3F1012-06, CR-3 Extended Power Uprate LAR Supplement and Response to Third Request for Additional Information to Support NRC Instrumentation and Controls Branch (Eicb) Technical Review

CR-3 Extended Power Uprate LAR Supplement and Response to Third Request for Additional Information to Support NRC Instrumentation and Controls Branch (Eicb) Technical Review

ML12279A383
Person / Time
Site:	Crystal River
Issue date:	10/04/2012
From:	Franke J Duke Energy Corp, Florida Power Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F1012-06, TAC ME6527
Download: ML12279A383 (20)

Text

PDuke WEnergy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 October 4, 2012 3F1012-06 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - CR-3 Extended Power Uprate LAR Supplement and Response to Third Request for Additional Information to Support NRC Instrumentation and Controls Branch (EICB) Technical Review (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" (ADAMS Accession No. ML112070659)

2. Email from S. Lingam (NRC) to D. Westcott (CR-3) dated August 9, 2012, "CR-3 EPU LAR - Additional draft RAIs from EICB (ME6527)"

3. NRC to CR-3 letter dated August 28, 2012, "Crystal River Unit 3 Nuclear Generating Plant - Request For Additional Information For Extended Power Uprate License Amendment Request (TAC No. ME6527)" (ADAMS Accession No. ML12240A015)

4. CR-3 to NRC letter dated March 19, 2012, "Crystal River Unit 3 - Response to Second Request for Additional Information to Support NRC Instrumentation and Controls Branch (EICB) Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)" (ADAMS Accession No. ML12081A293)

Dear Sir:

By letter dated June 15, 2011, Florida Power Corporation (FPC) 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). On August 9, 2012, via electronic mail, the NRC provided a draft request for additional information (RAI) needed to support the EICB technical review of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR) (Reference 2). By teleconference on August 23, 2012, FPC discussed the draft RAI with the NRC to confirm an understanding of the information being requested. On August 28, 2012, the NRC provided a formal RAI required to complete its evaluation of the CR-3 EPU LAR (Reference 3).

Attachment A, "Response to Third Request for Additional Information - Instrumentation and Controls Branch Technical Review of the CR-3 EPU LAR," provides the CR-3 formal response to the RAI.-A )C) (

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

U.S. Nuclear Regulatory Commission Page 2 of 3 3F1012-06 Attachment B, "Inadequate Core Cooling Mitigation System EMF/RFI Emissions and Susceptibility Testing Summary Report," provides a summary of the results associated with the Inadequate Core Cooling Mitigation System electromagnetic and radio frequency interference emissions and susceptibility testing. Transmittal of this document partially satisfies the regulatory commitment required by November 9, 2012 included in the EICB RAI response letter dated March 19, 2012 (Reference 4).

Attachment C, "Inadequate Core Cooling Mitigation System Seismic Qualification Testing Report," provides the results associated with the Inadequate Core Cooling Mitigation System seismic qualification testing. Transmittal of this document partially satisfies the regulatory commitment required by February 28, 2013 included in the EICB RAI response letter dated March 19, 2012 (Reference 4).

This correspondence contains no new regulatory commitments.

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

Sin y J A. Franke ice President Crystal River Nuclear Plant JAF/gwe Attachments:

A. Response to Third Request for Additional Information - Instrumentation and Controls Branch Technical Review of the CR-3 EPU LAR B. Inadequate Core Cooling Mitigation System EMF/RFI Emissions and Susceptibility Testing Summary Report C. Inadequate Core Cooling Mitigation System Seismic Qualification Testing Report xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact

U.S. Nuclear Regulatory Commission Page 3 of 3 3F1012-06 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation; 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.

Jo.X Franke

/Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this LL day of 0 ýo .- ,2012, by Jon A. Franke.

