ML18058A286

Forwards Analysis & Schedule for Implementation of Flux Reduction Program to Address 10CFR61 Re PTS

ML18058A286
Person / Time
Site:	Palisades
Issue date:	03/13/1992
From:	Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9203230248
Download: ML18058A286 (14)

Text

consumers Power GB Slade General Manager POWERIN&

MICHl&AN'S PRD&RESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 MAR 13 1992 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - ANALYSIS AND SCHEDULE FOR IMPLEMENTATION OF A FLUX REDUCTION PROGRAM - 10CFR50.61 - PRESSURIZED THERMAL SHOCK In accordance with 10CFR50.6l(b)(4), each pressurized water reactor, for which the value of RTprs is projected to exceed the PTS screening criterion before the expiration date of the operating license, is required to submit, by March 16, 1992, an analysis and schedule for implementation of flux reduction programs as are reasonably practicable to avoid exceeding the PTS screening criteria set forth in paragraph (b)(2) of 10CFR50.61. Attachment 1 to this letter provides background information, describes present flux reduction efforts and discusses possible additional future flux reduction efforts and other methods of complying with 10CFR50. 61.

Our December 16, 1991 submittal which reported projected values of RTprs for reactor beltline materials estimated that, without additional fluence reduction, the screening criterion will be exceeded in 2005. However, the NRC staff has expressed concerns regarding the fluence calculation methodology and the determination of the best estimate weld material chemistry reported in that submittal. In responding to these concerns, Consumers Power Company has developed more extensive data which supports a conclusion that the screening criterion will not be exceeded before the end of licensed life in 2007. This information will be documented in a revision to the December 16, 1991 submittal and will be submitted by April 30, 1992. Until the NRC has had the opportunity to review this revised information and made a determination with respect to its acceptability, it would not be prudent to select and implement additional flux reduction methodology; therefore, the information in Attachment 1 should be considered preliminary. A final report will be submitted after the NRC determination has been made.

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Gerald B Slade cJ' -

General Manager CC Administrator, Region III, USNRC NRC Resident Inspector - Palisades Attachment

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ATTACHMENT 1 Consumers Power Company Pali sades Pl ant Docket 50-255 ANALYSIS AND SCHEDULE FOR IMPLEMENTATION OF A FLUX REDUCTION PROGRAM

• Page 1 INTRODUCTION Consumers Power Company's December 16, 1991 submittal indicated that the reactor vessel beltline circumferential weld is projected to exceed the PTS screening criterion in March 2007 and the axial welds are projected to exceed the screening criterion in November 2005. These dates assume Cycle 9 fluence is maintained until the screening criteria is exceeded or the expiration date of the operating license is reached. Since the PTS screening criteria will be exceeded prior to the 2007 expiration date of the Palisades Operating License, this information is submitted in accordance with the requirements of 10CFR50.61. Specifically, the information provides an analysis and schedule for implementation of potential flux reduction programs as deemed reasonably practicable to avoid exceeding the PTS screening criteria.

This information should be considered preliminary, in that the company is actively responding to NRC questions regarding vessel weld chemistry and fluence calculations used in our December 16, 1991 submittal. The outcome of these discussions will have a direct impact on the scope and schedule of the selected flux reduction program. A final report date will be established at the conclusion of the aforementioned discussions.

• Page 2 BACKGROUND In 1986, it was recognized that the Palisades reactor vessel base metal material would exceed the PTS screening criterion before the 2007 expiration date of the Palisades Operating License. This conclusion was based on the then existing requirements of 10CFRS0.61. Subsequent analyses based on the requirements of Regulatory Guide 1.99 Rev 2 reached the same conclusion; however, the limiting reactor vessel material has changed from base metal to the axial welds. The latest analysis, submitted on December 16, 1991, concluded that the limiting material continues to be the axial welds.

