ML18102A160

Application for Amends to Licenses DPR-70 & DPR-75,revising Rv Level Indication Sys Action Statements to Facilitate Actions Necessary for Channel Testing to Be Performed in Mode 3

ML18102A160
Person / Time
Site:	Salem
Issue date:	05/31/1996
From:	Storz L Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18102A161	List: ML18102A160 ML18102A162
References
LCR-S96-03, LCR-S96-3, LR-N96069, NUDOCS 9606110279
Download: ML18102A160 (17)

Text

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• Public Service I

Electric and Gas Company Louis F. Storz Public SeNice Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-5700 Senior Vice President - Nuclear Operations MAY 31 1996 LR-N96069 LCR S96-03 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS INSTRUMENTATION SYSTEM REQUIREMENTS SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 Gentlemen:

In accordance with 10CFR50.90, Public Service Electric & Gas (PSE&G) Company hereby requests.a revision to the Technical Specifications (TS) for the Salem Generating Station Unit Nos. 1 and 2. In accordance with 10CFR50.91 (b) (1), a copy of this submittal has been sent to the State of New Jersey.

The proposed TS changes contained herein represent changes to instrumentation specifications. These changes include: 1) revisions to Reactor Vessel Level Indication System (RVLIS)

Action Statements to facilitate actions necessary for channel testing to be performed in Mode 3, 2) revision to the Channel Calibration definition to better account for temperature detector channel calibration methodology and 3) deletion of a requirement to install a jumper in the Auxiliary Feedwater actuation logic since a design change will result in the jumper function being performed by a relay.

The proposed changes have been evaluated in accordance with 10CFR50.91(a) (1), using the criteria in 10CFR50.92(c), and PSE&G has concluded that this request involves no significant hazards consideration.

The basis for the requested change is provided in Attachment 1.

A 10CFR50.92 evaluation with a determination of no significant hazards consideration is provided in Attachment 2. The marked up Technical Specification pages affected by the proposed changes are provided in Attachment 3.

9606110279 960531 PDR ADOCK 05000272 p PDR 11>.. Printed on W Recycled Paper

Document Control Desk

• MAY 31 1996 LR-N96069 Based on a need to have this amendment approved prior to Unit 2 reaching Mode 3 for the RVLIS testing to be performed, PSE&G requests that this amendment be approved no later than July 15, 1996.

Upon NRC approval of this proposed change, PSE&G requests that the amendment be made effective on the date of issuance, but allow an implementation period of sixty days to provide sufficient time for completion of administrative activities associated with implementation of all of the changes included in this request.

Should you have any questions regarding this request, we will be pleased to discuss them with you.

Sincerely, Affidavit Attachments (3)

C Mr. T. T. Martin, Administrator - Region I U. S. Nuclear 'Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. Olshan, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. c. Marschall (X24)

USNRC Senior Resident Inspector - Salem Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625

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• REF: LR-N96069 LCR S96-03 STATE OF NEW JERSEY SS.

COUNTY OF SALEM L. F. Storz, being duly sworn according to law deposes and says:

I am Senior Vice President - Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Salem Generating Station, Units 1 and 2, are true to the best of my knowledge, information and belief.

Subscribed~nd Sworn to before me this 3\ 'S day of m~' 1996

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KIMBERLY JO BROWN NOTARY PUBLIC OF NEW JERSEY My Commission expires on ~~~~~M2y~C=om~m=is~sio~n~Ex~pi~res~A~pillri1~2~1.~rn~g~6~~-

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~ocument Control Desk Attachment 1

• LR-N96069 LCR S96-03 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-.311 CHANGE TO TECHNICAL SPECIFICATIONS INSTRUMENTATION SYSTEM REQUIREMENTS BASIS FOR REQUESTED CHANGE A. REACTOR VESSEL LEVEL INDICATING SYSTEM (RVLIS)

Requested Change and Purpose The current RVLIS Technical Specification (TS) action statements require shutdown in seven days if one channel is inoperable and shutdown in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if two channels are inoperable. This is being changed to require a special report if one channel is inoperable for thirty days or if two channels are inoperable for seven days. Unit shutdown will not be required for RVLIS inoperability.

To provide the necessary reporting requirements specified in the proposed action statements, a new administrative specification 6.9.4 is proposed. ITS specification 5.6.8 "Post Accident Monitoring (PAM) Report" was used as the model for the new reporting.requirements. The phrase "for inadequate core cooling" was added to the ITS paragraph to clarify what .moni tor*ing function is required, since only RVLIS is affected by the proposed change and RVLIS' function *is to monitor for inadequate core cooling.