Signature of Notary Public State of FFlorida CAROLYN E.PORTM*NN ag,,,""o, Iff: commission # DD 9 3 75 53 3 g;** Expires March 1,2014 I ",,IW ," M*,iTmy FainI Boqded eOV385-7019 WUrOn (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 A RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION - INSTRUMENTATION AND CONTROLS BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 1 of 10 RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION

- INSTRUMENTATION AND CONTROLS BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR By letter dated June 15, 2011, Florida Power Corporation (FPC) 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). On August 9, 2012, via electronic mail, the NRC provided a draft request for additional information (RAI) needed to support the Instrumentation and Controls Branch (EICB) technical review of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR). By teleconference on August 23, 2012, FPC discussed the draft RAI with the NRC to confirm an understanding of the information being requested. On August 28, 2012, the NRC provided a formal RAI required to complete its evaluation of the CR-3 EPU LAR. For tracking purposes, each item related to this RAI is uniquely identified as EICB X-Y, with X indicating the RAI set and Y indicating the sequential item number.

EICB REQUEST FOR ADDITIONAL INFORMATION

1. (EICB 3-1)

In order to comply with the regulatory requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Sections 50.55a(a)(1), 10 CFR 50.55a(h), 10 CFR 50.34(f)(2)(xxii) [II.K.2.9], and the general design criteria related to the control systems of a LAR, the licensee is required to complete appropriate analyses, such as independence and diversity analyses, and failure modes and effects analyses, for balance-of-plant control systems that were modified due to the replacement of components and control logics that could either increase challenges to the initiation of plant safety systems or otherwise impact the ability of plant systems to respond to events.

The licensee submitted the original LAR of CR-3 EPU on June 15, 2011. Section 2.4.4.2, "Integrated Control System (ICS)" of the Attachment 5 of the original LAR describes the ICS modifications. However, this section does not demonstrate how the licensee's analysis results verify that the modifications of the control systems will not impact the safety system functions or impact the ability of plant control systems to respond to the transient or accident events.

For the instrumentation and controls system and other balance-of-plant control system modifications that were implemented as part of the EPU, please provide a summary of analyses that were performed to ensure that:

a) the failure of any control system component or any auxiliary supporting system for control systems, such as loss of power sources, does not cause plant conditions more severe than those described in the analysis of anticipated operational occurrences in Chapter 15 of the safety analyses report (SAR).

b) the consequential effects of anticipated operational occurrences and accidents do not lead to control system failures that would result in consequences more severe than those described in the analysis in Chapter 15 of the SAR.

c) the independence of the normal control systems from safety system functions has been verified.

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 2 of 10 d) the design of the control system limits the potential for inadvertent actuation and challenges to safety systems.

e) the control systems are appropriately designed and are of sufficient quality to minimize the potential for challenges to safety systems.

f) ICS modifications do not change any conclusion of the original failure mode and effect analyses.

Response

CR-3 plant protective and control systems are described in Section 7.1, "Protection Systems,"

and Section 7.2, "Control Systems," of the CR-3 Final Safety Analysis Report (FSAR). The CR-3 protective and control systems include the Reactor Protection System, Engineered Safeguards Actuation System, Emergency Feedwater Initiation and Control (EFIC) System, Control Rod Drive (CRD) System, and ICS.

The non safety-related balance of plant (BOP) system instrumentation and control (I&C) changes required for EPU operation were evaluated to determine the impact of these I&C changes on CR-3 safety system functions and the impact on the ability of credited CR-3 plant control systems to respond to transient and accident events. The non safety-related BOP system I&C changes required for EPU operation consist of changes or additions to individual component controls; such as adjusting the moisture separator re-heater belly drain control valve setpoint and replacing variable speed condensate pumps with constant speed pumps and air operated flow control discharge valves. These type of plant modifications do not alter the protective or control systems described in Sections 7.1 and 7.2 of the CR-3 FSAR and malfunction of the new or modified BOP component controls will not result in plant conditions or event consequences more severe than those described for the plant anticipated operational occurrences (AOOs) and accidents described in either Chapter 14, "Safety Analysis," of the CR-3 FSAR or described in the CR-3 EPU Technical Report (TR) (Reference 1, Attachments 5 and 7). Also, the non safety-related BOP system I&C changes do not adversely impact the CR-3 safety system functions or impact the ability of credited CR-3 plant control systems to respond to the transient and accident events.

The ICS modification was reviewed and it was determined that no new failure modes were identified and the modification does not result in plant conditions or event consequences more severe than those described for the plant AOOs and accidents described in either Chapter 14 of the CR-3 FSAR or described in the CR-3 EPU TR.

