RS-02-080, Request for Technical Specifications Changes Related to Reactor Protection System Instrumentation

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Request for Technical Specifications Changes Related to Reactor Protection System Instrumentation
ML021190419
Person / Time
Site: Dresden  Constellation icon.png
Issue date: 04/15/2002
From: Jury K
Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-02-080
Download: ML021190419 (21)


Text

-1Exelon Exelon Generation www.exeloncorp.com Nucl ear 4300 Winfield Road WarTenville, IL60555 10 CFR 50.90 RS-02-080 April 15, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

Request for Technical Specifications Changes Related to Reactor Protection System Instrumentation (Reactor Vessel Steam Dome Pressure - High)

References:

(1) Letter from Joel P. Dimmette Jr. (Commonwealth Edison Company) to U. S. NRC, "Request for Technical Specifications Change Reactor Protection System Instrumentation Reactor Vessel Steam Dome Pressure - High," dated November 16, 1999 (2) Letter from U. S. NRC to 0. D. Kingsley (Commonwealth Edison Company), "Quad Cities - Issuance of Amendments on Replacement of Pressure Switches," dated January 28, 2000 In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company (Exelon), LLC, requests changes to the Technical Specifications (TS) of Facility License Nos. DPR-1 9 and DPR-25 for the Dresden Nuclear Power Station (DNPS), Units 2 and 3. Specifically, the proposed changes modify the reactor vessel steam dome pressure - high surveillance requirements (SRs) and allowable value (AV) specified in TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation." These proposed changes support a planned upgrade to the reactor vessel steam dome pressure-high instrumentation from pressure switches to analog trip units. Analog trip units use a proven technology that is more reliable than the existing pressure switches. Analog trip units are used in various applications at DNPS, including the reactor protection system low water level trip function. A similar TS change request for the Quad Cities Nuclear Power Station was submitted in Reference 1 and approved by the NRC in Reference 2.

Additionally, a new value is proposed for the AV in anticipated transient without scram (ATWS) surveillance requirement SR 3.3.4.1.4.b, Reactor Vessel Steam Dome Pressure High. The existing trip unit is being replaced with a functionally equivalent assembly from a different manufacturer to maintain diversity between ATWS and non-ATWS instrumentation.

The performance characteristics of the different trip unit result in the need for a revised AV.

ý, CA I

April 15, 2002 U. S. Nuclear Regulatory Commission Page 2 The subject analog trip units will be installed during the next refueling outage on Unit 3, scheduled to begin in October 2002. Therefore, Exelon requests approval of the proposed TS changes by September 9, 2002. Once approved, the amendment shall be implemented within 90 days for Unit 3. A similar instrumentation design change will be implemented on Unit 2 during its next refueling outage (i.e., D2R1 8), scheduled to begin in December 2003.

Therefore, implementation of this proposed change for Unit 2 is requested prior to startup from D2R18.

This request is subdivided as follows.

1. Attachment A provides a description and safety analysis of the proposed changes.
2. Attachment B provides the marked-up TS pages indicating the proposed changes. A marked-up copy of the affected TS Bases is also included for informational purposes.
3. Attachment C provides the revised TS pages for the proposed changes.
4. Attachment D provides our evaluation performed using the criteria in 10 CFR 50.91 (a), "Notice for public comment," paragraph (1), which provides information supporting a finding of no significant hazards consideration using the standards in 10 CFR 50.92, "Issuance of amendment," paragraph (c).
5. Attachment E provides information supporting an Environmental Assessment.

These proposed TS changes have been reviewed by the DNPS Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the Exelon Quality Assurance Program.

Exelon is notifying the State of Illinois of this request for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning his letter, please contact Mr. Allan R. Haeger at (630) 657-2807.

