L-97-034, Forwards Cycle 13 Reload & COLR, Updated to Include New Radial Peaking Factor at Rated Thermal Power Limits for Unrodded Core Planes.Cycle 13 Reload Design,Discussed

Forwards Cycle 13 Reload & COLR, Updated to Include New Radial Peaking Factor at Rated Thermal Power Limits for Unrodded Core Planes.Cycle 13 Reload Design,Discussed

ML20198B933
Person / Time
Site:	Beaver Valley
Issue date:	12/22/1997
From:	Jain S DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20198B937	List: L-97-034, Forwards Cycle 13 Reload & COLR, Updated to Include New Radial Peaking Factor at Rated Thermal Power Limits for Unrodded Core Planes.Cycle 13 Reload Design,Discussed ML20198B939
References
L-97-034, L-97-34, NUDOCS 9801070130
Download: ML20198B933 (5)

Text

. . - --.

@ @ % gv*r Powor station

$Nppinw> ort. PA 15077 0004

=

December 22,1997 L-97-034 U. S. Nuclear Regulatory Commission Document Control Pesk fttention:

Washington, DC 20555-0001

Subject:

Heaver Valley Power Station, Unit No.1 Docket No. 50-334, License No. DPR-66 Cycle 13 Reload and Core Operating Limits Report Beaver Valley Power Station, Unit No. I completed the twelfth cycle of operation on September 27,1997, with a burnup of 15148.40 MWD /MTU. This letter describes the Cycle 13 reload design, documents our review in accordance with 10 CFR 50.59 including our detennination that no unreviewed safety question is involved, and provides a copy of the Core Operating Limits Report (COLR) in accordance with Technical Specification 6.9.1.12.

The new core configuration is arranged in a low low leakage loading pattern and

involves replacing nin- (9) Region 1, eight (8) Region 10A and 10B, twenty-four (24)

Region 12, and forty.four (44) Region 13 fuel assemblies vith twelve (12) Region 10 fuel assemblies, eigh'. (8) Region 12 fuel assemblies, and forty (40) fresh Region 15A ZIRLO* fuel assemblies enriched to 4.2 weight percent and twenty-four (24) fresh Region ISB ZIRI.05 fuel assemblies enriched to 4.6 weight percent. The center fuel assembly from Region- I was replaced with another Region I fuel assembly from Cycle 1. In addition, two (2) Region 10 fuel assemblies and eight (8) Region 12 fuel assemblies have l'een reconstituted and will contain a total of 14 stainless steel rods.

WCAP-13060-P-A, " Westinghouse Fuel Assembly Reconstitution Evaluation i Methodology," provided the NRC approved codes and methods for evaluating the effects of the reconstituted fuel assemblies on the operation of the core in compliance with Technical Specification Design Feature 5.2.1.

i The mechanical design of the new Region 15A and ISB fuel assemblies is similar to the previous reload fuel except for the following features: ,

l 90010 I f ^70130 971222'"" 0*0@ !p OEllVElllRG e.,0033 '

1.1!l.ill.l!Ilj11Il.l H

^ ~

+  : Beaver Valley Pcwer Station, Unit No. l Cycle 13 Reload and Core Operating Limits Report Page 2 t

(1) To improve fuel assembly corrosion resistance and dimensional stability under irradiation, the fuel rod cladding will be ZlRLO and the fuel assembly skeleton .

will be fabricated with ZlRLO guide thimble tubes, ZlRLO instrument tubes and ZlRLO mid grids.

The chemical composition of the fuel rods and core components fabricated with ZlRLO alloy is similar to the Region 14 Zircaloy-4 fuel assemblies except for a slight reduction in the content of Tin _(Sn), Iron (Fe) and the climination of Chromium (Cr). The ZlRLO alloy also contains a nominal amount of Niobium (Nb). These composition changes, although small, are responsible for the improved corrosion resistance of ZlRLO* compared to Zircaloy-4.