The non safety-related ICS modification is designed to:

" the existing CR-3 plant design criteria and does not alter the independence between the ICS and safety actuation system and credited control system functions;

" limit the potential for inadvertent actuation and challenges to safety systems; and

" augmented quality standards, which impose safety-related design controls in accordance with the plant quality assurance program to provide sufficient quality to minimize the potential for challenges to safety systems.

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 3 of 10 Also, the ICS modification does not alter the conclusions of the generic Babcock and Wilcox (B&W) ICS reliability analysis (i.e., failure mode and effect analyses) performed to address Item II.K.2.9 of NUREG-0737, "Clarification of TMI Action Plan Requirements."

The following summarizes the evaluation performed for the ICS modification:

Integrated Control System As indicated in Section 2.4.4.2 of the CR-3 EPU TR, adjustments and rescaling will be made in the ICS to support operation at EPU conditions. Adjustments and rescaling the ICS modules to support higher flows and power level does not affect the ICS failure modes. The modules continue to fail in the same manner and, thus, the conclusions of the ICS reliability analysis are not adversely impacted. The ICS modification does not involve replacement of components or control logic. However, as indicated in Section 2.4.4.2 of the CR-3 EPU TR, the ICS modification removes asymmetric rod runback feature. Removal of the asymmetric rod runback feature involves minor internal changes in the ICS and does not impact the CRD System "out inhibit" which prohibits the control rods from being withdrawn during an asymmetric fault condition. The ICS modification does not alter the downstream control systems or control devices such as; the main feedwater (MFW) pump speed control, the CRD System, or the MFW startup control valves.

These ICS changes do not result in new ICS failure modes and do not cause plant conditions or event consequences more severe than those described for the plant AOOs and accidents described in either Chapter 14 of the CR-3 FSAR or described in the CR-3 EPU TR. The power sources for the ICS remain unchanged and the signals from the ICS to other systems remain unchanged. The signals to and from the ICS and the safety system functions remain isolated and independent from the ICS.

The plant response to selected transients was also evaluated at EPU conditions using the Digital Power Train (DPT) simulation code and the plant referenced simulator, both of which model the ICS. A description of the evaluations and analyses and the results of the plant operational transients performed at EPU conditions with the DPT simulation code are provided in Section 2.12.2, "Transient Performance," of the EPU TR.

The functions of the ICS and the CRD System during a dropped rod accident are described in Sections 7.2.2.3.4.2.1, "Asymmetric Rod Monitor," and 14.1.2.7, "Stuck-Out, Stuck-In, or Dropped Control Rod Accident," of the CR-3 FSAR. The ICS runback to 60% power associated with this event is being deleted as a result of the ICS modification. The dropped rod accident at EPU conditions was analyzed at EPU conditions with the ICS runback deleted using the DPT simulation code as documented in Appendix E, "Additional DPT Analysis," of AREVA Calculation 32-9129875-001, "CR-3 Large Transient Testing Evaluation with DPT." This AREVA calculation was transmitted in the FPC to NRC supplemental correspondence dated April 30, 2012 (Reference 2, Enclosure 3). The analysis results show that removing the ICS asymmetric rod runback does not adversely affect the dropped control rod event acceptance criteria.

To clarify, the 10 CFR 50.34(f)(2)(xxii) requirement does not apply to CR-3 as indicated in 10 CFR 50.34.(f); this regulation applies to light-water reactors whose construction permit applications were pending as of February 16, 1982. Rather, the NRC Safety Evaluation Report

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 4 of 10 (SER) associated with the FPC response to Item II.K.2.9 of NUREG-0737 (Reference 3) and supplemental correspondences address the acceptability of the CR-3 ICS reliability.

The CR-3 ICS reliability analysis was performed generically by B&W and documented in BAW-1564, "Integrated Control System Reliability Analysis," dated August 1979. As indicated in the NRC SER associated with the FPC response to Item II.K.2.9 of NUREG-0737 (Reference 3),

this generic B&W reliability analysis was considered applicable to the CR-3 ICS to resolve NUREG-0737 Item II.K.2.9.

The ICS modification required to support operation at EPU conditions has been evaluated and FPC has concluded that the ICS changes do not adversely impact the conclusions of the generic B&W ICS reliability analysis described in BAW- 1564.

