ML20039G329
| ML20039G329 | |
| Person / Time | |
|---|---|
| Site: | Bellefonte  |
| Issue date: | 03/26/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201180062 | |
| Download: ML20039G329 (11) | |
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w UNITED STATES
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't ATOMIC ENERGY COMMISSION EM WASHINGTON. D.C. 20545
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- "n Docket Nos. 50-438 and 50-439 2 2 s 574 Y. A. Moore, Assistant Director for Light Water Reactors Group 2 L NUCLEAR DESIGN & FUEL MECHANICAL DESIGN SER FOR BELLEFONTE UNITS 1 & 2 Plant Rame:
Bellefente Units 1 5 2 Licensing Stage:
CP Docket Nos.:
50-438 &'50-439 Responsible Branch LWR 2-3 and Project Manager:
D; Davis Requested Completion Date:
March 29,1974 Description of Review:
SER Input Attached is the Core Perfortnance Branch's writeup f r the Nuclear Design and Fuel Mechanical Design sections of the Bellefonte 1 & 2 plants.
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,j Victor Stello, Jr., Assistant Director for Reactor S'afety
, Directorate of Licensing
Attachment:
SER Input cc:
S. Hanauer J. Hendrie A. Giambusso W. Mcdonald A. Schwencer D. Ross D. Davis S. Varga E. Bailey L. Kopp F. Coffman i
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r201180062 010403 ND 80-S15 PDR t
t INPUT TO S.E.R. ON BELLEFONTE 4.2.1.1.
Fuel Mechanical Design The proposed Bellefonte reactor fuel elements to be provided by Babcock & Wilcox, will consist of Zircaloy-clad uranium dioxide fuel pellets. The fuel rod mechanical design is identical to that currently approved for use in Surry Units 3 & 4, with the exception of those items listed in table 4.2.1-1.
All Bellefonte design items listed exhibit larger engineering safety margins compared to the approved Surry Units 3 & 4 design.
All fuel rods will be internally prepressurized with helium during final welding to minimize cladding compr sive stresses during service. The level of prepressurization is designed to preclude any cladding tensile stresses throughout operations due to total internal pressure.
The staff assumes that densification of uranium dioxide fuel pellets may occur during irradiation in power reactors. The initial density of the fuel pellets and the size, shape, and distribution of pores within the fuel pellet influence the densification phenomenon.
The effect of densification on the fuel rod will increase the stored energy, increase the linear thermal output, increase the probability for local power spikes, and decrease the thermal conductance.
The primary effects of densification on the fuel rod mechanical design are manifested in calculations of time-to-collapse of the cladding and fuel-cladding gap conductance. Time-to-collapse calculations-predict the time required for unsupported cladding to become dimensionally unstable and to flatten into an axial gap caused by fuel pellet densification. Gap conductance calculations predict the decrease in l
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TABLE 4.2.1-1 FUEL MECHANICAL DESIGN COMPARISON BETWEEN BELLEFONTE & SURRY UNITS 384
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Assembly Bellefonte Surry Rod Array 17 x 17 15 x 15 Fuel Rods 264~
208 Rod-Rod Pitch
.501 inches
.568 inches Guide Tubes 24 16
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Fuel Rods Outside Diameter.
.379 inches
.430 inches Wall Thickness
.0235 inches
.0265 inches Average Specific Power 5.4 kw/ft 7.1 kw/ft 3.
Fuel Pellets Diameter
.324 inches
.370 inches 94%
91%
% Theoretical Density
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Stack Height
-143 inches 144 inches Diametral Gap
.008 inches
.007 inches a
Max. Temp 9 100% Power 3760 F 4410'F D
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thermal conductance due to opening of the fuel-clad radial gap.
On January 16, 1973 Babcock & Wilcox filed BAW-10054 entitled,
" Fuel Densification Report." This report is applicable to all B&W reactors beginning with Oconee Unit 1 and includes both Surry Units 3 & 4 and Bellefonte. The Staff's review and acceptance with
, modifications of the B&W fuel densification model was presented in its report " Technical' Report on Densification of Babcock & Wilcox Reactor Fuels," dated July 6.1973. This model also applies to Bellefonte.
The fabrication of the Bellefonte fuel is t planned until late 1978. Thus, it is quite likely that the as-manufactured fuel will reflect significant improvements in design and manufacturing processes. The staff will remain cognizant of any B&W fuel design and manufacturing process changes in its continuing review of both standard and specific designs.
On th'e basis of our review of the current analytical models and their confirmatory test results we have concluded that the Bellefonte 17xl7 fuel mechanical design provides for additional engineering safety margins compared to those provided in the approved design for Surry Units 3 & 4.~
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REFERENCES FOR 4.2 4
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PSAR on Surry Units 3 & 4, Docket No. 50-434.
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PSAR on Bel,lefonte, Docket No. 50-438.
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BAW-10054. Topical Report (Proprietary) " Fuel Densification Report" January 1973.
4.
" Technical Report on Densification of Babcock & Wilcox Reactor Fuels."
U.S. A.E.C. Regulatory Staff, July 6,1973.
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BELLEFONTE UNITS 1 & 2
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l 4.3 NUCLEAR DESIGN Nuclear Analysis Our review of the nuclear design of the Bellefonte 1 & 2 Nuclear Power Station was based on the information provided by the ap licant in Preliminary Safety Analysis Reports and revi-sions thereto, discussions with the applicant, and the results of independent calculations performed for us by the Brookhaven.
