ML20150B246

Amend 114 to License DPR-40 Updating RCS pressure-temp Limits for Heatup & Cooldown to Reflect New Fluence Prediction Developed & More Limiting Chemistry Factor Associated W/Weld 3-410

ML20150B246
Person / Time
Site:	Fort Calhoun
Issue date:	06/28/1988
From:	Calvo J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20150B228	List: ML20150B226 ML20150B246 ML20150B268
References
NUDOCS 8807110414
Download: ML20150B246 (9)

Text

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UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION o

B ,i i AsHINGTON. D. C. 20666

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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 114 License No. DPR-40

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by the Omaha Public Power District (the licensee) dated February 19, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as '

amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; I D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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8807110414 DR 880628 ADOCK 05000235 PDC

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2. Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8. of Facility Operating License No.

DPR-40 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A. as revised through Amendment No.114, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION gli , Dbvw Jose A. Calvo, Director Project Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 28, 1988 /

ATTACHMENT TO LICENSE AMENOMENT NO.114 FACILITY OPERATING LICENSE NO. DPR-40 ,

DOCKET NO. 50-285 ReviseAppendix"A"TechnicalSpebificationsasindicatedbelow. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insart Pages 2-4 2-4 2-5 . 2-5 2-6 2-6 Figure 2-1A Figure 2-1A Figure 2-1B Figure 2-1B Figure 2-3 Figure 2-3 I

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. . 1 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 temperature based on integrated fast neutron flux. If measurements on the irradiation specimens indicate a deviation from this curve, a new curve shall be constructed.

(b) The limit line on the figures shall be updated for a new integrated power period as follows: the total intograted reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (E21 MeV). For this plant, based upon surveillance materials tests, weld chemical composition data, and the effect of a reouced vessel fluence rate provided by core load designs beginning with fuel Cycle 8, the predicted surface fluence at the initial reactor vessel beltline weld material for 40 years at 1500 MWt and an 80%

load factor is 2.55x1019 n/cm2 The flux reduction applied to the fluence calculations was based on Cycle 1-7 average and Cycle 8 average azimuthal flux distribution plots I generated used DOT 4.3. The predicted transiti.on temperature -

l shift to the end of the new period shall then be obtained

! from Figure 2-3.

(c) The limi+ 'ines in Figures 2-1A and 2-1B shall be moved

! paralle ,o the temperature axis (horizontal) in the l direct.,a of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The boilup temperature limit line shall remain at 82 F as it is set by the NDTT of the reactor vessel flange and not subject to fast! neutron I flux. The lowest service temperature shall remain at 182 F because components related to this temperature are also not subject to fast neutron flux.

l (d) The Technical Specification 2.3(3) shall be revised each l time the curves of Figures 2-1A and 2-1B are revised.

I Basis All components in the reactor coolant system are designed to withstand theeffectsofcyclicgadsduetoreactorcoolantsystemtemperature These cyclic loads are introduced by normal and pressure changes 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 cool-down is based upon a rate of 100*F in any one hour period and for cyclic operation.

2-4 Amendment No. 22,47,64,74,77,100, 114 1

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2. 0 LIMITING CONDITIONS FOR OPERATION '

, 2.1- Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) ,

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

Theg)limitationsarederivedbyusingtherulescontainedinSection of the ASME Code including Appendix G, Protection Against Non-III 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 cooldowns.

The reactor vessel beltline material consists of six plates. The nil-ductility transition temperature (T N of each plate was established by drop weight tests. Charpytestsg)ethenperformedtodetermineat r -

what temperature the plates exhibited 50 ft-lbs. absorbed energy and 35 mils laterial expansion for the longitudinal direction. NRC technical position MTEB-5-2 was used to establish a reference temperature for transverse direction (RTNDT) f -12*F.