2. (EICB 3-2)

In various places (e.g., the last paragraph of page 2.4.2.3-1) of the Attachment 5 of the original LAR, the licensee mentions that Inadequate Core Cooling Monitoring System (1CCMS) receives inputs from "incore thermocouples" or "incore thermocouple temperatures," but the proposed CR-3 Improved Technical Specification Table 3.3.19-1 "Inadequate Core Cooling Monitoring System (ICCMS) Instrumentation," FUNCTIONS Id, 2c, and 3c (Attachments 2 and 3 of the original LAR) uses the term "Core Exit Thermocouples (CETs)" as ICCMS inputs.

Are those two terms the same? If they are the same, which one is the proper term to use? If they are different, please explain in detail.

Response

Yes, the term "incore thermocouples," mentioned in various places of the CR-3 EPU TR (Reference 1, Attachments 5 and 7), is equivalent to the term, "Core Exit Thermocouples (CETs)," identified in the proposed ICCMS Improved Technical Specifications (ITS) and Bases (Reference 1, Attachments 2, 3, and 4). The CETs are located in the core region; as opposed to the Reactor Coolant System loop; and therefore, are also referred to as incore thermocouples.

The term, "Core Exit Thermocouples (CETs)," is the most correct term to use.

3. (EICB 3-3)

During a loss of subcooling margin (LOSCM) event concurrent with a reactor trip, ICCMS will automatically trip the reactor coolant pumps within one minute and will additionally raise the steam generator (SG) secondary side water level to the inadequate subcooling margin setpoint within 10 minutes. How does the licensee ensure that the system meets these two response time requirements?

Response

The one-minute response time requirement for the reactor coolant pump (RCP) trip and the 10-minute response time requirement for raising the SG secondary side water level to the inadequate subcooling margin setpoint assumed in the safety analyses will be verified as part of the Channel Calibration testing of the ICCMS initiating channels. These response time requirements are verified by calibrating the associated time delay modules and verifying; by a

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 5 of 10 series of sequential, overlapping, or total steps; that the total time is within the required response times.

The calibration and functional testing of the ICCMS functions are included in the post EPU modification testing scope for the ICCMS modification described in Table 2.12.1-4 of Section 2.12.1, "Approach to EPU Power Level and Test Plan," of the CR-3 EPU TR (Reference 1, Attachments 5 and 7), including calibration of the associated time delay modules. Additionally, to ensure the ICCMS meets the one-minute and 10-minute response time requirements, the time delay modules will be calibrated per proposed Surveillance Requirement (SR) 3.3.19.3 in ITS 3.3.19, "Inadequate Core Cooling Monitoring System (ICCMS) Instrumentation," of the EPU LAR (Reference 1, Attachments 2 and 3).

Components downstream of the initiating channels required to actuate these ICCMS functions consist of instantaneous relays and contacts and have minimal impact on the associated response time requirement. The relays and contacts associated with these ICCMS functions (i.e., RCP trip and EFIC System inadequate subcooling margin setpoint actuation) will be functionally tested per proposed SR 3.3.20.2 and SR 3.3.20.3 in ITS 3.3.20, "Inadequate Core Cooling Monitoring System (ICCMS) Automatic Actuation Logic," of the EPU LAR (Reference 1, Attachments 2 and 3).

4. (EICB 3-4)

During a LOSCM event concurrent with inadequate high-pressure injection (HPI) flow, the existing emergency operating procedures direct the operator to manually perform a rapid primary system cooldown via SG pressure reduction. This action is being automated with the installation of the ICCMS and fast cooldown system (FCS). For EPU, FCS actuation is required within 10 minutes of LOSCM if HPI flow is inadequate. How does the licensee ensure that the system meets this response time requirement?

Response

The 10-minute response time requirement for the FCS assumed in the safety analyses will be verified as part of the Channel Calibration of the ICCMS initiating channels. This response time requirement is verified by calibrating the associated time delay modules and verifying; by a series of sequential, overlapping, or total steps; that the total time is within the required response time.

