status derived from independent instrument (continued) channels measuring the same parameter. CHANNEL FUNCTIONAL TEST of:' 0..// dQ.~ ~ t;.Q.r" I I~ ~~ a..~al1l1~ /

  . ~e,,~ "R~ ~,..

CAe:.iIlr'\ Q. J

      ~~~,q J9JL IT )0':

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose (continued) BWR/6 STS 1.1-2 Rev 1, 04/07/95

ToS TF-20~" . : Definitions f{e".,;~' 1.1 1.1 Definitions (continued) LOGIC SYSTEM FUNCTIONA A LOGIC SYSTEM FUNCTIONAL TEST s 11 be a test :_:i TEST of all required logic components (1 i , .~~ leq~;le8 pslays aAQ .BAiaiii, i~;~ YA;i'; &slii

     '7'~,,:rdl. ~               ~a!e legis elemsAis, iis,) of a logic circuit, from as close to the sensor as practicable up to, O,oe:T< J4 BJJ. rr't'       but not including, the actuated deVice, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MAXIMUM FRACTION The MFLPD shall be the largest value of the OF LIMITING fraction of limiting power density in the core. POWER DENSITY (MFLPD) The fraction of limiting power density shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type. MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core [for each class of fuel]. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. MODE A MODE shall correspond to anyone inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in fable 1.1-1 with fuel in the reactor vessel. OPERABLE -OPERAB ILITY A system, subsystem, division, component~ or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified . safety function(s) are also capable of performlng their related support function(s). (continued) BWR/6 STS 1.1-5 Rev 1, 04/07/95

RPS BASES SURVEILLANCE SR 3.3.1.1.3 (continued) REQUIREMENTS the total loop drive flow signals from the flow unit used to vary the setpoint are appropriately compared to a calibrated flow signal and therefore the APRM Function accurately ... ~~., reflects the required setpoint as a function of flow. Each flow signal from the respective flow unit must be 5 105% of the calibrated flow signal. If the flow unit signal is not within the 1imit, the APRMs that receive an- input from the inoperable flow unit must be declared inoperable. The Frequency of 7 days is based on engineering judgment, operating experience, and the reliability of this instrumentation. SR 3.3 . 1. 1. 4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel wi?l perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodo logy .. As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1 since testing of the MODE 2 required IRM and APRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9). SR 3 .3 . 1. 1. 5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the (continued) B~R/6 SiS Rev 1, 04/07/95

RPS Instrumentation B 3.3.1.1 nTF-2~ /r'ec..l'j' BASES SURVEILLANCE SR 3.3.1.1.5 {ContinUed)~Y)$e."'+ B] REQUIREMENTS intended Function.~ Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of Reference 9. (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.) SR 3.3.1.1.6 and SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status. The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a region without adequate neutron flux indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned from the SRMs to the IRMs. The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (initiate a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs concurrently have onscale readings such that the transition between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block. As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2). If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate (continued) BWR/6 STS B 3.3-27 Rev 1, 04/07/95

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued) REQUIREMENTS channel(s) that are required in the current MODE or condition should be declared inoperable. A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. SR 3 .3 . 1. 1. 8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 MWD/T Frequency is based on operating experience with LPRM sensitivity changes. SR 3.3.1.1.9 and SR 3.3.1.1.12 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be c en Wl e assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1.10 The calibration of trip units provides a checK of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but (continued) BWR/6 STS B 3.3-28 Rev 1, 04/07/95

SRM Instrumentation B 3.3.1.2 TSTF-;2.oS; Rev=-S BASES SURVEILLANCE SR 3.3. 1. 2.4 REQUIREMENTS (continued) This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate. 'rhis ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient nsourcen material, in the form of irradiated fuel assemblies, to establish the minimum count rate. ' To accomplish this, the SR is modified by a Note that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated quadrant, even with a control rod withdrawn the configuration will not be critical. The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours to 24 hours. SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is required in MODE 5, and the 7 day Frequency ensures that the channels are OPERABLE while core reactivity changes could be in progress. This 7 day Frequency is reasonable, based on operating experience and on other Surveillances (such as a CHANNEL CHECK) that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below and in MODES 3 and 4. Since core reactivity changes do not normally take place, the Frequency has been extended from 7 days to 31 days. The 31 day Frequency is based on operating experience and on other Surveillances (such as (continued) J BWR/6 STS B 3.3-40 Rev 1, 04/07/95

