L-MT-13-035, Extended Power Uprate: Response to NRC Requests for Additional Information and Supplemental Information

Extended Power Uprate: Response to NRC Requests for Additional Information and Supplemental Information

ML13154A011
Person / Time
Site:	Monticello
Issue date:	05/30/2013
From:	Schimmel M Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-13-035
Download: ML13154A011 (59)

Text

SXc1Energy Monticello Nuclear Generating Plant 2807 W County Rd 75 Monticello, MN 55362 May 30, 2013 L-MT-1 3-035 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed License No. DPR-22 Monticello Extended Power Uprate: Response to NRC Requests for Additional Information and Supplemental Information (TAC MD9990)

References:

1) Letter from T J O'Connor (NSPM) to Document Control Desk (NRC),

"License Amendment Request: Extended Power Uprate (TAC MD9990)," L-MT-08-052, dated November 5, 2008. (ADAMS Accession No. ML083230111)

2) Letter from T J O'Connor (NSPM) to Document Control Desk (NRC),

"License Amendment Request: Maximum Extended Load Line Limit Analysis Plus [MELLLA+]," TAC ME3145, L-MT-10-003, dated January 21, 2010. (ADAMS Accession No. MLI100280558)

3) Letter from M A Schimmel (NSPM) to Document Control Desk (NRC),

"Monticello Extended Power Uprate: Supplement for Gap Analysis Updates (TAC MD9990)," L-MT-12-114, dated January 21, 2013.

(ADAMS Accession No. ML13039A200)

4) Email from T Beltz (NRC) to J Fields (NSPM), "Monticello Nuclear Generating Plant - Draft Requests for Additional Information (SRXB) re: Review of Extended Power Uprate (MD9990)," dated March 28, 2013. (ADAMS Accession No. ML13137A103)

5) Email from T Beltz (NRC) to J Fields (NSPM), "Monticello Nuclear Generating Plant - Requests for Additional Information (SCVB)

Supporting the EPU and MELLLA+ Reviews (TAC Nos. MD9990 and ME3145)," dated April 24, 2013. (ADAMS Accession No. ML13137A102)

Aoo (

'a

Document Control Desk Page 2

6) Email from T Beltz (NRC) to J Fields (NSPM), "Monticello Nuclear Generating Plant -Requests for Additional Information (EMCB) re:

Extended Power Uprate License Amendment Request (TAC No.

MD9990)," dated May 10, 2013. (ADAMS Accession No. ML13136A012)

7) Letter from M A Schimmel (NSPM) to Document Control Desk (NRC),

"Monticello Extended Power Uprate: SECY 11-0014 Use of Containment Accident Pressure - Responses to Requests for Additional Information (TAC MD9990)," L-MT-13-033, dated March 21, 2013. (ADAMS Accession No. ML13085A033)

8) Letter from T J O'Connor (NSPM) to Document Control Desk (NRC),

"Monticello Extended Power Uprate: Updates to Docketed Information (TAC MD9990)," L-MT-10-072, dated December 21, 2010. (ADAMS Accession No. ML103570026)

9) Letter from M A Schimmel (NSPM) to Document Control Desk (NRC),

"Monticello Extended Power Uprate: Supplement to Revise Technical Specification Setpoint for the Automatic Depressurization System Bypass Timer (TAC MD9990)," L-MT-12-091, dated October 30, 2012.

(ADAMS Accession No. ML12307A036)

Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, requested in Reference 1 an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS) to increase the maximum authorized power level from 1775 megawatts thermal (MWt) to 2004 MWt. This is also known as an extended power uprate (EPU).

Also pursuant to 10 CFR 50.90, NSPM requested in Reference 2 an amendment to the MNGP Renewed OL and TS to allow operation within the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain.

On November 20, 2012, NSPM presented to the NRC the results of a Gap Analysis performed to verify the adequacy of the EPU documentation. Due to the delay in review of the MNGP EPU License Amendment Request (LAR), the NRC was concerned that various aspects of the NRC review were no longer applicable. Through the Gap Analysis review NSPM demonstrated that a small set of technical issues required revision and some design and licensing bases information had changed, but overall the body of EPU documentation was correct with the exception of the issues identified for correction. In Reference 3 NSPM provided to the NRC the results of many of the identified gaps and the associated corrections to EPU documentation.

In Reference 4, the Reactor Systems Branch of the NRC sent requests for additional information (RAI) concerning the letter sent by NSPM in Reference 3.

Document Control Desk Page 3 In Reference 5, the Containment and Ventilation Branch of the NRC sent a RAI concerning the letter sent in Reference 7. Enclosure 4 to this letter provides the NSPM response to the NRC RAI in Reference 5. , Part A, to this letter provides the NSPM response to the NRC RAI 1 from Reference 4. Question 6 will be addressed under a separate letter. In addition, , Part B, also contains responses to RAIs 1 - 4 from Reference 6. to this letter is General Electric-Hitachi (GEH) letter GE-MNGP-AEP-3284, which provides responses to NRC RAIs 2 - 5 from Reference 4 and RAI 10(b) from Reference 5. provides supplemental information in the form of a revised calculation to support the Automatic Depressurization System (ADS) bypass timer TS change for the EPU LAR. NSPM described the ADS bypass timer TS change in letters L-MT-12-091 (Reference 9) and L-MT-13-019 (ADAMS Accession No. ML13037A200). NSPM discovered an error in calculation 03-036, Revision 1, and notified the NRC of the error in a telephone conference call on April 5, 2013. More details concerning the calculation change are included in Enclosure 3 including calculation 03-036, Revision 2 which corrects the error.

The RAI responses and supplemental information provided herein do not change the conclusions of the No Significant Hazards Consideration and the Environmental Consideration evaluations provided in Reference 1 as revised by References 8 and 9 for the Extended Power Uprate LAR. Further, the RAI responses and supplemental information provided herein do not change the conclusions of the No Significant Hazards Consideration and the Environmental Consideration evaluations provided in Reference 2 for the MELLLA+ LAR.

In accordance with 10 CFR 50.91(b), a copy of this application supplement, without enclosures is being provided to the designated Minnesota Official.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

Document Control Desk Page 4 I declare under penalty of perjury that the foregoing is true and correct.

Executed on: May 30, 2013 Mark A. Schimmel Site Vice-President Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosures (3) cc: Administrator, Region Ill, USNRC (w/o enclosures)

Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC (w/o enclosures)

Minnesota Department of Commerce (w/o enclosures)

L-MT-13-035 ENCLOSURE I MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION FROM THE REACTOR SYSTEMS BRANCH AND THE MECHANICAL AND CIVIL BRANCH This enclosure provides responses from the Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, to requests for additional information (RAI) provided by the Nuclear Regulatory Commission (NRC)

Reactor Systems Branch on March 28, 2013 (Reference 1) from the Mechanical and Civil Branch dated May 10, 2013 (Reference 2).

References

1. Email from T Beltz (NRC) to J Fields (NSPM), "Monticello Nuclear Generating Plant

- Draft Requests for Additional Information (SRXB) re: Review of Extended Power Uprate (MD9990)," dated March 28, 2013. (ADAMS Accession No. ML13137A103)

2. Email from T Beltz (NRC) to J Fields (NSPM), "Monticello Nuclear Generating Plant

- Requests for Additional Information (EMCB) re: Extended Power Uprate License Amendment Request (TAC No. MD9990)," dated May 10, 2013. (ADAMS Accession No. ML13136A012)

PART A - Reactor Systems Branch RAI dated March 28, 2013 This section covers question 1 of the March 28, 2013 RAIs. Responses to questions 2

- 5 of the March 28, 2013 RAIs are provided in Enclosure 2. Question 6 will be addressed under a separate letter.

The NRC question is provided below in italics font and the NSPM response is provided in the normal font.

NRC Question

1. Page 16 of Enclosure I to the January21, 2013, letter (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13039A200) discusses Item 8 of the Extended Power Uprate (EPU) Gap Analysis, concerning emergency core cooling system (ECCS) pump flow rates. The response refers to additionalcorrespondence (ADAMS Accession Nos. ML12276A057 and Page 1 of 5

L-MT-13-035 Enclosure I ML12276A057) which, based on cursory review, appearto indicate that some assumptions and analyses credit revised ECCS pump flow rates that remain bounded by the SAFER ECCS evaluation.

Please confirm that the SAFER ECCS evaluation includes ECCS pump flow rates that are bounding of these revised ECCS pump flow rate assumptions.

NSPM Response:

NSPM provided Emergency Core Cooling System (ECCS) pump flow rates used for the Net Positive Suction Head (NPSH) Containment Accident Pressure (CAP) evaluation of the Design Basis Accident - Loss of Coolant Accident (DBA-LOCA) in L-MT-12-082 (Reference A-i) Tables 6.6.1-1 and 6.6.1-2. These flow rates were selected based on meeting SECY 11-0014, Enclosure 1, section 6.3.6 requirement that:

"The flow rate chosen for the NPSHa analysis should be greaterthan or equal to the flow rate assumed in the safety analyses that demonstrate adequate core and containment cooling. This ensures that the safety analysis and the NPSH analysis are consistent."

Thus the NPSHa flow rate values selected bound the safety analysis flow rate values used in L-MT-08-052 (Reference A-2), Enclosure 5, section 2.8.5.6.2.

The SAFER ECCS evaluation ECCS pump flow rates for flow delivered to the core are unchanged. The changes shown account for changes in pump flow due to the NPSH evaluation required by SECY 11-0014. Flow rates assumed for the NPSH evaluation are greater than or equal to the flow rates required for the SAFER ECCS evaluation (ie they are bounding).

References A-1 Letter from M A Schimmel (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address SECY 11-0014 Use of Containment Accident Pressure (TAC Nos. MD9990 and ME3145)," L-MT 082, dated September 28, 2012. (ADAMS Accession No. ML12276A057)

A-2 Letter from T J O'Connor (NSPM) to Document Control Desk (NRC), "License Amendment Request: Extended Power Uprate (TAC MD9990)," L-MT-08-052, dated November 5, 2008. (ADAMS Accession No. ML083230111)

Page 2 of 5

L-MT-13-035 PART B - Mechanical and Civil Branch RAIs dated May 10, 2013 This section covers responses to questions 1 -4 of the May 10, 2013 RAIs. The NRC question is provided below in italics font and the NSPM response is provided in the normal font.

NRC Question

1. With regard to condensate/feedwatermodification, letter L-MT- 12-114, Enclosure 1, page 56/80, Item 26, indicates that a "Complete discussion regardingjet impingement and pipe whip" were requested by the NRC.

Page 62/80 states that piping evaluations resulted in "a new limiting (for flooding) postulated 14"-line crack at the inlet to the 14 feedwater heater." Also, "The new crack did not result in any new jet impingement or pipe whip targets." Page 62/80 shows that "Pipewhip and jet impingement analyses are pending for the Condensate pump, Feedwaterpump, and piping replacement modifications."

If the statement above regardingpipe-whip and jet-impingement is accurate (i.e.,

analyses are still pending), then it will need to be discussed with the licensee.

NSPM Response The condensate pump, feedwater pump and piping replacements refer to modification work that is currently being installed in MNGP. In the statements above the use of the word "pending" was intended to clarify that the final as-built condition verification of calculation accuracy is pending until the completion of installation activities. Analyses related to design changes are approved and the new crack did not result in any new jet impingement or pipe whip targets.

To clarify our response, see Appendix A for revised pages from L-MT-12-114 (Reference B-i), Enclosure 2, Item 26 that provide clarifying information.

NRC Request

2. If the initial information in PUSAR regarding the paragraphthat discusses the 90%

and 63% of the RWCU HELB M&E releases is no longer accurate and has been deleted in the revised pages of the PUSAR, then the mark-up or revision to RAI-6(a)

Response needs to reflect that. As it currently exists in the submittal, it indicates that a question has been asked for the 90% and 63% increases and that an answer has not been provided.

NSPM Response Although the 90% and 63% increase noted did refer to EPU, those values have been superseded with subsequent mass and energy release values that incorporated Page 3 of 5

L-MT-13-035 enhanced characterization of actual releases. For RWCU the CLTP analysis was based on a single bounding break assumption. Whereas, the EPU analysis included consideration of multiple analyses for required break locations that included consideration of improved analysis assumptions such as double ended break flows and system depletions.

