Text

License Amendment Request to Revise ECCS TS 3/4.5.2 and FSAR Chapter 14 to Remove Charging
	ML16029A168
Person / Time
Site:	Millstone
Issue date:	01/25/2016
From:	Mark D. Sartain; Dominion Nuclear Connecticut
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	15-590
	Download: ML16029A168 (37)
	v • d • e

Dominion Nuclear Connecticut, Inc.

5000omoo DominionBularGeAlnV206

/ D Inn on Web Address: www.dom.com January 25, 2016 U.S Nuclear Regulatory Commission Serial No.15-590 Attention: Document Control Desk NSSL/LES R0 Washington, DC 20555 Docket No. 50-336 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 LICENSE AMENDMENT REQUEST TO REVISE ECCS TS 314.5.2 AND FSAR CHAPTER 14 TO REMOVE CHARGING Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests an amendment to Facility Operating License No. DPR-65 for Millstone Power Station Unit 2 (MPS2). The proposed amendment would revise MPS2 Technical Specification (TS) 3.5.2, "Emergency Core Cooling Systems, ECCS Subsystems - Tavg > 300°F," to remove the charging system and eliminate Surveillance Requirement 4.5.2.e from TSs.

The proposed amendment would also revise MPS2 Final Safety Analysis Report (FSAR) Chapter 14 relative to the long-term analysis in Section 14.6.1, "Inadvertent Opening of a Pressurized Water Reactor Pressurizer Pressure Relief Valve," and would clarify the existing discussion regarding the application of single failure criteria. An update to the associated TS Bases is included to address the proposed change. The scope of this License Amendment Request (LAR) was presented to the Nuclear Regulatory Commission (NRC) staff during a teleconference on December 10, 2015.

By letter dated April 29, 2015 (Reference 1), the NRC identified three apparent violations, two of which involved changes made by DNC to Section 14.6.1 of the MPS2 FSAR that removed credit for the chemical and volume control system charging pump flow in the mitigation of the event associated with the inadvertent opening of pressurizer pressure relief valves. These changes were made without obtaining prior NRC approval. An operability determination was completed to establish that the charging system was operable. On May 11, 2014, a standing order was implemented by DNC to establish requirements for the charging system similar to those that were in effect prior to License Amendment Number 283. The operability determination and standing order have been updated to ensure consistency With the August 26, 2015 Confirmatory Order EA-1 3-188 (Reference 2) actions related to this issue.

Confirmatory Order EA-13-188 required DNC to submit a LAR addressing the use of charging pumps in the analysis of the FSAR Section 14.6.1 event for the inadvertent opening of a pressurizer pressure relief valve. DNC reviewed and considered the need for the charging system in response to inadvertent opening of a pressurizer pressure relief valve. An analysis has been completed to show that the charging system is not needed to mitigate this transient.

Attachment I provides an evaluation of the proposed license amendment and FSAR change. Attachment 2 provides the marked-up TS page to reflect the proposed changes. Attachment 3 provides the marked-up TS Bases pages to reflect changes to t'

Serial No: 15-590 Docket No.: 50-336 Page 2 of 3 the TS Bases. The marked-up TS Bases pages are provided for information only and will be incorporated in accordance with the TS Bases Control Program upon approval of this amendment request. Attachment 4 provides the marked-up pages to reflect the proposed changes to the FSAR.

The proposed amendment does not involve a significant hazards consideration pursuant to the provisions of 10 CFR 50.92. The Facility Safety Review Committee has reviewed and concurred with the determination herein.

The NRC approved a similar LAR for Calvert Cliffs Units 1 and 2 (Accession No. ML031320507).

Issuance of this amendment is requested by January 31, 2017, with the amendment to be implemented within 90 days of NRC approval.

In accordance with 10 CFR 50.91(b), a copy of this request is being provided to the State of Connecticut.

Should you have any questions in regard to this submittal, please contact Wanda Craft at (804) 273-4687.

Si nce rely, M. D. Sartain Vice President - Nuclear Engineering l orNOTwAitY ofBLiro r oml~~i 4 31, z18 I

COMMONWEALTH OF VIRGINIA) "

)

COUNTY OF HENRICO)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief. 06 Acknowledged before me thi* _,5idayok 2016.;*#(

My Commission Expires: ,S "3i / */"* f /1 _*_

Notary P5ublic Commitments made in this letter: None.

Attachments:

1. Evaluation of Proposed License Amendment and FSAR Change

2. Marked-Up Technical Specifications Pages

3. Marked-Up Technical Specifications Bases Pages (For Information Only)

4. Marked-Up Chapter 14 ESAR Pages

Serial No: 15-590 Docket No.: 50-336 Page 3 of 3

References:

1. NRC Letter EA-13-188, "Millstone Power Station Unit 2 - Inspection Report 05000336/2015201, Investigation Report No. 1-2012-008, and Apparent Violations,"

dated April 29, 2015 (ADAMS Accession No. ML15119A028).

2. NRC Letter EA-13-188, "Confirmatory Order Related to NRC Report No.

05000336/2015201 and 0I Report 1-2012-008; Millstone Power Station Unit 2,"

dated August 26, 2015 (ADAMS Accession No. ML15236A207).

cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd Suite 100 King of Prussia, PA 19406-2713 Richard V. Guzman NRC Senior Project Manager U.S. Nuclear Regulatory Commission, Mail Stop 08 C2 One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No: 15-590 Docket No. 50-336 ATTACHMENT I EVALUATION OF PROPOSED LICENSE AMENDMENT AND FSAR CHANGE DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 1 of 21 Evaluation of Proposed License Amendment 1.0 Summary Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests an amendment to Facility Operating License No. DPR-65 for Millstone Power Station Unit 2 (MPS2). The proposed amendment would revise MPS2 Technical Specification (TS) 3.5.2, "Emergency Core Cooling Systems, ECCS Subsystems - Tavg - 3000 F, to remove credit for the charging pumps and eliminate Surveillance Requirement (SR) 4.5.2.e.

The proposed amendment would also revise Chapter 14, Section 14.6.1, "Inadvertent Opening of a Pressurized Water Reactor Pressurizer Pressure Relief Valve," of the MPS2 Final Safety Analysis Report (FSAR) to reflect the results of a new long-term analysis for the Inadvertent Opening of Pressurizer Pressure Relief Valve (IOPPRV) event that does not credit charging flow. The new long-term analysis demonstrates that for the IOPPRV event, the High Pressure Safety Injection (HPSI) pumps, with no credit for a charging pump, are sufficient to prevent a long-term core uncovery. Changes to the FSAR are proposed with respect to the FSAR as written prior to the 2009 change that was the subject of Confirmatory Order EA-13-188. The proposed amendment also includes a revision to Section 14.0.11 of the current MPS2 FSAR to clarify the existing discussion regarding the application of single failure criteria.

An update to the associated TS Bases is included to address the proposed TS change.

The TS Bases are being revised to identify that the charging pumps do not meet the four criteria of 10 CFR 50.36(c)(2)(ii) and should not be included in TSs.