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National Laboratory.
The proposed nuclear design of the Bellefogte 1 and 2 Nuclear Power Stations is the same as that reviewed and approved for the Surry Units 3 and 4 reactors, except that Bellefonte will use fuel assemblies with a 17 x 17 fuel element array while most earlier B & W reactors were designed for a 15 x 15 fuel element array.
The information available'from the applicants concerning the 17 x 17 fuel assembly design indicates that the change in fuel design will improve overall reactor safety by lowering the average and maximum linear heat generation rate.
The original Bellefonte 15 x 15 proposal, for example, was de-signed to operate at 3413 MWt with an average linear power density of 6.49 KW/ft as compared to the design heat output of 3600 MWt for the new 17 x 17 Bellefonte fuel assembly design with an average linear power density of 5.43 KW/ft.
The applicant has described the computer programs 9
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2 We have concluded that the information presented adequately demonstrates the ability of these analyses to predict reactivity and the physics' characteristics of the Bellefonte 1 & 2 Units.
Power Distribution Detailed three-dimensional power distribution measurements havebeenperformedattheBabcock&WilcoxCrigicalExperiments Laboratory.
The results of the applicant's calculations using PD007, a three-dimensional computer program, agree quite well with the measured power distribution.
PDQ07 as used by B & W incorporates a thermal feedback in obtaining radial and axial power distributions for operations involving (1) changes in con-trol rod positions,(2) various xenon stability and control condi-tions, and (3) various reactivity coefficients.
c Control Requirements To allow for changes of reactivity due to reactor heatup,
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operating conditions. fuel burnup and fission product buildup, a significant amount of excess reactivity is built into the core.
The applicant has provided substantial information relating to-core reactivity balances 'or'the first cycle and has shown that f
means have been incorporated into the design to control excess reac-s O
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- tivity at all times.
This is done through the use of soluble boron in the reactor coolant, movable control rods, and fixcd B C 4
burnable poison rod assemblies (BPRA).
The BPRA's are used 4
rather than increased scluble poison to prevent ths BOL moderator
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temperature coefficient from becoming more positive.
The applicant has shown that sufficient CRA worth is available to shut 1
down the reactor with at least a 1% ok/k subcritical margin in the hot condition at any time during the life cycle with the most reactive CRA stuck in the fully withdrawn position.
Equip-ment is also provided to add soluble boron to the reactor cool-1 ant to ensure a similar shutdown capability when the reactor is cooled to ambient temperatures.
We assume that the control requirements for cycles beyond the first will also be established and presented as the design becomes more hardened for the OL stage.
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On the basis of our review, we have concluded that the applicant's assessment of reactivity control requirements over the first core cycle is suitably conservative, and that adequate negative worth has been provided by the control rods, the soluble boron system, and the burnable poison rod assemblies to assure shutdown capability., Reactivity control requirements will be reviewed for additional cycles as this information becomes avail-able.
Stabilit;c The basic instrumentation for monitoring the nuclear power 4
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level and distribution in the Bellefonte 1 & 2 reactors is the same in principl as for all PWR plants recently licensed for operation.
Primary reliance is placed on four axially split, out-of-core detectors.
Also, 62 assemblies of self powered incore neutron detectors are available for incore mapping.
Each assembly can measure local neutron flux at seven eleva-tions in the core.
Test results showing that these incore detectors have a rated lifetime in excess of 5 years and a precision of + 5% in determinin~g relative power distribution are presented in B&W Topical Report 10001, "Incore Instrumenta-tion Test Program," (August 1969).
We have concluded that the out-of-core detectors are adequate for detecting power maldistributions originating from axial xenon instability and misplaced control rods if a power distri-bution mapping capability is provided by the incore detectors to calibrate tho'out-of-core detectors periodically and to investigate any power distribution anomalies detected by the c
out-of-core detectors.
We have reviewed the applicant's analyses of xenon-induced oscillations which are reported in three D&W Topical Reports BAW-10010, Part 1, " Stability Margin for Xenon Oscillations Model Analysis," August 1969, DAW-10010, Part 2,
" Stability Margin for Xenon Oscillation - One Dimensional Digital Analysis,"
February 1970, and DAW-10010, Part 3, " Stability Margin for Xenon Oscillations - Two and Three Dimensiona'l Analysis,"
April 1970.
Those analyses indicated that, while azimuthal and m
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radial xenon oscillations will not be divergent, axial xenon oscillations could be divergent at the beginning of the fuel cycle.
The analysis further indicated that axial xenon oscil-4 lations, which are slow changes taking place over,several hours,i.
can be controlled by having the reactor _ operator change the Position of the eight part-length APSR's.
Results from induced axial xenon oscillation tests during the initial startup of the Oconec Unit 1 have shown that good agreement exists between predicted and measured results and that the APSR's were effec-tive in damping the axial oscillations.
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REFERENCES FOR 4.3 1.
BAW-10001, Topical Report, "Incore Instrumentation Test Program,"
August 1969.
2.
BAW-10010, Part 1 Topical Report, " Stability Margin for Xenon Oscillations Model Analysis," August 1969.
3.
BAW-10010, Part 2, Topical Report, " Stability Margin for Xenon Oscillations - One Dimensional Digital Analysis," February 1970.
4.
BAW-10010. Part 3. Topical Report, " Stability Margin for Xenon Oscillations - Two and Three Dimensional Analysis," April 1970.
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