ML19290E315

Proposed Alternate Absorber Control Blade Test Program, Revision 1,Class 1

ML19290E315
Person / Time
Site:	Peach Bottom
Issue date:	01/31/1980
From:	Brugge R GENERAL ELECTRIC CO.
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Shared Package
ML19284A871	List: ML19284A870 ML19290E315 ML19290E318 ML19290E322 ML19294C288 ML19294C289
References
NEDO-24231, NUDOCS 8003100215
Download: ML19290E315 (23)

Text

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NED0-24231 Revision 1 CLASS 1 January, 1980 Proposed Peach Bottom Atomic Power Station Unit 2 Alternate Absorber Control Blade Test Program

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Approved: A //

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R. O. Brugge, Manager Services Licensing NUCLE AR ENERGY P90JECTS OlVISION

• GENE A AL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL h ELECTRIC 800310 0 El 5

NED0-24231 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT (Please Read Carefully)

This report was prepared by General Electric solely for Philadelphia Electric Company (PECo) for PECo's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending PECo's operating license of the Peach Bottom Atomic Power Station Unit 2. The information contained in this report is believed by General Electric to be an accurate and true repre-sentation of the facts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting information in this document are contained in the Test Agreement between Philadelphia Electric Company and General Electric Company and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

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NEDO-24231 TABLE OF CONTENTS Page

1. INTRODUCTION 1-1

2. TEST PROGRAM DESCRIPTION 2-1

3. TEST CONTROL BLADE 3-1

4. HAFNIUM ABSORBER R005 4-1 4.1 Corrosion Resistance 4-1 4.2 Radiation Growth 4-2 4.3 Nuclear Characteristics 4-2

5. SAFETY EVALUATION 5-1

6. PROPOSED TECHNICAL SPECIFICATION CHANGE 6-1

7. REFERENCES 7-1 iii

NED0-24231 LIST OF ILLUSTRATIONS Figure Title Page 3-1 Location of Hafnium Absorber Rods 3-2 4-1 Corrosion of Hafnium (Reference 1) 4-9 4-2 Effect of Nitrogen Content on Corrosion 4-10 of Hafnium 4-3 Corrosion of Hafnium (Reference 4) 4-11 LIST OF TABLES Table Title M 4-1 Chemical Analysis of Hafnium 4-4 4-2 Corrosion of Hafnium (Reference 1) 4-5 4-3 Corrosion of Hafnium (Reference 2) 4-6 4-4 Hafnium Absorber Rod Chemical Analysis 4-7 4-5 0xide Film Thickness on Irradiated Hafnium 4-8 5-1 Test Control Blade Scram Time Increase 5-3 iv

NED0-24231

1. INTRODUCTION It is proposed that an Alternate Absorber Control Blade Test Program be conducted at the Peach Bottom Atomic Power Station Unit 2 during fuel Cycle 5, (Reload 4), and Cycle 6 (Reload 5). The purpose of this test program is to obtain data on the performance of hafnium in a Boiling Water Reactor (BWR) environment. The primary parameters of interest are: (1) the corrosion rate, and (2) radiation growth. This series of tests is part of an improvement program to extend the useful operating life of BWR control blades.

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NEDO-24231

2. TEST PROGRAM DESCRIPTION The Alternate Absorber Control Blade Test Program consists of the following major activities.

Two standard control blades in the reactor core will be replaced with test control blades. The test control blades contain three solid hafnium rods in each wing (Figure 3-1), which replace three standard boron carbide (B4C) absorber rods.

The test control blade will be operated as a normal control blade during reactor operation.

After one fuel cycle, one of the control blades will be removed from service for destructive examination and replaced by a standard B 4C contral blade. The second test control blade will be removed after two fuel cycles and replaced by a standard 84C control blade.

The irradiated hafnium rods will be shipped to a hot cell for examination and data collection.

The performance of this test program at Peach Bottom 2 is desirable because of the high power density in the core, and the control cell core mode of operation. This mode of operation allows the maximum control blade burnup in the shortest period of time, since the control blades would almost continuously be in a high radiation field. This is expected to maximize the radiation growth of the hafnium absorber rods.