2.0 Proposed License Amendment 2.1 TS Change to Remove Credit for Charging from ECCS Subsystems The proposed amendment would revise MPS2 TS 3.5.2, to remove credit for the charging pumps and eliminate Surveillance Requirement 4.5.2.e in its entirety. TS Bases Section 3/4.5.2 will be revised to: 1) identify that charging pumps do not meet the four criteria in 10 CFR 50.36(c)(2)(ii) and therefore should not be included the TSs and,

2) to refer to the updated Probabilistic Risk Assessment (PRA) Maintenance Rule scoped functions, which specifies the charging pumps as not risk significant. The marked-up TS page and TS Bases pages are provided in Attachments 2 and 3, respectively.

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 2 of 21 2.2 FSAR Chanqes to Remove Credit for Charging from Chapter 14 and Clarify Single Failure Criteria Application DNC proposes to revise the MPS2 FSAR as written prior to the 2009 change to: 1) remove credit for charging pumps from the IOPPRV event from Section 14.6.1 based on a new long-term analysis of the event, 2) clarify the limiting event of the IOPPRV analysis, and 3) remove credit for charging pumps from Table 14.0.9-1 for the IOPPRV event. The proposed change also includes a revision to the current MPS2 FSAR Section 14.0.11 to clarify the discussion regarding application of single failure criteria.

The marked up ESAR pages are provided in Attachment 4.

3.0 Background 3.1 System Description 3.1.1 Emergency Core Cooling System FSAR Section 6.3 provides a description of the MPS2 safety injection system. As stated, the safety injection system is an integrated system comprised of high and low pressure centrifugal injection pumps and passive accumulators. The combined capabilities of these three subsystems meet the requirements of 10 CFR 50.46 and AEC General Design Criteria 35, 36, and 37.

Under the original plant design basis, charging flow was credited in the safety analysis for events requiring safety injection, and thus, the charging pumps were designed to automatically start on a safety injection actuation signal. Subsequent revisions to the analysis bases removed credit for the limited flow produced by the positive displacement charging pumps. FSAR Section 6.3 reflects this change. However, the automatic start signal for the charging pumps was retained as a defense-in-depth measure. As such, the charging pumps, not in pull-to-lock, continue to auto-start in response to a safety injection actuation signal.

3.1.2 Chargingl System The charging pumps are sub-components of the chemical and volume control system that is discussed in detail in FSAR Section 9.2. During normal operation, the charging pumps support the following functions: cleanup of the Reactor Coolant System (RCS),

RCS inventory control, and reactivity control. One pump is normally in service with two pumps aligned for standby operation. One of the standby pumps is normally in pull-to-lock to prevent simultaneous auto-start of three charging pumps on a safety injection actuation signal. Administrative controls for pump and subsystem functionality requirements are provided in the MPS2 Technical Requirements Manual (TRM). These controls specify the minimum complement of pumps and flow paths required to support the FSAR-described charging system functions. The TRM also prescribes periodic surveillance requirements to ensure the charging pumps are capable of performing their intended boration functions. The TRM includes surveillance requirements for

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 3 of 21 verification of pump flow capability in a manner identical to that prescribed in TS SR 4.5.2.e.

DNC has completed an analysis of the IOPPRV event for MPS2 using AREVA's NRC-approved S-RELAP5 small break Loss of Coolant Accident (LOCA) methodology (References 1, 2, and 3) to assess the long-term consequences of the postulated event.

The analysis results demonstrate that the HPSI pumps, with no credit for a charging pump, are sufficient to prevent a long-term core uncovery. In addition, a review of the MPS2 Analyses of Record (AOR) for the FSAR Chapter 14 events was performed and concluded that flow from the charging pumps is not credited for event mitigation. As a result, the charging pumps do not meet the first three criteria of 10 CER 50.36(c)(2)(ii) as design basis accident or transient mitigation equipment required to be controlled by TSs.

The Emergency Core Cooling System (ECCS) function of the charging pumps has been determined to be non-risk significant in accordance with the requirements of 10 CFR 50.65 (i.e., Maintenance Rule). As a result, the charging pumps do not meet the fourth criterion of 10 CFR 50.36(c)(2)(ii). Therefore, it is not necessary to include operability

requirements for the charging pumps in TSs.

3.2 Bases for Current TS Requirement and FSAR Chapter 14 Safety Analysis Flow from the charging pumps was previously credited for design basis accident mitigation; therefore, charging was included in TS 3.5.2 as a subsystem of the ECCS.

The current ESAR Chapter 14 safety analyses for the Main Steam Line Break (MSLB) and LOCA do not take credit for the charging pumps, as approved by the NRC in MPS2 TS Amendment 283, dated September 9, 2004. This amendment was based on DNC's application dated May 7, 2002, as supplemented by letters dated April 7, 2003 and July 19, 2004. In the original amendment request, DNC proposed retaining the requirement for charging pump operability due to risk significance in accordance with 10 CFR 50.36(c)(2)(ii) as determined in the then-current PRA model. While the changes proposed in DNC's application were under NRC review, revisions were made to the MPS2 PRA model that allowed the charging pumps to be reclassified as non-risk significant. This reclassification occurred late in the NRC review process and was discussed in DNC's supplement dated July 19, 2004. Since the NRC review was nearly complete at the time of the PRA model update, DNC elected to retain the requirements for charging pump operability in TS 3.5.2..

Subsequent to the issuance of MPS2 License Amendment 283 on September 9, 2004, DNC determined that credit for charging pumps was referenced in the long-term core uncovery portion of the IOPPRV analysis in ESAR Section 14.6.1. In 2009, Under the provisions of 10 CFR 50.59, DNC processed a change to this section to remove credit for charging pumps. In NRC Inspection Report 05000336/2015201, dated April 29, 2015, the removal of credit for charging from FSAR Section 14.6.1 was identified as an apparent violation of 10 CFR 50.59. Operability Determination 000582 was completed to establish that the charging system was operable. Standing Order 14-016, dated May

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 4 of 21 11, 2014, was implemented by DNC to establish requirements for the charging system similar to those that were in effect prior to License~Amendment 283. The operability determination and standing order have been updated to ensure consistency with the August 26, 2015 Confirmatory Order EA-1 3-1 88 actions related to this issue.

3.3 Reason for the Proposed Amendment The proposed amendment aligns TS operability requirements with the design basis of the plant and avoids unnecessary shutdown transients linked to Structures, Systems, and Components (SSCs) that are beyond the scope of 10 CFR 50.36. Removing the requirement for the charging pumps to be operable in TS 3/4.5.2 also clarifies actions associated with assessing ECCS subsystem operability and reportability when charging pumps are unavailable.

4.0 Technical Evaluation 4.1 Basis for the Proposed Amendment 4.1.1 Technical Review of FSAR Chapter 14 for Charging System A review of the M~PS2 AOR for FSAR Chapter 14 events was performed to determine if credit was taken for the charging system for event mitigation. The purpose of this review was to substantiate whether the charging system meets the criteria of 10CFR50.36(c)(2)(ii) for inclusion in the TSs.

In addition to the IOPPRV event discussed below, two other ESAR Chapter 14 events once credited the charging pumps for event mitigation, including the MSLB and the LOCA. Over time, credit for flow from the charging pumps has been removed from these safety analyses. The current MPS2 AOR MSLB (FSAR Section 14.1.5), small break LOCA (FSAR Section 14.6.5.2), and large break LOCA (FSAR Section 14.6.5.1) do not credit the charging system for event mitigation. DNC submitted a proposed License Amendment Request (LAR) to the NRC on September 1, 2015 for the small break LOCA reanalysis. This reanalysis does not take credit for the charging system for event mitigation (Reference 4).