If both CLTP and EPU HELB cases were run using similar analysis assumptions, the changes in mass and energy releases would be minor as a result of EPU. The minor impact on mass and energy releases is supported by the fact that there were no significant piping changes in RWCU and process temperatures only change by 0.5 0 F between CLTP and the constant pressure EPU. The results of the HELB calculations for RWCU included both improved analysis assumptions and EPU conditions. The results of these calculations were tabulated by volume and provided in Tables 26-1 and 26-2 of Reference B-I.

To clarify our response, see Appendix A for revised pages from L-MT-12-114 (Reference B-i), Enclosure 2, Item 26 that provide clarifying information.

NRC Request

3. Tables 26-1 and 26-2 contain data for HELB flood levels, HELB temperaturesand HELB pressures which show increases at EPU conditions compared to CLTP. The licensee needs to evaluate these data for the increases shown and determine whether the impacted SSCs are structurally adequate to perform their intended design functions for the increases in differentialpressures, temperaturesand flooding levels.

NSPM Response Each HELB analysis evaluated flood levels to verify no acceptance criteria were exceeded. Volume (room) temperature limits were verified accordingly. Pressure differentials across walls (block walls are most limiting) were also confirmed against established wall specific limits. The Table 26-1 and 26-2 (from Reference B-1) HELB analysis parameters for building volume flooding, temperature and pressure are all acceptable, and meet corresponding structural limits.

NRC Request

4. In the old response (2009 era) to RAI-6(b), NSPM stated that these parameterswere evaluated for plant areas and structures. At that time, though, the HELB analyses and modifications were incomplete. Now that they are completed, the licensee has shown how these parametershave increased for EPU. If the effects of the increasesin these parametershave been evaluated and shown to be acceptable, that is fine. The licensee needs to make that statement, since the staff does not make those determinations.

Page 4 of 5

L-MT-13-035 NSPM Response The response to RAI 6(b) in L-MT-09-044 (Reference B-2) stated that HELB analysis output parameters evaluated acceptably for plant areas and structures. Upon completion of all HELB analyses associated with the EPU, re-assessment of outputs (e.g., temperature, pressure, and water level) indicated the same conclusion; all parameters were found acceptable with regard to plant area and structural requirements.

References B-1 Letter from M A Schimmel (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate: Supplement for Gap Analysis Updates (TAC MD9990),"

L-MT-12-114, dated January 21, 2013. (ADAMS Accession Nos. ML13039A200 and ML13039A201)

B-2 Letter from T J O'Connor (NSPM), to Document Control Desk (NRC), " Monticello Extended Power Uprate: Response to NRC Mechanical and Civil Engineering Review Branch (EMCB) Requests for Additional Information (RAIs) dated March 28, 2009 (TAC MD9990)," L-MT-09-044, dated August 21, 2009. (ADAMS Accession No. ML092390332)

Page 5 of 5

L-MT-13-035 Appendix A Revised Pages from Docketed Correspondence Included within this appendix are revised pages from marked up page changes provided in L-MT-12-114 (Reference B-i) Enclosure 2. The following pages are included:

• Pages 7, 8 and 9 (including Insert A) of 46 from L-MT-09-044, Enclosure 1

• Pages 28 and 29 of 46 from L-MT-09-044, Enclosure 1 6 pages follow

L-MT-09-044 Page 7 of 46 EMCB RAI No. 6 (a)

The same paragraph on page 3-23, as above, in reference to the reactor water cleanup (RWCU), continues as follows:

"For the break location that was analyzed during Rerate, new mass and energy release calculations considered additional blowdown sources that had not been considered in the previous 1996 analysis. This resulted in an increase in integrated mass release of about 90% and an increase in integrated energy release of 63 percent."

Confirm that the 90% and 63% increases are referring to the proposed EPU.

iReplace this text with the applicable portions of Table 26-2 lReactor Building HELB Results. This table is provided in L-NSPM RESPONSE. *MT-12-114, Item 26. See also RAI response provided in L-o~ - ,.~ ' .o~ .". ~. .~ L Tk~ ~ 1 - . , M T ,"1,3- .03 5 , ,Enc lo sure 1,1Pa rt B - R A I 2 . . , , ; , , ÷,, k ,

Thel 99%' auld 639A- nfiRR.MVaF;A-A -;I-R nnR F5~

UtVIIU Wn theiv* PFP9e- F-R. It. 0.6 FefeFIIII9 WV*

If the CLTP HELB cases were run using similar assumptions, the changes in mass and energy releases would be minor as a result of EPU.

As noted on PUSAR page 2-21:

A review of the results from several recent EPU submittals concluded that, in most cases, environmental conditions are bounded by previous analyses, confirming that EPU produces relatively minor effects.

EMCB RAI RAI No. 6(b)

Please explain how the effects of the increased mass and energy release have been evaluated, include evaluations of pipe whip restraints and jet targets.

NSPM RESPONSE Changes in mass and energy were evaluated for impacts on HELBs using the GOTHIC code. This allowed a determination of time histories for all plant areas to evaluate effects on temperature, pressure and flooding. Differential pressures between plant areas verified acceptable margins for structures such as block walls. The effects of changes to temperature, pressure and flooding have been evaluated for impact on the environmental qualification (EQ) of equipment. Upgrades to EQ files to document this evaluation

Iltem 26 I L-MT-09-044 Enclosure 1 Page 8 of 46 is one RWCU pipe whip, jet impingement and safe shut down analyses folowing postulated pipe breaks or cracks are provided in USAR Appendix I. The RW U high energy lines are located in the RWCU compartment, steam chase; MG set roo, and the North West side of elevations 962' and 935' of the reactor building. There are-Re postulated breaks in the A.1G cot room and the reactor building elevations Q62' aRd 935' based on 6iORiGmis

.Re"'""".. There are no pipe whip targets for the RWCU piping in the steam chase. IH teriaI The safe shutdown evaluation for the RWCU compartment in Appendix I does not rely on pipe whip restraints or jet impingement shields to protect any equipment or structures. The effects of pipe whip and jet impingement in this area do not result in the loss of components required to mitigate the break and shut down the reactor. Therefore there is no impact on RWCU pipe whip and jet impingement due to EPU.

EMCB RAI No. 7 Page 2-37 states that: "The combination of stresses was evaluated to meet the requirements of the pipe break criteria. Based on these criteria, no new postulated pipe break locations were identified." For systems affected by the EPU, specifically steam (all EPU affected steam lines) and FW lines (including condensate), provide a pipe break analysis summary table (that includes the main steam increased turbine stop valve (TSV) closure transient loads in the analysis) which compares values at EPU and CLTP conditions and shows code equation stresses and CUFs compared to break limit for stresses and CUFs. Include pipe break locations and types selected for CLTP and EPU. Include lines inside and outside containment.

NSPM RESPONSE Systems that have piping meeting the MNGP design basis criteria for classification as "High Energy" include Main Steam, Condensate, Feedwater, Residual Heat Removal (RHR), Core Spray (CS), High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), Reactor Water Cleanup (RWCU), Off Gas, Control Rod Drive (CRD), Zinc Oxide Injection (GEZIP), and Standby Liquid Control (SLC). The parameters used for stress analysis in the high energy portions of these systems are unchanged due to EPU except in the Main Steam, Condensate, Feedwater, and GEZIP systems.

The Main Steam system analysis results including TSV closure loads are provided in the table below. The stress result for the Main Steam location with the maximum HELB break postulation equation result is also included in the table. The stress at that location does not meet (is less than) the current design basis criteria to require a postulated break. Hcno,, thc. or isen Main St.am brcak., .utcidctainm,,t

... poctulatod baccd on ctrocc itoria. Other postulated break locations are based on configuration (e.g., terminal en )which is not changed by EPU. Note that in the current design Evaluation of jet impingement from this new crack has been assessed and no safety related equipment is in the area. This new crack is bounded by other HELB cracks and breaks in the area for the impact from expected mass and energy release. Analyses related to design changes are approved and the new crack did not result in any new jet impingement or pipe whip targets.

Item 266 L-MT-09-044 Enclosure 1 Page 9 of 46 basis, specific HELB locations are not postulated inside containment. The current design basis does not include fatigue analysis of the Main Steam piping. Due to the revised analysis of the turbine stop valve closure loads, comparison to pre-EPU values is not meaningful.

The Main Steam evaluation results shown below are performed for the EPU pressure, temperature and flow parameters, including the TSV closure loads.

• Steam Outside Containment - Maximum EPU Results (Highest Interacti n R Lo Service Node Stress Allowable Rr--

Combnto -- Level psi ,S/Allow P+DW B --- *Z"A 67- 15000 0.46 TH Range B -Tt*ff -;4,941 22500 0.86 P+DW+TSV -_e 268 12236 1000

--- 0.68 DW+TSV+SRVt D 268 13795 26325 HEL +TH+OBE B TURB 27559 30000 0.92 Thc m ,aximumFo.dwater system ep..ating . tec..p.ratUe ... i.... at EP' o.,ditionc for the FeodWatr piping from the outboard containment iclto a othe con9t.A*Rimont- anRd inciqdoA cnt$aRRinment-. This vau cboun--ded by thoi Go@rinal analycifi

'Unch-angod by EPU-. Therforoff this piping ic,unafoetod by EPU rclative to HEL pestultieR.werer were The feedwater piping an densate piping from the condensa ump suction to the containment isolation valve* a re-analyzed during the Fe w ter and Condensate pump and heater replacement modification process. High F~rg( Line Breaks and pipe whip restraints in the high energy portion of this piping - ev.uated at that time.

GEZIP connections to the portion of the Feedwater system w analyzed as part of the modification process. Dotaile of the oedificatione to this pip*in are not yet fi.ali-cd.

The design w'-A ' ,i t:iR a. stresses in the condensate and FW piping within code allowable limits of ANSI- 1.1-1977, including Winter 1978 Addenda and the requirements of USAR Cha r 12 including USAR Appendix I Conf.ir..atio that the

..... !IA... .. are complete an eet the code allowables will be pro.ided to. the NC.

The FW an ondensate system difications are scheduled for co,,mpletio,-n du-riR,,g The calculations installation in EMCB RAI No. 8 2013.

Enclosure 5, PUSAR Section 2.2.1.2, Liquid Line Breaks, on page 2-23 states that:

"The mass and energy releases for HELBs in the RWCU, FW, Condensate, CRD, Standby Liquid Control, and Zinc Injection (GEZIP) systems and instrument and sample Following startup after installation of the new turbine and new FW heaters, the FW temperature increased by approximately 50 F for a portion of the FW piping, which was no longer bounded by the design temperature of 400°F for EPU operating conditions. Therefore, the affected FW piping design temperature was increased toi, 0ur- ran piping analyses were reperormedu to accuunrt 1r ute r'vv temperature change. All piping continues to meet code allowables.

Item 26 1 Insert A Maximum Pipe Stresses (Outside Containment)

Load Combination Service Node Stress Allowable Interaction Level (psi) (psi) Ratio P+DW A TURD 7650 15000 0.51 TH Range A TURB 16618 22500 0.74 P+DW+TSV B TURC 12288 18000 0.68 P + DW + OBE* B X7A 14289 18000 0.79 DW+SRSS(TSV, D X7A 21026 26325 0.80 SSE)*

HELB TH N/A TURB 16618 18000 0.92 HELB N/A TURD 32631 30000 1.09**

DW+TH+OBE I I I I

• Excluding seismic category II pipe between Stop Valves and Turbine

• • Indicates a HELB at this location, this load combination is used only to evaluate the need to assume a HELB and is not required to have an Interaction Ratio <1 to meet USAR requirements.