Respectfully, Keith R. Jury Director - Licensing Mid-West Regional Operating Group Attachments: Affidavit Attachment A: Description and Safety Analysis for Proposed Changes Attachment B: Marked-Up TS Pages for Proposed Changes Attachment C: Revised TS Pages for Proposed Changes

April 15, 2002 U. S. Nuclear Regulatory Commission Page 3 Attachment D: Information Supporting a Finding of No Significant Hazards Consideration Attachment E: Information Supporting an Environmental Assessment cc: Regional Administrator - NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station Office of Nuclear Facility Safety - Illinois Department of Nuclear Safety

STATE OF ILLINOIS )

COUNTY OF DUPAGE

)

IN THE MATTER OF )

EXELON GENERATION COMPANY, LLC ) Docket Numbers DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 ) 50-237 and 50-249

SUBJECT:

Request for Technical Specifications Changes Related to Reactor Protection System Instrumentation (Reactor Vessel Steam Dome Pressure - High)

AFFIDAVIT I affirm that the content of this transmittal is true and correct to the best of my knowledge, information, and belief.

Keith R. Jury Director - Licensing Mid-West Regional Operating Group Subscribed and sworn to before me, a Notary Public in and for the State above named, this /IS day of 1,20a.2.

Notary Public OFIIL SEAL ANESE L.GRIGSBY NOTARY PUBLIC, STATE OF ILLINOIS MY COMMISSION EXPIRES 3.1 3-2005

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES A.

SUMMARY

OF THE PROPOSED CHANGES In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company (Exelon), LLC, requests changes to Appendix A, Technical Specifications (TS), of Facility License Nos. DPR-19 and DPR-25 for the Dresden Nuclear Power Station (DNPS), Units 2 and 3. Specifically, Exelon proposes to revise the surveillance requirements (SRs) and allowable value (AV) in Table 3.3.1.1-1, "Reactor Protection System Instrumentation," Function 3, "Reactor Vessel Steam Dome Pressure High." These proposed changes support a planned upgrade to the reactor vessel steam dome pressure - high instrumentation from pressure switches to analog trip unit (ATUs). The AVs and SRs are revised to be consistent with those for currently installed ATU devices.

Additionally, Exelon proposes to revise the AV in SR 3.3.4.1.4.b of TS Section 3.3.4.1, "ATWS RPT Instrumentation," to allow the use of a different manufacturer's ATU. This change maintains diversity between anticipated transient without scram (ATWS) and non-ATWS related instrumentation.

Currently, four locally mounted, non-indicating pressure switches are used to monitor reactor pressure. The switches are arranged so that each pair provides an input to reactor protection system (RPS) trip systems A and B. The existing reactor vessel steam dome pressure - high scram pressure switches will be replaced with pressure transmitters (i.e., Rosemount Model No.

1153GB9P) that will utilize an ATU (i.e., Rosemount Model 71ODU) and a master trip relay to interface with the existing RPS logic. The new pressure transmitter trip units use proven technology, are highly reliable, and are currently used in various RPS applications at DNPS (e.g., reactor vessel water level - low).

To maintain diversity between ATWS and non-ATWS related pressure instrumentation, the existing Rosemount Model 51 ODU master trip units used in the ATWS instrumentation are being replaced with General Electric (GE) Model 184C5988G 131 master trip units. The GE Model 184C5988G131 master trip unit has been used at DNPS for many years for level instrumentation. It is functionally equivalent to the existing Rosemount 51 ODU master trip units.

The existing ATWS pressure transmitters are Rosemount Model 1151GP9E and will not be changed or replaced as part of this planned modification.

The related plant design changes will be installed during the next refueling outage on both Unit 3 (i.e., D3R17) and Unit 2 (i.e., D2R1 8), scheduled for October 2002 and December 2003, respectively.

A complete description of the proposed changes is given in Section E, "Description of the Proposed Changes," of this attachment. Attachment B provides the marked-up TS pages indicating the proposed changes. Attachment C provides the revised TS pages.

B. DESCRIPTION OF THE CURRENT REQUIREMENTS TS Section 3.3.1.1 provides the requirements for RPS instrumentation. The various functions of the RPS instrumentation are specified in Table 3.3.1.1-1, along with their applicable operational modes, SRs, and AVs. Function 3 of Table 3.3.1-1 specifies an AV of < 1058 psig Page 1 of 7

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES for the reactor vessel steam dome pressure - high, and the following applicable SRs for the existing pressure switches.

