ML20071D107

Forwards Util Application for Amend to License DPR-40, Revising RCS Heatup & Cooldown Limits for Operation Through End of Cycle 8.Certificate of Svc & Justification Encl.W/O Stated Application

ML20071D107
Person / Time
Site:	Fort Calhoun
Issue date:	03/07/1983
From:	LEBOEUF, LAMB, LEIBY & MACRAE, OMAHA PUBLIC POWER DISTRICT
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8303090181
Download: ML20071D107 (15)

Text

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L EBOEU F, LAM B, LElBY Ek. M Ac RAE A PAlb?NERS M:P INCLUDING PROFESSIONAL CORPOR ATIONS 1333 NEw HAMPSHIRE AVENUE, N.W.

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March 7, 1983 ao'-'55-**'*

• .O. SOX 750 ell PEQUOT AVENUE RA.EIGH, NC 27602 SOUTMPORT, CT 06490 919-833-9783 203-269-8383 Mr. Harold Denton United States Nuclear Regulatory Commission 7920 Norfolk Avenue Phillips Building Bethesda, MD 20014 Re: Omaha Public Power District

. Fort Calhoun Station, Unit No. 1 Docket No. 50-285

Dear Mr. Denton:

As counsel for Omaha Public Power District, we hereby submit three (3) signed originals and nineteen (19) copies of a document entitled " Application for Amendment of Operating License", together with forty (40) copies of the proposed Technical Specifications. The application seeks to amend Sections 2.1.2 and figures 2-1A, 2-1B, 2-2A, and 2-2B of the Technical Specifications set forth in Appendix A to revise the reactor coolant system heatup and cooldown limits for operation through the end of fuel Cycle 8 The proposed Amendment is deemed to be a Class III Amendment within the meaning of Section 170.22 of the regulations of the U.S. Nuclear Regulatory Commission.

Accordingly, a check for the appropriate fee of $4,000 is also enclosed.

A Certificate of Service showing service of these documents on the persons listed therein is also enclosed.

PDR_ 05000285 ADOCK

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l We respectfully request -that approval 'for this amendment be granted.by no later than June 1, 1983.

l l Very truly yours, h , ,

Y LeBoeuf, Lamb, Leiby . MacRae Attorneys for Omaha Public Power District Enclosures l

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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the matter of )

) Docket No. 50-285 OMAHA PUBLIC POWER DISTRICT )

(Fort Calhoun Station, )

Unit No. 1) ) -

CERTIFICATE OF SERVICE I hereby certify that I have served a document entitled " Application for Amendment of Operating License",

J together with proposed changes to Technical Specifications by mailing a copy thereof first class, postage prepaid, to the following persons this 7th day of March, 1983.

Mr. Frank Gibson Director W. Dale Clark Library 215 South 15th Street Omaha, NE 68102 Mr. Emmet Rogert Chairman, Washington County Board of Supervisors 16th & Colfax Streets Blair, NE 68008

/ Ik Marilyn Tebor Shaw v LeBoeuf, Lamb, Leiby & MacRae Attorneys for Omaha Public Power District

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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

(a) The curve in Figure 2-3 shall be used to predict the increase in transition tremperature based on integrated fast neutron flux.

If measurements on the irradiation specimens indicate a devi-ation from this curve a new curve shall be constructed.

(b) The limit line on the figures shall be updated for a new inte-grated power period as follows: the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (E>l MeV). For this plant, based upon surveillance materials tests and the reduced vessel fluence rata provided by core load designs beginning with fuel Cycle 8, the predicted surface fluence at the reactor vessel belt-line weld mater" 41 for 40 years at 1500 MWt and an 80% load factor is 3.1x10:Sn/cm2, The predicted transition temperature shift to the end of the rcw period shall then be obtained from Figure 2-3.

