ML11286A092

Response to Request for Additional Information to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 Extended Power Uprate LAR

ML11286A092
Person / Time
Site:	Crystal River
Issue date:	10/11/2011
From:	Franke J Florida Power Corp, Progress Energy Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F1011-05, TAC ME6527
Download: ML11286A092 (20)

Text

NProgress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 October 11,2011 3F1011-05 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)

References:

1. CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate" (Accession No. ML112070659)

2. Email from S. Lingam (NRC) to D. Westcott (CR-3) dated August 4, 2011, "CR-3 EPU LAR - Draft RAIs for Acceptance Review from SRXB (TAC No.

ME6527)"

3. Email from F. Saba (NRC) to D. Westcott (CR-3) dated September 22, 2011, "FW: ASME Service Level C"

Dear Sir:

By letter dated June 15, 2011, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt. On August 4, 2011, via electronic mail, the NRC provided a request for additional information (RAI) related to Extended Power Uprate (EPU) accident and transient analyses needed to support the Reactor Systems Branch acceptance review of the CR-3 EPU License Amendment Request (LAR). By teleconferences on August 16, 2011, September 20, 2011, October 3, 2011, and October 6, 2011, CR-3 discussed the RAI with the NRC to confirm an understanding of the information being requested. Additionally, a follow-up clarification regarding the requested information was provided by the NRC on September 22, 2011 via electronic mail. , "Response to Request for Additional Information to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 EPU LAR," provides the CR-3 formal response to the RAI as discussed with the NRC staff. , "List of Regulatory Commitments," includes a regulatory commitment to provide a summary of a feedwater line break overpressure protection analysis in support of an EPU acceptance review RAI response.

In support of the EPU acceptance review RAI responses, five enclosures are provided. , "AREVA 86-9159609-002 - CR-3 Steam Generator Tube Rupture Event for EPU Summary," provides a summary of the Steam Generator Tube Rupture thermal-hydraulic analysis assuming a loss of offsite power. Enclosure 2, "AREVA 86-9167251-001 - Summary of CR3 EPU Inadvertent Pressurizer Relief Valve Opening," provides a summary of an evaluation assessing the effects of inadvertent opening of the Reactor Coolant System (RCS) relief valves. Enclosure 3, "AREVA 86-9168766-001 - CR-3 EPU Inadvertent Engineered Safeguards Actuation (IESA) Analysis Summary," provides a summary of an evaluation assessing the effects of inadvertent operation of the Emergency Core Cooling System.

Progress Energy Florida, Inc.

Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428

U.S. Nuclear Regulatory Commission Page 2 of 3 3F1011-05 , "Markup Pages of Attachment 1 of CR-3 EPU LAR #309, Revision 0," provides an additional CR-3 licensing basis change request to support the CR-3 EPU and Enclosure 5, "Revised Pages of Attachment 1 of CR-3 EPU LAR #309, Revision 0," provides a clean typed revision of the affected pages with revision bars noting the changes and includes repagination.

This RAI response letter requests an additional licensing basis change to use an RCS pressure acceptance criterion for Rod Ejection Accident analyses consistent with the criterion in Section 15.4.8 of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition." The information provided herein does not change the intent or the justification for the requested EPU license amendment (Reference 1). Although minor editorial changes are proposed for the 10 CFR 50.92 evaluation associated with the CR-3 EPU LAR, FPC has determined that this supplement does not affect the conclusion that the proposed license amendment does not involve a Significant Hazards Consideration. Revised wording of the 10 CFR 50.92 evaluation provided in the June 15, 2011 submittal (Reference 1) is included in Enclosures 4 and 5.

If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.

Sin 0on A. Franke Vice President Crystal River Nuclear Plant JAF/gwe Attachments:

1. Response to Request for Additional Information to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 EPU LAR

2. List of Regulatory Commitments

Enclosures:

1. AREVA 86-9159609-002 - CR-3 Steam Generator Tube Rupture Event for EPU Summary

2. AREVA 86-9167251-001 - Summary of CR3 EPU Inadvertent Pressurizer Relief Valve Opening

3. AREVA 86-9168766-001 - CR-3 EPU Inadvertent Engineered Safeguards Actuation (IESA) Analysis Summary

4. Markup Pages of Attachment 1 of CR-3 EPU LAR #309, Revision 0

5. Revised Pages of Attachment 1 of CR-3 EPU LAR #309, Revision 0 xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact

U.S. Nuclear Regulatory Commission Page 3 of 3 3F1011-05 STATE OF FLORIDA COUNTY OF CITRUS Jon A, Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.

{ Jon A. Franke Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this - day of 2011, by Jon A. Franke.

Signature of Notary Public State .,,,* , * .

• .!*CAROLYN E.PORT-MA-NN--

-,:"*,";-Commission # DD 937-553

• .* SExpires March 1,2014 ]

(Print, type, or stamp Commissioned Name of Notary Public)

Personally Produced Known -OR- Identification

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 /LICENSE NUMBER DPR-72 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TO SUPPORT NRC REACTOR SYSTEMS BRANCH ACCEPTANCE REVIEW OF THE CR-3 EPU LAR

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 1 of14 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TO SUPPORT NRC REACTOR SYSTEMS BRANCH ACCEPTANCE REVIEW OF THE CR-3 EPU LAR By letter dated June 15, 2011, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt (Reference 1). The proposed license amendment is considered an Extended Power Uprate (EPU). On August 4, 2011, via electronic mail, the NRC provided a request for additional information (RAI) related to EPU accident and transient analyses needed to support the Reactor Systems Branch acceptance review of the CR-3 EPU License Amendment Request (LAR). By teleconferences on August 16, 2011, September 20, 2011, October 3, 2011, and October 6, 2011, CR-3 discussed the RAI with the NRC to confirm an understanding of the information being requested.