Surveillance Requirements Frequency SR 3.3.1.1.5, Functional Test of RPS Scram Contactor 7 days SR 3.3.1.1.8, Channel Functional Test 31 days SR 3.3.1.1.13, Channel Calibration 92 days SR 3.3.1.1.18, Logic System Functional Test 24 months SR 3.3.1.1.19, RPS Response Time Verification 24 months on a staggered test basis Note that a channel check SR is not applicable because the current pressure switch instrumentation is non-indicating and does not allow for a channel check.

TS Section 3.3.4.1 provides the SRs for ATWS - recirculation pump trip (RPT) instrumentation. SR 3.3.4.1.4.b currently specifies an AV of < 1231 psig for reactor vessel steam dome pressure - high.

C. BASES FOR THE CURRENT REQUIREMENTS The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. The RPS initiates a reactor scram, using a "one out of two, taken twice" logic, when one or more monitored parameters exceed their specified limits, to preserve the integrity of the fuel cladding and the reactor coolant system, and to minimize the energy that must be absorbed following a loss of coolant accident.

An increase in the reactor pressure during reactor operation results in a positive reactivity insertion, which can challenge the fuel cladding. Although no specific safety analysis takes direct credit for this function, the reactor vessel steam dome pressure - high function initiates a scram for transients that result in a pressure increase in order to rapidly reduce core power.

High reactor pressure signals are initiated from four pressure switches that sense reactor pressure.

To lessen the effects of an ATWS event, the ATWS - RPT System initiates an RPT. This RPT adds negative reactivity due to an increase in steam voiding. When the reactor vessel steam dome pressure - high setpoint is reached, the recirculation motor generator (MG) drive motor field breaker is tripped.

The ATWS - RPT consists of two independent trip systems, with two channels of reactor vessel steam dome pressure - high and two channels of reactor vessel water level - low low in each trip system. Each ATWS - RPT trip system is a two-out-of-two logic for each function. Thus, either two reactor water level - low low or two reactor pressure - high signals are needed to trip a trip system. The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps by tripping the respective MG drive motor field breakers.

The ATWS - RPT is not assumed to mitigate any accident or transient in the safety analysis.

The ATWS - RPT initiates an RPT to aid in preserving the integrity of the fuel cladding Page 2 of 7

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES following events in which a scram should occur, but does not. Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4 of 10 CFR 50.36, "Technical specifications," paragraph (c)(2)(ii).

The TS required instrumentation, including RPS and ATWS - RPT, have SRs to ensure a high degree of safety system reliability. The SRs specified provide assurance that the reactor vessel steam dome pressure - high function of the RPS instrumentation will perform as required to shutdown the reactor on a high reactor pressure signal during reactor power operations. Likewise, the SRs provide assurance that the reactor vessel steam dome pressure - high function of ATWS - RPT instrumentation will perform as required to aid in preserving the integrity of the fuel cladding following events in which a scram should occur, but does not.

D. NEED FOR REVISION OF THE REQUIREMENTS The current reactor vessel steam dome pressure - high trip function instrumentation employs pressure switches, which are extremely sensitive to vibration. The pressure switches are also difficult to calibrate and have a tendency to drift. Since these pressure switches provide the logic actuation contacts for an RPS trip, any false actuation will initiate a spurious scram signal, which could result in a reactor scram. In addition, the existing pressure switches do not provide the necessary indication output to observe channel behavior during operation. For this reason, the current TS do not specify a channel check requirement for this RPS instrumentation function.

These proposed TS changes support a planned upgrade to the existing reactor vessel steam dome pressure - high scram pressure switches, which will be replaced with pressure transmitters that use an ATU and a master trip relay to interface with the existing RPS logic.

The ATUs replace the trip function at the sensor level with no change to the actuation logic.

These replacement pressure transmitters and trip units have a higher reliability and thus will result in decreasing unnecessary challenges to plant safety systems due to spurious actuation.