The particular core locations in the Peach Bottom 2 reactor core which would be most desirable for the test control blade are: 22-31, 38-31, 30-23, or 30-39. These are expected to be the locations of highest control blade exposure in the core. However, other high control blade exposure locations in the core are also acceptable.

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NED0-24231

3. TEST CONTROL BLADE The test control blade is a standard control blade which has been modified as described below.

Each test control blade contains twelve (12) solid hafnium rods (three per wing), and seventy-two (72) stainless steel tubes containing baron carbide (84C), as shown in Figure 3-1. The solid hafnium rods are the test material. The hafnium rods are positioned in the wings to obtain maximum data on irradiation growth due to fast neutron fluence. The test control blade mechanical design is the same as the standard 8 4C control blade currently in use.

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NED0-24231 l

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Figure 3-1. Location of Hafnium Absorber Rods 3-2

NED0-24231

4. HAFNIUM ABSORBER RODS Hafnium is a well known neutron absorbin9 material which has a long nuclear lifetime due to the characteristics of its neutron absorption products which decay into other high cross-section hafnium absorber isotopes. Most of the reactor experience with hafnium is in naval pressurized water reactors, where hafnium has been used extensively as the control blade material. It has also been used in the Shippingport Atomic Power Station pressurized water reactor.

4.1 Corrosion Resistance Almoct all data available on hafnium corrosion in a reactor environ-ment is under pressurized water conditions. Bare hafnium control blades were used in the Experimental Boiling Water Reactor (EBWR),

but no quantitative data on corrosion is readily available for these control blades.

Data on the corrosion of hafnium are presented in References 1 through 4. Tables 4-1 and 4-2 present data from Reference 1 on the corrosion of hafnium of various chemical compositions. Figure 4-1 provides a graphical representation of the data in Table 4-2. The data in Figure 4-1 show that hafnium has very good corrosion resis-tance. The minimal corrosion which occurs under such conditions has made detailed measurements difficult. The data show that the oxide layer formed is only 1 x 10

-6 inch thick after 200 days ir water at 540 F. In addition, after approximately 30 days at 540 F, the rate of oxide layer growth becomes small, and remains relatively constant with time.

Table 4-3 (from Reference 2) shows the superior corrosion resistance of hafnium compared to Zircaloy 2 at various temperatures in water and in steam. Figure 4-2, from the same reference, shows the effect of the nitrogen content in hafnium on its corrosion resistance 4-1

NED0-24231 in 680 F water and 750 F steam. The average nitrogen content of the hafnium to be used for the absorber rods in the test program at Peach Bottom is 26 ppm (Table 4-4), which is equal to the lowest value of nitrogen shown in the curves.

Figure 4-3 (from Reference 3) ; hows data reported by the Bettis Atomic Power Laboratory on the corrosion of hafnium at different temperatures in water and in steam. Table 4-5 (from Reference 4) gives irradiated corrosion data from the Shippingport Reactor.

While almost no quantitative data is available on hafnium under boiling water reactor conditions, it is expected that corrosion in a boiling water environment, even if significantly greater than under pressurized water raactor conditions, will be minimal, and fully acceptable. Corrosion of the hafnium rods to be used in the test program is expected to be low, particularly in view of the low average nitrogen content (26 ppm). However, one of the objec-tives of the test program is to quantitatively determine the actual corrosion rate under boiling water reactor operating conditions.

4.2 Radiation Growth Detailed radiation growth data for hafnium is not available except in nuclear naval reactor programs. The radiation growth of hafnium is expected to be less than that of Zircaloy. One of the objectives of the test program is to quantitatively determine the radiation growth of the hafnium absorber rods under boiling water reactor operating conditions.

4.3 Nuclear Characteristics Data on the nuclear characteristics of hafnium are available, includ-ing cross-section data. Hafnium cross-section data will be available in the ENDF/B-V library system maintained by the Brookhaven Naticnal 4-2

NED0-24231 Laboratory. Formal documentation of the ENDF/B-V data, which update the previous ENDF/B-IV data (Reference 5), will not be released until work on all of the libraries is completed. However, General Electric has the incomplete ENDF/8-V libraries, which include the completed cross-section data on the hafnium isotopes.

