ML031610859

TS Change No.03-08, Reactor Coolant System Heatup & Cooldown Curves

ML031610859
Person / Time
Site:	Sequoyah
Issue date:	06/05/2003
From:	Salas P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TVA-SQN-TS-03-08
Download: ML031610859 (26)

Text

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Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 June 5, 2003 TVA-SQN-TS-03-08 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

In the Matter of ) Docket No. 50-328 Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT (SQN) - UNIT 2 - TECHNICAL SPECIFICATION (TS) CHANGE NO. 03-08, "REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN CURVES"

Reference:

TVA letter to NRC dated September 6, 2002, "Sequcyah Nuclear Plant (SQN) - Units and 2 Technical Specification (TS) Change No. 00-14,

'Pressure Temperature Limits Report (PTLR) and Request For Exemption From The Requirements Of 10 CE'R 50, Appendix G'"

In accordance with 10 CFR 50.90, TVA is submitting a request for an amendment to SQN's License DPR-79 to change the TSs for Unit 2. The proposed amendment revises TS 3/4.4.9.1, "Pressure/Temperature Limits, Reactor Coolant System." The revision replaces the pressure-temperature (P-T) limits that are currently analyzed for 14.5 Effective Full Power Years (EFPYs) with new limits analyzed for 32 EFPY. In addition, the amendment includes corresponding changes to the TS figure associated with Low Temperature Over Pressure Protection (LTOP) and the TS Bases.

TVA s reference letter requests a TS amendment for SQN (both units) to incorporate a PTLR. The PTLR contains updated P-T limits applicable for 32 EFPYs. The limits are based on NRC approved methodology with two exceptions. One of the P.nted on cyd pape

U.S. Nuclear Regulatory Commission Page 2 June 5, 2003 exceptions eliminates the minimum temperature requirement evaluation for the reactor pressure vessel (RPV) closure flange region. Following recent discussions with NRC staff, TVA understands that review and approval of TVA' s requested change that eliminates evaluation of the flange region may not be complete before the applicability of SQN's present Unit 2 limits expire. The SQN Unit 2 P-T limits are currently applicable for 14.5 EFPYs which is projected to expire in early August 2003. Accordingly, TVA, as an interim measure, is submitting new Unit 2 P-T limits that include the minimum temperature requirement evaluation for the RPV flange closure region.

NRC approval of the new Unit 2 P-T limits will allow operation of Unit 2 beyond the August 2003 expiration date.

TVA understands that NRC will continue their ongoing review of TVA' s reference PTLR request. The next milestone associated with P-T limits is SQN Unit 1 with current limits applicable for 16 EFPY. The current Unit 1 limits are projected to expire in 2005.

It may be noted that the topical report that supports the enclosed amendment (Topical Report, WCAP-15321, Revision 1) utilizes alternatives to the requirements of 10 CFR 50, Appendix G as referenced by 10 CFR 50.60(a). In TVA's reference letter, TVA utilized exemptions from Appendix G.

The first exemption utilized American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI Code Case N-640, "Alternative Requirement Fracture Toughness for Development of P-T limit Curves for ASME Section XI, Division 1," in lieu of 10 CFR 50, Appendix G, paragraph IV.A.2.b. The second exemption utilized WCAP-15315, "Reactor

.Vessel Closure Head/Vessel Flange Requirements Evaluation for Operating PWR and BWR Plants" in lieu of 10 CFR 50, Appendix G, footnote 2 to Table 1. The first exemption associated with Code Case N-640 is required for the enclosed amendment. Accordingly, TVA requests that NRC apply the N-640 Code Case exemption (previously submitted by TVA's reference letter) to the enclosed amendment.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change is exempt from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). The SQN Plant Operations Review Committee and the SQN Nuclear Safety Review

U.S. Nuclear Regulatory Commission Page 3 June 5, 2003 Board have reviewed this proposed change and determined that operation of SQN Unit 2, in accordance with the proposed change, will not endanger the health and safety of the public. Additionally, in accordance with 10 CFR 50.91(b)(1),

TVA is sending a copy of this letter to the Tennessee State Department of Public Health.

