ML102440366

Request for Additional Information Related to a Modification That Replaces the Temperature-Based Isolation Instrumentation with Reactor Pressure-Based Isolation Instrumentation

ML102440366
Person / Time
Site:	Dresden
Issue date:	09/03/2010
From:	Gratton C Plant Licensing Branch III
To:	Pacilio M Exelon Nuclear
Gratton C, NRR/DORL/LPL3-2, 415-1055
References
TAC ME3354, TAC ME3355
Download: ML102440366 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 3, 2010 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO A MODIFICATION THAT REPLACES THE TEMPERATURE-BASED ISOLATION INSTRUMENTATION WITH REACTOR PRESSURE-BASED ISOLATION INSTRUMENTATION (TAC NOS. ME3354 AND ME3355)

Dear Mr. Pacilio:

By letter to the Nuclear Regulatory Commission (NRC) dated February 4, 2010 (Agencywide Documents Access and Management System Accession No. ML100470776), Exelon Generation Company, LLC submitted a request to revise Technical Specification (TS) 3.3.6.1, "Primary Containment Isolation Instrumentation," Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation," Function 6.a, "Shutdown Cooling System Isolation, Recirculation Line Water Temperature - High," to enable implementation of a modification that replaces the temperature-based isolation instrumentation with reactor pressure-based isolation instrumentation, for the Dresden Nuclear Power Station, Units 2 and 3.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on August 23, 2010, it was agreed that you would provide a response within 30 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of

M. Pacilio -2 efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-1055.

Sincerely, Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249 In reviewing the Exelon Generation Company's (Exelon's, the licensee's) submittal dated February 4, 2010 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML100470776), for the Dresden Nuclear Power Station (DNPS), Units 2 and 3, to change the sensors and Shutdown Cooling (SDC) system isolation logic that prevents exceeding the SDC system design temperature, the Nuclear Regulatory Commission (NRC) staff has determined that additional information is needed to evaluate Exelon's compliance with Title 10 of the Code of Federal Regulations, Section 50.36, and current review criteria that govern setpoints. This information is needed to verify the licensee's ability to identify inoperability and degradation of equipment based upon its setpoint methodology, and the calibration and surveillance check procedures associated with this license amendment request (LAR).

Also, additional information is required to evaluate the plant's ability to prevent a potential loss-of-coolant through means that ensure the SDC system's maximum design temperature of 350 OF is not exceeded. This information is needed to verify applicable portions of the NRC Standard Review Plan (SRP) (NUREG-0800), Branch Technical Position 7-1 "Guidance on Isolation of Low-Pressure Systems from the High-Pressure Reactor Coolant System" (ADAMS Accession No. ML070460345) and General Design Criteria (GDC) 15, as discussed in the Updated Final Safety Analysis Report, are met.

Setpoints - General: The following four Requests for Additional Information (RAls) (1-4) address the licensee's overall approach to meeting the current regulatory criteria for Technical Specification (TS) content in accordance with NRC Regulatory Issue Summary 2006-17 (ADAMS Accession No. ML051810077). The LAR proposes the following TS changes:

• Replacement of the "Recirculation Line Water Temperature-High" setpoint and its allowable value of "s 346°F" with one setpoint "Reactor Vessel Pressure-High" with two allowable values as follows:

o "S 114.1 psig (Loop 1, Reactor Wide Range Pressure)"

o "s 110.4 psig (Loop 2, Reactor Pressure Feedwater Control)"

1) Setpoint Calculation Methodology: In addition to the calculations, which were provided in the LAR, provide documentation of the setpoint methodology used for establishing the limiting setpoint (NSP) and the limiting acceptable values for the As-Found and As-Left setpoints as measured in periodic surveillance testing.

i) the limiting setpoint is referred to as calculated setpoints in the LAR.

ii) the limiting acceptable values for the As-Found setpoints are referred as "Expanded Tolerance" in the LAR.

Enclosure

-2 iii) the limiting acceptable values for the As-Left setpoints are referred as "Setting Tolerance" in the LAR.