These changes will permit data taking and normalization procedures to be implemented during the upcoming startup(s) as recommended by Westinghouse and will eliminate unnecessary shutdowns from RVLIS inoperability. The proposed change is needed to support the upcoming startup(s) since the planned evolutions cannot be completed within the specified time in the current action statement(s).

These changes are consistent with action statements and reporting requirements for RVLIS provided in the improved "Standard Technical Specifications for Westinghouse Plants" (ITS), NUREG 1431, Revision 1.

Background

RVLIS was upgraded during the Unit 1 and 2 refueling outages in 1992 and 1991 respectively. Amendment 95, for Salem Unit 2 was issued in September, 1990, to incorporate RVLIS into the TS and to provide action statements derived from Generic Letter 83-37.

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'Document Control Desk Attachment 1

• • LR-N96069 LCR S96-03 A similar amendment for Salem Unit 1, Amendment 117, was issued in February, 1991. Since the RVLIS was being upgraded, a temporary variance at each unit allowed both RVLIS channels to be inoperable as long as the required channels for Reactor Coolant System Subcooling Margin Monitor and the Core Exit Thermocouples were operable. This temporary variance has expired on both units.

RVLIS is part of the safety-related display instrumentation (UFSAR section 7.5). Its function is to display information for the operator "to enable him to perform required manual functions and to determine the effect of manual actions taken following a reactor trip due to operational occurrences or accident conditions discussed in Section 15." RVLIS performs no automatic functions designed to mitigate the consequences of any accident.

The RVLIS installed at Salem has three ranges of vessel level indication: 1) Dynamic head with any Reactor Coolant Pumps (RCPs) running, 2) Upper range with no RCPs running and 3) Full range for no RCPs running. The scheduled testing on RVLIS is to support the dynamic head indication. The other two indications are unaffected and are expected to read correctly.

The RVLIS dynamic head indication gives the control room operators vessel level indications when any of the RCPs are running. This indication is obtained by sensing the differential pressure (d/p) developed from the flow through the reactor core and also from the static head of the Reactor Coolant System (RCS) level in the core. At hot zero power, with four RCPs running, the RVLIS Dynamic Head should read 100%; indicating a water solid system. Any reduction in the d/p compared to the normal operating condition is an indication of voids in the vessel. If less than four RCPs are operating, the display would read a lower d/p. The correct d/p readings are determined by taking empirical data during plant heat up.

During the system readiness review, a review of vendor documentation revealed that certain data had not been obtained during initial system startup following upgrade. This data needs to be taken during the startup from the current outage(s).

Without this data, the dynamic head indications from the RVLIS may not have the desired accuracy. Westinghouse, the RVLIS vendor, has recommended that PSE&G normalize the RVLIS dynamic head indication to 100% at hot zero power conditions and take heat up data to determine plant specific dynamic head zero void setpoints.

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Attachment 1 *

• Document Control Desk

• LR-N96069 LCR S96-03 This data collection will start at the end of Mode 5 and continue through Mode 3. RVLIS is required to be operable in Modes 1, 2 and 3. Without the proposed change, entry into Mode 3 would be permissible, but TS Table 3.3-11 Action 2 would require return to Mode 4 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if RVLIS is inoperable. This evolution is expected to last more than a week after entering Mode 3 which is longer than the existing Allowed Outage Times (AOTs).

Justification of Requested Changes The existing Salem TSs for RVLIS were developed from Generic Letter 83-37. The proposed changes reflect a reduction in the relative significance provided in Generic Letter 83-37; however, this is consistent with the system's design basis as a post-accident indication system. No longer is it deemed necessary or appropriate to shutdown the plant due to RVLIS inoperability.

Instead, alternate monitoring for inadequate core cooling is used and a Special Report is forwarded to the NRC which outlines the preplanned alternate method of monitoring for inadequate core cooling, reasons for RVLIS inoperability and plans to restore RVLIS operability.

Since RVLIS does not initiate a transient nor would its unavailability cause the failure of any automatic accident mitigating system actuation the RVLIS is not directly modeled in the Salem Probabilistic Risk Assessment (PRA) . Therefore, RVLIS is considered to have an insignificant impact on the plant's Core Damage Frequency (CDF). Because of its low risk significance, the AOT should not be based on Salem specific impact on core damage.

These proposed AOTs and associated Special Reporting requirements are consistent with equivalent statements used for RVLIS in the ITS. The revised ITS was issued in April 1995. These ITS were developed over a number of years with input from industry, vendor and NRC personnel and represent current evaluations of the significance of components and systems to be retained in the TS.