TVA requests approval of this TS change by August 1, 2003, and that implementation of the revised TS be within 15 days of NRC approval.

There are no regulatory commitments contained in this submittal. This letter is being sent in accordance with NRC RIS 2001-05. If you have any questions about this change, please telephone me at (423) 843-7170 or J. D. Smith at (423) 843-6672.

I declare under penalty of perjury that the oregoing is true and correct. Executed on this 53 day of lJIIJ, L..

s

• fsing and Industry Affairs Manager

Enclosures:

1. TVA Evaluation of the Proposed Changes

2. Proposed Technical Specifications Changes (mark-up)

3. Changes to Technical Specifications Bases Pages cc: See page 4

U.S. Nuclear Regulatory Commission Page 4 June 5, 2003 Enclosures cc (Enclosures):

Framatome ANP, Inc.

P. 0. Box 10935 Lynchburg, Virginia 24506-0935 ATTN: Mr. Frank Masseth Mr. Michael L. Marshall, Jr., Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop O-8G9A One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. Lawrence E. Nanney, Director Division of Radiological Health Third Floor L&C Annex 401 Church Street Nashville, Tennessee 37243-1532

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 2

1.0 DESCRIPTION

OF THE PROPOSED CHANGE This letter is a request to amend Operating License DPR-79 for SQN Unit 2. The proposed change would revise the Operating License for SQN Unit 2 Technical Specification (TS) 3/4.4.9.1, "Pressure/Temperature Limits, Reactor Coolant System." The proposed amendment deletes two figures (Figures 3.4-2 and 3.4-3) that are referenced in Limiting Condition for Operation (LCO) 3.4.9.1. and provide pressure-temperature (P-T) limits for reactor coolant system (RCS) heatup and cooldown. The figures are currently applicable for 14.5 Effective Full Power Years (EFPYs) and are being replaced with updated figures analyzed for 32 EFPYs.

In addition, the proposed change includes a revision to TS 3/4.4.12 and the associated Figure 3.4-4 that provides low temperature overpressure protection (LTOP) setpoints. The proposed revision extends the applicability of the setpoints to 32 EFPYs and revises the Appendix G limit to reflect the updated 32 EFPY cooldown limit.

The proposed changes described above include appropriate revisions to the associated TS Bases sections.

2.0. PROPOSED CHANGE In summary, TVA's proposed change updates the SQN Unit 2 RCS heatup and cooldown curves (P-T limits) and extends the life of the Unit 2 limits from 14.5 EFPYs to 32 EFPYs.

In addition, the amendment includes a change to the TS figure associated with the Low Temperature Over Pressure Protection (LTOP) system and provides corresponding changes to the TS Bases.

3.0. BACKGROUND By letter dated September 6, 2002, TVA requested a TS amendment to update SQN's current P-T limits and incorporate a Pressure Temperature Limits Report (PTLR) within the SQN TSs for both units. The updated P-T limits provided in the SQN PTLR contain heatup and cooldown limits applicable for 32 EFPYs. The limits are based on the latest analytical methodology with some exceptions.

One exception included the elimination of the minimum El-i

temperature requirement evaluation for the reactor pressure vessel (RPV) closure flange region. Following recent discussions with NRC staff, TVA understands that review of the change may not be complete before applicability of the present Unit 2 P-T limits expires.

The Unit 2 P-T limits are currently applicable for 14.5 EFPYs and are projected to expire in early August 2003. Accordingly, TVA, as an interim measure, is submitting new Unit 2 P-T limits that account for the minimum temperature requirements for the RPV closure flange region. NRC approval of the new Unit 2 P-T limits will allow operation of Unit 2 beyond the projected August 2003 expiration date.

4.0 TECHNICAL ANALYSIS

Title 10 of the Code of Federal Regulations, Part 50, Appendix G, requires the establishment of P-T limits for specific material fracture toughness requirements of the reactor coolant pressure boundary materials. The 10 CFR 50, Appendix G establishes an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME)

Code,Section XI, Appendix G.