This documentation should:

a) Provide NES-EIC-20.04, "Analysis of Instrument Channel Setpoint Error and Instrument Loop Accuracy," Revision 5, which is Reference 3.1.2 of Attachment 4, Setpoint Calculation No. DRE09-0041, "Shutdown Cooling Reactor High Pressure (Cut-in Permissive) Setpoint Calculation," of the LAR.

b) Explicitly identify the methodology and its current revision that is referenced within the TS Design Bases statements:

i) "Any setpoint adjustment shall be consistent with the assumptions of the current plant-specific setpoint methodology" (reference "Dresden 2 and 3 Technical Specification Bases", page B 3.3.6.1-6, Revision 0) ii) "there is a plant-specific program which verifies that the instrument channel functions as required, by verifying the as-left and as-found settings are consistent with those established by the setpoint methodology" (see proposed "Dresden 2 and 3 Technical Specification Bases", page B 3.3.6.1-26, Revision 0).

iii) Or, provide a confirmatory statement that this methodology is identical to the methodology identified in 1.a above.

2) Safety Limit (Sl)-Related Determination: Provide a statement as to whether or not the setpoint is a Limiting Safety System Setting (LSSS) for a variable on which a SL has been placed as discussed in 10 CFR 50.36(c)(1)(ii)(A), so as to represent a "SL-Related" setpoint.

Such setpoints are described as "SL-Related" in the discussions that follow. In accordance with 10 CFR 50.36(c)(1)(ii)(A), the following guidance is provided for identifying a list of functions to be included in the subset of LSSSs specified for variables on which SLs have been placed as defined in Standard Technical Specifications, Sections 2.1.1, "Reactor Core SLs," and 2.1.2, "Reactor Coolant System Pressure SL." This subset includes automatic protective devices in TSs for specified variables on which SLs have been placed that:

(1) initiate a reactor trip; or (2) actuate safety systems. As such, these variables provide protection against violating reactor core safety limits, or reactor coolant system pressure boundary safety limits.

Examples of instrument functions that might have LSSSs included in this subset in accordance with the plant-specific licensing basis, are pressurizer pressure reactor trip (pressurized-water reactors), rod block monitor withdrawal blocks (boiling-water reactors),

feedwater and main turbine high water level trip (boiling-water reactors), and end of cycle recirculation pump trip (boiling-water reactors). For each setpoint, or related group of setpoints, that you determined not to be SL-Related, explain the basis for this determination.

3) For setpoints that are determined to be SL-related: The NRC letter to the NEI Setpont Methods Task Force dated September 7,2005 (ADAMS Accession No. ML052500004),

describes Setpoint-Related TS (SRTS) that are acceptable to the NRC for instrument settings associated with SL-related setpoints. Specifically: Part "A" of the Enclosure to the letter provides limiting condition for operation notes to be added to the TS, and Part "B"

-3 includes a check list of the information to be provided in the TS Bases related to the proposed TS changes.

a) Describe whether and how you plan to implement the SRTS suggested in the September 7,2005, letter. If you do not plan to adopt the suggested SRTS, explain how you will ensure compliance with 10 CFR 50.36 by addressing items 3.b and 3.c, which follow.

b) As-Found Setpoint evaluation: Describe how surveillance test results and associated TS limits are used to establish operability of the instrument channels that are used for initiating the applicable safety system functions. Show that this evaluation is consistent with the assumptions and results of the setpoint calculation methodology. Discuss the plant corrective action processes (including plant procedures) for restoring channels to "operable" status when channels are determined to be "inoperable" or "operable but degraded." Describe the processes that will be used to track corrective actions required for channels whose performance has been identified as "operable but degraded." If the criteria for determining operability of the instrument channel being tested are located in a document other than the TS (e.g. plant test procedure), explain how the requirements of 10 CFR 50.36 are met.

c) As-Left Setpoint control: Describe the controls employed to ensure that the instrument setpoint is, upon completion of surveillance testing, consistent with the assumptions of the associated analyses. If the controls are located in a document other than the TS (e.g. plant test procedure), explain how the requirements of 10 CFR 50.36 are met.

4) For setpoints that are not determined to be SL-related: Describe the measures to be taken to ensure that the associated instrument channel is capable of performing its specified safety functions in accordance with applicable design requirements and associated analyses. Include in your discussion information on the controls you employ to ensure that the as-left trip setting after completion of periodic surveillance is consistent with your setpoint methodology. Also, discuss the plant corrective action processes (including plant procedures), for restoring channels to operable status when channels are determined to be "inoperable" or "operable but degraded." If the controls are located in a document other than the TS (e.g., plant test procedure), describe how it is ensured that the controls will be implemented.