(c) The limit lines in Figure 2-1A through 2-2B shall be moved parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The boltup temoerature limit line shall i remain at 82 F as it is set by the N3TT of the reactor vessel flange and not subject to fast neutron flux. The lowest service temperature shall remain at 162 F because components related to this temperature are also not subject to fast neutron flux. ,

(d) The minimum temperature at which the 100*F/hr cooldown rate curve may be used is defined by the LPSI pumps outlet pressure to provide for protection against low temperature overpressuri-zation per Technical Specification 2.3(3). The Technical Specification 2.3(3) shall be revised each time the curves of Figures 2-1A through 2-2B are revised.

Basis All components in the reactor coolant system are designed to withstand theeffectsofcygljcloadsduetoreactorcoolantsystemtemperatureand pressure changes.l l> These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.

During unit startup and shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown is based upon a rate of 100 F in any one hour period and for cyclic operation.

Amendment No. 22, 47, 64 2-4 Attachment A i

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' 2.0 LIMITING. CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

The maximum allowable reactor coolant system pressure at any temperature is based upon the stress limitations for brittle fracture considerations.

Thege)

IIIt2 of the limitations ASME Codeare derived byAppendix including using the G, rules contained Protection Against in Section Non-ductile Failure and the rules contained in 10 CFR 50, Appendix G, Fracture Toughness Requirements. This ASME Code assumes that a crack 10-11/16 inches long and 1-25/32 inches deep exists on the inner surface of the vessel. Furthermore, operating limits on pressure and temperature assure that the crack does not grow during heatups and cuoldowns.

The reactor vessel belt-line material consists of six plates. The nil-ductility transition temperature (TNDT) of each plate was established by drop weight tests. Charpy tests were then performed to determine at what temperature the plates exhibited 50 ft-lbs. absorbed energy and 35 mils l lateral expansion for the longitudinal direction. NRC technical position MTEB 5-2 was used to establish a reference temperature for transverse direction (RTNDT) Of -12 F-The mean RTNDT value for the Fort Calhoun submerged arc vessel weldments was determined to be -56*F with a standard deviation of 17*F. In accordance with the methods identified in "NRC Staff Evaluation of Pressurized Thermal Shock", SECY 82-465, Appendix E, a weld material reference temperature (RTNDT) was established at -22*F based on a mean value plus two standard deviations.

Similar testing was not perfonned on all remaining material in the reactor coolant system. However, sufficient im to meet appropriate design code requirementsand (3)apact conservative testing was performed RTNDT of 50 F has been established.

As a result of fast neutron irradiation in the region of the core, there will be an increase in the TNDT with operation. The techniques used to predict the integrated fast neutron (E11 MeV) fluxes of the reactor except that the vessel integrated are fast described in flux neutron Section 3.4.6 of (Ell MeV) is the 3.1 USAR,19n/qm2 x 10 tolerance, over the 40 year design life of the vessel.(5), including Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calibrated azimuthal neutron flux variation. The maximum integrated fast neutron (E>l MeV) exposure of the reactor vessel -

including tolerance is computed to be 3.1 x 1019n/cm2 for 40 years l Amendment No. 22, #7, 64 2-5

'2.0 LIMITING. CONDITIONS FOR OPERATION 2.1 Reactor Conlant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) operation at 1500 MWt and 80% load factor.(5) The predicted TNDT shift for an integrated fast neutron (E>l MeV) exposure of 3.1x1019n/cm2 is 321*F, the value obtained from the curve shown in Figure 2-3. The actual shift in TNDT will be re-established periodically during the plant operation by testing of reactor vessel material samples which are irradiated cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.5.3 and Figure 4.5-1 of the USAR. To compensate for any increase in the TNDT caused by irradiation, l limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown. Analysis of the first removed irradiated reactor vessel surveillance specimen, combined with a new core loading design for Cycle 8, indicates that the fluence will be 1.05 at xthe end ofI 8 1019n/cm 49 Effective no the Full Power inside surface Years vessel.

of the reactor (EFPY)5)at 1500 M(Wt This results in a total shift of the RTNDT of 260*F for the area of greatest sensitivity (weld metal) at-the 1/4t location as determined from Figure 2-3. Operation through fuel Cycle 10 will result in less than 8.49 l EFPY. .

