ML12088A170

Response to NRC Second Request for Additional Information Regarding Application to Modify Technical Specifications for Use of Areva Advanced W17 Htp Fuel (TS-SQN-2011-07)

ML12088A170
Person / Time
Site:	Sequoyah
Issue date:	03/23/2012
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TS-SQN-2011-07
Download: ML12088A170 (36)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 March 23, 2012 10 CFR 50.4 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

References:

Printed on recycled paper Response to NRC Second Request for Additional Information Regarding Application to Modify Technical Specifications for Use of AREVA Advanced W17 HTP Fuel (TS-SQN-2011-07)

1. Letter from TVA to NRC, "Application to Modify Technical Specifications for Use of AREVA Advanced W1 7 HTP Fuel (TS-SQN-2011-07)," dated June 17, 2011

2. Letter from TVA to NRC, "Response to NRC Request for Supplemental Information Regarding Application to Modify Technical Specifications for Use of AREVA Advanced W1 7 HTP Fuel (TS-SQN-2011-07)," dated July 27, 2011

3. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Units 1 and 2 -

Request for Additional Information Regarding the Propopsed [sic]

Technical Specification Changes to Allow Use of AREVA Advanced W17 High Thermal Performance Fuel (TAC Nos. ME6538 and ME6539)," dated October 14, 2011

4. Letter from TVA to NRC, "Response to NRC Request for Additional Information Regarding Application to Modify Technical Specifications for Use of AREVA Advanced W1 7 HTP Fuel (TS-SQN-2011-07)," dated November 14, 2011

ý)o

U.S. Nuclear Regulatory Commission Page 2 March 23, 2012

5. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Units 1 and 2 -

Request for Additional Information Regarding the Propopsed [sic]

Technical Specification Changes to Allow Use of AREVA Advanced W17 High Thermal Performance Fuel (TAC Nos. ME6538 and ME6539), dated February 8, 2012 By letter dated June 17, 2011 (Reference 1), the Tennessee Valley Authority (TVA) submitted a request for amendment to the Technical Specifications (TSs) for Sequoyah Nuclear Plant (SQN), Units 1 and 2. The amendment request proposed to allow use of AREVA Advanced W17 High Thermal Performance (HTP) fuel at SQN. In a letter dated July 27, 2011 (Referen'ce 2), TVA provided supplemental information requested by the Nuclear Regulatory Commission (NRC) on July 21, 2011, in a telephone call between TVA and NRC representatives. By letter dated October 14, 2011 (Reference 3), the NRC forwarded a request for additional information (RAI) regarding the proposed changes to the TSs. TVA provided the response to the NRC's RAI in a letter dated November 14, 2011 (Reference 4). A second RAI regarding the proposed TS changes was forwarded to TVA in a letter dated February 8, 2012 (Reference 5). The letter noted that TVA had agreed to provide the response within 30 days from receipt of the RAI letter.

On February 15, 2012, in a phone call with the NRC Project Manager for SQN, TVA requested a 14 day extension to the original 30 day due date in order to revise the previously submitted Core Safety Limit Lines (CSLLs) in TS Figure 2.1-1 and provide the additional supporting analytical basis information. The CSLLs in TS Figure 2.1-1 have been determined to be non-conservative for the core peaking limits used in the current operating cycles for SQN, Units 1 and 2. Administrative controls have been imposed consistent with the SQN Corrective Action Program and the guidance in Administrative Letter 98-10 to ensure the CSLLs are not violated. The change to the CSLLs in TS Figure 2.1-1 is needed because the CSLLs proposed in Reference 1 would have resulted in the core peaking limits being overly restrictive. Accordingly, the NRC agreed to the 14 day extension, and revised the due date to March 23, 2012 for submittal of the RAI response. The enclosures to this letter provide the requested response to the NRC's second RAI and the revised TS Figure 2.1-1 for SQN, Units 1 and 2. Associated changes to the TS Bases are also provided for NRC review. contains TVA's responses to the Reference 5 RAI, and includes information provided by AREVA NP Inc. (AREVA NP) which they consider to be proprietary in nature, pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding," paragraph (a)(4). It is requested that this information be withheld from public disclosure. Enclosure 2 provides a redacted version of the proprietary Enclosure 1 with the proprietary material removed, and which is suitable for public disclosure. Enclosure 3 to this letter contains the affidavit supporting the request for withholding the information from public disclosure.