To reflect the ATU design features and performance, a TS change is required to revise the instrumentation AV, as well as to add new SRs and change existing surveillance frequencies to those appropriate for this type of instrument.

To maintain diversity between ATWS and non-ATWS related instrumentation, the existing ATWS - RPT ATU is being replaced with a functionally equivalent ATU from a different manufacturer. The performance characteristics of the different trip unit result in the need for a revised AV.

The planned instrumentation upgrade will maintain compliance with the commitments identified in the Updated Final Safety Analysis Report for RPS and analog trip system instrumentation.

The change provides increased reliability and better overall performance of the trip function while maintaining the identical RPS function.

Page 3 of 7

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES E. DESCRIPTION OF THE PROPOSED CHANGES There are a total of six changes to be made to the existing TS. While two portions of the TS are revised, for clarity the changes are numbered one through six within this section. The six changes are as follows:

The five proposed changes to Function 3 in TS Table 3.3.1.1-1 are as follows.

1. Add a new SR (SR 3.3.1.1.1) to perform a channel check every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2. Add a new SR (SR 3.3.1.1.12) to calibrate the individual ATUs every 92 days.
3. Revise the frequency of the channel functional test from 31 days (SR 3.3.1.1.8) to 92 days (SR 3.3.1.1.11).
4. Revise the frequency of the channel calibration from 92 days (SR 3.3.1.1.13) to 24 months (SR 3.3.1.1.17).
5. Revise the AV from < 1058 psig to < 1045 psig.

The sixth proposed change is to revise SR 3.3.4.1.4.b of TS Section 3.3.4.1 as follows.

6. Revise the AV from < 1231psig to < 1241 psig.

The proposed TS changes are reflected on a marked-up copy of the affected TS pages in Attachment B. A marked-up copy of the affected page from the TS Bases is also provided for information in Attachment B. Following NRC approval of this request, Exelon will revise the TS Bases, in accordance with the TS Bases Control Program of TS Section 5.5.10, "Technical Specifications (TS) Bases Control Program," to incorporate the changes identified in Attachment B. The revised TS pages are provided in Attachment C.

F. SAFETY ANALYSIS OF THE PROPOSED CHANGES Currently, four locally mounted, non-indicating pressure switches are used to monitor reactor pressure for the RPS reactor vessel steam dome pressure - high trip function. The switches are arranged so that each pair provides an input to RPS trip systems A and B. The existing reactor vessel steam dome pressure - high scram switches will be replaced with pressure transmitters that will utilize an analog trip unit (i.e., Rosemount Model No. 1153GB9P transmitters and Rosemount Model 71 ODU ATU) and a master trip relay to interface with the existing RPS logic. The new pressure transmitter trip units use proven technology, are highly reliable, and are being used in various applications at DNPS (e.g., reactor vessel water level low).

To maintain diversity between ATWS and non-ATWS related instrumentation, the existing Rosemount Model 51 ODU master trip units used in the ATWS instrumentation are being replaced with General Electric (GE) Model 184C5988G131 master trip units. The GE Model 184C5988G131 master trip unit has been used at DNPS for many years for level instrumentation. It is functionally equivalent to the existing Rosemount 51 ODU master trip units.

Page 4 of 7

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES The existing ATWS pressure transmitters are Rosemount Model 1151GP9E and will not be changed as part of the planned modification.

The discussion below corresponds to the numbering of the changes within Section E above.

