ML20096F325

Advises That Detailed Fluence Analysis for Fracture Toughness Requirements for Protection Against PTS Events Completed & Schedule Developed for Implementation of Further Improvements in Flux Reduction Program

ML20096F325
Person / Time
Site:	Fort Calhoun
Issue date:	05/15/1992
From:	Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-92-175R, NUDOCS 9205210034
Download: ML20096F325 (4)

Text

f Omaha Public Power District 444 South 16th Street Mall Omaha, Nebraska 68102 2247 h

2 402/636-2000 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station Pl-137 Washington, DC 20555

References:

1.

Docket No. 50-285 2.

Letter from LeBoeuf, Lamb, Leiby & MacRae (Attorneys for OPPD) to NRC (H. R. Denton Desk) dated Decembe(R).

dated July )17,1986 3.

Letter from OPPD r 21,1987 (LIC-87-0692)(Document Control L. Andrews to NRC 4.

Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988 5.

Letter from OPPD (K. J. Morris) to NRC (Document Control Desk) dated May 18, 1989 (LIC-89 284) 6.

Letter from OPPD16(,1991 (LIC-91-327R)(Document Control Desk)

W. G. Gates) to NRC dated December 7.

LetterfromOPPD(W.G. Gates)toNRC(DocumantControlDesk) dated March 17, 1992 (LIC-92-093R)

Gentlemen:

SUBJECT:

Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock (PTS) Events, 10 CFR 50.61(b), for fort Calhoup i

Station (FCS)

In accordance with the Omaha Public Power District 7, this letter is submitted to notify the NRC tha(0 PPD) commitment of Reference t a detailed fluence analysis for FCS has been completed.

A schedule for implementation of further improvements to the flux reduction program has also been developed and is discussed later in this letter. These actions have been taken in accordance with 10 CFR 50.61(b)(4) and OPPD commitments made in Reference 6.

The FCS operating license currently expires in June 2008.

dased upon 6

conservative fluence extrapolations, Reference 6 concluded that the PTS screening i

criterion would not be reached until the year 2009.

This follow-up submittal is required per 10 CFR 50.61 due to OPPD's request (Reference 2) for a five year extension of the license until the year 2013. As noted in Reference 5, OPPD has been closely monitoring the regulatory isscas associated with mitigating PTS events.

Initial PTS flux reduction efforts included the implementation of a low radial leakage fuel management strategy in 1983 Cycle 8 and improvements have continued during subsequent cycles. For examp(le, duri)g Cycle 10, FCS utilized n

a non-optimized extreme low radial leakage fuel management strateg' strategy for y.

FCS has implemented an optimized extreme low radial leakage fuel management the current fuel cycle (Cycle 14).

9205210034 920515 i

PDR ADOCK 05000285 i'

P POR

)

l

@ 0 ()B ~*

'* x-='a *=x

< s-s

• LIC 92 175R

Page 2 Core loading for Cycle 14 includes the use of hafnium flux suppression rods in reactor vessel welds twelve peripheral fuel assemblies located near the limitinfmiting reactor vessel four natural uranium fuel assemblies located near the l welds and integral fuel burnable absorbers. This type of core loading scheme and I

continued improvements in core periphery flux reduction are the basis for OPP 0's

[

fuel management program.

To comply with the latest 10 CFR 50.61 requirements the Reference 6 assessment used best estimate neutron fluence values for Cycleg 11 through 13.

The accumulated surface fluence was estimated at 9.03 x 10 n/cm' for the 60' and 300* 3-410 course) longitudinal weld locations.

The 180' 3 410 longitudinal (lowerweld was determined to be less limiting than the 60' and 300' 3 410 longitudinal welds.

Future cycles were assumed to use core loading patterns equivalent to Cycle 14.

Cycle 14 fluence data did not exist prior to the Reference 6 submittal date.

Therefore, Reference 6 conservatively assumed that Cycle 10 fluence data would be representative of fuel cycles following Cycle 13. Based upon a revious Cycle 10 DOT 4.3 analysis performed by ABB/ Combustion Engineering ABB/CE), the incremental fluence per effective full power year (hrough Cycle 13 combined with EFPY) was estima ed to be 0.41 x 10" n/cm,.

Using best estimate fluence values t extrapolations based on the Cycle 10 DOT 4.3 analysis, it was determined in Reference 6 that-the PTS screening criterion of 270*F would be reached in the year 2009.

To obtain a current fluence evaluation of the FCS reactor vessel and update future projections, OPPD recently contracted with ABB/CE to perform this work.

Detailed flux calculations were performed using the DOT 4.3 computer code to determine the fast neutron fluence for Cycles 11-13 and to project the impact of implementing the Cycle 14 optimized extreme low radial leakage fuel management strategy.

OPPD has received the final results from the new ABB/CE DOT 4.3 analysis as described below.