Additionally, a follow-up clarification regarding the requested information was provided by the NRC on September 22, 2011 via electronic mail.

NRC Request for Additional Information Please provide your responses, that are required for our acceptance review, for the following draft RAIs from NRC's Reactor Systems Branch (SRXB):

1. Please supplement Technical Report (TR) Section 2.8.4.2.2 with a discussion of each of the limiting overpressure events and its adherence to the Standard Review Plan (SRP)

Acceptance Criteria discussed in Section 5.2.2, "Overpressure Protection," of NUREG 0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." Where analyses of the limiting events do not adhere or conform to the SRP acceptance criteria, please provide a justification that explains how the plant meets applicable regulatory requirements despite not conforming to the SRP acceptance criteria.

2. Please clarify whether separate analyses for the limiting Decrease in Secondary Heat Removal transient were performed to maximize both the peak reactor coolant system and the main steam system pressure.

3. Please provide steam generator tube rupture analyses that consider loss of offsite power conditions.

4. TR Section 2.8.5.6.2, "Steam Generator Tube Rupture," discusses that the steam lines are hydraulically analyzed for water-filled conditions to the main steam line isolation valves.

Address whether the main steam system pressure relief valves upstream of the main steam line isolation valves are qualified to remain closed under water-filled conditions associated with a steam generator tube rupture, or for liquid water relief.

5. An analysis for the inadvertent opening of a reactor coolant system relief valve has not been provided. Please provide.

6. An analysis for the inadvertent emergency core cooling system actuation has not been provided. Please provide.

7. Please provide sensitivity studies demonstrating that the following assumptions have an insignificant effect on the results of analyzed transients:

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 2 of 14

a. Use of nominal reactor coolant system pressure and pressurizer level

b. Complete mixing in asymmetric transients.

8. The staff is unable to determine the applicability of the 3200 psi acceptance criterion to the control rod ejection event. It is not clear that the Service Level C limit for the CR-3 reactor vessel and connected components is 3200 psi, or that the Service Level C limit is an acceptable acceptance criterion for the control rod ejection accident. Notably, the CR-3 licensing basis (Page 47 of Chapter 14 of the UFSAR) indicates that the acceptance criterion is 110% of the vessel design pressure. Please explain and justify the application of the 3200 psi acceptance criterion.

CR-3 Responses:

1. Please supplement Technical Report (TR) Section 2.8.4.2.2 with a discussion of each of the limiting overpressure events and its adherence to the Standard Review Plan (SRP) Acceptance Criteria discussed in Section 5.2.2, "Overpressure Protection," of NUREG 0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." Where analyses of the limiting events do not adhere or conform to the SRP acceptance criteria, please provide a justification that explains how the plant meets applicable regulatory requirements despite not conforming to the SRP acceptance criteria.

The EPU overpressure protection analyses for CR-3 operational transients were performed in accordance with the CR-3 current licensing basis (CLB) to ensure the maximum primary system pressure remains within 110% of design pressure. Neither the current CR-3 nor the EPU overpressure protection analyses assumptions are fully consistent with the criteria of SRP Section 5.2.2. These analyses demonstrate compliance with the reactor coolant pressure boundary (RCPB) acceptance criterion of 110% of design pressure in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," (SRP) Acceptance Criterion 3.B of Section 5.2.2, "Overpressure Protection," for the CR-3 most severe anticipated operational occurrences.

For feedwater line break (FWLB) accidents, bounding analyses were performed for the most limiting postulated FWLB accident at EPU conditions. These bounding analyses demonstrate compliance with the RCPB acceptance criterion of 120% of design pressure.

Limiting Primary System Overpressure Operational Transient Section 2.8.4.2, "Overpressure Protection During Power Operation," of the CR-3 EPU Technical Report (TR) (Reference 1, Attachment 7) lists the two events that have historically been the limiting over-pressurization events for CR-3. Those events are:

• Loss of Load (Turbine Trip) for secondary system pressure; and

• Uncontrolled Control Rod Assembly Withdrawal from Low Power (Startup Accident) for primary system pressure.

Section 2.8.4.2 of the CR-3 EPU TR incorrectly identifies the Startup Accident as the limiting primary system pressure event at EPU conditions. As indicated in Section

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 3 of 14 2.8.5.2.3, "Loss of Normal Feedwater," of the CR-3 EPU TR (Reference 1, Attachment 7), the Loss of Feedwater (LOFW) event is the limiting primary system overpressure event at EPU conditions; approximately 4 psi higher than the peak pressure from the Startup Accident. Given the small difference in the peak pressure values, both events (LOFW and Startup Accident) were considered when further evaluating the limiting primary system overpressure event.