Control Rod Block Instrumentation B 3.3.2.1 T~ TF - ;2o~ tfJ(;v . .3' BASES SURVEILLANCE associated Conditions and Required Actions may be delayed REQUIREMENTS for up to 6 hours, provided the associated Function (continued) maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 8) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary. SR 3.3.2.1.1. SR 3.3.2.1.2. SR 3.3.2.1.3. and SR 3.3 .2. 1. 4 The CHANNEL FUNCTIONAL TESTS for the RPC and RWL are performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying that a control rod block occurs. Any setpoint adjustment shall be conS1S en Wl e assumptions of the current plant specific setpoint methodology. As noted, the SRs are not required to be performed until I hour after specified conditions are met (e.g., after any control rod is withdrawn in MODE 2). This allows entry into the appropriate conditions needed to perform the required SRs. The Frequencies are based on reliability analysis (Ref. 7). SR 3.3.2.1.5 The LPSP is the point at which the RPCS makes the transition between the function of the RPC and the RWL. This transition point is automatically varied as a function of power. This power level is inferred from the first stage turbine pressure (one channel to each trip system). These power setpoints must be verified periodically to be within the Allowable Values. If any LPSP is nonconservative, then the affected Functions are considered inoperable. Since this channel has both upper and lower required limits, it is not allowed to be placed in a condition to enable either the RPC or RWL Function. Because main turbine bypass steam flow can affect the LPSP nonconservatively for the RWL, the RWL is considered inoperable with any main turbine bypass valves (continued) BWR/6 STS B 3.3-49 Rev 1, 04/07/95

Control Rod Block Instrumentation B 3.3.2.1

                                                         ~Tr-;LoS; ttL.S BASES SURVEILLANCE REQUIREMENTS SR   3.3.2.1.8   (continued)                bL   VlS ""+ A} \

withdraw any control rod with the reactor mode switch in th~ shutdown position and verifying a control rod block occurs. As noted in the SR, the Surveillance is not required to be performed until 1 hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable limits. This allows entry into MODES 3 and 4 if,the 18 month Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when perfonled at the 18 month Frequency. SR 3.3.2.1. 9 LCO 3.1.3 and LCD 3.1.6 may require individual control rods to be bypassed in RACS to allow insertion of an inoperable control rod or correction of a control rod pattern not in compliance with BPWS. With the control rods bypassed in the RACS, the RPC will not control the movement of these bypassed control rods. To ensure the proper bypassing and movement of those affected control rods, a second licensed operator or other qualified member of the technical staff must verify the bypassing and movement of these control rods. Compliance with this SR allows the RPC to be OPERABLE with these control rods bypassed. REFERENCES 1. FSAR, Section [7.6.1.7.3].

2. FSAR, Section [15.4.2].

(continued) BWR/6 STS B 3.3-51 Rev 1, 04/07/95

                                                                                       ", :~:,' :~~;;::i~'{

EOC-RPT Instrumentation** ;./ ;,:*~t~-&J:

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B 3.3.4.1 ......"~'}".'t.~ 7 ~7 '::::<AC ..,~~:::,,~) BASES ~ -5' . (.,/.-/.. . ACTIONS C.1 and C.2 (continued) .-\ ...\<.. experience~ to reduce THERMAL POWER to < 40% RTP from full power conditions in an orderly manner and without challenging plant systems. .: :' .;: :' tt.;~~..* SURVEILLANCE Reviewer's Note: Certain Frequencies are based on approved] REQUIREMENTS topical reports. In order for a licensee to use these [ Frequencies, the licensee must justify the Frequencies as required by the staff SER for the topical report. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains EOC-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary. SR 3.3.4.1. 1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended f ction. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology .. The Frequency of 92 days is based on reliability analysis (Ref. 5). SR 3 .3 .4. 1. 2 The calibration of trip units provides a check of the act~al trip setpoints. The channel must be declared inoperable 1f the setting is discovered to be less conservative than the (continued) BWR/6 SiS B 3.3-78 Rev 1, 04/0i/95

BASES SURVEILLANCE SR 3.3.4.2.1 (continued) ...* ~*~1***tr:*** REQUIREMENTS ~*:**~Mj instrumentation continues to operate properly between each CHANNEL CALIBRATION. . ;;"~!f~:';.:

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Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a 'channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO. SR 3.3.4.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodo logy. , The Frequency of 92 days is based on the reliability analysis of Reference 2. SR 3.3.4.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in SR 3.3.4.2.4. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of t~e plant safety analysis. Under these conditions, the setpolnt must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. (continued) BWR/6 STS B 3.3-89 Rev 1, 04/07/95

ECCS Instrumentation 8 3.3.5.1 7~' ~~{Jr BASES 1('SI-'"' S : SURVEILlANCE SR 3 .3 . 5 . 1. 1 REQUIREMENTS

 *.(continued.)  Performance of the CHANNEL CHECK once every 12 hours ensures that.a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the 'channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3 . 5. 1. 2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function~ Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analyses of Reference 4. (continued) BWRj6 SiS 8 3.3-126 Rev 1, 04/07/95

RCIC System Instrumentation B 3.3.5.2

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SURVEILLANCE SR 3.3.5.2.1 (continued) .

t.* ....