The components of the RCS at SQN are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.

The P-T limits, as established by the requirements of 10 CFR 50, Appendix G, are periodically reanalyzed and revised as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. The P-T limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. Periodic adjustment of the P-T limits is needed to account for these time-dependent parameters.

Adjustment of the limits is acceptable if performed in accordance with methodology approved by the NRC. TVA's proposed change utilizes analytical methods approved by NRC as prescribed in Westinghouse Topical Reports WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996." In addition, a plant specific analyses for SQN's Unit 2 P-T limits is provided in WCAP-15321, Revision 1, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation, April 2001."

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The updated limits are within applicable plant design assumptions as discussed in Section 5.2.4.3 of the SQN Final Safety Analysis Report (FSAR). These safety analyses demonstrate that SQN's Unit 2 reactor vessel is adequately protected against brittle fracture when operated within these limits.

Each P-T limit curve defines an acceptable region for normal plant operation. The curve is used for operational guidance during heatup and cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

TVA' s proposed TS change updates current RCS P-T limit curves for Unit 2 and extends the applicability of these limits from 14.5 EFPYs to 32 EFPYs. The proposed change is based on analyses documented in WCAP-15321, Revision 1.

The revised P-T limits include the minimum temperature requirements for the RPV closure flange region as prescribed in 10 CFR 50, Appendix G. The 10 CFR 50, Appendix G requirements address the metal temperature of the reactor vessel in the closure head flange and vessel flange regions. The requirement limits the normal operating metal temperature in this region to temperatures which exceed the material unirradiated nil-ductility transition reference temperature (RTNDT) by at least 120 degrees Fahrenheit (F) when the system pressure exceeds 20 percent of the pre-service hydrostatic test pressure.

For SQN Unit 2, the limiting unirradiated RTNDT of -13°F occurs in the reactor vessel closure head flange. In addition, 20 percent of the pre-service hydrostatic test pressure (3106 pounds per square inch [psi]) is 621 psig.

Based on this data, the minimum allowable temperature of the flange region is 107°F at pressures greater than 621 psig. When instrumentation margins of 10°F and 60 psig are considered (reference WCAP-15321, Revision 1), the allowable temperature in this region becomes 117°F at pressures greater than 561 psig. Accordingly, this limit has been included in the revised Unit 2 P-T limit curves.

The LTOP system setpoints and the RCS vent size are evaluated for compliance each time the P-T limit curves are revised based on the results of the vessel material surveillance. SQN's current power-operated relief valve setpoints for LTOP were evaluated with the updated P-T limits to ensure these setpoints provide sufficient margins against overpressure transients. The evaluation show that new setpoints for SQN's Unit 2 LTOP system are not required for the revised 32 EFPY P-T limits. This is E1-3

based on the current LTOP setpoints being more restrictive than the setpoints developed for the 32 EFPY P-T limits.

In addition, the analyses show that SQN's current vent size of 3.0 square inches is capable of mitigating a LTOP transient. The capacity of a vent this size is greater than the flow of the limiting transient for the LTOP configuration. Accordingly, SQN's current LTOP system setpoints and SQN's current RCS vent size are capable of mitigating LTOP transients.

5.0 REGULATORY SAFETY ANALYSIS In accordance with 10 CFR 50.36, TVA proposes to amend the Sequoyah Nuclear Plant (SQN) Unit 2 Technical Specification (TS) 3/4.4.9.1, "Pressure/Temperature Limits, Reactor Coolant System." The proposed amendment deletes the current figures referenced in Limiting Condition for Operation 3.4.9.1 (Figures 3.4-2 and 3.4-3) and replaces the figures with updated pressure-temperature (P-T) limit curves for reactor coolant system (RCS) heatup and cooldown. The applicability of the updated limits is extended from the current 14.5 Effective Full Power Years (EFPYs) to 32 EFPYs.

5.1. No Significant Hazards Consideration Determination TVA has concluded that operation of SQN Unit 2 in accordance with the proposed change to the TSs, does not involve a significant hazards consideration.

TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed revision does not affect plant equipment, test methods or operating practices.

The modification to SQN TSs is consistent 10 CFR 50, Appendix G in conjunction with alternative methods provided in American Society of Mechanical Engineers (ASME) Code Case N-640, "Alternative Requirement Fracture Toughness for Development of P-T Limit Curves for ASME Section XI, Division 1."

The proposed change continues to provide controls for safe operation within the required limits.

The proposed changes do not contribute to events or assumptions associated with postulated design basis accidents (DBA). The proposed revisions E1-4

continue to maintain the required safety functions. Accordingly, the probability of an accident or the consequences of an accident previously evaluated is not increased.

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

Response: No.

The proposed revision is not the result of changes to plant equipment, test methods, or operating practices. The proposed revision to the SQN Unit 2 P-T limits continues to ensure that conservative fracture toughness margins are maintained to protect against reactor pressure vessel failure.

In addition, SQN s current setpoints for low-temperature overpressure protection were evaluated and are bounding for the proposed 32 EFPY P-T limits. The updated P-T limits are based on NRC approved methodology in conjunction with alternative methods provided in American Society of Mechanical Engineers (ASME) Code Case N-640, "Alternative Requirement Fracture Toughness for Development of P-T Limit Curves for ASME Section XI, Division 1."

The reactor vessel P-T limits are operational limits and are not considered to be contributors to the generation of postulated accidents. The safety functions of the associated systems remain unchanged and do not affect the assumptions of DBAs. The operational limits continue to be governed within the TSs. Accordingly, the proposed change does not create the possibility of a new or different kind of accident.

3.Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

TVA's proposed TS amendment provides revised reactor pressure vessel P-T limits that are within the design capabilities of the pressure control systems for protection of the RCS. The limits are based on conservative design margins that ensure that plant operation is within the design capacity of the reactor vessel materials.

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Accordingly, the function of the RCS to provide a fission product barrier is not compromised.

TVA's proposed change to revise P-T limits does not result in a change to system design features.

The proposed change does not affect plant conditions that result in precursors to accidents or cause degradation of accident mitigation systems. The plant system safety functions are not altered by the proposed change.

The proposed changes allow plant operation with different P-T limits while continuing to retain conservative margins for assuring integrity of the reactor vessel and the RCS. Consequently, the proposed TS revisions do not significantly reduce the margin of safety.

5.2. Applicable Regulatory Requirements/Criteria The pressure-temperature (P-T) limits are established by requirements defined in 10 CFR 50, Appendix G, entitled "Fracture Toughness Requirements." These limits are periodically reanalyzed and revised as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation.

The P-T limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. Adjustment of the limits is acceptable if performed in accordance with methodology approved by the NRC. The NRC approved methodology for Westinghouse Electric Company's pressurized water reactor plants is prescribed in Westinghouse Topical Report WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996."

Alternatives to the 10 CFR 50, Appendix G requirements for development of P-T limits are provided by American Society of Mechanical Engineers (ASME) Code Cases (such as Code Case N-640) and must be approved for use. TVA' s proposed change to update SQN's Unit 2 P-T limits utilizes ASME Code Case N-640 as documented in WCAP-15321, Revision 1, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation, April 2001." This Unit 2 specific analysis updates these E1-6

limits within applicable plant design assumptions as stated in the SQN Final Safety Analysis Report (FSAR). These safety analyses demonstrate that SQN' s Unit 2 reactor vessel is adequately protected against brittle fracture when operated within these limits.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

6.0. ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

7.0. REFERENCES

1. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements."

2. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996."

3. WCAP-15321, Revision 1, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation, April 2001."

4. TVA letter to NRC dated September 6, 2002, "Sequoyah Nuclear Plant (SQN) - Units 1 and 2 Technical Specification (TS) Change No. 00-14, 'Pressure Temperature Limits Report (PTLR) and Request for Exemption From the Requirements of 10 CFR 50, Appendix G.'"