The LCOs, actions and Surveillance Requirements in the ITS are consistent with the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132).

The type and number of channels that provide RVLIS at Salem Units 1 and 2 are similar to the RVLIS provided for the standard Westinghouse plant. Therefore, it is justifiable for Salem to utilize the AOTs provided for RVLIS in the ITS.

Action statements for Reactor Coolant System Subcooling Margin Monitor and the Core Exit Thermocouples are not being changed.

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'Document Control Desk Attachment 1

• LR-N96069 LCR S96-03 The respective action statements will continue to require plant shutdown if these independent, diverse instrumentation indications of an approach to inadequate core cooling are unavailable. This is also consistent with NUREG 1431.

Deletion of the footnote *** and existing ACTION 8 are considered administrative changes since they were temporary changes that have since expired.

Conclusions The proposed AOTs and associated Special Reporting requirements are consistent with the safety significance of the system for accident monitoring and with the design basis of the system. The proposed AOTs are consistent with equivalent statements that are used for RVLIS in the ITS. The Salem LCO and Surveillance Requirements for RVLIS are consistent with the ITS and are not being changed.

B. CHANNEL CALIBRATION DEFINITION Requested Change and Purpose Salem Units 1 and 2 TS Definition 1.4 pertaining to Channel Calibration is to be revised to replace the existing definition with the corresponding definition from NUREG 1433 "Standard Technical Specifications for General Electric Plants," Revision 1.

The definition for Channel Calibration from NUREG 1433 "Standard Technical Specifications for General Electric Plants," Revision 1, was used rather than from NUREG 1431 "Standard Technical Specifications for Westinghouse Plants," Revision 1. The NUREG 1433 definition is more inclusive, more closely aligns with other existing Salem TS definitions, and was consistent with existing procedures and practices at Salem. Use of the NUREG 1431 definition would have required additional changes to other existing TS definitions since it relies on other definitions to completely describe channel calibration requirements.

The proposed change ensures the required testing methodology aligns with standard industry methodology for instrument channels having a thermocouple (T/C) or Resistance Temperature Detector (RTD) as a sensor in order to prevent unnecessary removal of these sensors.

Currently, channel calibrations for instrument channels having RTD or T/C sensors are completed by performing an inplace Page 4 of 7

• Document Control flPk

• LR-N96069 LCR S96-03 qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. This test methodology is consistent with standard industry practice and was considered to satisfy the surveillance requirements.

A similar change to the Channel Calibration definition was submitted for Hope Creek and approved as Amendment 90. Other Westinghouse plants' TS having the same definition for Channel Calibration normally have added a qualifying note, in individual TS instrumentation sections, that exempts RTDs and T/Cs from the sensor calibration requirement. Salem has no such footnote(s).

Justification of Requested Changes The intent of the surveillance requirements which calibrate instrument channels is to ensure the channel accurately reflects and responds to the actual state of the monitored parameter.

Most instrument channels identified in the TS have a sensor that may vary its output with time without a corresponding change in the state of the monitored parameter. This is known as sensor drift. Periodic calibration of these sensors is necessary to ensure necessary accuracy levels are maintained.

RTDs and T/Cs, however, are relatively insensitive to sensor drift and are considered to be either accurate or not accurate.

Any change to their output, independent of the state of the monitored environment, will usually be observable by comparison with other devices measuring the same environment. Failures of these devices tend to be gross and readily observable.

In addition, it is difficult if not impossible to calibrate RTDs and T/Cs inplace. Removal and subsequent re-installation of the sensors introduces a potential for an unidentified failure that outweighs the benefits of the sensor calibration.

Calibration may also result in additional personnel radiation exposure which is inconsistent with ALARA goals. Deleting the requirement to calibrate RTDs and T/Cs prevents the diversion of plant personnel and resources for unnecessary testing.

For these reasons, IEEE Standard 338-1977, "IEEE Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems," provides the guidance that when complete checks, including those of the sensor, are not practicable, an analog or digital input for partial testing should be introduced and varied as appropriate.

In addition to deleting the requirement to calibrate RTD and thermocouple sensors, incorporation of the NUREG 1433 definition Page 5 of 7

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• LR-N96069 LCR S96-03 added a new requirement to include channel displays in the calibration requirements. Inclusion of channel displays is, also, consistent with current standard practices and is conservative with respect to the existing definition.

Changes to Definition 1.4, other than as noted above, are considered grammatical in nature and do not affect channel calibration methodology.