Due to the longer fuel rod end plugs in the protective bottom grid described below, the Region 15 ZIRLO fuel rod length is longer than the Zircaloy-4 clad fuel rods  ;

used in Region 14. The Region 15 plenum spring is a variable pitch spring but is similar to the previous Zircaloy-4 design and provides a comparable holddown force. The variable pitch design provides additional space for gas release and regains sufYicient plenum volume while not sacrificing perfonnance.

(2) The design of the Beaver Valley Unit 1 Cycle 13 Region 15 fuel assemblies

-incorporates _ debris mitigating features that include one additional grid (protective -

bottom grid), modified fuel rod end plugs and a protective oxide coating along the bottom section of the fuel rod clad.

The protective bottom grid is a partial height grid similar in configuration to the mid grid, but fabricated ofinconel without mixing vanes. It is positioned on the top plate of the bottom nozzle. The fuel rod end plug positioned within the protective bottom grid is an elongated version of the current fuel rod end plug design. In conjunction with the protective bottom grid and the new elongated bottom end plug, the fuel rod top end plug was elongated and fitted with an extemal gripper to assist

! in repositioning the fuel rod during fabrication and also to facilitate fuel rod removal during reconstitution. The bottom Inconel grid was raised and the first ZIRLO grid was lowered to reduce the lowermost span in the fuel assembly. This resulted in Span No. I being shonened and Span No. 2 being lengthened. These i changes increase the resistance of the fuel assembly to flow induced fuel rod i- vibration to preclude excessive clad wear. The protective bottom grid retains the original function as a debris mitigation feature and also its benefit of increased

. support to-the fuel rod end. All other grid elevations remain the same as the previous region.

~

L

~ _ _ . ._.__ _ _ . _ _ .._ _ .___ __ _ -__ ___

" Beaver Valley Power Station, Unit No.1 Cycle 13 Reload and Core Operating Limits Report Page 3 ,

The hydraulic effects of the protective bottom grid are minimized by positioning the fuel rods 0.085 inch above the bottom nozzle in a manner that reduces the pressure -

drop at the_ fuel rod ends. Component testing indicated that the design of the protective bottom grid and the elongated end plug causes no significant effect on fuel assembly hydraulic performance. The combination of the lowered fuel rod position and the longer fuel rod end plug results in no change to the axial fuel stack height from the previous fuel region.

As an additional level of debris defense, a hardened coating of zirconium oxide shields the bottom section of the fuel rod clad, increasing wear resirtance by more than 'a factor of ten over the current design. Fonned by exposing rods and end plugs to heating in an induction fumace, this zirconium oxide layer is at least twice as hard as the most common type of debris. Should debris pass through the bottom nozzle,. and progress past the lower dimples of the bottom grid, this coating provides an added measure of protection.

(3) The reconstituted fuel assemblies replaced damaged fuel rods with stainless steel filler rods. Stainless steel filler rods have been used as replacements for fuel rods in previous cycles as well as in other plants-and do not adversely affect the mechanical integrity of the fuel assembly. Thc core loading pattern satisfies all fuel assembly design criteria with the reconstituted fuel assemblies inserted.

1 (4) The Region 15 Integral Fuel Burnable Absorber (IFBA) rods will have a rod internal gas pressure decrease from 200 to 100 psig. This was done to provide 3 improved margins to the rod internal pressure criteria. Evaluations performed I demonstrated that the change to 100 psig rod internal gas pressure satisfied the fuel design criteria.

( Four Region 10 and eight Region 12 assemblies will utilize damper rods. Damper  !

rods were used in Cycle 12. The purpose of the damper rod assembly is to reduce the l

flow induced vibration of the older VANTAGE SH assemblies with non-rotated grids to

! acceptable levels when the fuel assemblies are located at the core periphery. Each i damper assembly consists of twenty-four (24) solid Zircaloy-4 damper rodlets attached E to a holddown assembly. The damper rods are inserted into the fuel assembly guide thimbles and the damping devices are designed to be used in fuel assemblies without rotated grids that may be required to be placed in the baffle region of the core. The i

. pertinent thermal-hydraulic and boiling criteria were evaluated and found to be y acceptable. The use of vibration damping assemblies does not compromise the ,

performance of any' safety-related system nor result in any adverse effect on any l analysis. since this change does not affect the normal plant operating parameters, the

! l

Beaver Valley Power Station, Unit No.1 ,

Cycle 13 Reload and Core Operating Limits Report Page _4 4 .