Limiting Primary System Overpressure Infrequent Event As described in Section 2.8.5.0, "Non-LOCA Analyses Introduction," of the CR-3 EPU TR (Reference 1, Attachment 7), CR-3 does not consider any CLB accident or transient an infrequent event (i.e., American Nuclear Society (ANS) Condition III). The CR-3 CLB FWLB accident is considered an ANS Condition IV accident; an event not expected to occur, but postulated because the consequences would include the potential for the release of significant amounts of radioactive material. FWLB accident analyses were performed at EPU conditions for the most limiting postulated FWLB accident assuming a double-ended guillotine break and confirm that the total relieving capacity of the pressurizer safety valves (PSVs) is sufficient to limit the maximum Reactor Coolant System (RCS) pressure to less than 120% of design pressure. The peak primary system pressure criterion of 120% of design pressure is consistent with the NRC Staff feedback received during the October 6, 2011 teleconference; SRP 5.2.2 Criterion 3.B related to the most severe infrequent event appears to be inconsistent with the preponderance of SRP 5.2.2 and should be 120% of design pressure. A summary of the limiting FWLB analysis is provided in Section 2.8.5.2.4, "Feedwater System Pipe Breaks Inside and Outside Containment," of the CR-3 EPU TR (Reference 1, Attachment 7), including key analysis input assumptions and RCS pressure results.

In addition, CR-3 will provide a summary of a FWLB overpressure protection analysis and will include key analysis input assumptions and RCS pressure results. This additional analysis will be provided by November 11, 2011.

SRP Section 5.2.2 Comparison The CR-3 EPU analyses for the LOFW, FWLB, and Startup Accident events were performed considering the acceptance criteria of SRP Section 5.2.2 with the following exceptions:

The analyses do not include uncertainty limits on initial operating conditions as specified in SRP Section 5.2.2, Criterion 3.B.ii. Nominal values are assumed for initial operating conditions except for pressurizer level as allowed by the NRC approved methodology; BAW-10193NP-A "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water Reactors," (Reference 2). As described in BAW-10193NP-A, Reactor Protection System (RPS) trip setpoints and Engineered Safeguards Actuation System (ESAS) trip setpoints in the analyses are set conservatively high or low depending upon which setting will provide the most conservative result. For example: Instrument uncertainty and margin are added from the Improved Technical Specifications (ITS) RPS RCS High Pressure Allowable

,Value to establish a conservative RCS high pressure reactor trip analysis value and equivalent instrument uncertainty and margin are subtracted from the ITS RPS RCS High Pressure Allowable Value to establish the RPS RCS high pressure trip setpoint.

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 4 of 14 This net effect, coupled with maximum lift tolerance conservatively added to the relief valve lift settings and other conservative boundary conditions (e.g., reactivity coefficients and inserted rod worth), obviates the need for initial RCS pressure to be adjusted for instrument uncertainty. Pressurizer level included instrument uncertainty high or low as indicated by BAW-10193NP-A to maximize system pressure. The NRC approved the use of RELAP5/MOD2-B&W safety analyses code for non-Loss of Coolant Accident (LOCA) events as stated in the associated Safety Evaluation dated October 15, 1999 (Reference 3) and, as allowed by 10 CFR 50.34(h)(2), is considered an acceptable method of complying with applicable regulatory requirements.

CR-3 does not have a second safety-grade signal from the RPS to initiate a reactor trip in the RCS overpressure analyses as specified in SRP Section 5.2.2, Criterion 3.B.iii. The limiting overpressure analyses assume a reactor trip on a RCS high pressure signal. CR-3 overpressure protection analyses do not postulate a failure of an RPS trip on RCS high pressure and is beyond CR-3 design and licensing basis.

However, since the initial design and licensing of CR-3, several related plant improvements have been made for defense-in-depth:

1. Safety-related Anticipatory Reactor Trip System (ARTS) RPS trips were added to address NUREG-0737, Action Item II.K.2.10, "Safety-Grade Anticipatory Reactor Trip." These anticipatory trips are neither credited nor characterized as first safety-related RPS trips, but will provide anticipatory RPS trip signals to the RCS high pressure condition. CR-3 ITS Table 3.3.1-1, Function 9, Main Turbine Trip (Control Oil Pressure), is required to be Operable > 45% Rated Thermal Power (RTP) and Function 10, Loss of Both Main Feedwater Pumps (Control Oil Pressure) is required to be Operable > 20% RTP. Although not explicitly credited, the ARTS RPS trips are considered first reactor trip signals for RCS overpressure protection and, as allowed by 10 CFR 50.34(h)(2), are considered acceptable methods of complying with applicable regulatory requirements.

2. In response to implementation of the Anticipated Transient Without Scram (ATWS) rule requirements, 10 CFR 50.62(c)(2), CR-3 added a Diverse Scram System (DSS). The DSS is independent of RPS and serves as a backup to the RPS RCS high pressure trip ensuring a reactor trip via a diverse method on overpressure ATWS transients. The DSS trips only regulating rods (i.e., Rod Groups 5-7) and is not safety-related. Although the DSS is not designed to maintain the peak pressure below either 110% or 120% of the RCPB design pressure, the NRC concluded in 1989 (Reference 4) that the DSS design was in compliance with the ATWS rule. The DSS does not satisfy the SRP as a means of complying with the acceptance criterion of a second safety-grade RPS signal.

However, the postulated failure of the RPS is fundamentally equivalent to an ATWS event. Thus, as allowed by 10 CFR 50.34(h)(2), the DSS constitutes an acceptable method of complying with the applicable regulatory requirements of providing backup RCS overpressure protection. Analysis results of the Startup and LOFW events, assuming a failure of the RPS RCS high pressure trip, indicate a peak primary system pressure greater than 110% of the RCPB design pressure but less than 120% of the RCPB design pressure and less than the American Society of Mechanical Engineers (ASME) Service Level C stress limits.

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 5 of 14

2. Please clarify whether separate analyses for the limiting Decrease in Secondary Heat Removal transient were performed to maximize both the peak reactor coolant system and the main steam system pressure.