Using the hafnium data, the reactivity worth of an unirradiated test control blade was determined to be equivalent to a standard unirradiated B4C control blade. Since the test control blades are designed to the same end of life criteria as the standard B C 4

control blade, there will be no change in the shutdown margin calculations, scram reactivity, and transient analyses during the test program. Evaluations have been performed to assess the impact of control blade depletion on ACPR. These evaluations demonstrate that over the control blade lifetime limit of 10% change in relative control blade worth, the aCPR results remain unchanged. Also, since the worth of the test control blades are equivalent to the worth of the standard 84 C control blade, the control rod drop accident and rod withdrawal error evaluations are not impacted.

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NED0-24231 TABLE 4-1 CHEMICAL ANALYSIS OF HAFNIUM

% PPM Hf Zr W Fe Si Cu Ti Al N U 2

Bal. 1.1 100 190 35 10 8 25 - -

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(95.3) (4.5) (150) (500) -

(100) (100) (100) (100) (200) min max max max max max max max max

• Wet chemical analysis

() Values in parentheses represent the military specifications for reactor grade hafnium 4-4

NEDO ' '?31 TABLE 4-2 CORROSION OF HAFNIUM (REFERENCE 1)

DAYS -- Mg/Dm No. 7 14 28 42 46 70 84 98 112 140 163 168 191 196 540 F 1 1. 4 2.0 1.7 1. 7 2.0 2.6 -

2.3 2.8 2.3 2.0 2.8 Water 5 1.1 1. 7 2. 3 2. 3 1. 7 2.3 -

2.6 2. 3 2.3 2. 3 2. 8 9 0.9 1. 4 1.7 1. 4 1. 7 1.1 -

2.3 2.0 2.3 1. 7 2.8 11 0.6 1.4 1. 4 1.1 1.2 0.9 -

1. 7 0.9 1.1 1.1 2.0 Avg. 1.0 1.6 1. 8 1. 6 1. 6 1.7 -

2.2 2.0 2.0 1. 8 2.6 680 F 2 2.8 4.0 4.8 4.8 5.1 5.7 -

6.8 7.7 6.5 6.8 8.3 Water 6 2. 3 3.7 4.6 4.9 4.6 5.4 -

5. 7 6.8 6.0 6. 8 8.0 7 2. 8 3.4 4.3 4.9 4.8 5.4 -

6.3 7.4 6.3 6.8 8.0 12 2.0 3.4 4.3 4.0 4.3 4.6 -

5.4 6.8 6.0 6.0 7.4 Avg. 2. 5 3.6 4.5 4.7 4.7 5.3 -

6.1 7.2 6.2 6.6 7.9 750 F 3 4.5 7. 7 13.1 14.5 17.4 20.8 26.5 30.8 Steam 4 4.8 8.5 12.8 14.2 16.4 19.8 25.5 29.8 8 3. 7 7.7 12.8 13.9 16.7 20.1 26.1 30.9 10 3.1 7.1 11.4 11.9 14.5 17.9 24.7 29.0 Avg. 4.0 7.8 12.5 13.6 16.3 19.7 25.7 30.1 4-5

NED0-24231 TABLE 4-3 CORROSIO*! 0F HAFNIUM * (REFERENCE 2)

(Weight Increase in Mg/Dm )

TIME (DAYS) 600 F WATER 680 F WATER 750 F STEAM HF ZR-2 HF ZR-2 HF ZR-2 44 5 14 6-9 27 4-7 48 195 6 21 7-9 65 7 210 234 -

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NED0-24231 20 g g j g j;;

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NED0-24231

5. SAFETY EVALUATION The results of the safety analysis presented in the Peach Bottom 2, Reload 4, licensing amendment submittal (Reference 6) are unchanged by the use of the test control blade in the Reload 4 core during Cycle 5 operation.

As previously mentioned in Section 4.3, the worth of test control blades are equivalent to the original worth of the standard 8 C4 blades being replaced. Therefore, the control rod drop accident, and the rod withdrawal error event results previously reported for the Reload 4 core (Reference 6) are unchanged with the test control blades in the Reload 4 core.