The limit lines in Figures 2-1A through 2-2B are based on the following:

A. Heatup and Cooldown Curves - From Section III of the A3ME Code Appendix G-2215.

KIR = 2 KIM + KIT KIR = Allowance stress intensity factor at temperatures related to RTNDT ( ASME III Figure G-2110.1).

KIM = Stress intensity factor for membrane stress (Pressure). The 2 represents a safety factor of 2 on pressure.

KI r = Stress intensity factor radial thennal gradient.

The above equation is applied to the reactor vessel belt-line. For I plant heatup the thermal stress is opposite in sign from the pressure stress and consideration of a heatup rate would allow for l a higher pressure. For heatup it is therefore conservative to consider an isothermal heatup or KIT = 0.

l Amendment No. 22, #7, 64 2-6

r 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

For plant cooldown thermal and pressure stress are additive.

P KIM " MM _R t

MM = ASME III, Figure G-2214-1 P = Pressure, psia R = Vessel Radius - in, t = Vessel Wall . Thickness - in.

KIT = M TA Tw MT = ASME III, Figure G-2214-2 ATw = Highest Radial Temperature Gradient Through Wall at End of Cooldown KIT is therefore calculated at a maximum gradient and is considered a constant = A for cooldown and zero for heatup.

My R is also a constant = B.

t Therefore:

KIR = AP + B P = Kip - B A

KIR is then varied as a function of temperature from Figure G-2110-1 of ASME III and the allowable pressure calculated. Hydrostatic head (48 psi) and instrumentation errors (12 F and 32 psi) are considered when plotting the curves.

B. System Hydrostatic Test - The system hydrostatic test curve is developed in tne same manner as in A above with the exception that a

, safety factor of 1.5 is allowed by ASME III in lieu of 2.

l C. Lowest Service Temperature = 50 F + 100 F + 12 F = 162 F. As indicated previously, an RTNDT for all material with the exception of the reactor vessel belt-line was established at 50 F. ASME III, Art. NB-2332(b) requires a lowest service temperature of RTNDT +

100 F for piping, pumps and valves. Below this temperature a pressure of 20 percent of the system hydrostatic test pressure

(.20)(3125) 32 psi = 545 psia cannot be exceeded.

Amendment No. 22, #7, 64 2-7 ,

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2.0 LIMIT NG CONDITIONS FOR OPEr<ATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

D. Boltup Temperaturc = 10 F + 60*F + 12*F 82*F. At pressure below 545 psia, a minimum vessel temperature must be maintained to comply with the manufacturer's specifications for tensioning the vessel head. This temperature is based on previous NDTT methods. This temperature corresponds to the measured 10 F NDTT of the reactor vessel flange, which is not subject to radiation damage, plus 60*F data scatter in NDTT measurements, plus 12 F instrument error.

E. Reactor Critical Heatup and Cooldown Figures. During low power physics testing, the reactor may be made critical at reduced temperature and pressure. To provide for heatup and cooldown during testing, Appendix G requires that the RCS temperature be increased an additional 40 F beyond heatup and cooldown curves for the non-critical reactor. Also, Appendix G requires that the RCS temperature must be greater than the minimum temperature, 348*F, required for the 2310 psia hydrostatic testing to 110% of the 2100 psia RCS operating pressure, in accordance with Article IWB-5000 of the ASME Boiler and Pressure Vessel Code,Section XI.

F. Minimum Temperature for 100 F/hr Cooldown Rate = 153 F. This limit provides protection against low tem rature overpressurization during operation of the LPSI pumps. 1 This temperature corresponds to a pressure of 200 psia on the 100 F/hr curve, which is the LPSI pump dead head and minimum flow pressure. For temperatures of 153*F or less, a cooldown rate of 20 F/hr maximum will allow unrestricted operation of the LPSI pumps so that shutdown cooling may be utilized.