The calibration and functional testing of the 10-minute response time requirement is included in the post EPU modification testing scope for the ICCMS and FCS modifications described in Table 2.12.1-4 of Section 2.12.1 of the CR-3 EPU TR (Reference 1, Attachments 5 and 7),

including calibration of the associated time delay modules. Additionally, to ensure the ICCMS and FCS meets the 10-minute response time requirement, the time delay modules will be calibrated per proposed SR 3.3.19.3 in ITS 3.3.19 of the EPU LAR (Reference 1, Attachments 2 and 3).

Components downstream of the initiating channels required to actuate the FCS consist of instantaneous relays and contacts and have minimal impact on the associated response time requirement. The relays and contacts associated with the automatic actuation of the FCS and opening the atmospheric dump valves will be functionally tested per proposed SR 3.3.20.1 in

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 6 of 10 ITS 3.3.20 and SR 3.7.20.5 in ITS 3.7.20, "Fast Cooldown System (FCS)," of the EPU LAR (Reference 1, Attachments 2 and 3).

5. (EICB 3-5)

Clause 5.8.3 of the Enclosure 3, "IEEE 603-1991 and IEEE 279-1971 Compliance Matrix" of the licensee's supplemental letter dated August 18, 2011, states:

"System Spec 5.9.2.1 Fifteen (15) TRIP/AUTO/BYPASS Switches shall be provided to allow bypassing or tripping each output trip function in each initiation channel and bypassing or tripping each output trip function in each actuation train. The TRIP/AUTO/BYPASS Switches shall be located in the channel enclosure."

a) Where are channel cabinets located?

b) What controls and indications are available to the operator?

c) Please provide detailed description of the initiation channel TRIP/AUTO/BYPASS switches (including diagrams, function, and operation).

Response

To clarify; during development of the ICCMS modification, the fifteen (15) testing/maintenance switches have been modified slightly from the initial design described in the FPC to NRC supplemental letter dated August 18, 2011 (Reference 4).

Each of the three ICCMS initiation channels utilizes one channel trip module (CTM) for each of the three safety functions for a total of nine (9) initiation CTMs. The nine (9) ICCMS initiation channel testing/maintenance switches maintain the initially described TRIP and BYPASS functionality, but the AUTO position is labeled NORMAL. Each CTM has one LOCAL TRIP/NORMAL/BYPASS testing/maintenance switch located on the front panel of the module.

Each of the CTMs provides an input to the two-of-three voting logic in both ICCMS actuation trains.

Each of the two ICCMS actuation trains utilize one train trip module (TTM) for each of the three safety functions for a total of six (6) actuation TTMs. These six (6) testing/maintenance switches are revised to eliminate the TRIP position; thereby, minimizing the possibility of inadvertently actuating safety functions due to improperly positioning the switch; and the AUTO position is relabeled NORMAL. Each TTM has one NORMAL/BYPASS testing/maintenance switch located on the front panel of the module.

a) The three ICCMS cabinets are located in separate engineered safeguards (ES) switchgear rooms of the control complex building. The ICCMS Initiation Channel l/Actuation Train

'A' cabinet is located on the 108-ft. elevation in the Train 'A' 4160V ES switchgear room.

The ICCMS Initiation Channel 2/Actuation Train 'B' cabinet is located on the 108-ft.

elevation in the Train 'B' 4160V ES switchgear room. The ICCMS Initiation Channel 3 cabinet is located on the 124-ft. elevation in the Train 'B' 480V ES switchgear room.

b) The ICCMS initiating channel testing/maintenance switches and the ICCMS actuation train testing/maintenance switches are not used by operators for operation of the ICCMS. Main control room annunciators and ICCMS status lights provide indications when one or more ICCMS initiating channel or actuation train testing/maintenance switch is out of the normal

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 7 of 10 position. Drawings of the ICCMS and FCS controls and indications available to the control room operator were provided in the FPC to NRC EICB RAI response letter dated January 5, 2012 (Reference 5, Enclosure) and an updated ICCMS and FCS annunciator drawing was provided in the FPC to NRC EICB RAI response letter dated March 19, 2012 (Reference 6, Enclosure 1 of Attachment A). As noted in the EICB RAI Background section of Attachment A to the March 19, 2012 letter, the precise annunciator locations and labeling may change as the ICCMS and FCS modifications are finalized.

c) Each ICCMS CTM, as shown in Figure 1, "Channel Trip Module," has a three-position toggle switch with NORMAL being the center position, BYPASS being the up position, and LOCAL TRIP being the down position. The BYPASS position will not allow the initiation CTM to trip even if a valid trip signal is present. The LOCAL TRIP position will immediately place the module in a trip state. Indicating lights on the front panel show the state of the bypass and local trip signals. The BYPASS and LOCAL TRIP positions are used for maintenance and testing purposes and the NORMAL position is used for normal operation.