Conclusions The proposed change ensures the required testing methodology aligns with standard industry methodology. The submitted definition is a replicate of the definition for Channel Calibration in ITS, NUREG 1433. Use of the ITS definition was approved by the NRC for Hope Creek in TS Amendment 90. In addition, for instrument channels having a thermocouple (T/C) or Resistance Temperature Detector (RTD) as a sensor, the proposed change will prevent unnecessary removal of these sensors.

C. AUXILIARY FEEDWATER ACTUATION SYSTEM JUMPER Requested Change and Purpose A requirement to install a jumper was added to Action Statement 21.b of Table 3.3-3 via Amendment 39 in Unit 1 and Amendment 116 in Unit 2 to ensure the actuation of Auxiliary Feedwater would occur upon loss of the Feedwater Pumps. Since a permanent design modification is being implemented, a jumper is no longer required to ensure Auxiliary Feedwater initiation. Action Statement 21.b of Table 3.3-3 is being deleted.

Justification of Requested Changes Design changes are being installed in both Salem Units to provide failsafe (deenergize to actuate) relays which automatically perform the same function as the manual installation of the jumpers upon a Steam Generator Feedwater Pump (SGFP) Trip. The relay actuation condition is displayed to the operators by illumination of control console pushbutton light "SGFP TRIP-AFP AUTO ARMED" (one for each SGFP) . This relay actuation provides the same function for the Auxiliary Feedwater Pump (AFP) auto-start circuitry as the jumpers; that is, when the affected SGFP is tripped AFPs will start upon the loss of the remaining SGFP.

This automatic and failsafe permanent hardware modification provides an enhanced method of enabling the AFP auto-start circuitry and obviates the need for manually installing a jumper.

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• Document Control Attachment 1

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• LR-N96069 LCR 896-03 Conclusions The automatic and failsafe installation of the permanent design change via a relay provides an enhanced method, both for equipment protection and personnel protection, of enabling the AFP auto-start circuitry. The new circuitry obviates the need to manually install jumpers.

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• Document Control Attachment 2

• LR-N96069 LCR S96-03 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS INSTRUMENTATION SYSTEM REQUIREMENTS 10CFR50.92 EVALUATION Public Service Electric & Gas (PSE&G) has concluded that the proposed changes to the Salem Generating Station Unit Nos. 1 and 2 Technical Specifications (TS) do not involve a significant hazards consideration. In support of this determination, an evaluation of each of the three standards set forth in 10CFR50.92 is provided below for each of the three changes proposed.

A. REACTOR VESSEL LEVEL INDICATING SYSTEM (RVLIS)

Requested Change This submittal changes the RVLIS action statements to require a special report if one channel is inoperable for thirty days or if two channels are inoperable for seven days. Shutdown will not be required for RVLIS inoperability.

The proposed amendment will align the action statements with those provided in improved "Standard Technical Specifications for Westinghouse Plants", NUREG 1431, Revision 1 (ITS.) and will permit the performance of vendor recommended data taking and normalization evolutions while in Mode 3. Use of ITS action statements for Salem is supported by a comparison of the use of RVLIS at Salem against the use of RVLIS in the standard Westinghouse plant.

Basis

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

RVLIS is part of the safety-related display instrumentation (UFSAR section 7.5). Its function is to display information for the operator "to enable him to perform required manual functions and to determine the effect of manual actions taken following a reactor trip due to operational occurrences or accident conditions discussed in Section 15."

RVLIS performs no automatic functions designed to mitigate the consequences of any accident.

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• Document Control Ilk Attac:f.i...ment 2

• LR-N96069 LCR S96-03 Since no hardware changes are being made by this proposal and since the RVLIS is a post-accident monitoring system, no increase in the probability of any evaluated accident will occur as a result of implementation of the proposed change.

Other redundant, diverse instrumentation is available to operators to indicate inadequate core cooling.

Since RVLIS indication has limited use under normal conditions, performs no automatic function to mitigate an accident, and since it is augmented during emergency conditions by other independent indications of inadequate core cooling, its increased AOT does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

RVLIS is a Post Accident Monitoring System which does not initiate a transient or initiate any mitigating function.

RVLIS's function is to assist the operator once an accident occurs.

Since no hardware changes are being made by this proposal and since the RVLIS is utilized as a post-accident monitoring system and is not considered a contributor to an accident, implementation of the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

The proposed amendment to TS Table 3.3-11 will permit vendor recommended preventive maintenance-type activities to be performed on RVLIS following startups from extended outages.