Babcock & Wilcox (B&W) plants do not consider a single limiting Decrease in Secondary Heat Removal transient for purposes of primary and secondary overpressure analyses. As stated in Section 2.8.5.2, "Decrease in Heat Removal by the Secondary System," of the CR-3 EPU TR (Reference 1, Attachment 7), the loss of load (turbine trip) event is the limiting secondary system overpressure transient. As discussed previously, the LOFW event is the limiting primary system overpressure transient.

A single LOFW analysis model and a single turbine trip analysis model were performed for the CR-3 EPU, each using conservative boundary conditions to evaluate system over-pressurization. These boundary conditions are consistent with the NRC approved methodology for performing non-LOCA safety analyses; BAW-10193NP-A (Reference 2).

The limiting Decrease in Secondary Heat Removal transient for peak primary pressure transient is selected by considering the event with the largest mismatch between the short-term integrated reactor power and the secondary heat sink, which for CR-3 at EPU conditions is the LOFW event. The mismatch is larger than other events because it produces the fastest reduction in main feedwater flow (3 seconds) and the longest delay before the reactor trip (15 seconds).

LOFW EPU Analysis Considerations A summary of the LOFW event analysis is provided in Section 2.8.5.2.3 of the CR-3 EPU TR (Reference 1, Attachment 7). In the LOFW analysis timeline, the turbine trip occurs following the reactor trip. Consequently, secondary heat removal from the LOFW is further reduced shortly after the turbine trip. As indicated in Figures 2.8.5.2.3-1, "LOFW RCS Pressures - Condition A," and 2.8.5.2.3-4, "LOFW RCS Pressures -

Condition B," of the CR-3 EPU TR. (Reference 1, Attachment 7), this further reduction in secondary heat transfer exacerbates the rate of pressure increase in the primary system.

The LOFW analysis includes several conservative boundary conditions to maximize primary system pressure response, including raising the RCS high pressure reactor trip setpoint above the ITS RPS RCS High Pressure Allowable Value to account for instrument uncertainty and increasing the PSV lift setpoint above the nominal value to account for a maximum setpoint tolerance of 3%. Assuming a reactor trip at a higher RCS pressure, along with biasing initial pressurizer level high and not crediting pressurizer spray, maximizes the RCS pressure rate of increase. In addition, delaying the opening of the PSVs and not crediting the power operated relief valve (PORV) exacerbates the pressure response and results in a conservatively high primary system peak pressure. The synergistic effect of these boundary conditions eliminates the need for additional analytical cases.

The limiting peak secondary pressure transient is selected by considering the event with the longest delay before a reactor trip following a turbine trip. Delaying the reactor trip maximizes the heat transferred to the secondary system, which results in a higher secondary system pressure rate of increase and peak pressure. The only event with a

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 6 of 14 reactor trip occurring following a turbine trip is the turbine trip event. Thus the turbine trip event is considered the limiting peak secondary pressure event.

Turbine Trip EPU Analysis Considerations A summary of the turbine trip analysis is provided in Section 2.8.5.2.1, "Loss of External Load, Turbine Trip, Loss of Condenser Vacuum, and Steam Pressure Regulatory Failure," of the CR-3 EPU TR (Reference 1, Attachment 7). In the turbine trip analysis, all steam flow to the main turbine is terminated upon a turbine trip at the onset of the transient while the reactor continues to operate. The secondary system inventory is relieved by the main steam safety valves (MSSVs) resulting in secondary pressure maximizing and stabilizing as indicated in Figure 2.8.5.2.1-5, "Turbine Trip Steam Line Pressure versus Time," of the CR-3 EPU TR. (Reference 1, Attachment 7). Also in the turbine trip analysis, the initial pressurizer level is biased low and pressurizer spray is modeled with design flow to maximize secondary system pressure response.

Specifically, the analysis model assumptions for initial pressurizer level and pressurizer spray, along with assuming a reactor trip at a higher RCS pressure than nominal, were selected to reduce the RCS pressure rate of increase to delay the reactor trip. Additional conservative boundary conditions to maximize secondary system pressure response include accounting for MSSV setpoint tolerance and valve accumulation and not crediting the Integrated Control System and associated power runback or the opening of the main turbine bypass valves. The synergistic effect of these boundary conditions eliminates the need for additional analytical cases.

3. Please provide steam generator tube rupture analyses that consider loss of offsite power conditions.

Section 2.8.5.6.2, "Steam Generator Tube Rupture," of the CR-3 EPU TR (Reference 1, Attachment 7) provides a summary of the steam generator tube rupture (SGTR) thermal hydraulic (T-H) analysis performed at EPU conditions consistent with the CR-3 CLB, which assumes offsite power is available.

An additional SGTR T-H analysis was performed for CR-3 at EPU conditions assuming a loss of offsite power (LOOP) and a single active component failure that results in the most limiting radiological consequences consistent with Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," (Reference 5) to ensure this event is not limiting for CR-3 control room habitability (CRH) dose at EPU conditions. As stated in Section 2.9.2, "Radiological Consequences Analyses," of the CR-3 EPU TR (Reference 1, Attachment 7), radiological evaluations confirmed that the CRH dose from a LOCA bounds the dose from a SGTR with a LOOP. Enclosure 1, "AREVA 86-9159609-002 -

CR-3 Steam Generator Tube Rupture Event for EPU Summary," provides a summary of the SGTR T-H analyses at EPU conditions (LOOP and non-LOOP) that were used to evaluate dose consequences, including analysis input assumptions, methods, and results.