Based on the conclusion that the screening criteria would be exceeded prior to the end of licensed life, Palisades initiated a fuel management program in 1988 with Cycle 8. The purpose of this program was to reduce neutron fluence at the vessel inside diameter (ID). Due to the long lead times involved, manufacturing of special assemblies was not possible at that time; however, 16 fuel assemblies were reconstituted. Those fuel assemblies had four rows of stainless steel rods on the outside of the fuel assembly and were placed at the outer perimeter of the core. The selected positions were chosen to maximize flux reduction at the axial weld positions which were limiting.

These reconstituted fuel assemblies were used for Cycle 8 only. In 1991, flux reduction efforts continued with Cycle 9 which utilized 16 hafnium inserts placed in thrice burned fuel. These fuel assemblies were placed on the core perimeter to maximize flux reduction at the axial welds. In 1992, Cycle 10 fuel will use eight shielded fuel assemblies manufactured by the fuel vendor.

They will have two rows of stainless steel rods on opposite sides of the fuel assembly and can be reused for six cycles. These fuel assemblies will have low enrichment and, therefore, will provide improved flux reduction. The 16 hafnium insert fuel assemblies used in Cycle 9 will be reused in Cycle 10.

Additionally, development of the PIDAL incore monitoring methodology has allowed greater flexibility in core design by allowing 1/4 core rotational symmetry (versus INCA's 1/8 core reflective symmetry) which provides 22 additional core locations for new fuel. Cycle 10 will be the first core to take advantage of this methodology.

Figures 1 through 4 provide a pictorial of core peripheral patterns for Cycles 1-7 and Cycles 8-10. Table 1 provides neutron flux data versus azimuth for 1/8 core symmetry for Cycles 1 through 10. The axial welds are at the 0° and 30° locations while the peak fluence for the circumferential weld occurs at the 16°location. The data illustrates the success of our ongoing fuel management program to significantly reduce vessel fluence. Cycle 11 will be discussed later in this enclosure.

While not part of the Palisades flux reduction program, an extensive dosimetry program was initiated in 1988 corresponding to the beginning of life (BOL) of Cycle 8. The purpose of this program is to develop plant specific benchmarks to verify the accuracy of DOT calculations used to predict reactor vessel fluence. This program utilizes both in-vessel and ex-vessel dosimetry. Ex-vessel dosimetry was removed at the end of Cycle 8 and verified that fluence calculations were conservative when compared to measured values. In-vessel dosimetry was installed at the beginning of Cycle 9. During the present plant refueling outage, both the in-vessel and ex-vessel dosimetry have been removed

• Page 3 for analysis. It is estimated that results of the analysis will be available in mid-1992. These results will provide a direct comparison of the calculated and measured fluxes at both the ID and OD of the reactor vessel for the same fluence history. This is considered the best data available to confirm plant specific accuracy of the calculational DOT model. While this program will not necessarily show a reduced vessel flux, it will significantly increase the confidence level of the DOT model and will form the bases to preclude any fluence bias or uncertainty penalties.

Present and Future Flux Reduction Efforts Methods to achieve further flux reduction to the critical reactor vessel welds include:

- Neutron Pads

- More Aggressive Fuel Management Core Barrel Replacement Reactor Vessel Replacement

- *Reactor Vessel Annealing

- *Integrated Surveillance Program

• These are considered equivalent or alternatives to flux reduction.

Neutron Pads Neutron pads installed on the core barrel were selected for further study based on cost/benefit, availability of technology, and implementation schedule. A preliminary neutron pad analysis was conducted by Westinghouse.

The basic conclusions of the analysis were:

• A flux reduction factor of two could be achieved at the axial locations.

Similar reductions could be achieved on the circumferential weld.

• Stainless steel or tungsten encapsulated in stainless steel cans are suitable shield material.

• Detailed engineering, manufacturing, tooling and installation would require approximately 18-24 months.

• Pads could provide a minimum of six years additional vessel lifetime if installed not later than BOL for Cycle 12.

Figure 5 illustrates the neutron pad design for axial welds. Figure 6 shows the expected reduction in neutron flux as a fraction of pad thickness for stainless steel. Tungsten is approximately 50% more effective than stainless steel.