While not associated with the LOCAL TRIP/NORMAL/BYPASS switch, Figure 1 also shows the TRIP CONDITION DELAY adjustments and a SYSTEM TRIP indicator. The TRIP CONDITION DELAY adjustments provide an adjustable time delay to the CTM input to delay the module trip output signal for a preset amount of time based on the CTM function (i.e., RCP trip, EFIC System inadequate subcooling margin setpoint actuation, or FCS actuation). Also, each ICCMS initiation channel has modules which are critical to the functionality of the initiation channel trip string. These critical modules have micro switches, wired in series, which open when a critical module is removed from its chassis; thereby placing the associated CTM in the trip condition. The SYSTEM TRIP indicator on the front panel of the CTM denotes a trip of the CTM when a critical module within the associated ICCMS initiation channel trip string is withdrawn from its chassis.

Figure I -Channel Trip Module

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 8 of 10 Figure 2, "CTM Simplified Schematic," shows how the LOCAL TRIP/NORMAL/BYPASS switch electronically interfaces with the CTM logic circuitry and the local indicators that show the status of the CTM.

TPIP Figure 2 - CTM Simplified Schematic Each ICCMS TTM, as shown in Figure 3, "Train Trip Module," has a two-position toggle switch with NORMAL being the down position and BYPASS being the up position. The BYPASS position will not allow the actuation TTM to trip even if a valid trip is present.

Indicating lights are provided for the three initiation channel trip input signals, the train trip signal, maintenance bypass, and the 2/3 logic signal. The BYPASS position is used for maintenance and testing purposes and the NORMAL position is used for normal operation.

Figure 3 - Train Trip Module

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 9 of 10 Figure 4, "TTM Simplified Schematic," shows how the NORMAL/BYPASS switch electronically interfaces with the TTM logic circuitry and the local indicators that show the status of the TTM.

A ISCHANNELS CHA*NNEL2 CHIS TRI IN ESPA* TId I Figure 4 - TTM Simplified Schematic References

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

2. FPC to NRC letter dated August 30, 2012, "Crystal River Unit 3 - Response to Second Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch (EMCB) Technical Review of the CR 3 Extended Power Uprate LAR (TAC No.

ME6527)." (ADAMS Accession No. ML12254A241)

3. NRC to FPC letter dated April 19, 1982, "Item II.K.2.9, NUREG-0737 - Crystal River Unit 3 Nuclear Power Plant (Crystal River Unit 3)." (ADAMS Accession No. 8205040770)

4. FPC to NRC letter dated August 18, 2011, "Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Instrumentation and Controls Branch Acceptance Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)." (ADAMS Accession No. MLC1234A427)

5. FPC to NRC letter dated January 5, 2012, "Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Instrumentation and Controls Branch Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)." (ADAMS Accession No. ML12030A209)

6. FPC to NRC letter dated March 19, 2012, "Crystal River Unit 3 - Response to Second Request for Additional Information to Support NRC Instrumentation and Controls Branch

U.S. Nuclear Regulatory Commission Attachment A 3F1012-06 Page 10 of 10 (EICB) Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)."

(ADAMS Accession No. ML12081A293)

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT B INADEQUATE CORE COOLING MITIGATION SYSTEM EMF/RFI EMISSIONS AND SUSCEPTIBILITY TESTING

SUMMARY

REPORT

1. EMI/RFI Test Results Testing measured the electromagnetic interference / radio frequency interference (EMI/RFI) radiated and conducted susceptibility of, and the EMI/RFI radiated and conducted emissions from the ICCMS. Testing was conducted in accordance with NUS-A321JA, "ICCMS EMI/RFI Test Requirements", according to the guidelines published in NRC Regulatory Guide 1.180 (Revision 1, dated October 2003), which endorses EPRI report TR-1 02323 Rev 3 as modified by USNRC RG 1.180 Rev 1.