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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 2 Proposed Technical Specification Changes (mark-up)

I. AFFECTED PAGE LIST Unit 2 3/4 4-29 3/4 4-30 3/4 4-35 B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 B 3/4 4-14 II. MARKED PAGES See attached.

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IPPLICABLE REPLACE FIGURE 3.4-2 WITH ATTACHMENT FOR HEATUP RATES UP r FOR THE SERVICE PERIOD UP TO 444. EFPY. MARGINS OF 60 PSIG I

AND 10°F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERRORS.

a. = - CRITICALITY LIMIT FOR 60-F/HR HEATUP w _ fl- (SEE T.S. BASES)

U, rACCEPTABLE -

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0.

0 w

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ATERIAl- OPERTY BASIS- SON UNIT 2

)NTROLLIN ATERIAL: WLDS

)PPER CONT 0.13 WT%

rCKEL CONTEN 0. II WT%

4ITIAL RTNOT: -4 F T TAFTERI 1_/4T 142F A I 3/4T. 14F INICATED TEMPERATURE F)

FIGURE 3.4-2 SEQUOYAH UNIT 2 R OOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO EFPY I April 30, 2002 SEQUOYAH - UNIT 2 3/4 4-29 Amendment No. 138,148,264 E2-2

REPLACE FIGURE 3.4-3 CURVES APPLICABLE FOR COOLDN RATES UP TO 100°F/HR WITH ATTACHMENT 2 FOR THE SERVICE PERIOD UP TO 444 EFPY. MARGINS OF 60 PSIG I AND 10°F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERRORS.

N cn a-w I-0:

w C

z INDICATED TEMPERATURE (F)

FIGURE 3.4-3 SEQUOYAH UNIT 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 444 EFPY I April 30. 2002 SEQUOYAH - UNIT 2 3/4 4-30 Amendment No. 138, 148,264 E2-3

ATTACHMENT 1 2500 prim Ve on:5. R n5 4 Leak TestLit 220. _............... E It 200Q* lOperatlor l . f f I 00 1250 ........

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0 50 100 150 2 25 300 no 400O45t 15 Q0 TEMPERATUREd 'F PORV NOMINAL LIFT SETTINGS - APPLICABLE UP TO UrE4t*;@' I FIGURE 3.4-4 Apnl3D,2002 I SEQUOYAH - UNIT 2 3/4 4-35 Amendment No. 147, 264 E2-6

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIMTY (Continued)

Reducing T. to less than 500°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limit e consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section UI,Appendix G.

1) The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the first full-power service period.

a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2) These limit lines shall be calculated periodically using methods provided below.

3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 700F.

4) The pressurizer heatup and cooldown rates shall not exceed 100°F/ hr and 2000 /hr respectively.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 5600 F.

June 25, 1985 SEQUOYAH - UNIT 2 B 3/4 4-6 Amendment No. 32 E2-7

REACTOR COOLANT SYSTEM BASES PRESSURETEMPERATURE LIMITS (Confinued)

5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section Xl.

10 CFR 50, Appendix G, addresses metal temperature of the closure head flange and vessel regions. Appendix G states that the minimum metal temperature of the closure flange region should be at least 120 degrees Fahrenheit (F) higher than the limiting RTNDT for this region when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (561 pounds per square inch gauge (psig) for Wesffnghouse Electric Corporaton plants). For SQN, Unit 2, the minimum temperature of the closure flange and vessel flange regions is 117 degrees F since the limitfng initial RTNDT for the closure head flange is -13 degrees F (see Table B 3/4.4-1). These numbers (56 es F) include instrumentaion error of 10 dearees F and 60 i il N URit 2 hcat up and Goeldown cures shown n FiUres 3.4 2 and 3.1 3 aro 00t v w .. ,-.

ir. - - Ir_._ ,waTrzT Thc fracturo toughnec properics of the frritic matcrials in the reactor essecl arc determined in DELETE ccGdn it RC Stanidard Reiew Plan, and ASTM E85 82, aRd in arcerdance with additenal ator vesol eqUirents. These prFpeFresio ar then ealuated in aecordance ith-AppeRdix G to 10 CFR 50 and Appendik G ef the 1986 ASME BoileF and PFecSUFe Vessel Code,-

SeGoR Il, Di'.io I ad the calculateRn Faetheds desGFIbed in WCAP 70924 A, "Basis forHcFu and Ceoldown Umit Gupves, April 1075."