The new analysis determined that the accumulated fast neutron fluence at the reactor vessel surface for the 60* and 300' 3-410 longitudinal weld locations is 9.57 x 10" n/cm' through the end of Cycle 13. The 180' 3-410 longitudinal weld was again determined to be less limiting than the 60' and 300* 3-410 longitudinal welds. Based upon cycle specific D0T 4.3 calculations for Cycles 11 through 13 and design DOT 4.3 calculations f or Cycle 14, the projected accumulated fluence at the end of Cycle 14 is 1.00 x 10" n/cm'. The projected Cycle 14 incremental fluence per EFPY for the 60* and 300* 3 410 longitudinal weld locations is 0.34 x 10" n/cm'.

As a result of this new analysis, OPPD has extended its estimate for reaching the PTS screeniag criterion of 270*F.

Using the best estimate fluence values from the new ABB/CE DOT 4.3 analysis and an estimated load capacity factor of 0.77, the PTS s<;reening criterion of 270'F is now estimated to occur late in the year 2010 or early 2011.

i l

l

1 LIC 92-175R Page'3 1

OPPD has projected values for RT,3. for FCS reactor vessel beltline materials based upon Reference 3 data and on the accumulated fluence data through Cycle 13 l

from the new ABB/CE DOT 4.3 analysis.

Table 1 (attached) summarizes the chemistry factors and adjusted reference temperatures of nil ductility transition

( ARTu,,) obtained through the application of 10 CfR 50.61(b)(2)(iv).

The chemistry factors remain unchanged from those previously submitted in Table 2 of Reference 3.

of 270*F will be /CE DOT 4.3 analysis predicts that the PTS screening criterion Since the new ABB exceeded prior to the year 2013 (final year of the requested operating license extension), OPPD plans to continue evaluating additional flux reduction options. Currently identified flux reduction options include the use of staihless steel replacement pins in perioheral fuel assemblies and the use of full core fuel management rather than quarter core reflective fuel management to minimize thermal margin losses. Prior to Cycle 15 startup (currently scheduled for November 1993)

O management strategies. PPD intends to implement additional flux reduction fuel Therefore, OPPD requests that the five year license extension application remain active.

Additional information to justify the proposed lic(TAC 82834) ion will be provided within ninety (90) days ense extens

-following Cycle 15 startup.

If you should have any questions, please contact me.

1 Sincerely, AV. 5]. h tw W. G. Gates Division Manager Nuclear Operations WGG/sel Attachment c:

LeBoeuf, Lamb, Leiby & MacRae R. D. Martin, NRC Regional Administrator, Region IV R. P. Mullikin, NRC Senior Resident inspector D. L. Wicqinton, NRC Senior Project Manager S. D. Bloom. NRC Project Engineer i

i l

l l

-.u.

. _ ~ -. _,, _.. -

7 Attachment to LIC-92-175R 4

Table 1 ARTnD7 or Fort Calhoun Beltline Materials f

Using Reg. Guide 1.99, Rev. 02, and 10 CFR 50.61 60/300 Degree Angle, DOT 4.3 Results Incorporated Weld Cu NI Chemistry ARTuo7('F) ARTuo7('F) ARTuor('F)

Seam (w/o) (w/o)

Factor Cycle 14 2008 2013 2-410 0.17 0.17 89.45 99.42 107,66 109.85 (longitudinal) 3-410 0.22 1.02 234.50 244.4 266.02 271.78 (longitudinal) 8-410 0.21 0.73 185.45 195.40 212.47 217.02 (circumferential) 9-410 0.21 0.74 187.10 197.05 214.28 218.86 (circumferential)

D-4802 0.12 0.56 82.20 118.18 125.75 127.76 (intermediate shell-plate)

D-4812 0.12 0.60 83.00 118.90 126.62 128.65 (lower shell-plate Regulatory Guide 1.99, Rev. 02 Equation:

ARTNOT = Adjusted Reference Temperature of = 1 + M + (CF) f $28-0.10og0 Nil Ductility Transition Where; CF = Chemistry Factor determined from Tables in Regulatory Guide 1.99 and 10 CFR 50.61 f = t'c!culated value of neutron fluence at the reactor vessel / clad interface divided by 1010 For Weld Material:

I = Generic mean value of initial reference temperature = -56 F for welds rnade with Linde 1092 and 124 fluxes.

M = Margin to cover uncertainties in initial RTNDT = 66 F since generic value of I was used.

For Plate Material:

I = Initial reference temperature of irradiated material as defined in the ASME Code = -12 F for reactor vessel beltline plate material.

M = Margin to cover uncertainties in initial RTnor = 48 F since a measured value of I was used.

The proposed PTS criteria applied to the vessellD for longitudinal weld scarns and plate l

materialis RTPTs = 270*F and for circumferential weld seams is RT s = 300 F PT l