Iltem 26 1 L-MT-09-044 Enclosure 1 Page 28 of 46 Maximum Support Loads MS Relief Valve Discharae Line SuDoort RV25A-H1 (sorina hanaer)

Max Min Service Node Load Allowable IR Load Allowable IR Load Condition Level lb lb Max/Allow lb lb Allow/Min DW+TH+

SRSSTVSRVOBE) B 285 [1341 11344 ]0.998 1162 780 0.671 am Steam Outside Containment u EPU Results (Highest Interaction Ratio): IDeleted per Item 11 Maxlmu i e Stresses Service Node Stress Allowable Ratio C bin n Levelpsi S/Allow P+DW N, B X7A 6877 15000 0.4 TH Range B TURB 19441 22500 .6 P+DW+TSV B 268 12236 18000 0.68 DW+TSV+SRV+SSE D 268 13795 263 0.52 HELB DW+TH+OBE TURB 27559 000 0.92 Maximum Turbine Loads "* z Load Service Node Mx able Ratio Mz Allowable Ratio Combination Level ft-lb ft- MxIAllow ft-lb ft-lb Mz/AlIow DW+T B

• 322 413000 0.078 171446 722000 0.237 DW+TH I B ;7-321 430 .57 302310 ,722000 0.419

• Note: Loads from all turbine nodes re combined Maximum Support Loa; _

Main Steam Line Su ott PS16 , Node28 Max Service Load Allowable IR Load Co On Level Component lb lb Max/Allow DW+)4+SRSS(TSV,SRV,OBE) B IAnchor bolt 20026 I 20731 0.966 Response to Part b The maximum FecdWater System operating tomerf~aturo is 89E.0 at EPU conditiont P - .0" - LMU Mmpr-q Mr I M ri L""I" ri "Mygn ft" rs "M -rm -rm

--- " r "r -" -* ..........

oontailmcItI andIV InI Id oVIntalnnIItIVI. lh IcIIAlulcV I* PVoIuInd bY Vd the oriain1al anIalYtiS P IN

• I P dl I tempfiWee4O900F. The desig

• rc,*ccuo I fo ticG .o..... .. th*o 1,vcmcr G-Yoo'tcr m ý i imý 1 ýý ý -^ ^

AI 1 IU m,3 Um . mI If I II , 'V U It, .IIII I

,e,,,,+,ie.; The current design basis for Feedwater piping analysis does not include fluid transient analysis. The stress analyses for the Feedwater piping from the outboard

Iltem 26 1 LFollowing startup after installation of the new0 turbine and new FW heaters, the FW LMT-09044 o temperature increased by approximately 5 F, which was no longer bounded by EnlsueIthe design temperature of 4000 F. Therefore, the FW design temperature was

ý M ^ Iincreased to 41 0°F and piping analyses were reperformed to account for the FW Page 29 of 46 temperature change. All piping continues to meet code allowables.

containment isol tion valve to the containment and inside containment are therefore unaffected by E U. have been re-analyzed for The feedwater piping and condensate piping, rom the condensate pump suction to the containment isolation valves -Ail ber al-,--yz-d durin-,g the Feedwater and Condensate system modifications (reference response to RAI 7). Installation of these modifications is

/ý currently in progress.

EMCB RAI No. 18 In accordance with Section 2.2.2 of the PUSAR, the main steam and associated piping system structural evaluation was performed to justify the operation of these systems at EPU conditions. This evaluation showed that one small bore branch line did not meet the displacement criteria. PUSAR further states that, "Additional detailed analysis will be performed to qualify this line or the piping modified prior to operation at EPU conditions."

a) Provide identification of the small bore branch line (size, system, location, function).

b) Describe the required displacement limits and their bases.

c) Since this piping analysis, with potential piping and or support modifications, is required for EPU, please discuss the reasoning for not including this information in your application. Also, indicate when necessary modifications, as needed, will be completed.

NSPM RESPONSE a) The branch line is a 1 inch instrument sensing line located inside the primary containment. The line connects one of the differential pressure sensing ports on the D steam line flow restrictor to a containment instrument piping penetration. This line is used for flow sensing in main steam line D and serves a safety related input function to the high flow Group 1 Containment Isolation logic that will automatically isolate the MSIV's in the event of a main steam line break.

b) A differential displacement of 1/16 inch for branch connection points was used as screening criteria in the piping analysis. Those in excess of 1/16 inch were noted as outliers needing further evaluation. The basis for the 1/16 inch criteria is:

1. The 1/16 inch displacement produces an insignificant stress in the branch line which is typically supported by a standard deadweight span (span length from run pipe nozzle connection to first support on the branch).

L-MT-13-035 ENCLOSURE 2 GE-MNGP-AEP-3284, ENCLOSURE 1 - NON-PROPRIETARY GEH RESPONSES TO RAIS SUPPORTING THE EPU AND M+ REVIEW REACTOR SYSTEMS BRANCH QUESTIONS 2 - 5 CONTAINMENT AND VENTILATION BRANCH QUESTION 10(b) 10 pages follow

ENCLOSURE 1 GE-MNGP-AEP-3284 GEH Responses to RAIs Supporting the EPU and M+ Review GEH Non-Proprietary - Class I (Public)

GE-MNGP-AEP-3284 Non-proprietary Information Page 2 of 10 NRC RAI #2:

Page 20 of Enclosure I to the January 21, 2013, letter discusses Item 10 of the EPU Gap Analysis, concerning the effects of a final feedwater (FW) temperature change. It states,

"[General Electric-Hitachi (GEH)] performed a study and determined that the impact of the FW temperature change on anticipated operational occurrences (AOOs) was negligible."

Please describe how the study was performed and provide additional information regarding the basis for this determination.

GEH Response:

GEH performed a dual reload license for Monticello Cycle 26 at 1775 MWt (CLTP) and 2004 MWt (EPU). At CLIP conditions, the limiting AOO events were evaluated using ODYN assuming feedwater temperature input of 383.07F and 388.07F. The impact on corrected Change in Critical Power Ratio (ACPR) was determined to be < 0.0045 for all events. The increase in FW temperature was shown to benefit several transients, with the difference being > -0.0048.

Additionally, the limiting transient with respect to ACPR decreased by 0.0019, resulting in no impact to the calculated Operating Limit MCPR. All thermal and mechanical overpower results demonstrated margin to the limits.

At EPU conditions, TRACG is the licensing basis code for AOOs. Instead of explicit analysis, the TRACG AOO LTR was leveraged to quantify the sensitivity of transients to a change in FW temperature. Table 8-10, Reference 1 documents the FW temperature sensitivity on the TTNBP event with respect to DCPR/ICPR. Specifically, a change in FW temperature of -56K (-l007) resulted in a DCPR/ICPR effect of-0.013. The DCPR/ICPR is scaled by the ratio of FW temperature change (5/100) to obtain a more realistic value of 0.00065 DCPR/ICPR. Section 8.2.1, Reference I states a 0.005 DCPR/ICPR is considered 'insensitive".

The nominal FW temperature is provided on a cycle-specific basis as input to the reload licensing evaluations. Therefore, all AOOs for future reloads will be evaluated with the appropriate FW temperature.

References:

1. NEDE-32906P-A R3, Licensing Topical Report, TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses, September 2006.

GE-MNGP-AEP-3284 Non-proprietary Information Page 3 of 10 NRC RAI #3:

Page 20 of Enclosure I to the January 21, 2013, letter discusses the effects of a final FW temperature change. It states, "GEH further concluded that sufficient margin remains in the peak dome pressure safety limit and ASME upset condition limit when accounting for this small FW temperature change."

Describe how this conclusion was reached. Explain how much margin is required to offset the effects of the final FW temperature change, how the amount of margin remaining in these limits was determined, and how MNGP will ensure that adequate margin is maintained in cycle-specific safety analyses.

GEH Response:

A qualitative evaluation was performed and the conclusion was reached using a combination of sensitivity results for non-limiting pressurization transients (e.g. turbine trip no bypass) and the margin to the dome pressure safety limit and ASME code upset condition limit. The +57F FW temperature change increased dome and vessel bottom pressure for the non-limiting pressurization transients by -5 psi. The FW temperature increase would impact the limiting vessel overpressure event (MSIVF) by a similar magnitude. The pressurization rate increase due to the FW temperature increase would be similar to the other non-limiting transients; however, the high-pressure RPT would occur earlier in the event offsetting some of pressure increase. The Cycle 26 EPU results for vessel overpressure demonstrated 10.9 psi margin to the dome pressure safety limit and 30.5 psi margin to the ASME code upset condition limit.

The nominal FW temperature is provided on a cycle-specific basis as input to the reload licensing evaluations. Therefore, the ASME overpressure event for future reloads will be evaluated with the appropriate FW temperature.

GE-MNGP-AEP-3284 Non-proprietary Information Page 4 of 10 NRC RAI #4:

Page 20 of Enclosure I to the January 21, 2013, letter discusses the effects of a final FW temperature change. The applicable section describes and evaluation of the design basis accident (DBA) - loss of coolant accident (LOCA) containment response. The section does not describe the effects that the final FW temperature change could have on the DBA-LOCA ECCS evaluation.

Please explain how the EPU ECCS evaluation accounts for the final FW temperature change.

GEH Response:

Feedwater temperature changes impact the Monticello ECCS LOCA response by directly affecting the initial core coolant energy content. With higher feedwater temperature expected at EPU power (2004 MWth) and MELLLA+ flow (46.1 Mlbm/hr) conditions (e.g. 5 0F above the analysis-basis value of 395.8°F to 400.8°F [Reference 1]), a corresponding increase in feedwater enthalpy yields a small increase in core coolant inlet enthalpy (less than 2.0%). A postulated large break LOCA may then cause the core to enter boiling transition at slightly earlier times along the fuel axial length, whereas a small break LOCA would see effectively no change in boiling transition behavior. Any LOCA scenario evaluated with a small increase in feedwater temperature would also experience a small increase in the cladding heatup rate early in the event due to a minor reduction in the coolant inventory heat absorption capacity. The additional energy from the higher feedwater temperature yields slightly higher cladding temperatures until ECCS provides effective cooling and inventory makeup.

The limiting ECCS LOCA scenario for Monticello is a large break in the recirculation suction line evaluated with Appendix K assumptions at Current Licensed Thermal Power (CLTP, 1775 MWth) and MELLLA flow (47.4 Mlbm/hr) conditions with LPCI injection valve failure

[Reference 2, Reference 3]. The resulting Licensing Basis Peak Cladding Temperature (LBPCT) for this scenario is 2140'F.

Heat balance assessments demonstrate lower feedwater temperature values at the CLTP power level and MELLLA flow conditions when compared to EPU power and MELLLA+ or rated flow conditions. Generally, the temperature of feedwater delivered to the reactor vessel is predominately dependent upon reactor power but weakly dependent on core flow, such that the higher EPU power level amplifies the impact of the minor core enthalpy increase on the LOCA response.

An assessment performed for the Appendix K large break LOCA scenario with a 5°F increase in feedwater temperature at EPU power and rated flow conditions shows an insignificant PCT change of approximately 6°F. A similar assessment performed for the limiting case at CLTP power and MELLLA flow conditions yields a negligible difference. Therefore the LBPCT does not change and the ECCS LOCA response for Monticello is not affected by a 5°F increase in feedwater temperature at EPU power and rated, MELLLA, and MELLLA+ flow conditions.

Additionally, operation at MELLLA+ conditions requires a larger setdown in the linear heat

GE-MNGP-AEP-3284 Non-proprietary Information Enclosure I Page 5 of 10 generation rate as compared to the setdown applied for operation with MELLLA conditions

[Reference 3], thus assuring the limiting LOCA scenario defining LBPCT remains at CLTP power and MELLLA flow conditions.

References I, Letter from M.A. Schimmel (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate (EPU): Supplement for Gap Analysis Updates (TAC M1D9990),"

L-MT-12-114, dated January 21, 2013.

2, GEH Nuclear Energy, "Safety Analysis Report for Monticello Constant Pressure Power Uprate," NEDC-33322P, Revision 3, October 2008.

3, GEH Nuclear Energy, "Safety Analysis Report for Monticello Maximumr Extended Load Line Limit Analysis Plus," NEDC-33435P, Revision 1, December 2009.

GE-MNGP-AEP-3284 Non-proprietary Information Page 6 of 10 NRC RAI #5:

Page 40 of Enclosure I to the January 21, 2013, letter discusses Item 15 of the EPU Gap Analysis, concerning a change in the turbine bypass valve capacity value, which apparently amounted to a slight reduction, i.e., from 11.6% to 11.5%. The section states that "the evaluation of plant transients is performed on a cyclic basis for MNGP and has been completed for EPU core design using a value of 11.5% for the evaluation of transients... the results of this...

evaluation are available in the MINGP cycle 26 supplemental reload licensing report..."

Please address the effects of this change with respect to the limiting ATWS overpressure events.

GER Response:

A change in the turbine bypass capacity could only impact the Pressure Regulator Failed Open (PRFO) event. However, plants that have small TCV and turbine bypass capacities will likely not be able to depressurize the reactor down to the low pressure isolation setpoint, which would then trigger MSIV closure. Procedurally, GEH sets the demand to -120% of rated conditions to drive plants to these pre-isolation vessel conditions. Therefore, the ATWS overpressure analysis assumes a turbine bypass capacity above the actual value. Note the turbine bypass system is not part of the ATWS overpressure mitigation. As a result, there would be no impact to the ATWS safety analysis due to the change in the turbine bypass capacity.