1. Additional SR 3.3.1.1.1 The addition of a channel check provides for an appropriate SR for the modified instrumentation. The modified instrumentation provides indication and thus allows for a channel check to occur.
2. Additional SR 3.3.1.1.12 The addition of an SR to calibrate the individual trip units provides an appropriate SR for the modified instrument. The frequency of 92 days is consistent with the frequency for existing trip units. Review of historical surveillance data for both the Rosemount Model 710DU and GE Model 184C5988G 131 master trip units found that both have provided acceptable performance at DNPS. The TS currently require a 92-day calibration frequency for both trip units. Thus, the revised 92-day frequency for the reactor scram function is appropriate, and the continued use of the 92-day frequency for the modified ATWS-RPT instrumentation remains appropriate.
3. Revised frequency for Channel Functional Test The change of the surveillance frequency of the channel functional test from 31 days to 92 days (SR 3.3.1.1.11) for ATUs was evaluated by DNPS in Reference 1. Reference 1 proposed TS changes which, in part, increased the surveillance test intervals (STIs) and permitted specified channel functional tests to be extended to quarterly versus monthly or weekly. These STIs were based on GE licensing topical reports (LTRs) previously reviewed and approved by the NRC. The NRC approved the Reference 1 proposed TS changes in Reference 3.

The relevant LTR for this proposed change is NEDC-30851 P-A (Reference 2), which evaluated the impact of extending functional testing requirements from monthly to quarterly.

The proposed design of using a pressure transmitter, ATU, and a relay is the generic design evaluated by NEDC-30851P-A. Section 5.6.2 of NEDC-30851P-A notes that, "For each of the initiating events, the RPS unavailability was determined to be insensitive to the changes in component failure rates." Thus, the use of the proposed design is acceptable.

Section 5.7 of NEDC-30851 P-A includes a general discussion on the acceptability of extending testing requirements for ATUs. For the RPS scram function, the use of a quarterly frequency for channel functional tests is consistent with current TS for instruments of a similar design. Current TS for ATWS - RPT uses a quarterly frequency for channel functional tests and this frequency is not altered by this amendment request. For these reasons, the proposed functional test frequency is consistent with previously approved evaluations for DNPS.

NEDC-30851 P-A, however, did not contain specific instrument drift assumptions. DNPS evaluated the setpoint drift associated with these ATUs, consistent with the current Exelon setpoint methodology, NES-EIC-20.04, "Analysis of Instrument Channel Setpoint Error and Instrument Loop Accuracy," and confirmed that extending the calibration frequency from 31 to 92 days is acceptable and within proposed setpoint allowances. This methodology was Page 5 of 7

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES approved by the NRC in Reference 5. Therefore, increasing the channel functional test interval from 31 days to 92 days is appropriate.

4. Revised frequency for Channel Calibration The use of a pressure transmitter, ATU, and relay in place of the existing pressure switches allows the use of a 24-month frequency (SR 3.3.1.1.17) for channel calibration, provided that the ATU is calibrated every 92 days. Reference 2 evaluated the use of a pressure transmitter, ATU, and relay for the reactor vessel steam dome pressure scram function and allowed for channel calibration of the pressure transmitters on a 18-month frequency when the ATUs are calibrated on a 92-day frequency. The further extension of this surveillance frequency from 18 months to 24 months for existing station instruments that use the pressure transmitter, ATU, and relay design was evaluated by DNPS in a submittal to the NRC (Reference 4). The evaluation justified the extension of the channel calibration surveillance frequency to a maximum of 30 months for many instruments, including similar pressure transmitters, ATU, and relay designs. The rationale used in those submittals also applies in this case. The NRC approved this extension from 18 months to 24 months in Reference 5.

The reactor vessel steam dome pressure - high function will be provided by Rosemount Model No. 1153GB9P transmitters and Rosemount Model 71ODU master trip units. The Rosemount trip units functional check and setpoint verification frequency will remain unchanged at 92 days. Therefore, an increase in the surveillance interval to accommodate a 24-month fuel cycle does not affect the Rosemount trip units with respect to drift.

The Rosemount transmitter drift is determined by the vendor using quantitative analysis.

The drift value determined was used in the development of the proposed plant setpoint and TS AV. The results of this analysis support a 24-month surveillance interval. DNPS has quantitatively evaluated the historical performance of both Rosemount Model 1151 GP9E and Model 11 53GB7P transmitters and found that they can support a 24-month surveillance interval. Since the Model 1153GB9P has not been installed at DNPS, no historical quantitative evaluation of DNPS data is possible. Review of historical surveillance data of similar pressure transmitters found no failures that would invalidate this conclusion. In addition, the proposed 24-month SR, if performed at the maximum interval allowed (i.e., 30 months in accordance with SR 3.0.2), does not invalidate any assumptions in the plant licensing basis.