PEF continues to consider the existing CR-3 SGTR T-H analysis (assuming offsite power is available and no single failure) as the CR-3 CLB and design basis.

4. TR Section 2.8.5.6.2, "Steam Generator Tube Rupture," discusses that the steam lines are hydraulically analyzed for water-filled conditions to the main steam line isolation valves. Address whether the main steam system pressure relief valves

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 7 of 14 upstream of the main steam line isolation valves are qualified to remain closed under water-filled conditions associated with a steam generator tube rupture, or for liquid water relief.

The MSSVs are designed for steam service and not for liquid service. However, the MSSV vendor (Dresser) confirmed that initial unseating of the MSSVs will not occur until the inlet pressure reaches the set-pressure regardless of the fluid condition in the main steam lines (liquid or gaseous).

The CR-3 SGTR T-H analyses do not explicitly prevent steam generator overfill. The CR-3 SGTR T-H analyses terminate upon isolation of the faulted once through steam generator (OTSG). Consequently, CR-3 allows the faulted OTSG to fill after the OTSG is isolated. To avoid re-initiation of the radioactive release and resulting dose consequences via the MSSVs during a SGTR event due to OTSG overfill, the RCS is depressurized and cooled prior to isolating the faulted OTSG to assure the secondary pressure in the OTSG remains well below the pressure required to lift the MSSVs.

Following isolation, the faulted OTSG will eventually achieve water solid conditions, but the T-H analysis indicates that, even when using the most conservative method of RCS depressurization (i.e., pressurizer high point vent valve), RCS pressure will be appreciably lower than any MSSV setpoint when water solid conditions are reached in the OTSG. Therefore, the need for an MSSV to provide liquid relief is not credible. The MSSVs are assumed to remain closed following isolation of the OTSG, and are not evaluated for liquid water relief.

5. An analysis for the inadvertent opening of a reactor coolant system relief valve has not been provided. Please provide.

As indicated in Section 2.8.5.6.1, "Inadvertent Opening of Pressurizer Pressure Relief Valve," of the CR-3 EPU TR (Reference 1, Attachment 7), the inadvertent opening of the RCS relief valves is not a CR-3 design or licensing basis accident or transient. Consistent with SRP Section 15.6.1, "Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve or a BWR Pressure Relief Valve," an analysis assessing the effects of the inadvertent opening of an RCS relief valve has been performed and confirmed that RCS pressure will not exceed 110% of RCS design, departure from nucleate boiling ratio (DNBR) will remain above the 95/95 DNBR limit, and the event will not develop into a more serious plant condition without other faults occurring independently. A more serious plant condition, such as core uncovery, is not considered credible since the High Pressure Injection System would be available for loss of inventory mitigation. Also, the equivalent RCS break size of a fully opened RCS relief valve is at the lowest end of the spectrum for small break LOCAs allowing time for operators to isolate the break before core uncovery is a concern. Therefore, this event will not evolve into a more serious plant condition. Enclosure 2, "AREVA 86-9167251-001 - Summary of CR3 EPU Inadvertent Pressurizer Relief Valve Opening," provides a summary of an evaluation assessing the effects of inadvertent opening of the RCS relief valves at EPU conditions.

This information is being provided only to aid the NRC review of the CR-3 EPU LAR.

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 8 of 14

6. An analysis for the inadvertent emergency core cooling system actuation has not been provided. Please provide.

As indicated in Section 2.8.5.5, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory," of the CR 3 EPU TR (Reference 1, Attachment 7), the inadvertent operation of the Emergency Core Cooling System (ECCS) is not a CR-3 design or licensing basis accident or transient.

Currently, CR-3 Abnormal Procedures (APs) provide requirements to mitigate this event and will continue to address this event at EPU conditions. In the event of an inadvertent automatic actuation of the ECCS, the control room operators will be immediately alerted to the condition by annunciator alarms and engineered safeguards (ES) control panel status light changes. The control room operator will terminate the event by bypassing the ES automatic actuation logic with ESAS test switches located on the ES control panel.

This action allows manual shutdown of the makeup pumps (i.e., high pressure injection pumps) and other ES actuated components. PEF performed time validation of the actions to terminate an inadvertent automatic actuation of the ECCS on the CR-3 simulator. It has been demonstrated that the operator can accomplish the actions in less than two minutes from event initiation. Operation at EPU conditions will not alter the sequence or timing of the current AP. Therefore, termination of an inadvertent actuation of the ECCS at EPU conditions will continue to be feasible in less than two minutes.

Consistent with SRP Section 15.5.1, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory," a T-H evaluation assessing the effects of an inadvertent operation of the ECCS at EPU hot full power (HFP) conditions has been performed and confirmed that RCS pressure will not exceed 110% of RCS design, DNBR will remain above the 95/95 DNBR limit, and pressurizer level will not reach water solid conditions for at least 10 minutes following event initiation. The event analysis assumes a single failure of one emergency feedwater train and the initial pressurizer level is biased high at 240 inches to maximize the potential for pressurizer overfill. Consistent with NRC Regulatory Issue Summary 2005-29, "Anticipated Transients That Could Develop into More Serious Events," (Reference 6), the evaluation did not credit operation of the PORV for mitigation of the inadvertent operation of the ECCS event. Sensitivity cases were performed with and without the PORV modeled. These sensitivity cases determined that the modeling of the PORV resulted in a faster increase in pressurizer liquid level. Therefore, the PORV was modeled to produce a conservative evaluation of the necessary operator response time, but the PORV is not assumed for event mitigation. Enclosure 3, "AREVA 86-9168766-001 - CR-3 EPU Inadvertent Engineered Safeguards Actuation (IESA) Analysis Summary," provides a summary of the evaluation assessing the effects of an inadvertent operation of the ECCS at EPU conditions. This information is being provided only to aid the NRC review of the CR-3 EPU LAR.