Future Fuel Management While the neutron pad option is feasible, it would require significant development and installation costs and an extensive outage to accomplish. For

• Page 3 for analysis. It is estimated that results of the analysis will be available in mid-1992. These results will provide a direct comparison of the calculated and measured fluxes at both the ID and OD of the reactor vessel for the same fluence history. This is considered the best data available to confirm plant specific accuracy of the calculational DOT model. While this program will not necessarily show a reduced vessel flux, it will significantly increase the confidence level of the DOT model and will form the bases to preclude any fluence bias or uncertainty penalties.

Present and Future Flux Reduction Efforts Methods to achieve further flux reduction to the critical reactor vessel welds include:

Neutron Pads

- More Aggressive Fuel Management Core Barrel Replacement Reactor Vessel Replacement

- *Reactor Vessel Annealing

- *Integrated Surveillance Program

• These are considered equivalent or alternatives to flux reduction.

Neutron Pads Neutron pads installed on the core barrel were selected for further study based on cost/benefit, availability of technology, and implementation schedule. A preliminary neutron pad analysis was conducted by Westinghouse.The basic conclusions of the analysis were:

• A flux reduction factor of two could be achieved at the axial locations.

Similar reductions could be achieved on the circumferential weld.

• Stainless steel or tungsten encapsulated in stainless steel cans are suitable shield material.

• Detailed engineering, manufacturing, tooling and installation would require approximately 18-24 months.

• Pads could provide a minimum of six years additional vessel lifetime if installed not later than BOL for Cycle 12.

Figure 5 illustrates the neutron pad design for axial welds. Figure 6 shows the expected reduction in neutron flux as a fraction of pad thickness for stainless steel. Tungsten is approximately 50% more effective than stainless steel.

Future Fuel Management While the neutron pad option is feasible, it would require significant development and installation costs and an extensive outage to accomplish. For

• Page 4 this reason, it was decided to further investigate more aggressive fuel management concepts. The following peripheral fuel management options are presently being considered for Cycle 11 and beyond.

Number and Type of Fuel Assembly Option Hafnium Shielded Twice Burned Fuel A 16 16 16 B 16 24 8 c 24 24 Added options are possible by reaching the above configurations in two cycles versus one cycle. Basically this would depend on whether the shielded assemblies are manufactured or reconstituted.

Initial scoping studies indicate the benefit of this more aggressive fuel management concept is approximately equivalent to the flux reduction achieved for neutron pads. Option B provides the greatest flux reduction of the options being considered.

The scoping calculations indicate that a fuel management concept can provide a flux reduction estimated to extend lifetime by 5-6 years, operate within peaking factor limits, and have a cycle lifetime comparable to Cycle 10 (350 EFPD). The selected option should be available for Cycle 11, but not later than Cycle 12.

Core Barrel/Reactor Vessel Replacement These options are not considered to be economically feasible.

Reactor Vessel Annealing While not an option that reduces neutron flux, annealing is considered an equivalent option since it has been shown to eliminate embrittlement and return the vessel material to nearly its original condition. We have held exploratory discussions on annealing with MPR Associates, Inc. and Westinghouse. MPR would utilize proven Russian expertise and technology.

Presently, we are evaluating feasibility study proposals from both MPR and Westinghouse. These studies would evaluate conceptual equipment designs, thermal and stress analyses, implementation procedures, recovery and re-embrittlement rates, and associated costs and schedule. We expect to make a decision on the feasibility study within the next few months. The study is expected to take 7 to 10 months to complete. The decision to pursue annealing as an option, depends on decisions related to lifetime extension, regulatory requirements and economic risks.

• Page 5 Integrated Surveillance Program We are presently a participant in a Combustion Engineering sponsored reactor vessel owners group. The purpose of this group is to develop a comprehensive data bank of surveillance specimen information to identify vessels with similar materials and develop the methodology to allow use of surveillance data from one vessel to be used to determine embrittlement properties in a similar vessel. This approach would allow Palisades to use Section 2 of Regulatory Guide l.gg Rev 2, which allows a reduction in the margin term (used to account for uncertainty) when surveillance data is available. If successful, this integrated surveillance program could be used to demonstrate compliance with PTS screening criteria.