With a new analysis for the long-term response for the IOPPRV event in FSAR Section 14.6.1 that does not credit flow from the charging pumps, MPS2 does not credit the charging pumps for mitigation of any FSAR Chapter 14 events. Details of the new analysis of the long-term response to the IOPPRV event, where no credit is taken for charging pump flow, is provided in Section 4.1.4 below.

4.1.2 Revised PRA Model The charging pumps were once considered risk significant in accordance with 10 CFR 50.36(c)(2)(ii) and were therefore retained in TS 3.5.2. The term "risk significant" is

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 5 of 21 used to categorize SSCs in accordance with the requirements of 10 CFR 50.65 (Maintenance Rule). NUMARC 93-01, endorsed by Regulatory Guide 1.160, defines risk significant as those SSCs that are significant contributors to risk as determined by PRAllndividual Plant Examination or other methods. At the time of the LAR submittal that resulted in License Amendment 283, the charging pumps met the quantitative criteria established in NUMARC 93-01 for categorization as a risk significant SS0 using the approved PRA model.

Subsequent to the LAR submittal that resulted in License Amendment 283, the PRA model was revised and the charging pump risk importance measures no longer met the NUMARC 93-01 quantitative criteria. This change in risk categorization was reflected in the response to an NRC Request for Additional Information (RAI), which stated ". .. risk evaluations have been performed to demonstrate that the charging system is not risk significant as defined in 10 CFR 50.36(c)(2)(ii) Criterion 4." In the response to the RAI, DNC did not request to remove the charging pump requirements from TS 3.5.2, since it may have delayed amendment approval or required submittal of a new LAR with another full review cycle. Therefore, the charging pump operability requirements were retained in TS 3.5.2, as approved by the NRC in License Amendment 283.

The current maintenance rule risk-ranking evaluation of the ECCS function of the charging system confirms that the charging system remains non-risk significant, and therefore does not meet Criterion 4 of 10 CFR 50.36(c)(2)(ii) for inclusion in the TSs.

4.1.3 DNC Analysis of the Short-Term Portion of the IOPPRV Event DNC provided an analysis of the short-term portion of the IOPPRV event (Reference 5, as supplemented by Reference 6), in accordance with the AREVA NRC-approved methodology (Reference 7). The short-term analysis remains unchanged. The results of the IOPPRV analysis are presented in FSAR Section 14.6.1.6. The IOPPRV event is a moderate frequency (i.e., Condition II) event analyzed to ensure that the Departure from Nucleate Boiling Ratio (DNBR) and fuel centerline melt design limits are not violated as a result of the rapid RCS depressurization caused by the IOPPRV. The analysis of this short-term portion of the event assumed that the charging pumps were disabled and letdown was maximized at the initiation of the event to maximize the RCS depressurization and therefore maximize the challenge to the fuel design limits. As such, the short-term response of the plant to an IOPPRV does not credit charging pumps. NRC approval of this analysis was based on the demonstration that the DNBR and fuel centerline melt specified acceptable fuel design limits were not violated (Reference 8).

4.1.4 DNC Analysis of the Long-Term Portion of the IOPPRV Event DNC provided an analysis of the long-term portion of the IOPPRV event (Reference 5, as supplemented by Reference 6), in accordance with the AREVA NRC-approved methodology (Reference 7). Charging and HPSI flow were previously credited for mitigating the long-term portion of the IOPPRV event for MPS2. This previous long-

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 6 of 21 term analysis demonstrated that the capacity of the charging and HPSI pumps was sufficient to compensate for the loss of primary coolant mass through the IOPPRV such that there was no core uncovery. With no core uncovery, there is no long-term challenge to the DNBR and fuel centerline melt specified acceptable fuel design limits.

This MPS2 long-term analysis has been revised using the NRC-approved S-RELAP5 small break LOCA methodology (References 1, 2, and 3). The charging system is not credited in this analysis. The revised analysis assumed an open Pressurizer Safety Valve (PSV) flow area of 0.01767 ft 2 , which bounds the physical area of two pressurizer Power-Operated Relief Valves (PORVs). The results demonstrate that flow from two HPSI pumps, with no credit for a charging pump, is enough to keep the reactor core covered with a two-phase liquid/vapor mixture and preclude the heatup of the fuel rods.

The IOPPRV event is postulated to occur as a result of an electrical or mechanical failure. Table 1 provides the sequence of events for the long-term IOPPRV event. The opening of a pressurizer pressure relief valve or safety valve results in a blowdown of primary coolant as steam through the faulted valve(s). Primary system pressure drops rapidly to a pressure determined by the saturation curve at the temperature of the coolant in the upper vessel head. A reactor scram will occur on Thermal Margin/Low Pressure (TM/LP) before any significant loss of primary system inventory, terminating the challenge to specified acceptable fuel design limits. The variable portion of the TM/LP reactor trip is not modeled in this long-term analysis, and a floor value of 1700 psia is conservatively credited with a harsh containment environment uncertainty. The rapid RCS depressurization results in the generation of a safety injection actuation signal. HPSI flow is subsequently delivered to the RCS when RCS pressure decreases below the shutoff head of the HPSI pumps (approximately 1200 psia) at 178.5 seconds.

RCS pressure remains above the safety injection tank pressure and the shutoff head of the low pressure safety injection pumps. The minimum liquid inventory in the reactor vessel is 151,649 pounds-mass (Ibm) at 898 seconds into the transient. After this time, there is a steady increase in the vessel inventory. A two-phase mixture is present over the entire length of the active fuel for the duration of the transient.

As indicated in Figure 1, the pressurizer pressure relief valve opens at t=0 seconds and initiates a subcooled depressurization of the primary system. Pressurizer level indication is shown in Figure 2. As indicated, the pressurizer water level increases as RCS pressure decreases and reaches the approximate elevation of the upper level tap, causing an indication of about 100% pressurizer level. A two-phase mixture covers the entire length of the active fuel as shown in Figure 3. The liquid inventory in the reactor vessel continues to drop until a minimum mass of 151,649 Ibm is reached at 898 seconds. The collapsed liquid level above the core, as a function of time after the initiation of the transient, is shown in Figure 4. HPSI flow is sufficient to keep the boil-off from uncovering the core. Since the reactor core does not become uncovered, all parts of the core are quenched by a low quality mixture, and no fuel rod heatup occurs as indicated in Figure 5.

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 7 of 21 As depicted in Figure 2, the pressurizer fills following the IOPPRV event. The MPS2 PORVs and PSVs are not qualified for two-phase liquid/vapor mixture or water relief at high pressure (Reference 9). However, since the RCS pressure at the time the pressurizer fills is less than 1200 psia, the remaining PSVs and PORVs would not open and relieve two-phase liquid/vapor mixture or liquid. As such, there would be no further challenge to this RCS fission product barrier filling the pressurizer beyond that caused by the initiating event, the IOPPRV.

The analysis results demonstrate that in the case of an IOPPRV, flow from two HPSI pumps, with no credit for the charging pumps, is sufficient to prevent a long-term core uncovery. Therefore, this analysis demonstrates that charging flow is not needed to prevent long-term core uncovery for the lOP PRV event.