GE-MNGP-AEP-3284 Non-proprietary Information Page 7 of 10 NRC RAI on GEH Response - RAI 10(b)

Please provide additional information in reference to NSPM letter dated March 21, 2013 (Agencywide Documents Access and Management System Accession No. M1L13085A033), , GEH Response - RAI 10(b).

Refer to NEDC-33322P, Revision 2, Section 2.6.3.1.1 Section 2.6.3. 1.1- Short Term Gas Temperature Response The drywell air space temperature limit is specified in Table 2.6-1. The limit is increased for EPU from 335°F to 340'F.

The GEH response to RAI 10(b) states the following:

The peak drywell temperatures reported tinder EPU/MELLLA conditions in Table 2.6-1 of NEDC-33322P were obtained from a long-term containment response calculation for a small steam line break accident (SBA) at 102% of EPU power and 100% core flow with the SHEX code.

NRC Staff Comment The GEH response to RAI 10(b) appears to conflict with the NEDC-33322P Revision 2, Section 2.6.3.1.1.

The RAI response implies that the drywell gas temperature values in Table 2.6-1 of NEDC-33322P are based on a long term SBA analysis using SHEX code.

Section 2.6.3.1.1 of NEDC-33322P is referring to the short term drywell gas temperatures listed in Tables 2.6-1.

Pleaseprovide clarificationas to whether the peak drywell temperatures of 335°F,336TF, and 338°F listed in Table 2.6-1 are based on short termn SBA analysis or long term SBA analysis with SHEX code.

Pleaseprovide additionalclarification(e.g., footnote (s)) in Table 2.6-1 to further differentiate between long terni and short termn SBA analyses.

GEH Response - Question 1 The peak drywell temperatures of 335°F, 336'F, and 338°F, which are reported in Table 2.6-1 of NEDC-33322P, Revision 2 (Reference 1), were determined from long term containment analyses for a small steam line break (SBA) performed with the GEH SHEX code.

GE-MNGP-AEP-3284 Non-proprietary Information Enclosure I Page 8 of 10 GEH Response - Question 2 A revision to Table 2.6-1 of NEDC-33322P is included for this response which expands the discussion in footnote 5 to Table 2.6-1. The expanded footnote clarifies the analysis basis for the calculated peak drywell temperatures reported in this table. Note that the footnote now identifies that the analysis basis for peak reported drywell atmosphere temperatures in this table is the same as the basis for the peak drywell wall temperatures. The expanded footnote also now reports the peak drywell temperatures obtained from the short-term DBA-LOCA recirculation suction line break analyses performed with the GEH M3CPT code.

GE-MNGP-AEP-3284 Non-proprietary Information Page 9 of 10 Table 2.6-1 Containment Performance Results CLTP from CLTP with EPU Parameter USAR MethodI EPU Limit Peak Drywell 39.5 43.4 44.1' 563 Pressure (psig)

Peak Drywell 3355 336' 3382,5 3404 Temperature (°F)

Peak Drywell Wall 2735 2775 2785 281 Temperature (fF)

Peak Bulk Suppression Pool 194.2 1936 203/2077 2088 Temperature (F_)

Peak Wetwell 31.2 31.3 32.7 56' Pressure (psig)

Notes:

1. The EPU Method, which was used for the EPU analysis, uses the EPU RTP analysis method with CLTP inputs. The EPU Method includes a more bounding initial containment pressure of 3.0 psig as compared with the CLTP of the USAR, which assumed an initial containment pressure of 2.0 psig. The EPU method also assumes the initial reactor power is at 102% of the RTP.

2. Includes an increase in the assumed initial containment pressure from 2.0 psig of the method of the USAR analysis to 3.0 psig for the EPU Method.

3. The design pressure for the drywell and wetwell is 56 psig. Maximum internal pressure is 62 psig, as shown in USAR Table 5.2-1.

4. Limit for the drywell environmental temperature is increased for EPU from 335°F shown in USAR Table 5.2-8 to 340'F.

5. Peak drywell atmosphere temperatures and peak drywell wall temperatures are calculated assuming a 0.50 sqft steam break into the drywell with UCHIDA condensing heat transfer to the drywell wall to the saturation temperature at the drywell pressure, and initiation of drywell sprays at 10 minutes. The peak drywell atmosphere temperatures obtained from the short-term DBA-LOCA recirculation suction line break analysis are 285.5'F (CLTP from USAR), 290'F (CLTP with EPU Method) and 291'F (EPU).

6. Reduction in peak bulk pool temperature from 194.2°F shown in USAR Table 5.2 to 1937F shown above for CLTP with EPU Method is primarily due to use of a K-value that increases with increasing hot inlet water temperature.

7. The first value is the peak suppression pool temperature for the DBA LOCA with direct suppression pool cooling, 90'F service water temperature, and an RHR heat exchanger K-value that increases with increasing hot inlet water temperature. The second number is the peak suppression pool temperature for the same DBA LOCA and 90'F service water temperature, but with containment cooling using containment sprays and a constant K-value of 147 BTU/sec°F, used for NPSH evaluation.

8. The limit for peak bulk pool temperature, determined as the design temperature for the torus-attached piping, is increased for EPU from 196.7°F (Reference 19) to 208'F.

GE-MNGP-AEP-3284 Non-proprietary Information Page 10 of 10

Reference:

1. GE Nuclear Energy, "Safety Analysis Report For Monticello Nuclear Generating Station Extended Power Uprate," NEDC-33322P, Revision 2, October 2008.

L-MT-13-035 ENCLOSURE 3 MONTICELLO NUCLEAR GENERATING PLANT MODIFICATION TO CALCULATION 03-036, REVISION 2 INSTRUMENT SETPOINT CALCULATION - REACTOR LOW PRESSURE PERMISSIVE BYPASS TIMER Recently the Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy discovered an error in calculation 03-036, Revision 1, "Instrument Setpoint Calculation - Reactor Low Pressure Permissive Bypass Timer."

Calculation 03-036, Revision 1 was provided to the NRC in NSPM letter L-MT-13-019, (Reference E3-1).

The identified error concerns the accuracy of the instrument loop under normal (and trip) conditions. The instrument loop accuracy under normal (and trip) conditions should have used a value of +/- 1.93 minutes. This value should have been applied in Section 6.5.1 for calculation of the Allowable Value. Instead, a value of 1.0 minutes was incorrectly used.

The Allowable Value calculated in 03-036, Revision 1 is thus incorrect and non-conservative, as calculation of the Allowable Value using the corrected 1.93 minutes random error term for loop accuracy would result in an Allowable Value (AV) <= 17.5 minutes (less than the upper AV of 18.0 minutes given in 03-036, Revision 1).

NSPM revised the calculation as follows: The Analytical Limit of <= 19.3 minutes used in 03-036, Revision 1 results in a peak clad temperature of 1500 OF. A higher Analytical Limit allows the necessary calculation revision without impacting the Allowable Value and nominal trip setpoint determined in Revision 1 and does not change the Technical Specification setpoint. Therefore, the Analytical Limit was changed to a peak clad temperature of 1700 OF. 1700 OF is well below the maximum permitted peak clad temperature of 2200 OF and is therefore acceptable. Figure 3e in NEDC-33800P (Ref.

E3-1, Enclosure 2) provides the basis for the revised Analytical Limit. See the attached calculation for more details.

References E3-1 Letter from M A Schimmel (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate (EPU): Response to Request for Additional Information related to Automatic Depressurization System Bypass Timer Setting (TAC MD9990)," L-MT-13-019, dated January 31, 2013. (ADAMS Accession No. ML13037A200) 31 pages follow

QF-0549 (FP-E-CAL-01), Rev. 7 Page 1 of 4 SXcelEnergy" Calculation Signature Sheet Document Information NSPM C alculation (Doc) No: 03-036 NRevision: 2

Title:

Instrument Setpoint Calculation - Reactor. Low Pressure Permissive Bypass Timer Facility: M] MT L PI Unit: E 1 E] 2 Safety Class: E SR El Aug Q LI Non SR Special Codes: EL Safeguards EL Proprietary Type: Calc Sub-Type:

NOTE: Print and sign name in signature blocks, as required.

Major Revisions I N/A EC Number: 20651 EL Vendor Calc Vendor Name or Code: Vendor Doc No:

Description of Revision: Revision 2 of this calculation is being performed to correct the calculation errors identified per Attachment C of the new revision (see AR #01377658).

The following calculation and attachments have been reviewed and deemed acceptable as a legibleQA recorý Prepared by: (sign) / (print) Rhon Sanderson Date: e,_ / -

Reviewed by:(sign) -a,// 13 .,P / (print) Joel Beres Date: - /

Type of Review: [Z Design Verification EL Tech Review LI Suitability Review Method Used (For DV Only): Z1 Review LI Alternate Calc L] Test Approved by: (sign) M P. I (print) Ed Watzl Date: i, 2/-

Minor Revisions . N/A EC No: EL Vendor Caic:

Minor Rev. No:

Description of Change:

Pages Affected:

The following calculation and attachments have been reviewed and deemed LI acceptable as a legible QA record E Prepared by: (sign) I (print) Date:

Reviewed by: (sign) /(print) Date:

Type of Review: EL Design Verification EL Tech Review LI Suitability Review Method Used (For DV Only): EL Review LI Alternate Calc LI Test Approved by: (sign) /(print) Date:

Record Retention: Retain this form with the associated calculation for the life of the plant.

This reference table is used for data entry into the PassPort Controlled Documents Module reference tables (C012 Panel). It may NOTE: also be used as the reference section of the calculation. The input documents, output documents and other references should all be listed here. Add additional lines as needed by using the "TAB" key and filling in the appropriate information in each column.

Reference Documents (PassPort C012 Panel from C020)

Document Doc Ref Type**

1. Controlled* Document Name Doc?+ Type Number Rev INPUT OUTPUT 1 x PROC APPENDIX I (GE METHODOLOGY INSTRUMENTATION & ESM-03.02-APP-I 4 X CONTROLS) 2 x CALC INSTRUMENT DRIFT ANALYSIS, AGASTAT ETR14D3 TIME 03-054 0 X DELAY RELAYS 3 X CALC DETERMINATION OF INSTRUMENT SERVICE CONDITIONS95-027 2 X FOR INPUT INTO SETPOINT CALCULATION 4 x DRAW CORE SPRAY SYSTEM SCHEMATIC DIAGRAM NX-7833-21-1 78 X 5 x DRAW ELEMENTARY DIAGRAM RESIDUAL HEAT REMOVAL SYSTEM NX-7905-46-2. 81 X 6 x PROC ADS SYSTEM 20 MINUTE TIMER TEST 0113-02 11 X 7 x PROC STOPWATCH FUNCTIONAL TEST 1318 04 X 8 -K' CALC AUTOMATIC DEPRESSURIZATION SYSTEM BYPASS TIMER 12-046 0 X 9 X CALC INSTRUMENT SETPOINT CALCULATION - AUTO BLOWDOWN 03-037 0 X INITIATION TIME DELAY RELAY 10 x LIC PLANT SAFETY ANALYSIS - ACCIDENT EVALUATION USAR-14.07 29 X METHODOLOGY 11 x DBD CORE SPRAY SYSTEM DBD-B.03.01 04 12 x DBD RESIDUAL HEAT REMOVAL SYSTEM DBD-B.03.04 06 13 x DRAW S/D RESIDUAL HEAT REMOVEL SYSTEM NX-7905-46-3 76 14 X DRAW SCHEMATIC DIAGRAM RESIDUAL HEAT REMOVAL SYSTEM NX-7905-46-7 76 15 x LIC MNGP TECHNICAL SPECIFICATIONS (AMENDMENT 171) TECH-SPECS 168 X 16 x CALC SEISMIC ANALYSIS OF AGASTAT RELAY 95-035 0 X 17 x CALC AUTOMATIC DEPRESSURIZATION SYSTEM BYPASS TIMER 12-050 X 17_ X CAL CURRENT LICENSED THERMAL POWER Record Retention: Retain this form with the associated calculation for the life of the plant.

Controlled Doc marked with an "X" means the reference can be entered on the C012 panel in black. Unmarked lines will be yellow. If marked with an "X", also list the Doc Type, e.g., CALC, DRAW, VTM, PROC, etc.

