ML20198F907
| ML20198F907 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 08/05/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0987, CON-WNP-987 NUDOCS 8605290137 | |
| Download: ML20198F907 (6) | |
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Docket No. 50-450 V. A. Moore, Assistant Director for Light Water Reactors Group 2. L NUCLEAR DESIGN SER FOR WPPSS-1 Plant Name:
Washington Public Power Supply System, Unit 1 Licensing Stage:
CP Docket No.:
50-460 Responsible Branch:
LWR 2-3 and Project Manager:
T. Cox Technical Review Branch Involved:
Core Perfonnance Branch Requested Cocpletion Date:
August 5,1974 DescrtDtion of Review:
SER Input Enclosed is the Core Perfonnance Branch's writeup for the Fuels Section (4.2.1), Huclear Design Section (4.3) and the Control Rod Ejection Accident (15.1.18) of the WPPSS-1 plant. These are in acco. dance with the latest version of the Standard Review Plan.
Orfsfnni stang g v m.- e, Victor Stello,.n, Jr.,
ssistant Director for Reactor Safety Directorate of Licensing
Enclosure:
As State Above cc:
S. Hanauer Distribution; F. Schroeder Central File A. Giambusso CPB Reading W. Mcdonald L Reading A. Schwencer T. Cox D. F. Ross P. Check L. Kopp i, l'in p#2g;ggeg;;o E
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INPUT TO SER ON WPPSS-1 4.2.1 Fuel The proposed reactor fuel elements for WPPSS-1, provided by Babcock & Wilcox, will enploy Zircaloy-clad fuel rods containing uranium dioxide pellets.
The WPPSS-1 fuel mechanical design is identical with the design currently approved for use in Bellefonte Units 1 & 2 and Greenwood Units 2 & 3 as shown in Table 4.2.1-1.
All fuel rods will be internally prepressurized with helium during final welding to minimize cladding compressive 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 dinxide fuel pellets may occur during irradiation in power reactors.
The initial density of the fuel nallats end the si<e, thepe, er.d di:tributien cf pcrec within the fuel pellet influence the densification phenomenon. The effects 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 conductancc calculations. predict the decrease in thermal conductance due to opening of the fuel-clad radial gap.
The analytical models employed by the applicant to account for these effects are currently being modified and updated. The Staff has agreed in principle on the validity of the models.
The proposed testing program for the 17x17 fuel rod bundle has also been reviewed and the Staff is generally in agreenent with the planned studies.
Further discussion will continue on these items.
The fabrication of the WPPSS-1 fuel is not planned until 1979.
- Thus, it is quite likely that the as manufactured fuel, which will incorporate the results of the 17x17 bundle research program, will reflect signi-ficant improvenents in design and manufacturing processes.
The Staff will remain cognizant of any B&W fuel design and manufacturing process
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changes in its continuing review of both standard and specific designs.
On the basis of our review of the proposed analytical models and e
their confirmatory test re:;ults, the Staff has concluded that the WPPSS-1 17x17 fuel mechanical design provides for conservative engineering safety margins and thus, is approved.
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Table 4.2.1-1 Mechanical Design of 8ellefonte (Units 1 & 2),
Greenwood (Units 2 & 3) and WPPSS-1 e
Design Parameter Value
- 1. Rod array per assembly 17x17
- 2. Rods per assembly 264
- 3. Fuel weight (1bs) 233,844
- 4. Guide thimbles per assembly 24
- 5. Rod outside di,ameter (in.)
0.379
- 6. Clad wall thickness (in.)
0.0235
- 7. Retin of diameter-to-wall thickness 16.1
- 8. Pellet theoretical density 94%
- 9. Fuel stack height (in.)
143
- 10. Fuel-clad diametral gap (in.)
O.008 t
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, 4.3 Nuclear Design The review of the nuclear design of the Washington Public Power Supply System fluclear Project flo.1 (WPPSS-1) is based on the information provided by the applicant in the PSAR and revisions thereto, discussions with the applicant, and the results of independent calculations performed by the Staff or its consul-tants.
The applicant has described the computer programs and calcula-tional techniques used to predict the nuclear.characte.ristics of the reactor design and has provided examples to demonstrate the ability of these methods to predict experimental results.
The Staff concludes that the information presented adequately demonstrates the ability of these analyses to predict reactivity and the physics characteristics of the WPPSS-1.
To allow for changes of reactivity due to reactor heatup, changes in operating conditions, fuel burnup and fission product build-up, a significant amount of excess reactivity is built into the core. The applicant has previoed substantial information re-lating to core reactivity balances for the first cycle and i
has shot!n that means have been incorporated into the design tc control excess reactivity at all times. The applicant has
- hown that sufficient control rod worth is awilable tu siat down the reactor with at least a 1% ak/k suberitical margin in the hot condition at any time during the cycle with the most reactive control rod stuck in the fully withdrawn position.
- 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 system to assure shutdown capability.
Reactivity control requirements will be reviewed for additional cycles as this information becomes available.
15.1.18 Rod Ejection Accident The Staff has evaluated the applicant's analysis of the assumed red ejection accident and finds the assumptions, calculational te:hniques, and consequences acceptable.
Since the calculations re.;ulted in peak fuel enthalpies less than 280 cal /gm, prompt fut;l rupture with consequent rapid heat transfer to the coolant from finely dispersed molten U0 was assumed not to occur. The 7
pre.tsure surge was, therefore, calculated on the basis of con-ventional heat transfer from the fuel and resulted in a pressure increase below the ASME Emergency Condition stress lim'it for the i
maximum rod worths assumed. The Staff believes that the calcula-tions contain sufficient conservatism both in the initial assump-tions and in the analytical models to ensure that primary system integrity will be maintained.
REFERENCES
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1 l
l.
" Standard Format and Content of Safety Analysis Reports for Nuclear i
Power Plants," (Revision 1) USAEC, Oct.1972.
I 2.
Clark, R.H.,'and T.G. Pitts, " Physics Verification Experiments, Core I," BAW-TM-455, June 1966.
3.
Clark, R.H., " Physics Verification Experiments, Cores II and III,"
l BAW-TM-458, July 1966.
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4.
Hassan, H.A., and W. A. Wittkopf, "Three-Dimensional Power Distri-bution Analysis," BAW-10061. June 1973.
5.
Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident in PWR's," May 1974.
6.
Lellouche, G.S., " Rod Ejections at Full Power," BNL Memo, l
June 09,1971.
7.
Le11ouche, G.S., " Rod Ejections at Full Power, II," BNL Memo, j
June 30, 1971.
8.
Scott, J.F., " Calculation of Maximum ~ Fuel Enthalpy During Rod Ejection Accident," BAW-10081, Feb.1974.
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