32 ant rimirli rnniiown imitrurve-am-Iculated using the most limifing value of the nil-ducility reference temperature, RTNDT at the end f effective full power years of service life. The EFPY service life period ischosen such that the limting RTNDT at the I/4T location in the core region is greater than the RTNDT of the limitng unirradiated mateial. The selecton of such a limiting RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The rcastorevccel materals have been tested to determninc their iPal F T .; the rcults of thec REPLACE tcts are chown in Table B 314.4 1. Reactor operation and rcsultant fast neutron (E greater than WITH INSERT A teaperatue, based upon the fluene mateal in qui, Rfhe ha _bn pr dcted -SIR Regulato' Guide 1.00, R svi-ion 2and a peak curfaco fluence of 0.864 x 104-RM for46 effoce;ve full power years (Referenee WcAP 12071, "Heatup and Coeldewn 1Umit Cures for Normal OpeFatn," JuR 1001. The heatup and Goldewn limit eursc of Figurec 3.1 2 an 3.1 3 s~~Wl- di-te-adst R-t-;fA -- i-_94 r aVtteeR f 4 EFPY, wel as9__

adju -

for possibloorrei h preccUFe and temperatUre 6eRsin irumAents. The heatup and-seeldowR limts i WCAP 12071 were based on a core thermal power of 3111 MM. The GuruFec have been evaluated in WCAP 16725 to be effestie for operation through the end of 14.6 EF=PY for the UpratedorFe theFrAal per F 3455 MM.

April 30,2002 SEQUOYAH - UNIT 2 B 3/4 4-7 Amendment No. 148, 264 E2-8

INSERT A The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence of the material in question, has been predicted using Regulatory Guide 1.99, Revision 2 and a peak surface fluence of 1.82 x 10 19 n/cm2 for 32 effective full power years (reference WCAP-1 5321, Revision 1, uSequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," April 2001).

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNDT at the end of 32 EFPYs, as well as adjustments for possible errors in the pressure and temperature sensing instruments. The heatup and cooldown limits in WCAP-15321, Revision 1 were based on a core thermal power of 3411 MWt. The curves have been evaluated in WCAP-15725 to be still effective for operation through the end of 32 EFPYs for the uprated core thermal power of 3455 MWt.

E2-9

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

Values of ART deteFRmiRed OR thi FaaRnr may b ud ti the FeSult foma the Faatcrial cuveillanoc program, ev'aluated acording to ASTM El 85, arc available. The fAirt capsule wi-Ll IDLT Femoved at the eRd of the fiFrAetFer cycle. Sussessive capulo wil e FeMeyed i arcGeFdanGe wth the rqUirFeRRtS f ASTI El 2 and 10 GFR 50,C5 AppeRdi H. The heatup aRd oldw-curvcs and the low temperature overprcssurc protcction setpointr must be recalculated when thc Adotcrmi-'nedfrm th srveillanGe apsule exseeds the alculated ART forFthe cquvalcnt capsulc radiation cxposuro.

Allowable pressure-temperature relationships for various heatu Pnd nnlown ratpC re calculated using methods derived frorn Appendix G in Section U-of the ASME Boiler and Pressure Vessel Code as required by Appendix to 10 CFR Part 50 nd thoso methods ar discussed in detail in WCAP 7921 A I the Summer 1996 Addenda of The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 312T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimenstions of this postulated crack, referred to in Appendix G of ASME If as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. eretore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RTNDT, is used and this includes the radiation induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

March 31, 1992 SEQUOYAH - UNIT 2 B 3/4 4-8 Amendment Nos. 147,148 E2-10

REACTOR COOLANT SYSTEM BASES PRESSURETEMPERATURE LIMITS (Continued) 7SERTB ME approach for calculatng the allowable limit curves for various heatup and cooldown rates specifies that the- t s intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cani erter than the reference stress intensity factor, KIR, for the metal temperature at that time. KIR is obtained fromi nce fracture toughness curve, defined in Appendix G to the ASME Code. The KIR curve is given by KIR = 26.78 + 1.223 exp 0.0145(T-RTNDT + 160)1 (1) where KR is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C KIM + KftSJKE (2)

IC Where, KM is the stress intensity factor caused by membrane (pressure) stress.