Section 3 of Enclosures 1 and 2 provided the responses to the first NRC RAI (Reference 4) regarding the proposed use of AREVA Advanced W1 7 HTP fuel.

Section 4 of these enclosures provides the responses to the second NRC RAI (Reference 5) regarding the subject use of AREVA HTP fuel.

U.S. Nuclear Regulatory Commission Page 3 March 23, 2012 Modified proposed TS Mark-Up Insert pages are provided in Enclosure 4 to incorporate the revised TS Figure 2.1-1 CSLLs for SQN, Units 1 and 2. The modified Final Typed pages incorporating the revised TS Figure 2.1-1 CSLLs are provided in. The details regarding the reason for the minor changes to the previously submitted TS Figure 2.1-1 CSLLs are included in the response to RAI Question 8, which is provided in Subsection 4.8 of Enclosures 1 and 2 to this letter. The modified figure comparing the CSLLs for the current Mark-BW fuel with that of the proposed Advanced W1 7 HTP fuel, previously presented as Figure 1 of Reference 1, is included in Subsection 4.8 of Enclosures 1 and 2 as Figure 5. Modified Mark-Up and Final Typed pages for TS Bases Section 2.1.1 are provided In Enclosures 5 and 7, respectively. No other changes have been made to the previous submittals regarding the proposed TS amendment to allow use of AREVA Advanced W1 7 HTP fuel at SQN.

Consistent with the standards set forth in 10 CFR 50.92(c), TVA has determined that the proposed minor changes to the TS Figure 2.1-1 CSLLs provided in this letter do not affect the no significant hazards considerations associated with the proposed TS amendment provided in Reference 1. TVA has further determined that the proposed amendment still qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and the enclosures to the Tennessee State Department of Environment and Conservation.

There are no regulatory commitments associated with this submittal. Please address any questions regarding this request to Clyde Mackaman at 423-751-2834.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2 3 rd day of March, 2012.

Res c ully, Mner,Corporate Nuclear Licensing

Enclosures:

1. ANP-3053(P), Revision 2, Sequoyah HTP Fuel Transition - NRC RAls and Responses, March 2012 (Proprietary Version)

2. ANP-3053(NP), Revision 2, Sequoyah HTP Fuel Transition - NRC RAls and Responses, March 2012 (Non-Proprietary Version)

3. AREVA NP Affidavit

4. Modified Proposed Technical Specification Figure 2.1-1 (Mark-Up Insert)

5. Modified Technical Specification Bases Section 2.1.1 (Mark-Up)

6. Modified Proposed Technical Specification Figure 2.1-1 (Final Typed)

7. Modified Technical Specification Bases Section 2.1.1 (Final Typed) cc: See page 4

U.S. Nuclear Regulatory Commission Page 4 March 23, 2012 cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS I AND 2 AREVA NP Affidavit Attached is the affidavit supporting the request to withhold the proprietary information included in Enclosure 1 from public disclosure in accordance with 10 CFR 2.390, "Public inspections, exemptions, requests for withholding," paragraph (a)(4).

AFFIDAVIT STATE OF WASHINGTON

)

) ss.

COUNTY OF BENTON

)

1.

My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2.

I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3.

I am familiar with the AREVA NP information contained in the report ANP-3053(P), Revision 2, "Sequoyah HTP Fuel Transition - NRC RAIs and Responses," dated March 2012 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6.

The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a)

The information reveals details of AREVA NP's research and development plans and programs or their results.

(b)

Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c)

The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d)

The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e)

The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

. 7.

In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8.

AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this

/ 6 day of //

C-.

2012.

NOTA Susan K. McCoy e,,

NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 1/14/2016

ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS I AND 2 Modified Proposed Technical Specification Figure 2.1-1 (Mark-Up Insert)

This enclosure provides the modified proposed Technical Specification Figure 2.1-1 mark-up insert pages for Units 1 and 2. These pages replace the corresponding pages originally provided in Attachment 1 to the enclosure of the letter from TVA to NRC, "Application to Modify Technical Specifications for Use of AREVA Advanced W17 HTP Fuel (TS-SQN-2011-07)," dated June 17, 2011 (Reference 1 of the cover letter for this submittal).