Hafnium is heavier than 8 4C. This results in a 16 lbm. increase in the total test control blade weight. The effect of this increased weight compared to a standard 4B C blade is a slight increase in scram time.

The calculated increase in scram time of the test control rod for 5%,

20%, 50%, and 90% insertion is given in Table 5-1. The effect of the slight increase in the scram times for two test control blades on the core average scram times employed in the Reload 4 licensing amendment submittal (Reference 6) is negligible. Evaluations demonstrate that the critical power ratio (ACPR) results for all of the abnormal operational transients reported in the Reload 4 licensing amendment submittal are unchanged.

The proposed test program could be implemented under the provisions of 10CFR50.59, except that a change to the Technical Specifications is required. This is discussed in Section 6. However, the other require-ments of 10CFR50.59 are satisfied by the proposed test program, as discussed below.

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NE00-24231 The first requirement of 10CFR50.59 is that a proposed change in a facility, or the operating procedures, shall not increase the proba-bility of occurrence or the consequences of an accident or malfunction of equipment important to safety. The proposed test program meets this requirement, since it has been shown above that the Reload 4 safety analysis (Reference 6) is not affected by operation with the test control blades.

The second requirement of 10CFR50.59 is that the possibility of an accident or a malfunction of a different type than any evaluated pre-viously in the safety analysis report shall not be created. This requirement is also satisfied for the proposed test program, since the test control blade mechanical design is the same as the standard 8 C blade and the 4

nuclear design is equivalent.

The third requirement or 10CFR50.59 is that the margin of safety as defined in the basis for any technical specification is not reduced.

This requirement is also satisfied. The operating limits defined in the Reload 4 licensing amendment submittal (Reference 6) are unaffected by the use of the test control blade. The shutdown margin and control rod drop accident results are also unaffected. Therefore, the margins of safety as defined in the basis for the Technical Specifications are not reduced by the test control blades.

However, changes can only be implemented under 10CFR50.59 if, in addition to satisfying the above requirements, no changes to the Technical Specifications are required. The proposed test program does require a change to the Peach Bottom 2 Technical Specifications, as discussed in Section 6.

The effect of the hafnium test control blades on reload analyses is generic in nature and the foregoing safety evaluation applies equally to operation with the test blades during the test program period (i.e.,

Cycle 5 and 6).

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NE00-24231 TABLE 5-1 TEST CONTROL BLADE SCRAM TIME INCREASE SCRAM TIME

% INSERTION INCREASE (SEC) 5 0.005 20 0.009 50 0.018 90 0.03 5-3

NED0-24231

6. PROPOSED TECHNICAL SPECIFICATION CHANGE To conduct the proposed test program a temporary change must be made in Section 5.2 of the Peach Bottom 2 Technical Specifications.

Specification 5.2.B currently limits the control material to boron carbide powder. The specification should be modified to read as follows:

The reactor core shall contain 185 cruciform-shaped control rods.

The control material shall be boron carbide powder (8 C)4 compacted to approximately 70% of the theoretical density or hafnium metal.

It is proposed that this specification be changed, as showr, to permit operation with two control blades (the test control blades) containing up to 12 solid hafnium absorber rods each in place of a corresponding number of B4 C absorber rods. The test control blades will replace two standard control blades in the core.

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NEDO-24231

7. REFERENCES

1) 8.E.

Dearing and A.M. Andrako,

" Corrosion Behavior of Hafnium in Aqueous Media as a Function of Temperature", KAPL-M-BED-3 August, 1961.

2) 0.E. Thomas and E.T. Hayes, "The Metallurgy of Hafnium", Naval Reactors, Division of Reactor Development, AEC.

3) "A Compilation of the Properties of Hafnium", Bettis Atomic Power Laboratory, WAPD-TM-528.

4) G.J. Salvaggio, " Hafnium Control Rod Behavior in the Shippingport Pressurized Water Reactor", Nuclear Applications, Vol. 5, July,1968.

5) " Data Formats and Procedures for the Evaluated Nuclear Data File, ENDF", BNL-NCS-50496 (ENDF 102), TID-4500, October 1974.

6) Supplemental Reload Licensing Submittal For Peach Bottom Atomic Power Station Unit 2 Reload No. 4, NED0-24237, January 1980.

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