KIt is the stress intensity factor caused by the thermal gradients.

KKlis provided by the code as a function of temperature relative L-.J to the RTNDT of the material.

C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

SEQUOYAH - UNIT 2 B 3/4 4-11 E2-11

INSERT B The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K,, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kc, for the metal temperature at that time. Kc is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Altemative Reference Fracture Toughness for Development of PT Limit Curves for Section Xl" [1 22 of the ASME Appendix G to Section Xl. The K,c curve is given by the following equation:

Kc = 32.2 + 20.734 exp [0.02(T-RTNDT)1 E2-12

REACTOR COOLANT SYSTEM BASES PRESSURETEMPERATURE LIMITS (Confinued)

At any time during the heautp or cooldown transient, etermined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve.

The thermal stresses resulfing from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed. From Equaton (2) the pressure stress intensity factors are obtained and from these the allowable pressures are calculated.

COOLDOWN For the calculaton of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relabons composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limifing pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 114T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the delta T developed during cooldown results in a higher value oLKv; at the 114T location for finite cooldown rates than for steady-state operation. Furthermore, if cond"rs exist such that the increase in i4Eeceeds KIT. the calculated allowable pressure during cooldow be greater than the steady-state value SEQUOYAH - UNIT 2 B 34 4-12 E2-13

REACTOR COOLANT SYSTEM BASES PRESSURETEMPERATURE LIMITS (Continued)

The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

HEATUP Three separate calculations are required to d termine the limi s for finite heatup rates. As is done in the cooldown analysis, allowable pressure- mperature relationship eveloped for steady-state condifions as well as finite heatup rate cond' ns assuming the presence of a defect at the inside of the vessel wall. The thermal gradients during eatup produce compressive stresses a side of the wall that alleviate the tensile stresses produced interal pressure. The metal temperature atrack tip lags the coolant temperature; therefore, the 4w for the 1/4T crack during heatup is lower than the Kp for the 114T crack during steady-state conditions at the same coolant temperature. During heatup, especially the end of the transient, conditions may exist such that the effects of compressive thermal stresses and diffW6 K~e for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state condifions no longer represents a lower bound of all similar in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second porton of the heatup analysis concems the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis.

SEQUOYAH - UNIT 2 B 3/4 4-13 E2-14

REACTOR COOLANT SYSTEM BASES PRESSURETEMPERATURE LIMITS (Continued)

Following the generaton of pressure-temperature curves for both the steady-state and finite heatup rate situatons, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

The leak test limit curve shown on Figure 3.4-2 represents the minimum temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of Branch Technical Position MTEB 5-2 and 10 CFR 50, Appendix G.

The criticality limit curve shown in Figure 3.4-2 specifies pressure-temperature limits for core operation to provide additional margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) require the reactor vessel to be at a temperature equal to or higher than the minimum temperature required for the in-service hydrostatic test, and at least 40 degrees F higher than the minimum pressure-temperature curve for heatup and cooldown. The maximum temperature for the in-service hydrostatic test for the SQN Unit 2 reactor vessel is 24egrees F. A vertical line at degrees F on the pressure-temperature curve, intersectng a curve 40 degree higher than the pressuretemperature limit curve, constitutes the limit for core operation for the reactor vesse l\@

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in temperature ranges above those for which there is reason for concem of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

3/4.4.10 DELETED August22, 1995 SEQUOYAH - UNIT 2 B 3/4 4-14 Amendment No. 148, 198 E2-15