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 9 of 14

7. Please provide sensitivity studies demonstrating that the following assumptions have an insignificant effect on the results of analyzed transients:

a. Use of nominal reactor coolant system pressure and pressurizer level

b. Complete mixing in asymmetric transients Use of Nominal Initial RCS Pressure and Pressurizer Level Values The non-LOCA analysis input assumptions are consistent with NRC approved methodology; BAW-10193NP-A (Reference 2). The NRC approved the use of RELAP5/MOD2-B&W safety analyses code for non-LOCA events as stated in the associated Safety Evaluation dated October 15, 1999 (Reference 3) and is considered an acceptable method to meet applicable regulatory requirements. As stated in Section 2.8.5.0, "Non-LOCA Analyses Introduction," of the CR-3 EPU TR (Reference 1, Attachment 7), hot channel DNBR and core neutron kinetic calculations were performed using other NRC approved computer codes (i.e., LYNXT and NEMO-K). DNBR sensitive parameters (e.g., pressure and temperature) included instrument uncertainties to ensure a conservative DNBR result consistent with BAW-10193NP-A. Additionally, core kinetics parameters and reactivity feedback coefficients are chosen to provide a conservative core response.

Consistent with BAW-10193NP-A, the initial RCS pressure in analyzed non-LOCA transients was set to a nominal value. Maximum lift tolerance was applied to safety and relief valve lift setpoints and instrument uncertainties were applied to the RPS trip setpoints and the ESAS setpoints to yield conservative predictions of system response.

Specifically, the RCS high pressure reactor trip assumed in the RCS overpressure and pressurizer level analyses includes considerable margin above the RPS RCS high pressure trip setpoint: instrument uncertainty and margin are added from the ITS RPS RCS High Pressure Allowable Value to establish a conservative RCS high pressure reactor trip analysis value and equivalent instrument uncertainty and margin are subtracted from the ITS RPS RCS High Pressure Allowable Value to establish the RPS RCS high pressure trip setpoint. This net effect, coupled with other conservative boundary conditions (e.g., reactivity coefficients and inserted rod worth), obviates the need for initial RCS pressure to be adjusted for instrument uncertainty.

Consistent with BAW-10193NP-A, the initial pressurizer level in analyzed non-LOCA transients was set to the nominal value except for transients that cause an increase in pressurizer liquid level and/or RCS pressure; in which case the initial pressurizer level was set to a nominal value adjusted conservatively for instrument uncertainty. The reason for adjusting for instrument uncertainty for these types of analyses is that the pressurizer level has an effect on pressure that is not addressed by adding uncertainties to RPS trip setpoints or PSVs setpoints. Thus, the nominal value of 220 inches is adjusted for instrument uncertainty, depending on the transient, to 240 or 200 inches to maximize system pressure and/or pressurizer level. Additionally, key analyses most sensitive to pressurizer level increase were performed using an initial level more conservative than adjusted nominal conditions.

Use of Complete Mixing The appendix, "Non-LOCA Analysis Methodology for B&W-Designed Plants," of BAW-10193NP-A describes the large detail model used for B&W plant safety analyses.

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 10 of 14 The reactor vessel downcomer, core, or core exit plenums are modeled using a single volume for each region consistent with the previously approved TRAP2 model (Reference 7). Consequently, the model assumes complete mixing even for asymmetric transients, most notably the main steam line break (MSLB).

BAW-10193NP-A explains that investigations were performed to verify that using a complete mixing model yields conservative core power predictions during steam line break events. The investigation was performed using TRAP2 to determine the effects of limited thermal mixing at the core inlet on the predicted consequences of a Steam Line Break Accident on a B&W designed PWR. It was shown that when no fluid mixing was allowed in the reactor vessel, the peak power due to subcritical multiplication was a factor of 1.07 times greater than the predicted peak power with perfect mixing. However, that same study showed that the peak core power predicted using multi-dimensional neutron kinetics and thermal-hydraulic models was forty (40) percent of the value predicted by the point kinetics core model. Consequently, it is concluded from that study that the point kinetics solution utilized by RELAP5/MOD2-B&W with perfect thermal mixing will provide a conservative prediction of core power during an asymmetric overcooling event.

The analyses referred to in BAW-10193NP-A were performed in the early 1980's as a part of a large analytical effort to demonstrate that MSLB analyses based on perfect thermal mixing and point kinetics provide a conservative result when compared to analyses based on limited thermal mixing and three-dimensional reactivity effects.

Summary Performing sensitivity studies for analyzed transients to show insignificant effect on results would require resource intensive and time consuming efforts beyond the design and licensing basis of CR-3. BAW- 101 93NP-A (Reference 2) allows nominal values to be used for initial operating conditions except in cases where the results would be significantly altered or margin to the limit is small. BAW-10193NP-A methods also allow the use of complete mixing in asymmetric transients. Non-LOCA analyses for EPU conditions were performed in accordance with BAW- 101 93NP-A, which is an NRC approved methodology as stated in the Safety Evaluation dated October 15, 1999 (Reference 3) and, as allowed by 10 CFR50.34(h)(2), is considered an acceptable method of complying with applicable regulatory requirements. As such, sensitivity studies to demonstrate that use of nominal RCS pressure and pressurizer level and complete mixing in asymmetric transients have an insignificant effect on the results of analyzed transients are not provided.