Table 1 Sequence of Events for Long-Term IOPPRV Event with Two HPSI Pumps Event Time (s)

Event initiation (break open) 0 Pressurizer pressure reached low pressure trip setpoint (1700 psia) 62.2 Reactor trip, MFWV terminated, and turbine tripped. 63.6:

SIAS issued 83.31 HPSI flow available 93.3 HPSI flow. began . ... 178.5 Loss of Subcooling Margin 212.2 RCPs tripped 332.2 Motor-driven AFW delivery began 472 Minimum RV mass occurred 898 Calculation terminated 10,000 MFW: Main Feedwater AFW: Auxiliary Feedwater RCP: Reactor Coolant Pump RV: Reactor Vessel

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 8 of 21 Figure 1 System Pressure (Two HPSI Pumps) 2500.0

--- Presaunz.er Pressure

...-- = BG-1 Prenaur

- - -.

SG-2 Pressure 2000.0 S1500.0 10000.0 "lime(seec)

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 9 of 21 Figure 2 Pressurizer Level Indication (Two HPSI Pumps) 100.0 v r," i* v"' "

-- Pressurizer Level 80.0 N

80.0 40.0 0"0.0 2000.0 4000.0 8000.0 8000.0 10000.0 Time (sac)

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 10 of 21 Figure 3 Hot Assembly Collapsed Liquid and Mixture Levels (Two HPSl Pumps)

. . . .... ........ . . . . . ........ .... ... . ! -- --*-o.- * .... . . ... ;:o --- l ........ . ... .... .

11.0

-- 0 HA Collapsed Liquid L

,----:HAMbcturo Level 10.0 8.0 002000.0 4000.0 6000.0 8000.0 10000.0 Time (eec)

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 11 of 21 Figure 4 Collapsed Liquid Level Above Core (Two HPSI Pumps) 14.0

-- Liquid ~iCore Exit and Upper Plenum 12.0 S10.0 0

-J

8.0 I

4.o0 2000.0 4000.0 6000.0 8000.0 10000.0 "lime(sac)

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 12 of 21 Figure 5 Hot Rod Temperature (Two HPSI Pumps) 700.0 650.0

___ Hot Rod @ 9.52 ft 8 00.0 550.0 500.00.0-2000.0 4000.0 6000.0 ~

8000.0 8000.0100. 10000.0 Time (s)

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 13 of 21 4.1.5 Application of a Single Failure to the Long Term Portion of the IOPPRV Event The NRC-approved AREVA methodology for analyzing the IOPPRV event is documented in Appendix B, Section 15.6.1 of Reference 7. In addition to the analysis described above for short-term response to the IOPPRV event, this method identifies that the potential for core uncovery in the long-term is evaluated by considering the capacities of the charging and HPSI pumps and comparing these capacities against the inventory loss rate due to the stuck open valves. The methodology does not reference the postulation of a single failure. NRC approval of the methodology used in this analysis is documented in the safety evaluation dated July 13, 1990 and is included in Appendix B of Reference 7. For the long-term portion of the IOPPRV event, the NRC safety evaluation indicates that the possibility for core uncovery is calculated by considering the capacities of the charging pumps and HPSI pumps. The NRC approval indicates that single failure will be discussed on a plant-specific basis.

The NRC plant-specific safety evaluation for the MPS2 IOPPRV event is documented in Reference 8. This plant-specific safety evaluation does not reference prevention of core uncovery as part of the basis for approval. The IOPPRV event is a moderate frequency, or Condition II, event. All FSAR Chapter 14 events that result in a reactor trip assume that the reactor trip will occur considering a single failure in the Reactor Protection System (RPS). The IOPPRV event with the postulation of a single failure (other than within the RPS) would be categorized as an accident that would be bounded by the results of the small break LOCA presented in ESAR Section 14.6.5.

4.2 Proposed Chanqes to Technical Specification 3/4.5.2 The charging pumps do not meet any of the four criteria of 10 CFR 50.36(c)(2)(ii) and consequently do not need to be included in the MPS2 TSs. Therefore, the proposed change to TS 3/4.5.2 is to delete SR 4.5.2.e, which is the SR for flow from the charging pump.

Specifically, the current wording of SR 4.5.2.e:

"By verifying the delivered flow of each charging pump at the required discharge pressure is greater than or equal to the required flow when tested pursuant to Specification 4.0.5."

is being replaced with:

"Deleted" TS Bases Section 3/4.5.2, "ECCS Subsystems," will be updated to reflect the proposed TS change to remove credit for charging pumps for design basis accident mitigation.

The marked-up TS Bases pages are provided for information only and will be'

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 14 of 21 incorporated in accordance with the TS Bases Control Program upon approval of this amendment request. The current wording of TS Bases Section 3/4.5.2 in part states:

"Flow from the charging pumps is no longer required for design basis accident mitigation. The loss of coolant accident analysis has been revised and no credit is taken for charging pump flow. As a result, the charging pumps no longer meet the first three criteria of IOCFR 50.36 (c)(2)(ii) as design basis accident mitigation equipment required to be controlled by Technical Specifications. In addition, risk evaluations have been performed to demons trate that the charging system is not risk significant as defined in IOCFR 50.36(c)(2) (ii) Criterion 4. However, the charging system is credited in the PRA model for mitigating two beyond design basis events, Anticipated Transients Without Scram (A TWS) and Complete Loss of Secondary Heat Sink. On this basis, the requirements for charging pump OPERABILITY will be retained in Technical Specification 3.5.2. Consistent with the surveillance requirements, only the charging pump will be included in determining ECCS subsystem OPERABILITY.

As a result of the risk insight, the charging pump will be included as an Emergency Core Cooling System subsystem required by Technical Specification 3.5.2. That is, an ECCS subsystem will include one OPERABLE charging pump.

The charging pump credited for each ECCS subsystem must meet the surveillance requirements specified in Section 4.5.2. Consistent with the risk insights, automatic start of the charging pump is not required for compliance to TS 3.5.2. Thus, Section 4.5.2 does not specify any testing requirements for the automatic start of the credited charging pump. Similarly, since the ECCS flow path is not credited in the risk evaluation, there are no charging flow path requirementsincluded in TS 3.5.2.

The requirements for automatic actuation of the charging pumps and the associated boration system components (boric acid pumps, gravity feed valves, boric acid flow path valves), which align the boric acid storage tanks to the charging pump suction on a SIAS, have been relocated to the Technical Requirements Manual These relocated requirements do not affect the OPERABILITY of the chargingpumps for Technical Specification 3. 5.2."

will be replaced with:

"Charging pumps were originally classified as an ECCS subsystem, but over time, the flow from the pumps was removed from the safety analysis. Therefore, flow from the charging pumps is no longer required for design basis accident mitigation. The loss of coolant accident analysis has been revised and no credit is taken for chargingpump flow. As a result, the chargingpumps no longer meet the first three criteria of IOCFR 50.36(c) (2) (ii) as design basis accident mitigation equipment required to be controlled by Technical Specifications. In addition, risk evaluations have been performed to demonstrate that the charging system is not

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 15 of 21 risk significant as defined in IOCFR 50.36(c) (2) (ii) Criterion 4. Therefore, the chargingpumps are no longer included as an ECCS subsystem.