Thus, the proposed frequency for channel calibration is appropriate.

5,6 Allowable value changes The revised AVs in TS Table 3.3.1.1-1, Function 3 of < 1045 psig and TS SR 3.3.4.1.4.b of

_<1241 psig are based on calculations performed using the Exelon setpoint methodology previously approved by the NRC (Reference 5). No analytical limit is altered by any of the proposed changes. For this reason, the proposed change does not reduce plant safety.

Changes similar to the proposed changes described above were requested by the Quad Cities Nuclear Power Station, Units 1 and 2 in Reference 6 and approved by the NRC in Reference 7.

Page 6 of 7

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES G. IMPACT ON PREVIOUS SUBMITTALS Exelon has reviewed the proposed changes for impact on any submittals for DNPS currently being reviewed by the NRC, and has determined that there is no impact on any of these submittals.

H. SCHEDULE REQUIREMENTS We request approval of these proposed changes by September 9, 2002, to support activities in the next Unit 3 refueling outage (i.e., D3R17), scheduled to begin in October 2002. Once approved, the amendment shall be implemented within 90 days for Unit 3. A similar instrumentation design change will be implemented on Unit 2 during its next refueling outage (i.e., D2R18), scheduled to begin in November 2003. Therefore, implementation of this proposed change for Unit 2 is requested prior to startup from D2R1 8.

I. REFERENCES

1. Letter from J. M. Heffley (Commonwealth Edison Company) to U. S. NRC, "Proposed Technical Specifications Change Surveillance Test Intervals and Allowable Outage Times for Protective Instrumentation," dated January 11, 2000
2. General Electric Licensing Topical Report, NEDC 30851P-A, "Technical Specification Improvement Analysis for BWR Reactor Protection System," dated March 1988
3. Letter from L. W. Rossbach (U. S. NRC) to 0. D. Kingsley (Exelon), "Dresden - Issuance of Amendments Changing Allowable Out-Of-Service Times and Surveillance Test Intervals,"

dated August 2, 2000

4. Letter from R. M. Krich (Commonwealth Edison Company) to U. S. NRC, "Request for Technical Specifications Changes for Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, and Quad Cities Nuclear Power Station, Units 1 and 2, to Implement Improved Standard Technical Specifications," dated March 3, 2000
5. Letter from Stewart N. Bailey (U. S. NRC) to 0. D. Kingsley (Exelon), " Issuance of Amendments," dated March 30, 2001
6. Letter from Joel P. Dimmette Jr. (Commonwealth Edison Company) to U. S. NRC, "Request for Technical Specifications Change Reactor Protection System Instrumentation Reactor Vessel Steam Dome Pressure - High," dated November 16, 1999
7. Letter from U. S. NRC to 0. D. Kingsley (Commonwealth Edison Company), "Quad Cities Issuance of Amendments on Replacement of Pressure Switches," dated January 28, 2000 Page 7 of 7

Attachment B MARKED-UP TECHNICAL SPECIFICATIONS PAGES FOR PROPOSED CHANGES REVISED TS PAGES 3.3.1.1-9 3.3.4.1-3 REVISED BASES PAGE (PROVIDED FOR INFORMATION ONLY)

B 3.3.1.1-13

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)
c. Fixed Neutron 1 2 SR 3.3.1.1.1 < 122% RTP Flux-High SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.18 SR 3.3.1.1.19
d. Inop 1,2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.18
3. Reactor Vessel Steam 1,2 Dome Pressure-High SR .3 .1.1 SR 3..1.1.19
4. Reactor Vessel Water 1,2 2 SR 3.3.1.1.1 Ž 2.65 inches Level-Low SR 3.3..1.18 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.1 SR 3.3.1.1.18 SR 3.3.1.1.19 1, 2(c) 8 SR 3.3.1.1.15  ! 9.5% closed
5. Main Steam Isolation Valve-Closure SR 3.3.1.1.11 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 SR 3.3.1.1.13  ! 1.94 psig
6. Drywell Pressure-High 1,2 2 SR 3.3.1.1.11 SR 3.3.1.1.11 SR 3.3.1.1.18 SR 3.3.1.1.19 (c) With reactor pressure 2 600 psig.

sR'3s3{

sie I,.tl Dresden 2 and 3 3.3.1.1-9 Amendment No.

ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS


.NOTE ------------------------------------

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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains ATWS-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.4.1.2 Calibrate the trip units. 92 days SR 3.3.4.1.3 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.1.4 Perform CHANNEL CALIBRATION. The 24 months Allowable Values shall be:

a. Reactor Vessel Water Level-Low Low:

Ž -54.15 inches with time delay set to Ž 8.3 seconds and

  • 9.7 seconds; and
b. Reactor Vessel Steam Dome Pressure-High:
  • psig.

SR 3.3.4.1.5 Perform LOGIC SYSTEM FJNCTIONAL TEST 24 months including breaker actuAtion.

i12f I Dresden 2 and 3 3.3.4.1-3 Amendment No.

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 3. Reactor Vessel Steam Dome Pressure-High SAFETY ANALYSES, LCO, and An increase in the RPV pressure during reactor operation APPLICABILITY compresses the steam voids and results in a positive (continued) reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure-High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 2, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed Neutron Flux-High signal, not the Reactor Vessel Steam Dome Pressure-High or the Main Steam Isolation Valve-Closure signals), along with the safety valves, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor presssjare signals are initiated from four pressure j that sense reactor pressure. The Reactor Vessel Steam Dome Pressure-High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure-High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES I and 2 when the RCS is pressurized and the potential for pressure increase exists.

4. Reactor Vessel Water Level-Low Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at this level to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level-Low Function is assumed in the analysis of the (continued)

Dresden 2 and 3 B 3.3.1.1-13 Revi sion

Attachment C REVISED TECHNICAL SPECIFICATIONS PAGES FOR PROPOSED CHANGES REVISED TS PAGES 3.3.1.1-9 3.3.4.1-3

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)
c. Fixed Neutron 1 2 F SR 3.3.1.1.1 < 122% RTP Flux- High SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.18 SR 3.3.1.1.19
d. Inop 1,2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.18
3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.1 < 1045 psig Dome Pressure -High SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19
4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 > 2.65 inches Level - Low SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 F SR 3.3.1.1.5 < 9.5% closed
5. Main Steam Isolation 1, 2 (c) 8 Valve- Closure SR 3.3.1.1.11 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19
6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.5 < 1.94 psig SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 SR 3.3.1.1.19 (continued)

(c) With reactor pressure >. 600 psig.

Dresden 2 and 3 3.3.1.1-9 Amendment No.

ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS NOTE---------

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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains ATWS-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.4.1.2 Calibrate the trip units. 92 days SR 3.3.4.1.3 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.1.4 Perform CHANNEL CALIBRATION. The 24 months Allowable Values shall be:

a. Reactor Vessel Water Level- Low Low:

> -54.15 inches with time delay set to > 8.3 seconds and < 9.7 seconds; and

b. Reactor Vessel Steam Dome Pressure-High: K 1241 psig.

SR 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST 24 months including breaker actuation.

Dresden 2 and 3 3.3.4.1-3 Amendment No.

Attachment D INFORMATION SUPPORTING A FINDING OF NO SIGNIFICANT HAZARDS CONSIDERATION According to 10 CFR 50.92, "Issuance of amendment," paragraph (c) a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below regarding the proposed license amendment.