References (1) USAR, Section 4.2.2 l (2) ASME Boiler and Pressure Vessel Code,Section III (3) USfa, Section 4.2.4 (4) USAR, Section 3.4.6 (5) Omaha Public Power District, Fort Calhoun Station Unit No. 1, Evaluation of Irradiated Capsule W-225, Revision 1, August,1980 (6) Technical Specification 2.3(3)

(7) Article IWB-5000, ASME Boiler and Pressure Vessel Code,Section XI Amendment No. 22, 47, 64 2-7a

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DISCUSSION i This amendment application is required to allow for the safe operation of the Fort Calhoun Station reactor - and associated primary coolant system beyond the 6.1 Effective Full Power Years (EFPY) of operation for which the present Technical Specifications (TS) are valid. This application requests continued operation through 8 44 EFPY.

In detemining the limiting conditions for operation at this increased EFPY, (RTNDT

)the and impact of the associated initialshift RTNDT nil-ductility must betransition accountedreference for due totemperature the effect of neutron fluence on the reactor vessel belt-line region wel ds. For previous cycles, the District conservatively utilized 0*F as the initial RTNDT, as re-commended by Branch Technical Position MTEB 5-2. However, recent evaluations perfomed by the Commission demonstrated that this . initial RTNDT value ~ is over-ly conservative. Specifically, Appendix E of the report, "NRC Staff Evaluation of Pressurized Themal Shock: SECY 82-465", states " Estimates (i.e. initial RTNDT values) based on the 3 Charpy test results and MTEB 5-P. are not very satisfactory, because they' are overconservative for some cases." The recom-mended NRC methodology in this report for computing RTNDT, which utilizes a mean RTNDT value obtained from generic weld data and ensures conservatism by adding a two sigma uncertainty value, results in an initial RTNDT for a Com-bustion Engineering reactor vessel such as Fort Calhoun of -22 F. Therefore, this value will be utilized as the Fort Calhoun Station initial RTNDT value in this and future cycle analyses.

Additionally, commencing with start-up from the present (1983) refueling outage, the District will utilize a new core loading pattern which will result in a significant decrease in fluence to the critical belt-line wel ds. Further discussion of this core loading pattern is provided in the District's letter dated January 27, 1983. The combination of the lower initial RTNDT value and implementation of the reduced cycle fluence exteids the present Technical Specification heatup and cooldown limits applicability through 8.49 EFPY, with minor changes as indicated in the revised specification. This amendment will provide pressure - temperature operating limits through Cycle 10.

Regulatory Guide 1.99, Revision 1, provided the methodology used to determine' the nil-ductili ty transition reference temperature (RTNDT) shift reflected in the proposed heatup and cooldown limit curves. The fluence value for thepredicted li fe reactor vessel belt-line fluence weljgn/cm . material of 4.4x10 2 wasvalue This determined using the end-of-was calculated and approved for Cycle 6 operation using the Fort Calhoun Station first surveil-

lance capsule test data as detailed in the Combustion Engineeri ng report, i " Evaluation of the Irradiated Capsule W-225", Revision 1, dated August 1980.

In addition, the rate of fluence for Cycle 8 and future cycles was conserva-tively assumed to be 37% less than that of previous cycles due to the reduced l fluence core loading pattern to be implemented for Cycle 8. Thus, the heatup and cooldown pressure-temperature limit curves used for Cycle 7 were updated to 8.49 EFPY without a temperature shift and yet still ensure adequate fracture j toughness is maintained through all conditions of normal operation, including

! anticipated operation transients and system hydrostatic tests.

Attachnent B

The current Technical Specification to 6.1 EFPY, will provide operati ng limits for a period of 66 days, limited of full power operation (2,367,000 MW-HRS) after initial criticality is achieved for fuel Cycle 8. Therefore, Commission approval of the proposed Technical Specifications by no later than -

May 30, 1983 is requested.

1 i

4 Attachment B

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JUSTIFICATION FOR FEE CLASSIFICATION The proposed amendment is deemed to be Class III, within the

-meaning of 10 CFR 170.22, in that it involves a single safety-Concern.

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ATTAcifMENT C r

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