The requirements for automatic actuation of the charging pumps and the associatedboration system components (boric acid pumps, gravity feed valves, boric acid flow path valves), which align the boric acid storage tanks to the charging pump suction on a SIAS, and the requirement to periodically verify pump flow, have been relocated to the Technical Requirements ManuaL" In addition, the following paragraph will be deleted in its entirety to align with removing SR 4.5.2.e:

"Surveillance Requirement 4.5.2.e, which addresses periodic surveillance testing of the charging pumps to detect gross degradation caused by hydraulic component problems, is required by the ASME OM Code. For positive displacement pumps, this type of testing may be accomplished by comparing the measured pump flow, discharge pressure and vibration to their respective acceptance criteria. Acceptance criteria are verified to bound the assumptions utilized in accident analyses. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test point is greaterthan or equal to the performance assumed for mitigation of the beyond design basis events. The surveillance requirements are specified in the Inser'vice Testing Program. The ASME OM Code provides the activities and frequencies necessary to satisfy the requirements."

4.3 Proposed Changes to the FSAR As a result of the technical bases described above, this amendment proposes changes to FSAR Chapter 14, as published prior to the 2009 change, to remove credit for charging pumps from the IOPPRV event in ESAR Section 14.6.1.6 and FSAR Table 14.0.9-1 that were inappropriately incorporated under 10 CFR 50.59. The proposed ESAR changes are consistent with the description provided above and are included in .

The proposed changes prior to the 2009 version of the FSAR include:

FSAR Table 14.0.9-1, Section 14.6 for the IOPPRV event will be revised to replace "Charging and Safety Injection System" with "High Pressure Safety Injection System."

In FSAR Section 14.6.1.6, "Analysis Results" for the IOPPRV event, the sentences:

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 16 of 21 "The charging and SISs have been shown to have sufficient capacity to easily compensate for the loss of primary coolant mass through the inadvertentopening of the pressurizerpressure relief valves. Therefore, the core is not expected to uncover during this event."

are being changed to:

"The HPSI system has been shown to have sufficient capacity to compensate for the loss of primary coolant mass through the inadvertent opening of the pressurizerpressure relief valves. Analysis has shown that core uncovery does not occur during this event."

In addition, a change to the current version of FSAR Section 14.0.11 on application of the single failure is proposed. Specifically, the sentences:

"Except for the steam generator tube rupture, design basis event scenarios considered in the safety analysis depend on single failure criteria. The following single failure criteria are assumed in the safety analysis for Millstone 2:

1. The RPS is designed with redundancyand independence to assure that no single failure or removal from service of any component or channel of a system will result in the loss of the protection function.

2. Each ESF [Engineered Safety Feature] is designed to perform its intended safety function assuming a failure of a single active component.

3. The onsite power system and the offsite power system are designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in eitherpower system.

The safety analysis is structured to demonstrate that the plant systems design satisfies these single failure criteria. The following assumptions result:

1. The ESFs required to function in an event are assumed to suffer a worst single failure of an active component.

2. Reactor trips occur at the specified setpoint within the specified delay time assuming a worst single active failure."

are being changed to:

"All event scenarios considered in the safety analysis depend on the following single failure criteriain the RP'S:

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 17 of 21 The RPS is designed with redundancy and independence to assure that no single failure or removal from service of any component or channel of a system will result in the loss of the protection function. For each event, the reactortrips occur at the specified setpoint within the specified delay time assuming a worst single active failure.

Except for the steam generatortube rupture, design basis accident (limiting fault event) scenariosconsidered in the Millstone 2 safety analysis depend on one of the following additionalsingle failure criteria:

1. Each ESF is designed to perform its intended safety function assuming a failure of a single active component. Forthese events, the ESFs required to function in an event are assumed to suffer a worst single failure of an active component.

2. The onsite power system and the offsite power system are designed such that each shall independently be capable of providingpower for the 6SF assuming a failure of a single active component in eitherpower system."

FSAR Chapter 14 events that result in a reactor trip assume that the reactor trip will occur considering a single failure in the RPS. This change clarifies that the postulation of an additional single failure is applied to design basis accident scenarios or limiting fault event scenarios. It is not applied to all FSAR Chapter 14 event scenarios. Though this is the current intention of the FSAR, this proposed change clarifies that an additional single failure is not applied to moderate frequency, or Condition II, events.

Additionally, this amendment proposes to update the initiating event description of the IOPPRV event in the current version of FSAR Section 14.6.1.1 to identify that the maximum capacity of a single PSV is greater than the maximum capacity of two PORVs. The existing short-term analysis of the IOPPRV event is not affected by this change since that analysis assumed a capacity which bounds the maximum capacity of either one PSV or two PORVs. The revised analysis of the long-term portion of the IOPPRV event used the maximum capacity of a PSV which bounds the maximum capacity of two PORVs.

Specifically, the sentences:

"The event is postulated to occur as a result of the inadvertent opening of one or more pressurizerpressure relief or safety valves due to an electrical or mechanical failure. The limiting event is obtained by assuming the inadvertent opening of both pressurizerpower-operatedrelief valves (PORVs).

are being changed to:

"The event is postulated to occur as a result of the inadvertent opening of one or*

more pressurizerpressure relief or safety valves due to an electrical or

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 18 of 21 mechanical failure. The limiting event is obtained by assuming the inadvertent opening of a pressurizersafety valve which bounds the capacity of two pressurizerpower-operatedrelief valves (PORVs).

4.4 Safety Summary The proposed change to delete SR 4.5.2.e does not involve any modification to the function of the charging pumps or the method of operation for the charging system.

While the charging pumps will continue to start and run in the event of an accident, this feature is not required to be included in the ECCS TS because the function of the charging pumps is no longer credited in any analysis that relies on the EGGS for accident mitigation. Therefore, the deletion of SR 4.5.2.e does not prevent the EGOS subsystems from performing their intended safety function or affect any margins of safety. Additionally, the charging pumps are not significant to public health and safety.

4.5 Precedent By letter dated December 13, 2002, Galvert Gliffs Nuclear Power Plant, Inc. (CCNPP) requested an amendment to Operating License Nos. DPR-53 and DPR-69 to incorporate changes into the TS for Galvert Gliffs Unit Nos. 1 and 2. The proposed amendment revised TS 3.5.2, "EGGS - Operating," by removing the note that modifies the limiting condition for operation. The proposed change removed the requirement to have the charging pumps operable when thermal power is greater than 80% of rated thermal power. The proposed change also removed SR 3.5.2.4 for verifying the required charging pump flow rate. By a letter dated December 3, 2003, the NRG approved CCNPP's amendment request. CCNPP Units 1 and 2 are Gombustion Engineering plants, which are the same vintage as MPS2 and are nearly identical in design and construction.

4.6 Conclusion Flow from the charging pumps is not required for event mitigation. The new IOPPRV long-term analysis for FSAR Section 14.6.1 does not credit charging pump flow. As a result, the charging pumps do not meet the first three criteria of 10 CFR 50.36(c)(2)(ii) as design basis accident mitigation equipment required to be controlled by TSs. In addition, the EGGS function supplied by the charging system is not risk significant.

Gonsequently, the charging pumps do not meet Griterion 4 of 10 GFR 50.36(c)(2)(ii).

Because the charging pumps do not meet any of the four 10 GFR 50.36(c)(2)(ii) criteria, it is proposed that the requirements for charging pump operability be removed from TS 3/4.5.2.