REQUIREMENTS something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on* a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with the channels required by the LCO. SR 3,3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant spec ifi c setpo i nt methodology., The Frequency of 92 days is based on the reliability analysis of Reference 1. SR 3.3.5.2.3 The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be re-adjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. (continued) BWR/6 STS B 3.3-139 Rev 1, 04/07/95

Primary Containment Isolation Instrumentation B 3.3.6.1 7~ 71= 2.o~ ...... BASES !RQ;t,i;$ . r

SURVEILLANCE SR 3.3.6.1.1 (continued) REQUIREMENTS The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCD. SR 3 . 3 . 6 . 1. 2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency is based on reliability analysis described in References 5 and 6. SR 3.3 .6. 1. 3 The calibration of trip units consists of a test to provide a check of the actual trip setpoints. The channel must be" declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of References 5 and 6. (continued) BWR/6 STS B 3.3-174 Rev 1, 04/07/95

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES

                                                            /5 rF-20~ fev,1 SURVEILLANCE   This Note is based on the reliability analysis (Refs. 3 REQUIREMENTS   and 4) assumption of the average time required to perform (continued)  channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and the SGT System will initiate when necessary.

SR 3.3.6.2.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the indicated parameter for one instrument channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. (continued) BWR/6 STS B 3.3-186 Rev 1, 04/07/95

RHR Containment Spray System Instrumentation B 3.3.6.3

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BASES ~Q.Y..? *z SURVEILLANCE SR 3.3.6.3.2 REQUIREMENTS