Overview In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company (Exelon), LLC, is requesting a change to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. DPR-19 and DPR-25, for Dresden Nuclear Power Station, Units 2 and 3. The proposed change modifies the allowable value and surveillance requirements for reactor protection system instrumentation for the reactor vessel steam dome pressure - high function. Surveillance requirement changes include the addition of a channel check and a trip unit calibration, and the modification of required frequencies for the channel functional test and channel calibration to support a planned upgrade to the reactor vessel steam dome pressure - high instrumentation from pressure switches to analog trip units. The proposed change also modifies the allowable value for the anticipated transient without scram recirculation pump trip (ATWS - RPT) based on reactor vessel steam dome pressure - high. This ATWS -RPT TS change supports a planned replacement of a trip unit with an equivalent trip unit from a different manufacturer to maintain diversity between ATWS and non-ATWS instrumentation. No analytical limit is changed for any TS function.

The proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specifications (TS) changes support a design change that upgrades the existing reactor vessel steam dome pressure - high instrumentation from pressure switches to analog trip units. Analog trip units use a proven technology that is more reliable than the existing equipment. The proposed design is consistent with a generic design that has been previously reviewed and approved by the NRC. Analog trip units are currently used in various applications at Dresden Nuclear Power Station (DNPS), including the reactor protection system (RPS) low water level scram function.

The proposed TS changes add new channel check and trip unit calibration surveillance requirements (SRs), and modify other SRs in keeping with the use of pressure transmitters for the reactor vessel steam dome pressure - high function. The new SRs are not applicable to Page 1 of 2

Attachment D INFORMATION SUPPORTING A FINDING OF NO SIGNIFICANT HAZARDS CONSIDERATION the existing instrumentation because the current pressure switches are non-indicating and do not employ trip units.

TS requirements that govern operability or routine testing of plant instruments are not assumed to be initiators of any analyzed event because these instruments are intended to prevent, detect, or mitigate accidents. Therefore, these changes will not involve an increase in the probability of an accident previously evaluated. Additionally, these changes will not increase the consequences of an accident previously evaluated because the proposed change does not adversely impact structures, systems, or components. The planned instrument upgrade is a more reliable design than existing equipment. The proposed change establishes requirements that ensure components are operable when necessary for the prevention or mitigation of accidents or transients. Furthermore, there will be no change in the types or significant increase in the amounts of any effluents released offsite.

In summary, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes support a planned instrumentation upgrade by incorporating SRs required to ensure operability. The change does not adversely impact the manner in which the instrument will operate under normal and abnormal operating conditions. Therefore, these changes provide an equivalent level of safety and will not create the possibility of a new or different kind of accident from any accident previously evaluated. The changes in allowable values and surveillance requirements do not affect the current safety analysis assumptions.

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

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

The proposed TS changes support a planned instrumentation upgrade. The proposed changes do not affect the probability of failure or availability of the affected instrumentation. The revised allowable values, addition of a channel check and trip unit calibration, and revision of other SRs do not affect the analytical limit assumed in the safety analyses for the actuation of the instrumentation. Therefore, the proposed changes do not result in a reduction in the margin of safety.

Conclusion Based upon the above evaluation, Exelon has concluded that the criteria of 10 CFR 50.92(c) are satisfied and that the proposed TS changes involve no significant hazards consideration.

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Attachment E INFORMATION SUPPORTING AN ENVIRONMENTAL ASSESSMENT Exelon Generation Company (Exelon), LLC has evaluated these proposed changes against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." Exelon has determined that these proposed changes meet the criteria for a categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9), and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92, "Issuance of amendment," paragraph (b). This determination is based on the fact that these changes are being proposed as an amendment to a license issued pursuant to 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes a surveillance requirement (SR), and the amendment meets the following specific criteria:

(i) The proposed changes involve no significant hazards consideration.

As demonstrated in Attachment D, the proposed changes do not involve a significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The proposed changes, which revise the allowable value and SRs for the reactor vessel steam dome pressure - high function of the reactor protection system instrumentation, are consistent with the plant design basis. There will be no significant increase in the amounts of any effluents released offsite. The proposed changes do not result in an increase in power level, do not increase the production, nor alter the flow path or method of disposal of radioactive waste or byproducts.

Therefore, the proposed change will not affect the types or increase the amounts of any effluents released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed changes will not result in changes in the configuration of the facility.

There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

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