Licensee Setpoint Methodology: The following RAI (5) requests the basis for a specific aspect of the licensee's general setpoint methodology.

5) Explain how an expanded tolerance (ET) could be less than the setting tolerance (ST), as it is calculated and identified as a check criterion. This explanation should make it apparent the mechanism by which the calculated ET could result in a value less than the ST for the equation discussed in Section 2.4.4 of Attachment 4, Setpoint Calculation No. DRE09-0041, "Shutdown Cooling Reactor High Pressure (Cut-in Permissive) Setpoint Calculation," and provided as "ET = +/- [0.7 * (DTI - ST)] + ST."

LAR Setpoint Specific: The following RAI (6) addresses a specific aspect of the licensee's setpoint calculation.

-4

6) Provide a justification for excluding errors associated with 'dynamic effects.' Currently, the LAR does not discuss 'dynamic effects. '

Reference 3.1.1 of Attachment 4, Setpoint Calculation No. DRE09-0041, "Shutdown Cooling Reactor High Pressure (Cut-in Permissive) Setpoint Calculation" of the LAR is Part 1 of ANSIIISA-S67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation." Its Section 4.4(g) 'dynamic effects'states:

The behavior of a channel's output as a function of the input with respect to time shall be accounted for, either in the determination of the trip setpoint or included in the safety analyses.

Normally, these effects are accounted for in the safety analyses."

Regulatory Guide 1.105, "Setpoints for Safety-Related Instrumentation." Rev. 3 endorses-with exceptions and clarifications - Part 1 of ANSIIISA-S67.04-1994; however, contrary to this endorsement, the 'process effects' described in the LAR do not address time-dependency (Le.

dynamic effects) associated with the temperature to pressure transmitter change or any additional time-dependency deltas that have been introduced by differing measurement systems (in other words, the introduction of the feedwater control system into Loop 2). The LAR only addresses pressure transmitter static process errors, which the LAR deems as insignificant by engineering judgment. Therefore, the LAR does not presently address design modifications that change the instrument dynamic characteristics and relocate sensors.

Resolution of Inconsistencies: The following RAI (7) requests information to resolve apparent inconsistencies within the LAR and supporting documentation in order to clarify the scope and intent of the LAR instrumentation and control changes.

7) The licensee is requested to provide clarifications to apparent inconsistencies among the DNPS, Units 2 and 3, TS Bases and other information currently available to the NRC in order to clarify the full scope and nature of the proposed change. The licensee should address the inconsistencies to consistently describe the SDC isolation function, such that the maximum design temperature will not be exceeded. The licensee is requested to submit appropriate clarifications that address the following items 7.a through 7.f. These clarifications may include additional revisions/markups pages.

a) The continued use of the term 'temperature' for the isolation function: Specifically, the title on Bases Page B 3.3.6.1-18, for Function 6.a, should be changed from "Recirculation Line Water Temperature-High," to "Reactor Vessel Pressure - High," to match the proposed title forTS 3.3.6.1, Table 3.3.6.1-1, Function 6.a.

b) The continued use of the term 'bypass' in consideration of the proposed one-out-of-two taken-twice logic configuration: The licensee response should clearly, correctly, and consistently describe the sense-trip-Iogic-actuation sequence in order to evaluate the acceptability of the TSs, Table 3.7.6.1-1 entries for FUNCTION 6.a. under "Shutdown Cooling System Isolation," and in particular, the "REQUIRED CHANNELS PER TRIP SYSTEM," its referenced condition F, and any dependency of the function on the non-safety-related feedwater control system. The licensee response should consider whether 'bypassing' a failed channel, which is typical of one-out-of-four logic, remains appropriate, or rather forcing a one-half trip, which is typical of one-out-of-two-taken twice logic, is now appropriate. Currently, there is a lack of clarity, because the LAR indicates that one-out-of-four logic currently exists; however, the common DNPS,

-5 Units 2 and 3, TS Bases (see Revision 6, Page B 3.3.6.1-23) contains actions consistent with placing the failed channel in the tripped state (versus bypass-non tripped). The licensee response should fully resolve this inconsistency.

c) Definition and description of terms: The licensee should provide a definition for each of the following: 1) trip string, 2) trip channel, and 3) trip system, and a corresponding figure that identifies and shows each of these items in the configuration proposed by the LAR:

one-out-two-taken-twice logic using all pressure sensor inputs (for Loops 1 and 2). This information is required to provide the complete context of the proposed modification.