Table 1 Palisades Fluence Through Cycle g At the Reactor Vessel Clad-Base Metal Interface Cycle Cycle Flux Cycle Length (n/cm 2/s)

(EFPD) 0 Degrees 16 Degrees 30 Degrees 45 Degrees 1 37g,4 4.5gE+lO 6.03E+l0 4.70E+l0 2.g8E+lO 2 44g,1 4.5gE+lO 6.03E+l0 4.70E+l0 2.g8E+lO 3 34g,5 4.5gE+lO 6.03E+l0 4.70E+l0 2.g8E+lO 4 327.6 4.5gE+lO 6.03E+l0 4.70E+l0 2.g8E+lO 5 3g4_6 4.5gE+lO 6.03E+l0 4.70E+IO 2.g8E+lO 6 333.4 4.87E+l0 6.25E+l0 4,7gE+lO 3.03E+l0 7 35g,g 4.87E+l0 6.25E+l0 4.7gE+lO 3.03E+l0 8 373.6 2.lOE+lO 4.76E+l0 2.28E+l0 l.73E+l0 g 2g8,5 2.0gE+lO 3.05E+l0 1. ggE+lO l.14E+l0 10 350 ~2.0gE+lO ~3.05E+l0 ~1.ggE+lO ~l.14E+l0 The limiting axial welds are at 0° and 30° which corresponds to go 0

, 270° and 30°, 150°, 210°, 330° respectively in Figures 1-4.

The limiting circumferential weld locations are at 16° which corresponds to

+/-16° on each side of 0°, go 180° and 270° on Figures 1-4.

0

• Figure 1 TYPICAL CYCLES 1 THROUGH 7 PERIPHERAL LOADING PATTERN Number of Accumulated Cycles 0 0 0 0 0 0 0 0 0 0 0

0 90° 270° 0 0 0

0 0 0 0 0 0 0 0 0 0 0 i.

//30° Axial Weld Azimuthal Location ALL PERIPHERAL ASSEMBLIES ARE FRESH FUEL

• Figure 2 CYCLE 8 PERIPHERAL LOADING PATTERN Stainless Steel 180° Shield Rods Number of Accumulat d Cycles 2

0 0

3 3 270° 3

J 0

0 0 J 3 0 I

r.

Axial Weld Azimuthal Location

• Fignre 3 CYCLE 9 PERIPHERAL LOADING PATTERN Number of Accumulat d Cycles 2 3 3 2 2 2 2 2 3 2 2 2 2 2

2 2 3 3 90° 270° 3 3 2

2 2

3 2 ..,

2 2 2 3 3 2

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/30*

Axial Weld oo Azimuthal Location ASSEMBLIES REPRESENTED BY 3 ARE THRICE BURNED FUEL WITH HAFNIUM ABSORBERS

CYCLE 10 PERIPHERAL LOADING PATI'ERB

• • Figure 4 Stainless Steel Shield Rods Number of

• 2 2 2

2 2 2 2 2 0 0 90.J 0 0 0 270 2

2 2 2

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Axial Weld Azimuthal Location ASSEMBLIES REPDSERTED BY 4 ARE 4X BUR.lfED FUEL WITH HAJ'lllIUM ABSORBERS ASSEMBLIES UPUSEJIT!D BY 0 ARE HEW SHIELD FUEL WITH SS SHIELD RODS AND 1. 2 W/O ENRICBMDT

• 180

• Figure 5 150 210

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VESSEL 27C SURVEILLANCE*

CAPSULE HOLDER LOWER INTERNAL AXIAL WELD AZIMUTHAL LOCATION 30 0 330 S/G A Figure 4.3-5 Palisa~es Reactor Pressure Vessel - Lower Internals

Flux Reduction Factor

• Figure 6 10r-------------------------------------

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1~----..L-~--.l--~~l-----L-~--L-~__J 0 1 2 3 4 5 6 Inches of Stainless Steel Attached to ~rrel 0.5" gap Figure 4.3-1 Flux Reduction Benefits from the Addition of Steel