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 19 of 21 5.0 Regulatory Evaluation 5.1 No Significant Hazards Consideration According to 10 CFR 50.92(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 as they relate to this proposed license amendment, and the basis for the conclusion, is provided below.

Criterion I Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The FSAR Chapter 14 accident analyses for MPS2 do not take credit for the flow delivered by the charging pumps. Additionally, the proposed change does not modify any plant equipment or method of operation for any SSC required for safe operation of the facility or mitigation of accidents assumed in the facility safety analyses. Therefore, the proposed amendment will not significantly increase the probability or consequences of an accident previously evaluated.

Criterion 2 Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment does not modify any plant equipment or method of operation for any SSC required for safe operation of the facility or mitigation of accidents assumed in the facility safety analyses. As such, no new failure modes are introduced by the proposed change. Consequently, the proposed amendment does not introduce any accident initiators or malfunctions that would cause a new or different kind of accident.

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 20 of 21 Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3 Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment does not involve a significant reduction in a margin of safety since the proposed changes do not affect equipment design or operation, and no changes are being made to the TS-required safety limits or safety system settings. The proposed changes involve a new safety analysis for the long-term event response for FSAR Chapter 14.6.1, "Inadvertent Opening of a Pressurized Water Reactor Pressurizer Pressure Relief Valve." The analysis demonstrates that flow from two HPSI pumps, with no credit for the charging pumps, is sufficient to prevent long-term core uncovery, and thus there is no challenge to the specified acceptable fuel design limits.

By meeting the MPS2 FSAR Chapter 14 acceptance criteria for a moderate frequency event, there is no significant reduction in the margin of safety.

Conclusion Based upon this discussion, it is concluded that the proposed amendment does not involve a significant hazards consideration.

6.0 Environmental Considerations DNC has evaluated the proposed amendment for environmental considerations, and it is concluded that the change does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 References

1. Topical Report EMF-2328(P)(A), Rev. 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," AREVA, dated March 2001.

2. Topical Report EMF-2328(P), Rev. 0, Supplement 1, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," AREVA, dated March 2012.

3. Letter from M. Gavrilas (NRC) to P. Salas (AREVA), "Final Safety Evaluation by the Office of Nuclear Reactor Regulation for Topical Report EMF-2328(P)(A),

Serial No: 15-590 Docket No: 50-336 Attachment 1, Page 21 of 21 Revision 0, Supplement 1, Revision 0, 'PWR [Pressurized Water Reactor] Small Break LOCA [Loss-of-Coolant Accident] Evaluation Model, S-RELAP5 Based' (TAO NO. ME8227)," dated September 1, 2015. (ADAMS Accession No. ML15210A252).

4. Letter from M. Sartain (Dominion Nuclear Connecticut, Inc.) to U.S. NRC, "Millstone Power Station Unit 2 Proposed License Amendment Request: Small Break Loss of Coolant Accident Reanalysis," dated September 1, 2015. (ADAMS Accession No. ML15253A205).

5. Report EMF-87-161, Rev. 0, "Millstone Unit 2 Plant Transient Analysis Report:

Analysis of Chapter 15 Events", Advanced Nuclear Fuels Corporation, dated September 1988. Attachment 1 to Letter from E. J. Mroczka (Northeast Nuclear Energy Company) to U. S. NRC, "Millstone Nuclear Power Station, Unit No. 2 Disposition of Chapter 15 Events," dated September 19, 1988.

6. Report EMF-87-161, Rev. 0, Supplement 1, "Millstone Unit 2 Plant Transient Analysis Report: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation, dated October 1988. Attachment I to Letter from E. J. Mroczka (Northeast Nuclear Energy Company) to U. S. NRC, "Millstone Nuclear Power Station, Unit No. 2 Cycle 10 Analysis of Chapter 15 Events Supplement 1 TAO

68360," dated October 28, 1988.

7. ANP-84-73(P)(A), Rev. 5, Appendix B, "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Siemens Power Corporation - Nuclear Division, dated July 1990.

8. NRC No. 50-336, "Safety Evaluation by the Office of the Nuclear Reactor Regulation Related to Amendment No. 139 to Facility Operating License No.

DPR-65 Northeast Nuclear Energy Company, et al., Millstone Nuclear Power Station, Unit No. 2," dated March 1989.

9. Safety Evaluation Attached to NRC Letter from David H. Jaffe, "Millstone Unit 2 TMI Action Item II.D.I (Relief and Safety Valve Testing) - TAO No. 44594," dated March 1, 1988.

Serial No: 15-590 Docket No. 50-336 ATTACHMENT 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGE DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

/MLLd:U..IIIII*IIL L, rPi:ye;: I U.I September 9, 2004 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying each Emergency Core Cooling System manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

b. At least once per 31 days by verifying that the following valves are in the indicated position with power to the valve operator removed:

Valve Number Valve Function Valve Position 2-SI-306 Shutdown Cooling Open*

Flow Control 2-SI-659 SRAS Recirc. Open**

2-SI-660 SRAS Recirc. Open**

Pinned and locked at preset throttle open position.

- To be closed prior to recirculation following LOCA.

c. By verifying the developed head of each high pressure safety injection pump at the flow test point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5.

-cd. ---- By-veri-fying-the-developed-head-of-each-low-pressure-safety injection-pump at the -

flow test point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5. --Deleted

e. By verifying the delivre flo.. ofneac.* h charging pump..at the required discharge..

-Specification 41.0,5.

f. At least once per 18 months by verifying each Emergency Core Cooling System automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

g. At least once per 18 months by verifying each high pressure safety injection pump and low pressure safety injection pump starts automatically on an actual or simulated actuation signal.

MILLSTONE - UNIT 2 3/4 5-4 MILLTON

- NIT 3/5-4Amendment No. 5S2*, 4-59, 2-36, 283

Serial No: 15-590 Docket No. 50-336 ATTACHMENT 3 MARKED-UP TECHNICAL SPECIFICATIONS BASES PAGES FOR INFORMATION ONLY DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Attachment 3, Page 1 of 3 LBDCR 04-MP2-016 February 24, 2005 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

BASES 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS (continued)

Each Emergency Core Cooling System (ECCS) subsystem required by Technical Specification 3.5.2 for design basis accident mitigation includes an OPERABLE high pressure safety injection (HPSI) pump and a low pressure safety injection (LPSI) pump. Each of these pumps requires an OPERABLE flow path capable of taking suction from the refueling water storage tank (RWST) on a safety injection actuation signal (SIAS). Upon depletion of the inventory in the RWST, as indicated by the generation of a Sump Recirculation Actuation Signal (SRAS), the suction for the HPSI pumps will automatically be transferred to the containment sump. The SRAS will also secure the LPSI pumps. The ECCS subsystems satisfy Criterion 3 of 10 CFR 50.3 6(c)(2)(ii) as SSERT A]design basis accident mitigation equipment.

Flow from the charging pumps is no longer required for design basis accident mitigation. The loss of coolant accident analysis has been revised and no credit is taken for charging pump flow.