This clarification is required, because the DNPS, Units 2 and 3, TS Bases (see Revision 8, Page B 3.3.6.1-5) does not currently describe the use of 'trip strings' for the Recirculation Line Water Temperature-High Function, but rather only describes this aspect of the isolation function in terms of 'trip channels' and 'trip systems.'

d) Consistency of logic function description: The DNPS, Units 2 and 3, TS Bases (see Revision 8, Page B 3.3.6.1-5) describes the Recirculation Line Water Temperature-High isolation logic as one-out-of-four; however, it does not identify DNPS, Unit 2 as currently having a one-out-of-two-taken-twice logic configuration (reference the 4th paragraph on Page 4 of 13 of Attachment 1). It is noted that the cause for this difference between units is described in LAR (see the 4th complete paragraph of Page 6 of 13 of the Attachment 1) despite it not being reHected in the common TS Bases.

e) Clarification of the term "loop:" The DNPS, Units 2 and 3, TS Bases (see Revision 31, Page B 3.3.6.1-18), which currently describes the logic as one-out-of-four, includes a statement, "Only two channels (one channel from each loop) are required to be operable." Within the licensee's response, the licensee should clarify the term 'loop,'

which had been understood to reference the previous recirculation 'loops' where the previously relied-upon temperature sensors reside. Also, as currently written, the statements are consistent with one-out-of-four logic where a bypass may be permitted; however, the LAR now describes the proposed two pressure-based 'loops' to feed one-out-of-two-taken-twice logic. Therefore, the statement, as currently written, should be modified, as appropriate, in consideration of item 7.b.

f) Clarification of Loop 2 operability: The licensee should clarify its considerations of the adequacy of the TS surveillance requirements in order to address the insertion of the digital feedwater control system into the SDC isolation function. The clarification should address Loop 2's operability in a manner that considers failure modes of the digital feedwater control system for the "Reactor Vessel Pressure-High" function.

SDC Isolation Functions: The following RAI (8) requests information to support evaluation of the LAR against criteria applicable to the SDC isolation functions, including those related to the plant's diversity and defense-in-depth. This information is necessary because the LAR proposes to replace instrumentation used to perform the SDC isolation function (currently performed by analog safety-related instrumentation), with partial reliance upon the non safety-related digital feedwater control system.

8) Provide sufficient information to justify reliance upon the non-safety-related digital feed water control system to perform SDC Isolation functions. This information should:

-6 a} Address compliance with U.S. NRC SRP, Chapter 7, Section 7.6, "Interlock Systems Important to Safety" (ADAMS Accession No. ML070460348).

b} Demonstrate that any single-failure of the equipment used to support the SDC isolation functions, including the isolation devices and non-safety related digital feedwater system, does not result in a vulnerability to which either DNPS Unit 2 or Unit 3 has an inability to cope. This response should explain how the diversity and defense-in-depth that will remain following proposed LAR ensures reliable operation of the SDC Isolation functions (autoclosure and interlocks), so that the SDC system:

i} Isolates when required to prevent potential damage to the SDC components and possible radiological release; ii} Remains sealed-in and does not inadvertently isolate when needed, thereby interrupting shutdown cooling; and iii} Does not un-isolate when the temperature is above 350 of.

Within this response, the licensee should clarify all considerations made to address software common-cause failures that might affect the SDC Isolation functions.

The LAR proposes to use the non-safety digital feedwater control system to generate half of the actuation signals (Loop 2) that isolate the SDC system from the reactor pressure vessel.

However, SDC isolation functions are important to safety, because these functions are relied upon to meet portions of Appendix A to Part 50--General Design Criteria for Nuclear Power Plants: a} Criterion 14, "Reactor Coolant Pressure Boundary," Criterion 15, "Reactor Coolant System Design," and Criterion 34, "Residual Heat Removal," by preventing an improper connection of the SDC system to the Reactor Coolant System (RCS) when the RCS temperature is above 350 OF. The SDC isolation function is designed to prevent exceeding the SDC system design temperature of 350 OF, in part, to prevent subsequent equipment damage and resultant loss of coolant. SRP 7.6 addresses interlocks consistent with the type included in this LAR and identifies acceptance criteria

M. Pacilio -2 efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-1055.

Sincerely, IRA!

Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv DISTRIBUTION:

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