The results of the EMI/RFI testing are documented in Green Mountain Electromagnetics, Inc.

(GME) Report No. GM212033x, "EMC TEST REPORT FOR THE NUS INSTRUMENTS INADEQUATE CORE COOLING MITIGATION SYSTEM PART #NUS-A321 PA-4 REV B". The test report is in Attachment 5.

1.1. Unit Tested Testing was performed on the ICCMS Training Cabinet, NUS-A321PA-4, Serial Number 1103379, which includes an Initiation Channel, an Actuation Channel, and an online monitor.

The ICCMS equipment intended for MCB mounting was included with typical interconnecting wiring.

The cabinet was tested with the doors open. Closing the metal doors into their EMI/RFI gaskets would provide additional shielding of the internal cabinet equipment and would not increase the emissions from or susceptibility to external EMI/RFI to the ICCMS equipment. Therefore, testing with the doors open is the worst case configuration.

1.2. Functional Testing The ICCMS Training Cabinet unit was fully tested and configured and the ICCMS System Functional Test, NUS-A320TA was performed prior to EMI/RFI testing to rule out any damage during shipping.

Automatic test equipment was used to exercise the ICCMS Training Cabinet during the EMI/RFI test evaluation. The exercise consisted of an initialization section followed by an operation section which repeated itself until all other EMI/RFI testing was completed. During the EMI/RFI evaluation, the tester continuously varied the critical inputs to the cabinet and simultaneously collected responses from the cabinet. The testing performed is described in the test requirements document NUS-A321JA.

The ICCMS Training Cabinet unit was fully tested and configured and the ICCMS System Functional Test, NUS-A320TA was performed following the EMI/RFI testing to rule out any damage during testing.

1.3. Test Equipment The test equipment used to exercise the Training Cabinet was supplied by Scientech and consisted of National Instrument PXI chassis, Labview and test software, along with a 24 VDC external power supply.

Test instrumentation supplied and used by GME is listed in the test report. All test equipment used was verified to be within their calibration windows.

1.4. Testing Performed The following tests were performed:

" Low Frequency Conducted Emissions.

Low-frequency conducted emission measurement equipment and procedures were in accordance with MIL-STD-461F method CE101. The CE101 limit is per Figure 5-4 of TR-102323 from 30 Hz to 10 kHz, Curve >28V, no relaxation.

This test measured the harmonics emissions on the ICCMS power supply leads and verified that the ICCMS emissions are limited to levels that will not adversely affect the quality of the power source to which it will be connected.

• High Frequency Conducted Emissions.

High-frequency conducted emission measurement equipment and procedures were in accordance with MIL-STD-461F method CE102. The CE102 limit is per Figure 5-5 of TR-102323 from 10 kHz to 10 MHz. Basic Curve modified as per test plan Figure 3.

This test measured equipment emissions on ICCMS power supply leads and verified that the ICCMS emissions are limited to levels that do not adversely affect existing plant equipment.

" Low Frequency Radiated Emissions.

Low-frequency radiated emission measurement equipment and procedures were in accordance with MIL-STD-461F, method RE101. The RE101 limit is per figure 5-6 of TR-102323 from 30 Hz to 100 kHz.

This test measured the ICCMS magnetic field emissions and verified that the ICCMS emissions are limited to levels that do not adversely affect existing plant equipment.

Radiated emissions were measured relative to the boundary of the ICCMS.

• High Frequency Radiated Emissions.

High-frequency radiated emission measurement equipment and procedures were in accordance with MIL-STD-461F, method RE102. The RE102 limit is per Figure 5-7 of TR-102323 from 10 kHz to 1 GHz modified as per test plan Figure 5.

This test measured the ICCMS radiated electric field and verified that the ICCMS emissions are limited to levels that do not adversely affect existing plant equipment. Radiated emissions were measured relative to the boundary of the ICCMS.

• Low Frequency Conducted Susceptibility.