Mark with an "X"if the calculation provides inputs and/or outputs or both. Ifnot, leave blank. (Corresponds to. PassPort "Ref Type" codes: Inputs I Both =

"ICALC", Outputs =."OCALC", Other I Unknown = blank)

Other PassPort Data Associated System (PassPort C011, first three columns) OR Equipment References (PassPort C025, all five columns):

Facility Unit System Equipment Type Equipment Number MT I RHR Relay 10A-K95A MT 1 RHR Relay 10A-K95B MT 1 CSP Relay 14A-K27A MT 1 CSP Relay 14A-K27B Superseded Calculations (PassPort C019):

Facility Calc Document Number Title Description Codes - Optional (PassPort C018):

Code Description (optional) Code Description (optional)

Notes (Nts) - Optional (PassPort X293 from C020):

Topic Notes Text

[]Calc introduction []Copy directly from the calculation Intro, Paragraph or []See write-up below El (Specify)

Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-CAL-01), Rev. 7 Page 4 of 4 XceiEnergy Calculation Signature Sheet Monticello Specific Information Z YES F1 N/A Topic Code(s) (See MT Form 3805): RATE, NR737 El YES Z N/A Structural Code(s) (See MT Form 3805):

Does the Calculation:

[] YES [ No Require Fire Protection Review? (Using MT Form 3765, "Fire Protection Program Checklist", determine if a Fire Protection Review is required.) If YES, document the engineering review in the EC. If NO, then attach completed MT Form 3765 to the associated EC.

LI YES [ No Affect piping or supports? (If Yes, Attach MT Form 3544.)

[] YES [ No Affect IST Program Valve or Pump Reference Values, and/or Acceptance Criteria? (IfYes, inform IST Coordinator and provide copy of calculation.)

Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0527 (FP-E-MOD-07) Rev. 4 Page 1 of 1 d Xce Energy- Design Review Checklist EC Number or Document Number / Title / Revision Number: 03-036, Instrument Setpoint Calculation - Reactor Low Pressure Permissive Bypass Timer, Revision 2 Verifier's Name: Joel Beres t3 Discipline: Engineer DESIGN REVIEW CONSIDERATIONS: Yes No N/A

1. Were the inputs correctly selected and incorporated into design? n Li F1

2. Are assumptions necessary to perform the design activity adequately described and M El El reasonable? Where necessary, are the assumptions identified for subsequent re-verifications when the detailed design activities are completed?

3. Are the appropriate quality and quality assurance requirements specified? Li 11

4. Are the applicable codes, standards, and regulatory requirements including issue El ELi and addends properly identified and are their requirements for design met?

Li 0

5. Have applicable construction and operating experience been considered? Li EL Eli

6. Have the design interface requirements been satisfied? El

7. Was an appropriate design method used? EL El ELi

8. Is the output reasonable compared to inputs? EL El

9. Are the specified parts, equipment and processes suitable for the required Li application?

10. Are the specified materials compatible with each other and the design environmental Li Li 2

• conditions to which the material will be exposed?

11. Have adequate maintenance features and requirements been specified? EL Eli

12. Are accessibility and other design provisions adequate for performance of needed E] EL maintenance and repair?

13. Has adequate accessibility been provided to perform the in-service inspection expected to be required during the plant life?

14. Has the design properly considered radiation exposure to the public and plant Li Li []

personnel?

15. Are the acceptance criteria incorporated in the design documents sufficient to allow [] Li Li verification that design requirements have been satisfactorily accomplished?

16. Have adequate pre-operational, subsequent periodic test and inspection Li Li []

requirements been appropriately specified, including acceptance criteria?

17. Are adequate handling, storage, cleaning, and shipping requirements specified? E] Li

18. Are adequate identification requirements specified? ELi M] Eli Li

19. Are requirements for record preparation, review, approval, and retention adequately specified?

20. Have Design and Operational Margins been considered and documented? [ L Li COMMENTS: EL None J] Attached (Use Form QF-0528) Li In EC Topic Notes Form retained in accordance with record retention schedule identified in FP-G-RM-01.

QF-0528 (FP-E-MOD-07) Rev. 1 Sheet 1 of 1 DOCUMENT NUMBER/ TITLE: 03-036 Instrument Setpoint Calculation - Reactor Low Pressure Permissive Bypass Timer REVISION: 2 DATE: 04-08-13 ITEM REVIEWER'S COMMENTS PREPARER'S REVIEWER'S

1. RESOLUTION DISPOSITION 1 Add MELLLA+ to EPU operating Added MELLLA+ to conditions (i.e. EPU/MELLLA+) references to EPU / 11 consistent with 12-046 operating conditions.

2 In ALT computation, sigma value ALT computaton should be 2 sigma value corrected to be "2".

3 Add 12-050 to inputs and include 12-050 listed as Input statement that AL for EPU/MELLLA+ 4.18. Section 6.5.1 1400 seconds from 12-046 is discusses the fact that bounding for CLTP (12-050) in the EPU-MELLLA+

section 6.5.1 operating conditions are bounding.

4 Change QF-0549 revisions to Done.

passport values (add 12-050)

Reviewer: Date: .6'i/z/ Prep arer: Date: -

Page 1 of 1

Table of Contents - CA-03-036 - Revision 2 Document Title Number of Pages Q F-0549 ................................................. 4 Q F-0 5 27 ................................ ,......................... 1 Q F-0528 ..................................................... 1 Table of Contents ........................................... 1 Calculation Body ....................................... 16 Attachment A: Tyco Datasheet ................. 7 Attachment B: License Amendment 170 ....... 1 Total Pages ............................. 31

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 1 of 16

1. Purpose This calculation performs a setpoint calculation for the Reactor Low Pressure Permissive Bypass Timers 1OA-K95A, 10A-K95B, 14A-K27A, and 14A-K27B.

Revision 0 of this calculation was performed to support the extended calibration and surveillance intervals of the time delay relays as part of the 24-Month Fuel Cycle Extension project.

Revision 1 of the calculation was performed in accordance with License Amendment 170, Input 4.14, which removed the lower allowable limit for the Reactor Steam Dome Pressure Permissive - Bypass Timer (Automatic Depressurization system(ADS) bypass timer) of ">= 18 min", previously given in Table 3.3.5.1-1 of Technical Specifications.

Removal of this lower bound from Technical Specifications allowed revision i to derive a time delay setpoint to support both current (CLTP) and future (EPU/ MELLLA+) operating conditions based on information provided in design inputs 4.18 and 4.13. Revision 1 changed the nominal time delay setpoint from 20 minutes to 15 minutes, for the purpose of ensuring that peak cladding temperature remains well below the 10CFR50.46 limit of 2200 deg. F for both current (CLTP) and future (EPU/MELLLA+)

operating power levels.

Revision 2 of this calculation is being performed to correct calculation errors identified in revision 1. Revision 1 of this calculation had errors identified during the NRC review associated with EPU GAP Analysis review (MD9990). CAP AR # 01377658 was initiated*

to drive resolution of this issue (Reference 10.11). The correction of the revision 1 errors requires justification of a higher Analytical Limit for the time delay relays (20.0 minutes vs. 19.3 minutes).

2. Methodology

• This calculation is performed in accordance with ESM-03.02-APP-I (Input 4.1). The General Electric Setpoint Methodology is a statistically based methodology. It recognizes that most of the uncertainties that affect instrument performance are subject to random behavior, and utilizes statistical (probability) estimates of the various uncertainties to achieve conservative, but reasonable, predictions of instrument channel uncertainties. The objective of the statistical approach to setpoint calculations is to achieve a workable compromise between the need to ensure instrument trips when appropriate, and the need to avoid spurious trips that may unnecessarily challenge safety systems or disrupt plant operation.

• Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 2 of 16 Drift values for the time delay relays covered by this calculation were determined in Calculation CA-03-054 (Input 4.4).

The methodology for determining instrument setpoints is not described in the USAR or its references. However, USAR Section 7.1.2.2 does state that MNGP is committed to the GE Setpoint Methodology for instrument setpoint calculations associated with safety limits and Technical Specifications.

3. Acceptance Criteria The setpoint and instrument settings should be established such that there is a 95%

probability that the constructed Analytical Limit will envelope 95% of the instrument population of interest when all applicable instrumentation uncertainties are considered.

4. Design Inputs 4.1 Engineering Standards Manual ESM-03.02-APP-I, Appendix I (GE Methodology Instrumentation & Controls), Revision 4. The ESM provides plant specific guidance on the implementation of the General Electric guidelines (Reference 10.1) and methodology (Reference 10.2).

4.2 Deleted 4.3 Monticello Component Master List (CML). The CML contains instrument information relating to the installed equipment as listed in Section 6.2.

4.4 Calculation CA-03-054, Revision 0, Instrument Drift Analysis, Agastat ETR14D3 Time Delay Relays.

ADE.Random +1.7% Setpoint ADE.Bi a s +0.2% Setpoint Calibration Interval 24 months +25%

4.5 Deleted 4.6 Deleted

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 3 of 16 4.7 Calculation CA-95-027, Revision 2, Determination of Instrument Service Conditions for Input into Setpoint Calculations. Data obtained from this input is listed in Section 6.2. The relays included in this calculation are not listed in CA-95-027. Data for LIS-2-3-672A & C, which are also located in the Cable Spreading Room, is used for this calculation.

4.8 NX-7833-21-1, Revision 78, Core Spray System Schematic Diagram.

14A-K27A, B Agastat ETR14D3N.

4.9 NX-7905-46-2, Revision 80, Elementary Diagram Residual Heat Removal System.

10A-K95A, B Agastat ETR14D3N 4.10 Tyco Electronics, Agastat Nuclear Qualified Control Relays - Series EGP/EML/ETR, 4/24/2002 Edition (Attachment A).

I ETR14D3N relay 125 VDC, I to 30 minutes Repeat Accuracy - Normal Conditions +/-5% Setpoint Repeat Accuracy - Adverse Conditions +/-10% Setpoint The environments for which the relays are expected to trip are similar to the vendor defined normal operating conditions of the relay. However, the instrument operating range minimum temperature is 60 degrees F (Input 4.7) versus the vendor-specified normal environment minimum temperature of 70 degrees F. Therefore, for conservatism, the accuracy for adverse conditions (+/-

10% of setpoint) will be applied in this calculation.

4.11 0113-02, Revision 11, ADS System 20 Minute Timer Test.

As Found Range < 21.7 minutes As Left Range > 19 and < 21 minutes 4.12 1318, Revision 4, Stopwatch Functional Test.

Maximum Allowed Deviation in Test 0.1% Reading (0.06 Sec in 1 Min Test) 4.13 Calculation 12-046, Rev 0, MNGP Automatic Depressurization System Bypass Timer Extended Power Uprate. Data obtained from this input was used to determine an acceptable nominal setpoint to ensure peak cladding temperature

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 4 of 16 (PCT) was limited to well below the 10CFR5O.46 PCT of 2200'F (approximately 17007F) for current and future EPU operating conditions. Figure 3-e of calculation 12-046 shows fuel clad temperature vs. time for the limiting RWCU break analysis.

4.14 License Amendment 170 - Removal of lower allowable limit for 'Reactor Steam Dome Pressure Permissive - Bypass Timer (Pump Permissive)'. (Attachment B is the coversheet).

4.15 Calculation 03-037 Rev 0, Instrument Setpoint Calculation ADS Blowdown Initiation Time Delay Relay 4.16 USAR-14.07 Rev 29, Table 14.7 ECCS Injection Timing Parameters Used in ECCS Performance Evaluations 4.17 Calculation 95-035 Rev 0, Seismic Analysis of Agastat Relay 4.18 Calculation 12-050, Rev. 0, MNGP Automatic Depressurization System Byypass Timer Current Licensed Thermal Power (CLTP). Data obtained from Figure 3-e of 12-050 shows thatthe limiting RWCU break analysis for EPU has a fuel clad temperature time response that bounds the time response for CLTP operating conditions (see Input 4.13).

5. Assumptions None.

6. Analysis 6.1 Instrument Channel Arrangement Channel Dia..ram: L ...................... ......

Dagam Chane ILow-Low

. Water Level Reactor ":

Relay ea iCore

. I Spray/RHR oenitiation~pa/H

%..................~... H.........................r Definition of Channel: Each time delay relay is initiated by the one-of-two-twice low-low reactor water level signal. After the time delay, the relay provides a contact closure to the Core Spray and RHR systems (Input 4.8; References 10.5 and 10.6).