As a result, the charging pumps no longer meet the first three criteria of 10CFR 50.36 (c)(2)(ii) as design basis accident mitigation equipment required to be controlled by Technical Specifications.

qJSERT B3 In addition, risk evaluations have been performed to demonstrate that the charging system is not Srisk significant as dfndin 1OF 036(c)(2)(in) Crtro ..Ho.e.er, the..carginytem* is.....

credited-i-n~-t-hePRAmode! for miiatn wo beyond-design basis evns ntcptd Transients WNithout Sram (ATWS) and Complete oss ,of Seond,,a,.ry Heat Sink. On this basis, the requiremens o charging pump OPERA*B!LITY will be* retaoined in Tecoc,,Specficatonot,,

I determining, E~CCS subsystem OPERABILITY.

8,l*stem subsys..te required by*,Tehnical,* Specifica*tion 3.5.2. That is, an ECCS subsystem will

-- icludeone ,E-.,L cmhaorging p t-nmp7 ,:Tthaingpm q odinc, dl o eac flFFa subsystem g~ pm. Smlrysic ... eECflwptisntrdited in the risk evaluain hr

\JSE RT C The requirements for automatic actuation of the charging pumps and the associated boration system components (boric acid pumps, gravity feed valves, boric acid flow path valves), which align the boric acid storage tanks to the charging pump suction on a SIAS have-been relocate-d-to the Technical Requreens ana. These reocoated requiremenrts-do not af*tth

",..-'i i.;i',.*'*i./i-i--i i i i;i ;ill, i*ii*ii;*iil;* ;;*;iii¢*, i*;i ii.;,iiili;,*%i d* biii;,*l*il ,..',.g..*

MILLSTONE - UNIT 2 B 3/4 5-2a Amendment No. 641-, 7-2, 4-59, 2-1-7-, 2-, 36, Acknowledged by NRC letter dated 6/28/05

Serial No: 15-590 Docket No: 50-336 Attachment 3, Page 2 of 3 Millstone Unit 2 LBDCR 15-MP2-O11 Technical Specification 3.4/5.2 Bases Insertions INSERT A Charging pumps were originally classified as an ECCS subsystem, but over time, the flow from the pumps was removed from the safety analysis. Therefore, flow INSERT B Therefore, the charging pumps are no longer included as an ECCS subsystem.

INSERT C

,and the requirement to periodically verify pump flow, have been relocated to the Technical Requirements Manual.

Docket No: 50-336 Attachment 3, Page 3 of 3 June 19, 2007 LBDCR 07-MP2-014 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS')

BASES 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS (continued)

Surveillance Requirement 4.5 .2.a verifies the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths to provide assurance that the proper flow paths will exist for ECCS operation. This surveillance does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time.

This surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day frequency is appropriate because the valves are operated under procedural control and an improper valve position would only affect a single train. This frequency has been shown to be acceptable through operating experience.

Surveillance Requirement 4.5.2.b verifies proper valve position to ensure that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removing power to the valve operator ensures that the valves cannot be inadvertently misaligned or change position as the result of an active failure. A 31 day frequency is considered reasonable in view of other administrative controls ensuring that a mispositioned valve is an unlikely possibility.

Surveillance Requirements 4.5.2.c and 4.5.2.d, which address periodic surveillance testing of the ECCS pumps (high pressure and low pressure safety injection pumps) to detect gross degradation caused by impeller structural damage or other hydraulic component problems, is required by the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASMIE OM Code). This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the unit safety analysis. The surveillance requirements are specified in the Inservice Testing Program. The ASME OM Code provides the activities and frequencies necessary to satisfy the requirements.

Su..... anc Requiremen,*,*,,t 1 5 e hic;,h addresses...

'? perioi sur.eillac testing of the charging pupstodetec gross dgradaion cased b hyrali componet problems, is~ required:. by the ASME MCoder.. ,,0. For positive displacement pumps*, this...type. of testing may, .. be accomplish:ed by compring~~r th measuedr pump. flox di.srcoharge pressure andiratior;r~~n to their respective surveillance requirements are.. specifie in the In.....vie Testing Program. The ASM O Code.

provides t~he-activities and frequencies necessa3, to satisfyth rqirements.

MILLSTONE - UNIT 2 B 3/4 5-2b Amendment No. 4-5, 6-1-, 7-2,4-1-59, 4-8-,

24-5,21-6, 21-7, 22q, 227, -246,284t-

Serial No: 15-590 Docket No. 50-336 ATTACHMENT 4 MARKED-UP CHAPTER 14 FSAR PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No: 15-590 Docket No: 50-336 MPS-2 FSAR Attachment 4, Page 1 of 5 TABLE 14.0.9 - 1 OVERVIEW OF PLANT SYSTEMS AND EOUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT

. ..... .................. ....... i ......................................................

CONDITIONS Event Reactor Trip Funtions Other Signals and Equipment Inadvertent Operation of the ECCS / High- Power Trip Pressurizer Safety Valves CVCS Malfunction that Increases Therr aal Margin / Low-Pressure Trip Overpressurization Mitigation System Reactor Coolant Inventory High Pressurizer Pressure Trip

[High Pressure 14.6 Decrease in Reactor Coolant Inventory afty Injection System Inadvertent Opening of a PWR High- Power Trip Pressurizer Pressure Relief Valve Therr lal Margin / Low-Pressure Trip Pressurizer Heaters Steam Generator Tube Failure Therr aal Margin/Low-Pressure Trip Steam Generator Safety Valves Safer Injection Actuation Signal Main Steamline Isolation Valves Atmospheric Steam Dump Controller Steam Bypass to Condenser Controller Auxiliary Feedwater System Small-Break Loss-of-Coolant Therr tal Margin/Low-Pressure Trip Emergency Core Cooling System Accidents Resulting from a Spectrum Safet' Injection Actuation Signal Auxiliary Feedwater System of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary Low] '*eactor Coolant Flow Trip Containment Isolation Containment Spray and Air Cooler Rev. 24.10 Page 7 of 8

Serial No: 15-590 Docket No: 50-336 MPS-2 FSAR Attachment 4, Page 2 of 5 14.0.10 EFFECTS OF MIXED ASSEMBLY TYPES AND FUEL ROD BOWING To account for the possibility of a mixed core that includes both fuel assemblies with all HTP spacer grids and fuel assemblies with both HTP and HMP spacer grids in the core, a penalty was included in the AREVA MDNBR Calculations.

In accordance with AREVA rod bow methodology (Reference 14.0-12), the magnitude of rod bow for the AREVA assemblies has been estimated. The calculations indicate that 50% closure of the rod-to-rod gap occurs at an assembly exposure of about 76,450 MWd/MTU for the ARE VA 14 x odpl4 design. Significant impact to MDNBR due to rod bow does not occur until the gap closures exceed 50%. Since the maximum design exposure for AREVA reload fuel in Millstone Unit 2 is significantly less than that at which 50% closure occurs, rod bow does not significantly impact the MDNBR for AREVA fuel. Also, total peaking is not significantly impacted.

[INSERTI 14.0.11 PLANT LICENSING BASIS AND SINGLE FAILURE CRITERIA Except for the steam generator tube uptur, dsinbai evt cearos eonsc analysis depend on single failu' .ri.ri.- Th- folwn

e faiure "igl riri *arc -assumed-in the safcty analysis for Millstone 2:

--eha* assureJI v_ tha noL*LJ sVingl falr"rrmvlfo *evc of anyl~ component orLt Eah S i esgndtoprf.