Low-frequency conducted susceptibility measurement equipment and procedures were in accordance with MIL-STD-461F, method CS101. Applied CS101 levels are per Figure 5-1 of TR-102323 from 120 Hz to 150 kHz, Curve >28V.

This test verified the ability of the ICCMS to withstand signals coupled onto the input power leads.

" High Frequency Conducted Susceptibility.

High-frequency conducted susceptibility measurement equipment and procedures were in accordance with MIL-STD-461 F, method CS1 14. Applied CS1 14 levels are per Figure 5-2 of TR-102323, from 10 kHz to 200 MHz. Curves for power and signal lines modified as per test plan Figure 7.

This test verified the ability of the ICCMS to withstand radio-frequency signals coupled onto power and signal cables.

" Low Frequency Radiated Susceptibility.

Low-frequency radiated susceptibility measurement equipment and procedures were in accordance with MIL-STD-461F, method RS101. Applied RS101 levels are per Figure 5-3 of TR-102323 from 30 Hz to 100 kHz.

This test verified the ability of the ICCMS to withstand low frequency radiated magnetic fields.

" High Frequency Radiated Susceptibility.

High-frequency radiated susceptibility measurement equipment and procedures were in accordance with MIL-STD-461 F, method RS103. Applied levels are 10 V/m, per TR-102323, page 5-12 from 10 kHz to 10 GHz. An additional test was run at the 60-GHz ISM band.

This test verified the ability of the ICCMS to withstand high frequency radiated magnetic fields.

• Fast Transient Burst.

Electrical fast transient immunity measurement equipment and procedures were in accordance with MIL-STD-461 F, method CS1 15. Applied levels are +/-2-kV power/+/-1-kV signal for Category B, Medium Exposure, per USNRC RG 1.180 Section 5.3 and Table 22, with waveforms per Figures 5.4 and 5.5.

This test verified the ability of the ICCMS Training Cabinet to successfully withstand repetitive fast transient bursts on all of its input/output lines including the power loads and the AC power lines.

Surge.

Surge immunity measurement equipment and procedures were in accordance with IEC 61000-4-5, "EMC Part 4-5: Testing and Measurement Techniques - Surge Immunity Test

(2006)," and IEC 61000-4-12, "EMC Part 4-12: Testing and Measurement Techniques-Ring Wave Immunity Test (2006)." Applied levels are +/-4-kV power and +/-2-kV secondary/shields/grounds for Category B, Medium Exposure, per USNRC RG 1.180 Sections 5.1 and 5.22, and Table 22, with waveforms are per Figures 5.1, 5.2 and 5.3.

This test verified the ability of the ICCMS Training Cabinet to successfully withstand high-energy overvoltage circumstances on power and interconnection lines due to the effects of lightning transients.

Electrostatic Discharge.

Electrostatic discharge immunity measurement equipment and procedures were in accordance with IEC 61000-4-2, "Electromagnetic Compatibility (EMC) Part 4-2: Testing and Measurement Techniques - Electrostatic Discharge Immunity Test (2009)." Applied levels are +/-8-kV contact/+/-15-kV air.

This test verified the ability of the ICCMS Training Cabinet to successfully withstand repetitive electrostatic discharge bursts on all human-machine interface points.

1.5. Summary of Results The NUSI ICCMS, Part #NUS-A321PA-4 Rev B complies with the applicable TR-102323/USNRC RG 1.180 Electromagnetic Compatibility (EMC) qualification requirements.

The ICCMS complies with the MIL-STD-461F conducted and radiated emissions (CE101, CE102, RE101 and RE102) limits. The unit also passes MIL-STD-461F conducted and radiated susceptibility (CS101, CS1 14, CS1 15, RS101 and RS103 including 60-GHz ISM) tests.

The ICCMS was not affected by TR-102323/USNRC RG 1.180 levels for surge immunity per IEC 61000-4-5/61000-4-12, or by ESD strikes per IEC 61000-4-2.

For compliance with CE102, EMI filter Schaffner PN FN2070-6-06 was added to the power-input circuit and mounted within 10 cm of the cabinet penetration. The filter was installed in the ICCMS for all tests.

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT C INADEQUATE CORE COOLING MITIGATION SYSTEM SEISMIC QUALIFICATION TESTING REPORT