Sequoyah Nuclear Plant Unit I Proposed Technical Specification Figure 2.1-1 (Mark-Up Insert)

Insert SQN U1 Figure 2.1-1 680.

24040i 620 U..

I-580 ACPTABLE_____

CPERATJON 540 0.D 0.2 0.4 D.6.

8.'s 1.0

".2 FRACTION OF RATED THERMAL POWER

Sequoyah Nuclear Plant Unit 2 Proposed Technical Specification Figure 2.1-1 (Mark-Up Insert)

Insert SQN U2 Figure 2.1-1 UNACCEPTABLE 641 CPERATiC*4 52~

~

~

~

~

ACETBE 62*

1775 psia OPERATION 540 0.0 U.2 0.4 0.6 G..

10M 1.2 FRACTION OF RATED THERMAL POWER

ENCLOSURE 5 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS I AND 2 Modified Technical Specification Bases Section 2.1.1 (Mark-Up)

This enclosure provides the modified Technical Specification Bases Section 2.1.1 mark-up pages for Units 1 and 2. These pages replace the corresponding pages originally provided in Attachment 2 to the enclosure of the letter from TVA to NRC, "Application to Modify Technical Specifications for Use of AREVA Advanced W1 7 HTP Fuel (TS-SQN-2011-07)," dated June 17, 2011 (Reference 1 of the cover letter for this submittal).

Sequoyah Nuclear Plant Unit 1 Proposed Technical Specification Bases Section 2.1.1 (Mark-Up)

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE

-INSERT1 thit The restrictions of this safety limit prevent overheating of the fuel,

b cladding perforation whie* would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation tmINSERT 2corresponding significant Operation above the upper boundary of the nucleate boiling regime could result in excessive Gladdwng temperatures because of the onsqt of departure from nucleate boiling (DNB) and the FreA'I-'=t

_INSERT_3_

r-L hW reduction in heat transfer coefficientKD NB is not a directly measurable parameter during operation L*t eforiTHERMAL POWER and Reactor Coolant Temperature and Pressure have been related to lLJ DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the design DNBR limit.

I-INSERT 4 In meeting this design basis, uneortai~ntiacr plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is 9reater [INSERT 5' than or equal to the DNBR limit. I. h w.....int a---e plan, pa~aFRe

..; Used te deter,,ln5 the plant unccrtainty. This DNBR uner c

ed ýith the cr Io DBR limit, establishes a docig, D.

R Yalu. which mu.,.t be mc in.p.nt ty an.alyss ucing.aluc.

.f iRput paramete.. withou. t Theo c-UrPOPc Of Figure 2.1 1 chow the loci of poinitc of THERM-AL. POWE.9R, ReactorF CAol-Ant Syctom proccura arnd

'ercag tompertur for 4

"which the minimum DNBR c no les than the saf^'"

anolycic DNBR limit, or th1~rg nhalpy at the voccal exit is equal to the enthalpy of caturated liquid.

The eur~es of Figur~e 2.1 1 are based on an enthalpy rice hot channel "faet *-'-.;,- " nominal "'ue ha.. been.. r du d to iu de a R 0.4 tata, r.d poWeFr un, cortainty factor), and a fa tF,+W Aem na Yal*A

,.1 v,4,A

,Av F~l*

refeFrence esocine with a peak of 1.65 for axial power shape. An alieowanec ic inluedfr a;ninraen at"redu1based A the Axprecci'n:

79I 4

I1 f'

1'+3 THERMALB POWER

",eeP=

RATED THERMAL POWER 4.0-Mafk -

WFe Nemittal Vnaucj 44-AKcflinghuzFi SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19,114,138, 155, 223

SAFETY LIMITS BASES These limiting hoat flux conditionc aV highor than thoe* eealculated for the rango of all oontrVl F9616 fully I..4ithdrnwni to-the FRnaRMOum-R

-AlloW.Ablot oon-.trA-l rodt- *inco.F*in accuiming the axial poworibaaF sc within the Iimfitc, of the f, (Delta I) fun~tiOn Af the Gvei~empeffitur Dolte T-trp. WhoI h ~a oc FimbRAlAnc ic twithin the toloAnco, the axial poWo imaac foto h Oooprtr ot trips Will Froduc the

,etpeintl to prvi-dl protcvtin con*,

R,*,,tntwih iro,

,afo "Mot.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.11967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydro tested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Nominal Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Nominal Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Nominal Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Nominal Trip Setpoint and the Allowable Value is equal to or less than the rack allowance assumed for each trip in the safety analyses.