8. The staff is unable to determine the applicability of the 3200 psi acceptance criterion to the control rod ejection event. It is not clear that the Service Level C limit for the CR-3 reactor vessel and connected components is 3200 psi, or that the Service Level C limit is an acceptable acceptance criterion for the control rod ejection accident.

Notably, the CR-3 licensing basis (Page 47 of Chapter 14 of the UFSAR) indicates that the acceptance criterion is 110% of the vessel design pressure. Please explain and justify the application of the 3200 psi acceptance criterion.

CR-3 Final Safety Analysis Report (FSAR) Section 14.2.2.4 previously contained an incorrectly reported acceptance criterion of 110% of the RCS design pressure (i.e., 2750

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 11 of 14 psig) for the Rod Ejection Accident (REA). The CR-3 licensing basis acceptance criteria for a REA include a fuel enthalpy of 280 cal/gm and the amount of thermal energy release necessary to cause reactor pressure vessel (RPV) deformation; 3.37 X 108 cal.

CR-3 requests approval to use an RCS pressure acceptance criterion of 3200 psig for CR-3 REA analyses, which is less than the ASME Service Level C stress limit criterion specified in SRP, Section 15.4.8, "Spectrum of Rod Ejection Accidents (PWR),"

Revision 3. Using this RCS pressure acceptance criterion in REA analyses will continue to ensure the thermal energy release necessary to deform the RPV is not reached.

Enclosure 4, "Markup of Attachment 1 of CR-3 EPU LAR #309, Revision 0," provides this additional licensing basis change request including justification for the change and an editorial change to the No Significant Hazards Considerations evaluation. The editorial change to the 10 CFR 50.92 evaluation does not alter the conclusion that the proposed license amendment does not involve a Significant Hazards Consideration. Enclosure 5, "Revised Pages of Attachment 1 of CR-3 EPU LAR #309, Revision 0," provides a clean typed revision of the affected pages with revision bars noting the changes and includes repagination.

Pre-EPU Analyses Up to Revision 32.1 of the CR-3 FSAR Section 14.2.2.4, an acceptance criterion of 110%

of the design pressure of the reactor coolant system (2500 psig) for a peak pressure limit of 2750 psig was incorrectly reported for the REA. During development of the CR-3 EPU LAR, it was self-identified that this acceptance criterion was inappropriately added to the FSAR during a 1989 rewrite of FSAR Chapter 14. This was identified and investigated in accordance with the Progress Energy Corrective Action Program and the invalid statement has been removed from CR-3 FSAR Section 14.2.2.4. Beginning with Revision 32.2, the CR-3 FSAR no longer includes the erroneous acceptance criterion for the REA. It was determined that the 110% RCS pressure acceptance criterion is not applicable to the CR-3 REA based on the following:

" CR-3 FSAR Section 1.4, "Principal Architectural and Design Criteria," contains criteria relevant to the REA: Criteria 32 and 33. These criteria are the CR-3 equivalent to 10 CFR50, Appendix A, General Design Criterion (GDC) 28, "Reactivity limits." GDC 28 requires systems be designed to limit the rate of reactivity increase to, in part, prevent postulated reactivity accidents from resulting in damage to the RCPB greater than limited local yielding. Similarly, FSAR Sections 1.4.32 and 1.4.33 require the reactor and RCPB be designed such that; "Ejection of the maximum worth control rod will not lead to further coolant boundary rupture..."

• Section 14.2.2.4.2 of the FSAR at the time of CR-3 Operating License issuance stated that for the reactor protection criteria, a control rod ejection accident will not further damage the RCS. FSAR Section 14.2.2.4 discussed the margin between calculated rod worth during an REA and the rod worth that could result in a reactivity transient producing an explosive energy required to cause deformation of the Reactor Pressure Vessel (RPV). FSAR Section 14.2.2.4.5 indicated a reactivity addition acceptance limit of 1.52% Ak/k and the resulting REA reactivity addition was reported to be approximately one half of the limit.

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 12 of 14 In the original NRC Safety Evaluation (SE) (Reference 8) citing approval of the CR-3 license application, Chapter 15.0, "Accident Analysis," of the NRC SE provides offsite dose results and assumptions used in the offsite dose calculations of an REA.

RCS pressure was not included as an REA assumption in the SE and there was no reference explicitly or implicitly implied regarding an RCS pressure acceptance criteria of 110% for the REA and no resulting RCS pressure due to an REA included in the SE.

" The 1989 B&W REA analysis summary, which was the basis for the FSAR Chapter 14 rewrite, did not include a reference to RCS pressure in the analysis results for either the hot zero power condition or the HFP condition. Rather, the analysis summary concluded the reactivity insertion due to an REA continued to be approximately one half of the reactivity insertion required to deform the RPV. The FSAR Chapter 14 rewrite was performed utilizing the requirements of 10 CFR 50.59 and was not submitted as a LAR to the NRC for review and approval. Per the accompanying 10 CFR 50.59 evaluation, the revision to the REA section, including an addition of the erroneous 110% RCS pressure criterion, was made to clarify the description of the accident analysis. However, the rewrite continued to provide REA results related to energy deposition, the energy required to produce a pressure pulse of sufficient magnitude to result in RPV deformation (i.e., local vessel yielding), and radiological consequences. The revised FSAR section did not provide REA results associated with peak RCS pressure.