?

safeguards systems actuation, or the assumptions and input parameters used in these analyses.

As in the previous cycle,' eight Powe: Suppression Assemblies (PSAs) will continue .

to be utilized in fuel assemblies located on the core periphery. The PSAs contain hafnium rodlets to reduce local neutron leakage near the reactor vessel lower plate.

Pending the results of the Rod Cluster Control Assembly inspection; enhanced

-performance (EP-RCCAs) may be utilized in Cycle 13 or in later cycles as required.

. These have a thin chrome electroplate applied to a specified length of absorber rodlet cladding in contact with the reactor internal guides to provide increased resistance to cladding wear. . In addition, the absorber diameter is reduced slightly at the lower extremity of the rodlets in order to accommodate absorber swelling and minimize ,

cladding interaction.

These modifications meet fuel assembly and rod design criteria and will not  :

~

adversely affect the core safety considerations. Fuel rod design evaluaSons for the new fuel were performed using NRC approved methodology to demonstrate that the fuel rod design bases are satisfied.

Duquesne Light Company has performed a review of this reload core design including a review of the core characteristics to determine those parameters affecting the postulated accidents described in the Updated Final Safety Analysis Report (UFSAR).

The consequences of those accidents described in the UFSAR which could potentially be affected by the reload core characteristics were evaluated in accordance with the NRC approved methodology described in WCAP-9272-P-A " Westinghouse Reload Safety Evaluation Methodology." The effect of the reload design was either accommodated

- within the conservatisms of the assumptions used in the current analysis design basis, or

- it was demonstrated through evaluation that the reload parameters would not change the conclusions in the UFSAR.

No technical specification changes are required as a result of this reload design.

The NRC approved dropped rod methodology (WCAP-10298-A [non-proprietary],

June-1983) was used for this design evaluation and confirmed that the peaking factors

'did not exceed the safety analyses limits.

The reload core design will be verified by performing the standard Westinghouse

= reload core physics startup tests. The results of the. following startup tests will be submitted in accordance with Technical Specification 6.9.1.3:

N------ -i- - e,. , , .

- Beaver Valley Power Station 0 Unit No.1 .

- Cycle 13 Reload and Core Operating Limits Report Page 5 l.. Control rod drive tests and rod drop time measurements.

2.- Critical boron concentration measurements.

3. Control rod bank wodh measurements.

4. Moderator temperature coeflicient measurements.  ;

5. Stanup power distribution measurements using the incore flux mapping system.

The CO2 R (enclosed) has been updated for this cycle to include new radial peaking factor at rated thermal power [Fxy (RTP)] limits for unrodded core planes. Figure 4 has been replaced with a new figure to address these new limits.

.The Beaver Valley Onsite Safety Committee and the Duquesne Light Company Nuclear Safety Review Board'have reviewed the Reload Safety Evaluation and Core ,

Operating Limits Repon and determined that this reload design will not adversely affect the safety of the plant and does not involve an unreviewed safety question.

In addition, the fuel rod intemal pressure concem has been evaluated and the criterion remains valid to mid-cycle (8,000 MWD /MTU). A plant specific analysis is currently being performed to . validate that the 10 CFR 50.46 oxidation criterion will not be exceeded for the full operation of Cycle 13. The Commission will be notified should

-we find that the criterion may be exceeded for any part of the cycle and of our plans to ensure the plant remains in compliance with the criterion.

Sincerely, Sushil C. Jain Enclosure c: Mr. D. M. Kern, Sr. Resident Inspector Mr. H. J. Miller, NRC Region I Administrator Mr. D. S. Brinkman, Sr. Project Manager fT"- Y * ----p ..-> p en.e--- -e