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 5 of 16 6.2 Instrument Definition and Determination of Device Error Terms 6.2.1 Device 1 6.2.1.1 Instrument Definition Reference Component ID 1OA-K95A, B and 14A-K27A, B Location: Admin Building, 939', CSR 4.3 Panels C-32 and C-33 Manufacturer: Agastat 4.8, 4.9 Model Number: ETR14D3N 4.8, 4.9 Upper Range limit: 30 minutes 4.10 Adjustable Range: 1-30 minutes 4.10 Input Signal: Contact Closure 4.8, 10.5, 10.6 Output Signal: Contact Closure 4.8, 10.5, 10.6 6.2.1.2 Process and Physical Interfaces Calibration Conditions: Reference Temperature: 65 to 90°F 4.7 Surveillance Interval: 30 months 4.4 Calibration of the time delay relays is required every operating cycle per Input 4.4. A surveillance interval of 30 months (24 months + 25%) is used in accordance with the guidance in Generic Letter 91-04 (Reference 10.8).

Normal Plant Conditions: Reference Temperature: 60 to 1047F 4.7 Radiation: Negligible 4.7 Pressure: Ambient 4.7 Humidity: 0 to 90% 4.7 Trip Environment Conditions: Reference Temperature: 1047F 4.7 Radiation: Negligible 4.7 Pressure: Ambient 4.7

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 6 of 16 Humidity: 100% 4.7 Seismic Conditions: Reference OBE Prior to Function 1.476 g 4.17 OBE During Function 1.476 g 4.17 Process Conditions: Reference During Calibration: N/A N/A Worst Case: N/A N/A During Function: N/A N/A These relays are not subjected to process conditions (static pressure, overpressure, elevated temperatures, etc.) that would affect the accuracy of the instrument.

6.2.1.3 Individual Device Accuracy Term Value Sigma Reference VA: + 10.0% Setpoint (adverse) 2 4.10

+ 5.0 %Setpoint (normal) 2 ATE: 0 Note l OPE: N/A Note 2 SPE: N/A Note 5 SE: 0 Note 4 RE: 0 Note 7 HE: 0 Note 6 PSE: N/A Note 3 REE: N/A Note 3 Note 1: Accuracy Temperature Effect (ATE) data is not specified for these relays. The ATE is considered part of the Vendor Accuracy since the operating conditions are enveloped by the vendor's qualification limits for operation in adverse conditions.

Note 2: Overpressure Effects (OPE) are not applicable to relays.

Note 3: Error effects due to Power Supply Effects (PSE) and RFI/EMI Effects (REE) are considered negligible for bi-stable electro-mechanical devices (Reference 10.1).

Note 4: Seismic Effects (SE), Section 6.2.1.2 notes the seismic conditions for the relay. These conditions are bounded by the seismic qualified provided by the vendor as described by Input 4.10. Therefore, inaccuracies due to seismic effects are considered to be included in the VA for trip conditions.

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 7 of 16 Note 5: Static Pressure Effects (SPE) do not apply to bi-stable electro-mechanical devices (Reference 10.1).

Note 6: The normal operating conditions of the relays are within the vendor specified operating range of the relay (Input 4.10). Although Input 4.7 gives a humidity of 100% for trip conditions, this is based on assumption and for the applicable Cable Spreading Room environment humidity levels would not be expected to result in condensation. There are no significant instrument accuracy effects that would result from higher (non-condensing) levels of relative humidity. Therefore Humidity Effects (HE) are considered to be included in the VA for trip conditions.

Note 7: Radiation Effects (RE) is not specified for these relays, they are considered to be included in the VA for trip conditions VA Vendor Specifications (Adverse Conditions) = 10% of setpoint per Input 4.10 VA = +/-10 x 20.0 minutes = +/- 2.00 minutes ; Note that the setpoint is conservatively assumed to be the Analytical Limit of 20.0 minutes A~ 2 VAI + (ATE + (OPEjJ +SPEjJ + (E2+ __E) + + (PSEjJ+ _

AN 2x j2.00) +02+02 +b2+ 02+002+02

+ +-02 ALN = +/- 2.00 minutes ALT = +/- 2.00 minutes, as the vendor-specified adverse / abnormal environmental conditions bound the operating environment (with the exception of humidity, see discussion in Note 6 of this Section.

6.2.1.4 Individual Device Drift Term Value VD: Not Specified DIE: Not Specified Vendor drift (VD) is not specified. A Monticello specific drift analysis of Agastat ETR14D3 time delay relays was performed (Input 4.4) to determine the 30 month Analyzed Drift Value (AD) for these transmitters. The AD is used in place of both the VD and the DTE (Drift Temperature Effect):.

Monticello Nuclear Generating Plant CA-03-036

.Title: Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 8 of 16 ADE.Random = +/-1.7% Setpoint AD EBias = + 0.2% Setpoint DL.Random AD E.Random X Setpoint DL.Random z +/- 0.017 x 20.0 minutes DL.Random =+/- 0.34 minutes DL.Bias = ADE.Bias x Setpoint DL-Bias = + 0.002 x 20.0 minutes DL.Bias + 0.04 minutes 6.2.1.5 As-Left Tolerance (ALT)

Per Input 4.1, a suggested ALT is determined with the following equation:

2 VA

+ 2 + +CiSj ALT=+/-2

( (

÷1.00)2+ 0.02 0.0 2 ALT =+ 1.00 minutes Note that vendor accuracy (VA) used to calculate the As Left Tolerance is the typical 5% of setpoint accuracy specified by the vendor for normal operating conditions.

The setpoint is assumed to be the AL of 20.0 minutes.

The existing As-Left Tolerance specified in the surveillance procedure 0113-02 (Input 4.11) is +/- 1.0 minute; the existing ALT of 1.0 minute will remain unchanged. A value of 1.0 minute is reasonable considering the expected 5% of setpoint accuracy expected at typical calibration conditions.

As Left Tolerance (ALT) = +/- 1.0 minutes

Monticello Nuclear Generating Plant CA-03-036

Title:

J Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 9 of 16 6.2.1.6 Device Calibration Error 4

Term Value Sigma Reference C1: 0.02 minutes 3 Note 1 CISTD: 0.02 minutes 3 Note 2 ALT: 1 1 minute 2 Section 6.2.1.5 Note 1: The Calibration Tool Error (C1) is considered equal to the As Found tolerance from the functional test procedure (Input 4.12):

C1 = to0.1%xReading C1 = +/-0.001x20.0 minutes C1 = +0.02 minutes Note 2: In accordance with Input 4.1, the calibration standard error (CIsTD) is considered to be equal to C1.

Since calibration term values are controlled by 100% testing, they are assumed to represent 3-sigma values. Individual calibration error terms are combined usingthe SRSS method and normalized to a 2-sigma confidence level.

2

+I C1sTD2 ALT CL =+/-2x1E-C2- *_ n Z..n +-n CL 0.02 3 3 2 CL =I minute 6.3 Determination of Primary Element Accuracy (PEA) and Process Measurement Accuracy (PMA)

There are no PEA or PMA inaccuracies associated with these relays.

PMA =0 PEA = 0

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 10 of 16 6.4 Determination of Other Error Terms Term Value Indicator Readability/Operator 0.02 minutes Reading Error (ORE)

Resistors, Multiplexers, etc. 0 Software Errors 0 Degradation of Insulation. Resistance 0 (IRE)

An ORE equal to the Calibration Tool Error is applied for readability and operator reaction time.

6.5 Calculation of Allowable Value and Operating Setpoint 6.5.1 Allowable Value (AV):

From Input 4.13, it can be seen that for the bounding scenario of a RWCU break at EPU/MELLLA+ operating conditions with a gate valve a time of 1579 seconds is required to reach 22000 F. Input 4.13 also shows that it takes approximately 1400 seconds to reach 1700° F. Input 4.18 shows that the EPU / MELLLA+ case bounds current operating power (CLTP) conditions with respect to clad temperature vs. time for the limiting RWCU break. In order to provide sufficient time to cool the core, the actuation of ADS should occur prior to reaching 22000 F, therefore additional consideration is given for the following delays in the ADS initiation: from Input 4.13 a delay of 36 seconds from time 0 of the scenario is taken for initiation of the low low level signal, from Input 4.16 the time required for ECCS pumps to reach rated speed is 18 seconds, from Input 4.15 the ADS timer delay is 138 seconds. Therefore an analytical value of 1208 seconds will be used.

1400s - 36s - 18s - 138s = 1208s An Analytical Limit of 1208 seconds ensures actuation of ADS at approximately 1700 degrees F, well before reaching the 2200 degrees F limit. The Analytical Limit will be defined as 1200 seconds, or 20.0 minutes.

Analytical Limit (AL): < 20.0 minutes

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 11 of 16 Term Value (Minutes) Sigma Reference ALT 2.0 2 Section 6.2.1.3 CL 1.0 2 Section 6.2.1.6 PMA 0 2 Section 6.3 PEA 0 2 Section 6.3 IRE 0 N/A Section 6.4 ORE 0.02 2 Section 6.4 AV = AL - (.-5)(VALT2 + CL2 + PMA 2 + PEA2 + IRE 2 + ORE 2 ) + bias terms AV =20.0-(1.645 )(V2.02 +1.02 +02 +02 +02 +0.022)+0 2

AV = 20.0 -1.84 AV = 18.16 minutes As a result of CR 02001013 (Reference 10.9), a new Technical Specification Trip Setting is chosen to bound the As Found values (Refer to Section 6.5.5.).

Conservatively rounding down the calculated AV, the Technical Specification AV per this calculation will be:

AV <= 18.0 minutes 6.5.2 Nominal Trip Setpoint.(NTSP)

Term Valve (Minutes) Sigma Reference ALT + 2.0 2 Section 6.2.1.3 DL.Random + 0.34 2 Section 6.2.1.4 DL.Bias + 0.04 NA Section 6.2.1.4 CL + 1.0 2 Section 6.2.1.6 PMA 0 2 Section 6.3 PEA 0 2 Section 6.3 IRE 0 NA Section 6.4 ORE + 0.02 2 Section 6.4 RAVBias 0 NA Section 6.5.1

Monticello Nuclear Generating Plant CA-03-036.

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 12 of 16 2 +IRE 2)-RAVBia -DLBias NTSP1 = AL - ( 1 6 4 5 )(JALT 2+ CL2 -D L.Random 2 +PMA 2 + PEA 2 +ORE NTSP1 =20.0- (. 6 4 5 )(V2.02 +1.02 + 0.342 +02 +02 +0.022 +02 )-0-0.04 2

NTSP, = 20.0 -1.91 NTSP1 = 18.09 minutes 6.5.3 Licensee Event Report (LER) Avoidance Evaluation The purpose of the LER Avoidance Evaluation is to assure that there is sufficient margin provided between the AV and the NTSP to reasonably avoid violations of the AV. Any Z value greater than 1.29 provides sufficient margin between the NTSP and the AV. Therefore, NTSP 2 is calculated to provide an upper bound for the NTSP based on LER avoidance criteria.

2 2 Sigma' (LER)=+(I)(AJALN2 + CL + DL.Rmom ) + DL.Bia Sigma+ (LER)= +(2)(x/2.02 + 1.02 +0.342)+÷0.04 2

Sigma+ (LER) = +1.171 NTSP 2 = AV - (Z x Sigma' (LER))

NTSP, = 18.0 - (1.29 x 1.171)

NTSP 2 =16.48 Therefore, an NTSP 2 < 16.48, will result in a Z greater than 1.29 and provide sufficient margin between the NTSP and the Allowable Value.

6.5.4 Selection of Operating Setpoint TS = NTSP 2 -ALT TS = 16.48 - 1.0 TS = 15.48 The setpoint will be rounded down to 15 minutes for added conservatism.

NTSP = 15.0 minutes

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 13 of 16 6.5.5 Establishing As-Found Tolerance (AFT)

An As-Found Tolerance is calculated to provide suggested limits for use during the surveillance testing:

AFT =+/--3/VA2 + D L.Radom2 + DTE2 +D L.Bias 2

AFT=+3r1.02 +0.342 +02 + 0.04 2

AFT = -1.54 minutes, + 1.62 minutes The As Found Tolerance (AFT) range will be specified as:

AFT = +/- 1.5 minutes Note that vendor accuracy (VA) used to calculate the As Found Tolerance is the typical 5% of setpoint accuracy specified by the vendor for normal operating conditions. The setpoint is assumed to be the AL of 20.0 minutes.

A review of As-Found data (Input 4.4) shows that these relays have historically performed within the calculated AFT.