This will, potentially, enhance RVLIS reliability and availability and ensure that EOP data continues to be accurate.

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• LR-N96069 LCR S96-03 Since the RVLIS is a post-accident monitoring system that has no automatic initiation function, changing the AOT will have no significant impact on the margin of safety provided by RVLIS. In addition, since there are independent, diverse indications of inadequate core cooling available to the operator, changing the AOT for RVLIS will not significantly reduce the margin of safety provided by the post-accident monitoring system.

Conclusion Based on the above, PSE&G has determined that the proposed changes to RVLIS AOTs do not involve a significant hazards consideration.

B. CHANNEL CALIBRATION DEFINITION Requested Change Salem Units 1 and' 2 TS Definition 1.4 pertaining to Channel Calibration is to be revised to replace the existing definition with the corresponding definition from NUREG 1433 "Standard Technical Specifications for General Electric Plants,"

Revision 1.

Basis

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

Since no physical change is being made to the instrumentation channels, or to any system or component that interfaces with the instrumentation channels, there is no change in the probability of any accident analyzed in the UFSAR.

There is no change in the consequences of an accident. The proposed change continues to ensure the surveillance requirements meet the licensing basis. Also, the testing performed will continue to demonstrate the capability of the affected instrumentation channels to respond to changes in the state of the monitored parameters in a manner consistent with assumptions in the accident analysis.

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• Document Control LR-N96069 Attachment 2 LCR S96-03

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

The proposed change does not result in any design or physical configuration changes to the instrumentation channels. Operation incorporating the proposed change will not impair the instrumentation channels from performing as provided in the design basis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

Since the proposed change does not involve the addition or modification of plant equipment, is consistent with the intent of the existing TS, is consistent with the current industry practices as outlined in improved ITS and is consistent with the design basis of the Instrumentation Systems and the accident analysis, no action will occur that will involve a significant reduction in a margin of safety.

Conclusion Based on the above, PSE&G has determined that the proposed changes do not involve a significant hazards consideration.

C. AUXILIARY FEEDWATER ACTUATION SYSTEM JUMPER Requested Change Action Statement 21.b of Table 3.3-3 is being deleted based on a modification incorporating a relay to perform the function of the specified jumper.

Basis

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

Deletion of TS Table 3.3-11 Action 21.b is consistent with the modified Auxiliary Feedwater Actuation Logic. The modified design does not require a manual jumper to ensure proper response to a Feedwater Pump Trip. The new modification obviates the need for a manual jumper, and eliminates the possibility for human error.

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• Document Control Attachment 2

• LR-N96069 LCR S96-03 Since the installed relay actuation performs a function identical to the installation of a manual jumper, the probability and consequence of analyzed accidents is unchanged.

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

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

Deletion of TS Table 3.3-11 Action 21.b is consistent with the modified Auxiliary Feedwater Actuation Logic. The modified design does not require a manual jumper to ensure proper response to a Feedwater Pump Trip.

Since the installed relay actuation performs a function identical to the installation of a manual jumper, the probability of creating a new or different accident is unchanged.

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

3. The proposed change does not involve a significant reduction in a margin of safety.

The automatic relay replaces the manual jumper specified in the action statement and provides an equivalent margin of safety to manually installing a jumper, while reducing the potential for human error.

Therefore, the proposed amendment will not involve a significant reduction in a margin of safety.

Conclusion Based on the above, PSE&G has determined that the proposed changes do not involve a significant hazards consideration.

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• LR-N96069 LCR S96-03 SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 CHANGE TO TECHNICAL SPECIFICATIONS INSTRUMENTATION SYSTEM REQUIREMENTS TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-70 are affected by this change request:

Technical Specification Page Definition 1.4 1-1 Action 21.b 3/4 3-22 Table 3.3-11, item 19 3/4 3-55 Table 3.3-11, Notation 3/4 3-56a Special Reports 6-24 The following Technical Specifications for Facility Operating License No. DPR-75 are affected by this change request:

Technical Specification Page Definition 1.4 1-1 Action 21.b 3/4 3-23 Table 3.3-11, item 19 3/4 3-51A Table 3.3-11, Notation 3/4 3-51C Special Reports 6-24

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INSERT A 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 the remaining adjustable devices in the channel. 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.

INSERT B ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERABLE status within 7 days or submit a special report in accordance with Specification

6. 9. 4.

INSERT C 6.9.4 When a report is required by ACTION 8 or 9 of Table 3.3-11 "Accident Monitoring Instrumentation", a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.