The mean RT value for the Fort Calhoun submerged arc vessel weldments was determiNN to be -56*F with a standard deviation of 17*F. By apply-  :

ing the shift prediction methodology of the proposed Regulatory Guide 1.99, Revision 2, a weld material adjusted reference temperature (RT was establishedat10*FbasedonthemeanvalueplustwcstandardNN)iations.

The standard deviation was determined by using the root-mean-squares ,

method to combine the margin of 28*F for uncertainty in the shift equation with the margin of 17*F for uncertainty in the initial RT NDT value.

7 Similar testing was not performed on all remaining material in the reactor l coolant system. However, sufficient impact testing was performed to meet appropriate design code requirements (3) and a conservative RTNDT / 504 has been established.

l As a result of fast neutron irradiation in the region of the core, there will be an increase in the T with operation. The techniques used to predict the integrated fast NNtron (E21 MeV) fluxes of the reactor vessel are described in Section 3.4.6 of the USAR, except that the integrated fast neutron flux (E21 MeV) is 2.55x1018 n/cm2, including tolerance at the l inside surface of the critical reactor the 40 years design life of the vessel.gsel beltline weld material, over .

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 exp;sure by application of the calibrated azimuthal neutron flux variation. The maximum integrated fast neutron (E21 MeV) exposure of the reactor vessel at the critical reactor vessel beltline location including tolerance is computed to be l

2.55x1019 n/cm2 at the vessel inside surface for 40 years operation at [

2-5 Amendment No. 22,47,64,74,77,100, 114

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

1500 MWt and 80% load factor. The predicted shift at this location at the 1/4t depth from the inner surface is 332 F, including margin, and was calculated using the shift prediction equation of the proposed Regulatory Guide 1.99, Revision 2. The actual sh!ft in T will be re-established periodicallyduringtheplantoperationbyteskk[gofreactorvessel 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 T causedbyirradiation,limitsonthepressure-temperaturerelation$0Ip are periodically changed to stay within the stress limits during heatup and cooldown. Analysg)ofthesecondremovedirradiatedreactorvessel surveillance specimen , combined with weld chemical composition data and reduced core loading designs initiated in Cycle 8, indicated that the fluence at the end of 14.0 Effective Full Power Years (EFPY) at l l 1500 MWt will be 1.4x1019 n/cm2 on the inside surface'of the reactor l vessel. This results in a total shift of the RT of 285 F, including i margin,fortheareaofgreatestsensitivity(weEmetal)atthe1/4t

location as determined from Figure 2-3. Operation through fuel Cycle 15 will result in less than 14.0 EFPY.

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

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A. Heatup and Cooldown Curves - From Section III of the ASME Code,

, Appendix G-2215.

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

I K Stress intensity factor for membrane stress (pressure).

IM = The 2 represents a safety factor of 2 on pressure.

KIT = Stress intensity factor radial thermal gradient.

The above equation is applied to the reactor vessel beltline.

For plant heatup the thermal stress is opposite in sign fra the pressure stress and consideration of a heatup rate would allow for a higher pressure. For heatup it is therefore conservative to consider an isothermal heatup or KIT = 0.

For plant cooldown thermal and pressure stress are additive.

l 2-6 Amendment No. 22,47,64,74,77,100,114

RCS PRESS-TEMP LIMITS HEATUP 14 EFPY REACTOR NOT CRITICAL 1500 MWt 3200 75MER RESS PSIA) 3000 f

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TECHNICAL SPECIFICATIONS Amendment No. 75,77,100, 114 2-iA

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' I''I 82 0 50 100 ,150 200 250 300 350 400 450 500 Rt IMf( TE)F (DEG F) Tc FORTCALHOUN FIGURE l Amendment "o- 74 77 20$ 114 TECHNICAL SPECIFICATIONS 2-1B

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PREDICTED RADIATION INDUCED NDTT SHIFT FORT CALHOUN REACTOR VESSEL BELTLINE UN 500',

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FORT CALHOUN FIGURE TECHNICAL 2-3

. SPECIFICATIONS Amendment no. 7A,77,100,114

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