6.5.6 Required Limits Evaluation The purpose of a Required Limits Evaluation is to assure that the combination of errors present during calibration of each device in the channel is accounted for while allowing for the possibility that the devices may not be recalibrated. Since Leave Alone Zones are not used at MNGP, the devices are always verified or recalibrated to be within the As Left Zone. Therefore, a Required Limits Evaluation as discussed in the GE methodology is not applicable. Because the calibrated portion of this instrument loop consists only of the timers, the Loop As Found Tolerance is equal to the AFT from Section 6.5.5 above.

Loop AFT = AFT = +/- 1.5 minutes As a result of Condition Report (CR) 02001013 (Reference 10.9), the As Found values are reviewed to verify that the As Found value is not outside the Technical Specification range. The As Found limits are not outside the Technical Specification range, and are therefore acceptable as determined in Section 6.5.5.

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 14 of 16 6.5.7 Spurious Trip Avoidance Evaluation A spurious trip avoidance evaluation is performed to assure that there is a reasonable probability that spurious trips will not occur using the selected setpoint. The margin of the 15.0 minute setpoint to the minimum time delay of 10 minutes is large with respect to instrument accuracies. Spurious trip margin is more than adequate; no formal analysis is required.

6.5.8 Elevation Correction None.

6.5.9 Determination of Action Setpoint The nominal setpoint of 15.0 minutes will be used.

7. Conclusions The results of the calculations are as follows:

Term Value (minutes) Section ALN: + 2.0 6.2.1.3 ALT: + 2.0 6.2.1.3 DL.Random: + 0.34 6.2.1.4 DLBi a s: + 0.04 6.2.1.4 ALT: + 1.0 6.2.1.5 CL: + 1.0 6.2.1.6 PEA: NA 6.3 PMA: NA 6.3 AV (calculated): <= 18.0 6.5.1 NTSP 2: 16.48 6.5.3 Current Trip Setting: 20 + 1.0 4.13 Proposed Trip Setting: 15 + 1.0 6.5.1 AFT: +/- 1.5 6.5.5 AF Limits: >=13.5, < =16.5 6.5.5 Elevation Correction: NA 6.5.8

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 15 of 16

8. Future Needs

1. Revise procedure 0113-02 and supporting documentation as listed on the ADL of EC 20651 to reflect new setpoint. Revise EC 20651 as necessary due to revision 2 of this calculation. The instruments' nominal setpoints, setting tolerances, and the Allowable Value remain unchanged versus revision 1 of this calculation.

9. Attachments Attachment A: Agastat Datasheet for EGP/EML/ETR Series Relays Attachment B: Excerpt from License Amendment 170

10. References 10.1 GE-NE-901-021-0492, DRF AOO-01932-1, Setpoint Calculation Guidelines for the Monticello Nuclear Generating Plant, October 1992.

10.2 General Electric Instrument Setpoint Methodology, NEDC-31336P-A, September 1996.

10.3 DBD B.03.01, Revision 4, Design Bases Document for Core Spray System.

10.4 DBD B.03.04, Revision 6, Design Bases Document for Reactor Heat Removal.

10.5 NX-7905-46-3, Revision 76, Schematic Diagram Residual Heat Removal System.

10.6 NX-7905-46-7, Revision 76, Schematic Diagram Residual Heat Removal System.

10.7 NEDC-32514P, Revision 1, October 1997, Monticello SAFER/GESTER-LOCA Loss-of-Coolant Accident Analysis.

10.8 Generic Letter 91-04, Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.

10.9 Condition Report 02001013, Documentation of NRC Resident Question Regarding the Application of Tech Spec Deviations in As-Found Acceptance Criteria.

Monticello Nuclear Generating Plant CA-03-036

Title:

Instrument Setpoint Calculation Revision 2 Reactor Low Pressure Permissive Bypass Timer Page 16 of 16 10.10 Amendment No. 62 to DPR-22, Dated 03/31/89, Reactor Vessel Level Instrumentation, ADS Logic Changes and S/RV Discharge Pipe Pressure Switch Setpoints.

10.11 CAP AR # 01377658, Errors in Calc 03-036 ADS Bypass Timer Setpoint, 04-05-13

CA-03-034, KEVKý Ac-Tr. A.

4/2412002 Edition Nuclear Qualified Control Relays - Series EGP/EML/ETR SEISMIC AND RADIATION TESTED Inorder to satisfy the growing need for electrical to ANSI/IEEE C37.98 (formerly IEEE Standard control components suitable for class 1Eservice 501-1978, Standard for Seismic Testing of In nuclear power generating stations, AGASTAT Relays).

control relays have been tested for these The design of Series EGP, EML and ETR control applications. Series EGP, EML and ETR have relays has evolved over 20 years of continual use demonstrated compliance with the requirements In a wide range of industrial applications. Power 4, of IEEE Standards 323-1974 (Standard for qualifying Class 1E Equipment for Nuclear Power Relay, Magnetic Latch and Timing Relay versions are available for use with a choice of coil Generating Stations) and IEEE Standard 344-1975 voltages, as well as an Internal fixed or adjustable (Seismic Qualification for Nuclear Power potentiometer Inthe Series ETR time delay Generating Stations). Testing was also referenced version.

TEST PROCEDURE Test Procedure Seismic Aging voltage for AC units, and 80 and 120 percent of AGASTAP control relay Series EGP, EML and Sufficient Interactions were performed at levels rated voltage for DC units, with temperatures ETR were tested in accordance with the require- less than the fragility levels of the devices in ranging from 40'F to 172'F at 95 percent ments of IEEE STD. 323-1974 (Standard for order to satisfy the seismic aging requirements relative humidity.

Qualifying Class 1E Equipment for Nuclear Power of IEEE STI 323-1974 and IEEE STD 344-1975. Baseline Performance Generating Stations), IEEE STD. 344-1975 Seismic Qualification In addition to aging tests, a series of baseline (Seismic Qualification for Nuclear Power Artificially aged relays were subjected to simulated tests were conducted before, and Immediately Generating Stations) and referenced to ANSIAEEE seismic vibration, which verified the ability of the after each aging sequence, inthe following areas:

C37.98 (formerly IEEE Standard 501-1978, Individual device to perform its required function Pull-in Voltage Standard for Seismic Testing of Relays). The before, during and/or following design basis Drop-out Voltage relays were tested according to parameters earthquakes. Relays were tested In the non- Dielectric Strength at 1650V 60Hz which, In practice, should encompass the operating, operating and transitional modes. Insulation Resistance majority of applications. Documented data applies Operate Time (milliseconds) to relays which were mounled on rigid test Hostile Environment Since the relays are intended for use In auxiliary Recycle Time (milliseconds) fixtures. The following descriptions of the tests Time Delay (seconds) Series ETR performed are presented In their actual sequence. and control buildings, and not inthe reactor containment areas, a hostile environment test Repeatability (percent) f only Radiation Aging was performed in place of the Loss of Coolant Contact Bounce Relays were subjected to a radiation dosage of Accident (LOCA) test. Relays were subjected to (milliseconds at 28VDC, 1 amp.)

2.0 X101 Reds, which is considered to exceed combination extreme temperature/humidity plus Contact Resistance adverse plant operating requirements for such under/over voltage testing to prove their ability (milliohms at 28VDC, 1 amp.)

areas as auxiliary and control buildings. to function under adverse conditions even after having undergone all the previous aging Data was measured and recorded and used for Cycling with Load Aging The radiated units were then subjected to simulation and seismic testing. The devices comparison throughout the qualification test 27,500 operations at accelerated rate, with one were operated at minimum and maximum ' program in order to detect any degradation of set of contacts loaded to 120VAC, 60Hz at 10 voltage extremes: 85 and 120 percent of rated performance.

amps; or 125VDC at 1 amp, and the number of mechanical operations exceeding those PM=-Wn* C KWeECT~nRLMieP..a UnaGDEWTESa, experienced in actual service. an ~ ~

rc ý MnI Temperature Aging 100- Hal This test subjected the relays to a temperature I I II mill I I f Illll I11II IIIn ofl 00°C for 42 days, With performance I1 measured before and after thermal stress. -- __ . II In- -~ - --

The SRS shape (et 5 percent damping), Is defined by four points: ~ -

point A = 1.0 Hz and an acceleration equal to 25 percent of the Zero PeriodAcceieratioe (ZPA) point 0 . 4.0 Hzand 250 percent of the ZPA -r point E - 16.0 'z and 250 percent of tire ZPA point G - 33.0 -z and a level equal to the ZPA SPECIMEN13,15 & 1t (EGPSERIES)

RELAY STATE:NON-OPERATE MODE(iE-ENER.) - - eccacecace TESTRUNNO. one, 319, (205-206). (198-R99)

AXISH + V): -

COMPOSITEOF FEN-,SSN, FeN+ X .707 DUETO 45° INCLINATION OFTESTMACHINE.

Figure L. Model EGP,Response Seltrum, Novi-OperateMode AdditionalSeismic Response Curvesare available an request. ~~~1 10 .

• 00o 100o0 Relay State: Non-Operate Made (De-enee)

Test Run No. 31 . 319, (205-208). (198-199) tqCo/ Electronics Specifications subject to change Technical Support Center 1-800-522-6752 Dimensions are for reference only. www.tycoelectreonics.cor 1

CA-r -,A.

Nuclear Qualified Control Relays OPERATION Series EGP Power Relay m! Series ETR Time Delay Relay Applying a continuous voltage to the coil (B3- (Delay onEnerglzation)

B4) energizes the coil and instantaneously LE -- i .Applying a continuous voltage to the Input transfers the switch, breaking the normally terminals (B1-B4) starts a time delay lasting for closed contacts (M1-R1, M2-R2, M3-i3, M4- the preset time period. During this period the R4) and making the normally open contacts normally closed contacts (Four M-R sets)

(Mi1-T1, M2-T2, M3-T3, M4-T4). The contacts remain closed. At the end of the delay period, remain inthis transferred position until the coil - *the normally closed contacts break and the Isdeenergized, at which time the switch U -*-,U - nnnannlf, lI,"nr fl-T entoN mnieo instantaneously returns the contacts to their The contacts remain Inthis position until the original position. relay Is deenergized, at which time the contacts instantaneously return to their normal position.

Continuous Duty Wiring Deenergizing the relay, either during or after the Since the double wound coil does not have a delay period will recycle the unit within .075 continuous duty rating, voltage pulses to the second. Itwill then provide a full delay period coils should not exceed a ratio of 40% on, to upon reenergizatlon, regardless of how often 60% off, with maximum power-on periods not the voltage is Interrupted before the unit has to exceed 10 minutes. been permitted to "time-out' to its full delay INPUT If continuous energizing only is available, a setting.

resistor/capacitor network should be connected as shown below. In this case the shortest time between two operations must not te less than 5 seconds.

The relay will always assume the energized position in the event of both windings being energized simultaneously. ,PUT '

IN Series EML ItIs advisable not to put another load In Magnetic Latch parallel with the windings of the ML relay.

Application of a voltage to the latching input (B1-B4) will cause the relay to latch In(Make MLSeries Reloy far DCoperatlon with a the N.O. Contacts, break the N.C. Contacts). reslslor/capacitor network When this voltage Isremoved, the relay will remain Inthis "Latched" condition. Application of a voltage to the un-latching input (B3-B4) will cause the relay to dropout (Break the N.O.

Contacts, make the N.C. Contacts). When this PRESETTIMEDELAY voltage Is removed, the relay will remain Inthis "Unlatched" condition. i ENERGIZED INPUT TI L.--- EENERGIZED LATCH INPUT ENERGIZED I j r__CLO ED N.C.CONTACTS L-J---- OPEN (FOURM.-RSETS) I CLSE UNLATCH INPiUTI CNERGIZED Raslslarl B3-B4 OEENERGIZED Gagcllor L t--OPEN N.C.CONTACTS (FOURM-TSETS)

N.C.CONLACTS

• CLOSED (FOUR M-R SETS), OPEN R-C Values N.O.CONTACTS CLOSED (FOUR M-T SETS) OPEN Nominal - R C Voltage OHMS VDO +/-:5% Walts UF VMDC Wiring Diagram (Wiring and Connections)

The ML relay has three terminals for the 12 62 2 5000 15 windings: latching winding between terminals 24 240 2 2000 50 B1and B4, un-latching winding between terminals B3 and B4. 48 1000 2 500 100 The ML Relay is not symmetrical due to its 125 6200 2 150 150 three coil connections The relays are normally delivered polarized so that terminal B4 carries the negative voltage.