EPU Analysis For EPU, the CR-3 REA analysis was performed in accordance with the requirements of SRP Section 15.4.8, "Spectrum of Rod Ejection Accidents (PWR)," and provides a maximum pressure criterion: less than the value that result in stresses that exceed the "Service Limit C" as defined in the ASME Boiler and Pressure Vessel (BPV) Code.

The peak RCS pressure to preserve the Service Level C stress limit, as defined in the ASME BPV Code, is conservatively considered to be 3200 psig for B&W plants. BAW-1610, "Analysis of B&W NSS Response to ATWS Events," (Reference 9) evaluated the pressure retaining capability of RPV components for B&W plants, including CR-3. This B&W report indicates that the limiting RPV components can be subjected to pressures well above 3200 psig for ASME Service Level C stress conditions. During a 1974 comprehensive assessment confirming the structural adequacy of the RCPB during a postulated ATWS event, it was determined that the RCPB could withstand RCS pressures up to 3750 psig using ASME Emergency (Service Level C) stress levels. Major components included in this comprehensive assessment included the RPV, OTSGs, pressurizer, and RCS piping. Per CR-3 FSAR, Section 4.1.3, "Codes and Classifications," the CR-3 RPV was tested to ASME BPV Code,Section III, 1965 Edition up to and including the 1967 Summer Addenda (Reference 10) and RCS piping was analyzed to USA Standard (USAS) B31.7-1968 (Reference 11). A 1980 review of the previous assessment concluded that a majority of the components would meet ASME Service Level C stress requirements at an RCS pressure of 4000 psig and selected locations (e.g., RPV lower head, OTSG and pressurizer spray nozzles), would meet these requirements at a lower RCS pressure of 3750 psig. Further analyses were performed for the reactor coolant pumps and RCPB valves. The analyses utilized Service Level C stress criteria of the ASME BPV Code,Section III, 1979 Summer Addenda

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 13 of 14 (Reference 10). ASME stress reports, which were the basis for the results, were prepared using elastic, finite element models. The limiting component identified for CR-3 was the Byron Jackson RCP. The results, summarized in BAW-1610, indicated approximately 6% of the RCP pressure boundary area exceeds the primary membrane Level C stress limit when subjected to an internal pressure of 3500 psig. However, this overstress condition was determined to cause limited localized distortion and subsequent analysis indicated that it would not affect the Operability of the RCP. Based on these evaluations, 3200 psig is considered an acceptable RCS pressure limit for the CR-3 REA, which is less than the value that would result in exceeding the ASME Service Level C stress limits of any RCPB component, and therefore complies with the related acceptance criterion specified in SRP, Section 15.4.8.

Summary To summarize, CR-3 has previously concluded 110% of RCS design pressure (2750 psig)l is an erroneous acceptance criterion for the CR-3 REA. The CR-3 licensing basis acceptance criteria for the CR-3 REA include the amount of energy required to produce a pressure pulse of sufficient magnitude to cause rupture of the RPV. In addition, an RCS pressure of 3200 psig is an acceptable acceptance criterion for the CR-3 REA analyses.

As such, CR-3 requests approval to use an RCS pressure acceptance criterion of 3200 psig, which complies with the requirements of SRP, Section 15.4.8, for the REA analyses as a method of ensuring maximum rod worth during an REA will not lead to further coolant boundary rupture, a CR-3 design criterion.

References

1. CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate," (Accession No. ML112070659).

2. BAW-10193NP-A "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water Reactors," Framatome Technologies Group, Lynchburg, Virginia, January 2000.

3. NRC to Framatome Technologies letter dated October 15, 1999, Safety Evaluation of Topical Report BAW-10193P, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed PWRs," (TAC No. M93346).

4. NRC to CR-3 letter dated April 19, 1989, "Crystal River Unit 3 - Review of Final ATWS Design," (TAC No. 59085).

5. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

6. Regulatory Issue Summary 2005-29, "Anticipated Transients That Could Develop Into More Serious Events," December 14, 2005.

7. BAW-10128, "TRAP2 - FORTRAN Program for Digital Simulation of the Transient Behavior of the Once-Through Steam Generator and Associated Reactor Coolant System,"

August 1976.

U. S. Nuclear Regulatory Commission Attachment 1 3F1011-05 Page 14 of 14

8. AEC to FPC letter dated July 5, 1974, "Safety Evaluation by the Directorate of Licensing U.S. Atomic Energy Commission in the Matter of Florida Power Corporation Crystal River Unit 3 Docket No. 50-302."

9. BAW- 1610, "Analysis of B&W NSS Response to ATWS Events," January 1980.

10. ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Facility Components," 1965 Edition including 1967 and 1979 Summer Addenda.

11. USAS B31.7, "Nuclear Power Piping," February, 1968.

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT 2 LIST OF REGULATORY COMMITMENTS

U. S. Nuclear Regulatory Commission Attachment 2 3F1011-05 Page 1 of 1 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Florida Power Corporation (FPC) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please notify the Superintendent, Licensing and Regulatory Programs of any questions regarding this document or any associated regulatory commitments.

Regulatory Commitments Due Date/Event CR-3 will provide a summary of a FWLB overpressure November 11, 2011 protection analysis and will include key analysis input assumptions and RCS pressure results (Refer to Attachment 1, CR-3 Response #1).