  "(continued)   A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure the entire channel will perform the intended function.
~~~        I-----------J A            Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The Frequency of 92 days is based upon the reliability analysis of Reference 3. SR 3.3.6.3.3 The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.3-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based upon the reliability analysis of Reference 3. SR 3.3.6.3.4 and SR 3.3.6.3.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency of SR 3.3.6.3.4 is based on the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpaint analysis. (continued) BWR/6 SiS B 3.3-199 Rev 1, 04/07/95

SPMU System Instrumentation B 3.3.6.4

                                                             -{s-rt= :3. I) S BASES                                                           I<Q,.~  :!

SURVEILLANCE SR 3.3.6.4.1 (continued) REQUIREMENTS something even more serious. A CHANNEL CHECK will detect gross channel failure; thus it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel js outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of the LCD. SR 3.3.6.4.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodo logy .. The Frequency of 92 days is based on the reliability analysis of Reference 3. SR 3.3.6.4.3 The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.4-1. If the trip setting is discovered to be less conservative than . accounted for in the appropriate setpoint methodology but 1S not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to (continued) BWRj6 SiS 5 3.3-211 Rev 1, 04 /07/95

Relief and LlS Instrumentation B 3.3.6.5

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BASES Ra." j .t ACTIONS B.1 and B. 2 (continued) If the inoperable trip system is not restored to OPERABLE status within 7 days, per Condition A, or if two trip systems are inoperable, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE Reviewer's Note: Certain Frequencies are based on approved] REQUIREMENTS topical reports. In order for a licensee to use these [ Frequencies, the licensee must justify the Frequencies as required by the staff SER for the topical report. The Surveillances are modified by a Note to indicate that when a channel ;s placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains relief or LLS initiation capability, as applicable. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption of the average time required to perform channel surveillance. That analysis demonstrated the 6 hour testing allowance does not significantly reduce the probability that the relief and LLS valves will initiate when necessary. SR 3.3.6.5.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. (continued) BWR/6 SiS B 3.3-218 Rev 1, 04/07/95

CRFA System Instrumentation B 3.3.7.1

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BASES £~1oI f SURVEILLANCE SR 3.3.7.1.1 (continued) REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties including indication and readability. If a channel is ' outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO. SR 3 .3 . 7. 1. 2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intende function. Any setpoint adjustment shall be consistent with t e assumptions of the current plant

             !pecific setpoint methodology.

The Frequency of 92 days is based on the reliability analyses of References 4, 5, and 6. SR 3 .3 . 7. 1. 3 The calibration of trip units provides a check of the actual trip setpoints. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analyses of References 4, 5, and 6. (continued) BWR/6 SiS B 3.3-230 Rev 1, 04/07/95

LOP Instrumentation B 3.3.8.1

                                                                ,/S,"::2 os'         i BASES                                                             *1-11 .:;        j SURVEILlANCE        or expiration of the 2 hour allowance, the channel must be*-

REQUIREMENTS returned to OPERABLE status or the applicable Condition (contfnued) entered and Required Actions taken. SR 3.3.8.1. I Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of

               .' excessive instrument drift in one of the channel s or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of the LCD. SR 3 .3 .8. 1. 2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be conslstent with the assumptions of the current plant specific setpoint methodology. The Frequency of 31 days is based on plant operating. experience with regard to channel OPERABILITY and drlft that demonstrates that failure of more than one channel of a given Function in any 31 day interval is rare. (con t -j nued) BWR/6 STS B 3.3-238 Rev 1, 04/07/95

RPS Electric Power Monitoring B 3.3.8.2 T:; r F- ;20S; ,(Jev,;2 BASES ACTIONS 0.1. 0.2.1. and 0.2.2 (continued) In addition, action must be immediately initiated to either restore one electric power monitoring assembly to OPERABLE status for the inservice power source supplying the required instrumentation powered from the RPS bus (Required Action 0.2.1) or to isolate the RHR Shutdown Cooling System (Required Action-0.2.2). Required Action 0.2.1 is provided because the RHR Shutdown Cooling System may be needed to provide core cooling. All actions must continue until the applicable Required Actions are completed. SURVEILLANCE SR 3.3.8.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each overvo1tage, undervoltage, and underfrequency channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is only required to be performed while the plant is in a condition in which the loss of the RPS bus will not jeopardize steady state power operation (the design of the system is such that the power source must be removed from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of suffici~nt duration to allow for scheduling and proper performance of the Surveillance. The 184 day Frequency and the Note in the Surveillance are based on guidance provided in Generic Letter 91-09 (Ref. 2). SR 3.3.8.2.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. (continued) BWR/6 STS B 3.3-245 Rev 1, 04/07/95

RCS Leakage Detection Instrumentation B 3.4.7 BASES TSTF-..2os;tf'eu.S SURVEILLANCE SR 3.4.7.1 (continued) REQUIREM&NTS properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.7.2 This SR requires the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their function in the desired manner, The test also verifies the alarm setpoint and relative accuracy of the instrument string. The Fre uency of 31 da s considers i strument reliability, an opera 1ng experience as shown it proper for detecting degradation. SR 3.4.7.3 This SR requires the performance of a CHANNEL CALIBRATION of the required ReS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside the drywell. The Frequency of [18] months is a typical refueling cycle and considers channel reliability. Operating experience has proven this Frequency is acceptable. REFERENCES l. 10 CFR 50, Appendix A, GDC 30.

2. Regulatory Guide 1.45, May 1973.

3. FSAR, Section [5.2.5.2].

4. GEAP-5620, April 1968.

5. NUREG-75/067, October 1975.

6. FSAR, Section [5.2.5.5.3].

BWR/6 STS B 3.4-37 Rev 1, 04/07/95

Refueling Equipment Interlocks B 3.9.1 BASES IS T F,.:L 0 s: IP~~.~ LCO blocks to prevent operations that could result in (continued) criticality during refueling operations. APPLICABILITY In MODE 5, a prompt reactivity excursion could cause fuel damage and subsequent release of radioactive material to the environment. The refueling equipment interlocks protect against prompt reactivity excursions during MODE 5. The interlocks are only required to be OPERABLE during in-vessel fuel movement with refueling equipment associated with the interlocks. ' In MODES 1, 2, 3, and 4, the reactor pressure vessel head is on, and no fuel loading activities are possible. Therefore, the refueling interlocks are not required to be OPERABLE in these MODES. ACTIONS With one or more of the required refueling equipment interlocks inoperable, the unit must be placed in a condition in which the LCO does not apply. In-vessel fuel movement with the affected refueling equipment must be immediately suspended. This action ensures that operations are not performed with equipment that would potentially not be blocked from unacceptable operations (e~g., loading fuel into a cell with a control rod withdrawn). Suspension of in-vessel fuel movement shall not preclude completion of movement of a component to a safe position. SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. T e FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. (continued) BWR/6 STS B 3.9-3 Rev 1, 04/07/95

Refuel Position One-Rod-Out Interlock B 3.9.2 T~ TF-;2os, ~cJI::5 BASES ACTIONS A.l and A.2 (continued) containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted. SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper . positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when requi red. By "1 ocki ng the reactor ll mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation. The Frequency of 12 hours is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation. SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual 1rv)~u1 Al signal indicative of a required condition is injected into the 10gic~The CHANNEL FUNCTIONAL TEST may be performed by any serles of sequential, overlapping, or total channel steps so that the entire channel is tested. The 7 day Frequency is considered adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert the operator of control rods not fully inserted. To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.1 has been modified by a Note that states the (continued) BWR/6 STS B 3.9-7 Rev 1, 04/07/95}}

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