Technical specifications are required by 10 CFR 50.36 to contain Limiting Safety System Settings (LSSS) defined by the regulation as ".... settings for automatic protective devices.., so chosen that automatic protective action will correct the abnormal situation before a Safety Limit (SL) is exceeded."

The analytic limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the analytic limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protective devices must be chosen to be more conservative than the analytic limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

Sopt-mben 13, 2006 SEQUOYAH - UNIT 1 B 2-2 Amendment No. 310

INSERT 1 cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt), either of which could result in INSERT 2 Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs.

INSERT 3 from the outer surface of the cladding to the reactor coolant water. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

INSERT 4 To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertanties

INSERT 5 This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These lines are bounding for all fuel types. The curves in Figure 2.1-1 are based upon enthalpy rise hot channel factors that result in acceptable DNBR performance of each fuel type. Acceptable DNBR performance is assured by operation within the DNB-based Limiting Safety Limit System Settings (RPS trip limits). The plant trip setpoints are verified to be less than the limits defined by the safety limit lines in Figure 2.1-1 converted from power to delta-temperature and adjusted for uncertainty.

Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the acceptance criteria in the safety analysis.

The limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (Al), is within the limits of the f, (AI) function of the Overtemperature Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the fl(AI) trip reset function, the Overtemperature Delta-Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for centerline fuel melt are higher than those calculated for the range of all control rods from the fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (Al), is within the limits of the f2(AI) function of the Overpower Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the Overpower Delta-Temperature trip setpoint is reduced by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Sequoyah Nuclear Plant Unit 2 Proposed Technical Specification Bases Section 2.1.1 (Mark-Up)

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE Thi restrictions of this Safety Limit prevent overheating of the fuel 4,*d-pe', ""'cladding

'"'-" ' perforatio would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation

+nffl.

+,

r V1"'°'°u aU--INSERTT2 I 1corresponding significant 1

Operation above the upper boundary of the nucleate boiling regime could result in excessive eladdifcg temperatures because of the onset of departure from nucleate boiling (DNB) and the rceuIta IS shap reduction in heat transfer coefficient.kCNB is not a directly measurable parameter during operationIST and; therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the design DNBR limit.

,,4, meatri~g t.,,n* w,*,-,

,,"u

,a*tcin plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percept confidence level that the minimum DNBR for the limiting rod is greater lNSERT5 than or equal to the DNBR limit.J-"hZ uncrtARO.tRVA thA ahce plant para.eter-ar-ued W d-t the plant uncrtarinty-Rl.

Tic R uRtainty, e.mbin. d with the 8arr.lati. n DNR limit, e

.tablc*,he a dccig DNB

.a....heh muct be moet in plAnt safety anialycir, using Yaluerc of input paramoetreG witheut The ouvo f Figure 2.1 1 schcw;A the loci. of points of THERML POER, Roaotcr CAolan'At Syta przccur and avearag tempera~tura fory.whfich the m~inimum DNBR is no less than the safety ana!y'cic DNR limit, or the aveoao cnrthalpy at thoe "Voco oit ic equal to the enthalpy of

.atu.atod liqu.id.

The-Aura f Figuro 2.1 1 Wea based an an anthalpy rica hot channel faer

.1-TI,

,aluoc ha.ve beoo odurod to incudo a 4 1, total rod power uncitainty fartof), anda roforon-, -ocino with a peak of 1.55 f8r axial powE.

.hape.

An allowanc,,e ic n*,ludod far

_nRR *in"r i at raducod powor b-a-cod on. thc oA! reccin:,,,

THERM4AL POWAER, whiere-P

'n A-1r TD TIrS.