.i..nene safct "ucto "sunna "alue Thoniepwr ytmadth fst owrsse rcdsge uc htec shl neenetyb cpbeo pro Vidn oe the ESF assuming7 a "alr

_I- - ?- - - - - - - - ?

or a single acuve ~mponen1 in cimer power sysTem.

~1 tr~ n~frnt~' thnf ths' nlnnt ~ c1i~ioii ~ntv~ti~ fh"~ ninof p r~-~~- ~.'

failure criteria. The following assumptions result:

Ib, Ls requre, tluneton in n evet ares to. su..ff.,,,r ad worsJ,-t* ~ing1e

,ume ric-acitur Lrips ~c ir at the specified se4 mt within the specified delay time single active failure.

The assumptions for concurrent loss of offsite power are as follows:

1. The following postulated accidents are considered assuming a concurrent loss of offsite power: main steam line break, control rod ejection, steam generator tube rupture, and LOCA.

14.0-9 14.0-9Rev.

30.2

Serial No: 15-590 Docket No: 50-336 Attachment 4, Page 3 of 5 Millstone Unit 2 LBDCR 15-MP2-011 FSAR Chapter 14 Insertion INSERT "All event scenarios considered in the safety analysis depend on the following single failure criteria in the RPS:

The RPS is designed with redundancy and independence to assure that no single failure or removal from service of any component or channel of a system will result in the loss of the protection function.

For each event, the reactor trips occur at the specified setpoint within the specified delay time assuming a worst single active failure.

Except for the steam generator tube rupture, design basis accident (limiting fault event) scenarios considered in the Millstone 2 safety analysis depend on one of the following additional single failure criteria:

1. Each ESF is designed to perform its intended safety function assuming a failure of a single active component. For these events, the ESFs required to function in an event are assumed to suffer a worst single failure of an active component.

2. The onsite power system and the offsite power system are designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in either power system."

Form No. 731123 (Oct 2d

MPS-2 FSAR Serial No:

Docket No: 15-590 50-336 Attachment 4, Page 4 of 5 14.6 DECREASES IN REACTOR COOLANT INVENTORY 14.6.1 INADVERTENT OPENING OF A PRESSURIZED WATER REACTOR PRESSURIZER PRESSURE RELIEF VALVE 14.6.1.1 Event Initiator The event is postulated to occur as a result of the inadvertent opening of one or more pressurizer pressure relief or safety valves due to an electrical or mechanical failure. The limiting event is obtained by assuming the inadvertent opening of botha- ..... rzc powr........lofvavc 14.6.1.2 Event Description rlevaes(O s)

The opening of the pressurizer pressure relief valve or safety valve results in a blowdown of primary coolant as steam through the faulted valves. Primary system pressure drops rapidly until the pressurizer liquid is depleted, and then quite rapidly to a pressure determined by the saturation curve at the temperature of the coolant in the upper vessel head. Reactor scram will occur on thermal margin/low pressure (TM/LP) before the pressurizer liquid is depleted, terminating the challenge to Specified Acceptable Fuel Design Limits (SAFDLs). In this initial stage, pressurizer heaters would actuate in an attempt to maintain pressure, but would be turned off on a low-level signal before the heater elements were uncovered.

14.6.1.3 Reactor Protection The TM/LP trip provides initial protection against loss of thermal margin and possible fuel damage. Reactor protection for the Inadvertent Opening of a Pressurized Water Reactor (PWR)

Pressurizer Pressure Relief Valve event is summarized in TIable 14.6.1-i.

14.6.1.4 Disposition and Justification The event proceeds as a depressurization of the primary coolant system with a loss of inventory.

The core power and primary loop temperatures are relatively unaffected by the pressure drop.

Thus, a short term challenge to the SAFDLs exists due to the depressurization prior to scram..

There is also a long term concern in that if primary inventory cannot be restored and maintained, core uncovery may result.

The greatest challenge to core uncovery exists at rated power conditions when the core power and primary coolant stored energy are maximized. The greatest challenge to the SAFDLs occurs for the event initiated at rated power where the margin to Departure from Nucleate Boiling (DNB) is minimized.

An evaluation of the SAFDL challenge is also made for 5% power operating conditions in Mode 2 when the TM/LP trip may be bypassed. In this mode, the primary system may depressurize below the TM/LP setpoint pressure without an automatic reactor trip occurring. The Safety 14.6-1 14.6-1Rev.

32

Serial No: 15-590 MPS-2 FSAR Docket No: 50-336 Attachment 4, Page 5 of 5 Injection System (SIS) will, however, be available to inject boron and provide for inventory makeup.

The disposition of events for the Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve event is summarized in Table 14.6.1-2.

14.6.1.5 Definition of Events Analyzed As discussed above, this event is analyzed for Minimum Departure from Nucleate Boiling Ratio (MDNBR) for both Modes 1 (full-power), and 2 (startup). The startup power case is analyzed because the TMIlLP trip can be manually bypassed below 5% power.

The system response for the full-power case was evaluated by using PTSPWR2 (Reference 14.6-1). The full-power event MDNBR was calculated using XCOBRA-IIIC (Reference 14.6-2).

The system response for the startup case was determined by conservative problem constraints.

The maximum power was limited to 7% of the rated power. Above this power the assumed TM!l LP trip bypass is automatically removed. The system pressure is conservatively assumed to be at the core inlet saturation pressure. The core inlet temperature is assumed to be at a level consistent with a maximum power rise of 7% and a conservative time delay before the SIS terminates the event. XCOBRA-IIIC was used with these system responses to predict the hot channel mass flux required for the critical heat flux calculation. The thermal margin was conservatively determined by the Modified Barnett critical heat flux correlation (Reference 14.6-3), with the system pressure reduced to the 725 psia upper limit of the Modified Barnett correlation.

14.6.1.6 Analysis Results The sequence of events for the full-power analysis are given in Table 14.6.1-3. Figures 14.6.1-i to 14.6.1-6 show the transient response for key system variables. The MDNBR for this event initiated from full-power is above the CUF correlation limit. This event does not challenge the FCMLHIR limit. Therefore, LHR is not evaluated... .. .. .. .... . .. ... ...... .. .

The startup mode case resulted in a mimimum critical heat flux ratio of above 10, as calculated by the Modified Barneft correlation. The peak pellet LHR is less than the full-power value. Thus, the startup mode is bounded by the full-power mode.

!*, HPSI System has The e%4-*-t have been shown to have sufficient capacity to easily compensate for the loss of primary coolant mass through the inadvertent opening of the pressurizer pressure relief valves. Thrfrtocr sntcp~d~,"un.-*ox,*. duin th n.

14.6.1.7 Conclusions ,_Analysis has shown that core uncovery does not occur during this event.

The results of the analysis demonstrate that the event acceptance criteria are met since the MDNBR predicted for the full-power case is greater than the DNBR safety limit and the minimum Critical Heat Flux Ratio (CHFR) predicted for the startup mode case is greater than the Modified Barnett Critical Heat Flux (CHF) limit. The correlation limits assure with 95%

14.6-2 14.6-2Rev.

26.1

Dominion Nuclear Connecticut, Inc.

5000omoo DominionBularGeAlnV206

/ D Inn on Web Address: www.dom.com January 25, 2016 U.S Nuclear Regulatory Commission Serial No.15-590 Attention: Document Control Desk NSSL/LES R0 Washington, DC 20555 Docket No. 50-336 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.