To reverse the polarity, a deenergize/energize cycle should be carried out using a voltage 50%

greater than the normal rating.

tqo / E/ectronics Specificatlons subject to change Technical Support Center 1-800-522-6752 Dimensions are for reference only. 2 www.tycoelectronics.com

&-w, k NGAr i o Nuclear Qualified Control Relays SPECIFICATIONS Con...ac.. Ratings...

Contact Ratings The date of manufacture can be found in the first Series EGP/EML/ETR four (4) digits of the serial number on the nameplate Contact Capacity InAmperes (Resistive)

First two digits Indicate the XX XX Contact MlI.1,000,000 year.

Voltage Operations 24 vdc 10.0 amps Second two digits Indicate the 125 vdc 1.0 amp week.

120 vac, 60 Hz 10.0 amps 240 vsac,60 Hz 7.5 amps Example In the date code '7814" below:

"78" Indicates the year 1978; Contact Ratings, UL- Series EGP/EML Only "14" Indicates the 14th week Contact ratings as listed under the Underwriters (or April 3 through april 7).

Laboratory Component Recognition Program. S Medel (Two poles per load):

1/3 Horsepower, 120 vac 10 amps, General Purpose, 240 vac Colt 125 VDC 120 vdc, 1.0 amp Serial 78140028 Mechanical Life - Series EGP/EML/ETR 25,000 mechanical operations Approximate Weight - Series EGP/EML/ETR Note 1 lb. TycoFlectronles Corporation does not recommend the use ofits products Inthe containment areas ofNuclear Power Transient Protection - Series ETR Only Generating Stations.

A1500 volt.transient of less than 100 microseconds, or 1000 volts of less than 1 millisecond will not affect timing accuracy.

Timing Adjustment- Series ETR Only Internal Fixed Internal Potentiometer Time Ranges - Series ETR Only

.15 to 3 Sec. 4 to 120 Sec.

.55 to 15 Sec. 10 to 300 Sec.

1 to 30 Sec. 2 to 60 Min.

2 to 60 Sec. 1 to 30 Min.

Repeat Accuracy - Series ETR Only The repeat accuracy deviation (AR) of a time-delay relay is a measure of the maximum deviation In the time-delay that will be experienced In five successive operations at any particular time setting of the relay and over the operating voltage and temperature range specified. Repeat accuracy Isobtained from the following formula:

Replacement Schedule - Series = T,-Th)

EGPIEML/ETR 10 0

-Tif+ T')

The qualified life of these relays is Where -

25,000 electrical operations or 10 years T,= Maximum Time Delay.

from the date of manufacture, whichever Ti = Minimum Time Delay.

occurs first.

tqJco/ Electronics Specificatlons subject to change Technical Support Center 1-600-522-6752 Dimensions are for reference only. 3 www.tycoelectronlis.com

PkrT I.\

AGASTA- '1/2 Nuclear Qualified Control Relays OPERATING CHARACTERISTICS Environmental Cendilons (Qoualified Life)- Series EGP/EML/ETR Parameter Min. Normal Max.

Temperature ("F) 40 70-104 156 Humidity (R.H. %) 10. 40-60 95 Pressure AAtmospheric Radiation (rads) - 2.0 x 10' (Gamma)

Operating Conditions, Normal Environment Sernes EGP/EMLIETR Normal Operating Specillcatlons With DOCtals With ACCells EGP EML ETR EOP ETHI CollOperating Voltage, Nominal (rated)" As Spec. As Spec, As Spec. As Spec. As Spec.

Pull-in (% of rated value) 80% MIn. 85% Min. 80%Min. 85% Min. 85% MIn.

Drop-out (% of rated value) 5-45% 85% Min. 5-45% 5-45% 5-50%

Continuous (% of rated value) 110% Max. N/A 110% Max. 110% Max. 110% Max.

Power (Watts at rated value)

Pull-in 6 Apprx. 15 Apprx. 6 Apprx 6 Apprx. 6 Apprx Drop-out N/A 10 Apprx. N/A N/A N/A Relay Operate Time 30 ms Max. 25 ms Max. N/A 35 ms Max. N/A Withmin.

latch pulse ofa30 m.

Relay Release (Recycle) lime 25 ms Max. 20 ms Max. 75 ma Max. 85 ms Max. 75 ms Max.

With mnl.

latch pulse of 30 ms.

Contact Ratings, Continuous Reslsltve at 125 vdc 1.0 amp. 1.0 amp. 1.0 amp. 1.0 amp. 1.0 amp.

Resistive at 120 vac, 60 Hz 10.0 amp. 10.0 amp. 10.0 amp. 10.0 amp. 10.0 amp.

Insulatlon Resistance (in megohmo at 500 vdc) 500 Mn. 500 MIn. 500 MIn. 500MIn. 500 MIn.

Dielectric (voma,60 Hz)

Between Teormnals and Ground 1,500 1,500 1,500 1,500 1,500 Between Non-connected Terminals 1,500 1,500 1,500 1,500 1,500 Repeat Accuracy N/A N/A :5% N/A +/-5%

Operating Cortitionrs, Abnormal Environment- Series EGP/EML Adverse Operating Speeifications Normal 18 "A" 00 .'1 D, "C" DB "."

Temperature (°F)70-104 40 120 145 156 Humidity (R.H.%) 40-60 10-95 10-95 10-95 10-95 ColtOperating Voltage (% of rated)*

AC (Series EGPonly)85-110 85-110 85-110 85-110 85-110 0C (Series UOPonly)80-110 80-110 80-110 80-110 80-110 tC (Series EMLonly)85-110 85-110 85-110 85-110 85-110 Relay Operate Time (ms)

AC (Series EGPonly) 35 Max. 35 Max. 35 Max. 35 Max. 35 Max.

DC(Series tGP, Sodas EML) 30 Max. 25 Max. 37 Max. 40 Max. 40 Max.

Operating Conditlens, Abnormal Environment - Series ETR Adverse Operating Specitications With DCColns With ACColts CollOperating Voltage (rated)* As Spec, As Spec.

Pull-In (%of rated value) 80% Min. 85% MIn.

Continuous (%of rated value) 110% Max. 110% Max.

Drop-out (% of rated value) 5-45% 5-50%

Power (Watts at rated value) 6 Apprx. 6 Apprx.

Relay Release (Recycle) lime 75 ms Max. 75 mis Max.

Contact Ratings, Continuous Resistive at125 vdc 1,0 amp. 1,0 amp.

Resistive at 120 va0, 60 Hz 10.0amp. 10.0 amp.

Repeat Accuracy +/-10% +/-10%

'All colls may b0 operated on Iolermittent daty cycles at voltages 10% above blrtedmaximums (Intermittent Duty = Maximum 50% duty cycle and 30 minutes "ON" time.)

tqco/ Electronics Specficationso subject to change Technical Support Center 1-800-522-6752 Dimensions re for reference only. 4 www.tycoetectronisc.com

CA-0..:-

ýITTA- AI I,

PG&W Nuclear Qualified Control Relays DIMENSIONS AND MOUNTING Xaca11-M ýIl.

Series EGP,EMLand ETRAGASTAP control relays must be mounted In thehorizontalposition;,performance speclifcations of these unitsare valid only when they are mounted an Indicated In either of the above drawings.

tqc* / Electronics Specifications subject to change Technical Support Center 1-800-522-6752 Dimensions are for reference only. 5 www.tycoelectronlcs.com

A-TT .

Nuclear Qualified Control Relays ORDERING INFORMATION Catalog Number Code - Series EGP and EML El EI rI Nuclear AGASTATP Coll Configuration Safely Control Voltage Code*

Related Relay Model Code Code Code Code 004 SE GP - Power - A- 12VOC Relay B- 24 VOC ML - Magnetic DC C D--125 48 VDC VDC Latch L E-110VDC F -250 VDC(Series EGP Only)

G- 24 VAC 60 Hz (Series EGP Only)

H - 48 VAC60 Hz (Series EGP Only)

AC 1- 120 VAC60 Hz (Series EGP Only)

L J - 220 VAC60 Hz (Series EGP Only)

• Configuration Code The Conllguration Code Is a suffix to the Model Number which provides a means of Identlflcatlon. When a significant product change Is Introduced, the Configuration code and speciflcation sheets will be revised.

NJ El Ii I I.

Nuclear AGASTATO Operating Timing Time Configuration Safety Control

• Voltage Adjustment Range Cede Related Relay Model Code Code Code Code Code Code I- B - 24 VDC 1 - Internal A-.15to 3 se. 004 E TR14 - Time DC D - 125 VDC Fixed 8 -. 55 to 15 sec.

Delay 3 - Internal C- 1 to 30 sec.

Relay AC 1- 120 VAC60Hz Potenttometer D- 2 to 60 sec.

f.

(Delay E - 4 to 120 sec.

on G - 10 to 300 sec.

Pull-In) I - 2 to 60 min.

N- 1 to 30 min.

• Centlgurallon Code The Configuration Code Is a oufflx to the Model Number which provides a means ofIdentification. When a significant product change Is Introduced, the Configuration code and specification sheets wilt be revised.

tqco/ Electronics Specltications subject to change Technical Support Center 1-800-522-6752 DImensions are or reterence only. 6 www.tycoetectronlcs.com

0A-03-@34 vX L AvTT. A.

Relay Classifications Control Code Summary CONFIGURATION CONTROL

-aaa.rt.x r~L4t.-~rn2at.a.~ ~..z......~.

Product Code - 001 Code- 002 Code- 003 Code - 004 EGP Contains all materials present in Nov. 1981 - Material change to Dec. 1987 - Material change on Dec. 1995 - Material change on original qualification testing. coil wrapping tape and lead wire leaf spring from nickel copper bobbin from Nylon Zytel 101 to insulation to improve thermal to beryllium copper. Rynite FR530. Material change life. on base from Melamine Phenolic to Grion PMV-5HVO.

EML Contains all materials present in Nov. 1981 - Material change to Dec. 1987- Material change on Dec. 1995 - Material change on original qualification testing. coil wrapping tape and lead wire leaf spring from nickel copper bobbin from Nylon Zytel 101 to insulation to improve thermal to beryllium copper. Rynite FR530. Material change life. on base from Melamine Phenolic to Gnlion PMV-5HVO, ETR Contains all materials present in Nov. 1981 - Material change to Dec. 1987 - Material change on Dec. 1995 - Material change on original qualification testing. coil wrapping tape and lead wire leat spring from nickel copper bobbin from Nylon Zytel 101 to Insulation to Improve thermal to beryllium copper. Rynite FR530. Material change life. on base from Melamine Phenolic to Grllon PMV-5HVO.

ECRO001l Contains all materials present in June 1989 - Material change original qualification testing. from Noryl N-225 std. black to Noryl SE-1-701AA black.

ECROO02 Contains all materials present in original qualification testing.

June 1989 - Material change from Noryl N-225 std. black to ECROO95 Contains all materials present In Noryl SE-I-701AA black.

original qualification testing.

ECRO133 Contains all materials present in original qualification testing.

ECRO155 Contains all materials present In original qualification testing.

.1. ____________________ .1 .1.

Configuration Code: The Configuration code is a suffix to the Model Number which provides a means of Identification. When a significant product change is introduced, the Configuration code and specification sheets will be revised. (001, 002, 003, 004, etc.).

tqLcn/ Electronics Speclfications subject to change Technical Support Center 1-800-M22-6752 bDimensions are for reference only. www.tycoelectronlcs.com 7

A-I-. B.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 7, 2012 Mr. Mark A. Schimmel Site Vice President Monticello Nuclear Generating Plant Northern States Power Company - Minnesota 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT REGARDING THE AUTOMATIC DEPRESSURIZATION SYSTEM BYPASS AIMER (TAC NO. ME8345)

Dear Schimmel:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 170 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant.

The amendment consists of changes to the technical specifications (TSs) in response to your application dated April 5, 2012.

The amendment revises TSs to eliminate the lower allowable value limit of "> 18 minutes" for Functions I.e and 2.e, 'Reactor Steam Dome Pressure Permissive - Bypass Timer (Pump Permissive)," in Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation."

Please note that the NRC staff declined to issue this amendment by the licensee's requested issuance date of June 6, 2012; the application did not provide any reason for shortening the regulatory 60-day notice period during which interested parties may petition for a hearing.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly FederalRegister notice.

Sincerely, Ta , S ior Project Manager Plant Licensing Branch Il1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 170 to DPR-22

2. Safety Evaluation cc wlencls: Distribution via ListServ