4AIMA 1QULVF.

4 N&e-BA4Feel Nefitinal Va~e!ý 462-Westinghzuac Fuie SEQUOYAH - UNIT 2 B 2-1 Affpil 24, 4997 Amendment No. 21,104,130, 146, 214

SAFETY LIMITS BASES Theco limfiting hoat flux conditione4 aro highor than these ealeculatod 4Fo tho rango of All conrol9 rode fully withdr-awn to tho mflaximum all-w-ablo coentrol rod incortion accuming the axial poWoribaacoI Wiithin thec limite Of the flto Idet

,)

fuotia*n

-f the O.z

,.A.P...turc Delta T trip. When th** axial power imbalanro. ic-not within the tolorainco, the axial powor imaAncfffzct on thA 4e Ootefmponture dlelta T tripe Will rzduee the cetpefints to previde proteetien concictent with e9ro cofzty lifietc.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Nominal Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Nominal Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Nominal Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Nominal Trip Setpoint and the Allowable Value is equal to or less than the rack allowance assumed for each trip in the safety analyses.

Technical specifications are required by 10 CFR 50.36 to contain Limiting Safety System Settings (LSSS) defined by the regulation as "... settings for automatic protective devices.., so chosen that automatic protective action will correct the abnormal situation before a Safety Limit (SL) is exceeded."

The analytic limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the analytic limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protective devices must be chosen to be more conservative than the analytic limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

The Nominal Trip Setpoint is a predetermined setting for a protective device chosen to ensure automatic actuation prior to the process variable reaching the analytic limit and thus ensuring that the SL would not be exceeded. As such, the Nominal Trip Setpoint accounts for uncertainties in setting the device (e.g., calibration), uncertainties in how the device might actually perform (e.g., repeatability),

changes in the point of action of the device over time (e.g., drift during surveillance intervals), and any other factors which may influence its actual performance (e.g., harsh accident environments). In this manner, the Nominal Trip Setpoint plays an important role in ensuring that SLs are not exceeded. As such, the Nominal Trip Setpoint meets the definition of an LSSS in accordance with Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," and could be used to meet the requirements that they be contained in the technical specifications.

rptmber 13, 2006 SEQUOYAH - UNIT 2 B 2-2 Amendment No. 130, 146, 299

INSERT 1 cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt), either of which could result in INSERT 2 Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs.

INSERT 3 from the outer surface of the cladding to the reactor coolant water. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

INSERT 4 To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertanties

INSERT 5 This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These lines are bounding for all fuel types. The curves in Figure 2.1-1 are based upon enthalpy rise hot channel factors that result in acceptable DNBR performance of each fuel type. Acceptable DNBR performance is assured by operation within the DNB-based Limiting Safety Limit System Settings (RPS trip limits). The plant trip setpoints are verified to be less than the limits defined by the safety limit lines in Figure 2.1-1 converted from power to delta-temperature and adjusted for uncertainty.

Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the acceptance criteria in the safety analysis.

The limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (Al), is within the limits of the f, (AI) function of the Overtemperature Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f1(AI) trip reset function, the Overtemperature Delta-Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for centerline fuel melt are higher than those calculated for the range of all control rods from the fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (Al), is within the limits of the f2(AI) function of the Overpower Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the Overpower Delta-Temperature trip setpoint is reduced by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 Modified Proposed Technical Specification Figure 2.1-1 (Final Typed)

This enclosure provides the modified proposed Technical Specification Figure 2.1-1 final typed pages for Units 1 and 2. These pages replace the corresponding pages originally provided in Attachment 3 to the enclosure of the letter from TVA to NRC, "Application to Modify Technical Specifications for Use of AREVA Advanced W1 7 HTP Fuel (TS-SQN-2011-07)," dated June 17, 2011 (Reference 1 of the cover letter for this submittal).

Sequoyah Nuclear Plant Unit 1 Proposed Technical Specification Figure 2.1-1 (Final Typed)

Figure 2.1-1 Reactor Core Safety Limit - Four Loops in Operation ago UNACCEPTABLE 660 OPERATICON 6 40 5 p

-~620-5 ACCEPTABLE

SIN, OPERATION 540 0.G 13.2 0.4 "0G 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER SEQUOYAH - UNIT 1 2-2 Amendment No. 19,

Sequoyah Nuclear Plant Unit 2 Proposed Technical Specification Figure 2.1-1 (Final Typed)

Figure 2.1-1 Reactor Core Safety Limit - Four Loops in Operation UNACCEPTABLE 6eu -

OPERATICe t

640220pi

-620.

-96pa 58 ACCEPTABLE_____

OPERA-TON 50.

0.0 0.2

.0.4 0.6G U.8 1.0 1.2.

FRACTION OF RATED THERMAL POWER SEQUOYAH - UNIT 2 2-2 Amendment No. 21,

ENCLOSURE 7 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS I AND 2 Modified Technical Specification Bases Section 2.1.1 (Final Typed)

This enclosure provides the modified Technical Specification Bases Section 2.1.1 final typed pages for Units 1 and 2. These pages replace the corresponding pages originally provided in Attachment 4 to the enclosure of the letter from TVA to NRC, "Application to Modify Technical Specifications for Use of AREVA Advanced W1 7 HTP Fuel (TS-SQN-2011-07)," dated June 17, 2011 (Reference 1 of the cover letter for this submittal).

Sequoyah Nuclear Plant Unit I Proposed Technical Specification Bases Section 2.1.1 (Final Typed)

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt), either of which could result in cladding perforation that would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs.

Operation above the upper boundary of the nucleate boiling regime could result in excessive temperature because of the onset of departure from nucleate boiling (DNB) and the corresponding significant reduction in heat transfer coefficient from the outer surface of the cladding to the reactor coolant water. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

DNB is not a directly measurable parameter during operation and; therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the design DNBR limit.

To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These lines are bounding for all fuel types. The curves in Figure 2.1-1 are based upon enthalpy rise hot channel factors that result in acceptable DNBR performance of each fuel type. Acceptable DNBR performance is assured by operation within the DNB-based Limiting Safety System Settings (RPS trip limits). The plant trip setpoints are verified to be less than the limits defined by the safety limit lines in Figure 2.1-1 converted from power to delta-temperature and adjusted for uncertainty.

SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19, 114, 138, 155, 223,

SAFETY LIMITS BASES Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the acceptance criteria in the safety analysis.

The limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (Al), is within the limits of the f, (AI) function of the Overtemperature Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the fj(AI) trip reset function, the Overtemperature Delta-Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for centerline fuel melt are higher than those calculated for the range of all control rods from fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (AI), is within the limits of the f2(AI) function of the Overpower Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the Overpower Delta-Temperature trip setpoint is reduced by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

SEQUOYAH - UNIT 1 B 2-1a Amendment No.

Sequoyah Nuclear Plant Unit 2 Proposed Technical Specification Bases Section 2.1.1 (Final Typed)

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt), either of which could result in cladding perforation that would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs.

Operation above the upper boundary of the nucleate boiling regime could result in excessive temperature because of the onset of departure from nucleate boiling (DNB) and the corresponding significant reduction in heat transfer coefficient from the outer surface of the cladding to the reactor coolant water. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

DNB is not a directly measurable parameter during operation and; therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the design DNBR limit.

To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These lines are bounding for all fuel types. The curves in Figure 2.1-1 are based upon enthalpy rise hot channel factors that result in acceptable DNBR performance of each fuel type. Acceptable DNBR performance is assured by operation within the DNB-based Limiting Safety System Settings (RPS trip limits). The plant trip setpoints are verified to be less than the limits defined by the safety limit lines in Figure 2.1-1 converted from power to delta-temperature and adjusted for uncertainty.

SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21, 104, 130, 146, 214,

SAFETY LIMITS BASES Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the acceptance criteria in the safety analysis.

The limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods from fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (AI), is within the limits of the f1(AI) function of the Overtemperature Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the fl(AI) trip reset function, the Overtemperature Delta-Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for centerline fuel melt are higher than those calculated for the range of all control rods from fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance, or Delta-I (AI), is within the limits of the f2(AI) function of the Overpower Delta-Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the Overpower Delta-Temperature trip setpoint is reduced by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

SEQUOYAH - UNIT 2 B 2-1 a Amendment No.