ML19340A899
| ML19340A899 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/23/1965 |
| From: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19340A898 | List: |
| References | |
| NUDOCS 8009050630 | |
| Download: ML19340A899 (5) | |
Text
'O RAZARDS ANALYSIS BY THE RESEARCH AND POWER REACT 0k SAFETY BRANCH DIVISION OF REACTOR LICENSING IN THE MATTER OF COHMONWEALTH EDISON CGIPANY DOCKET NO. 50 _l0 PROPOSED CHANGE NO. 10 - TYPE III-F FUEL RELOAD Introduction Commonwealth Edison Company has requested, by application dated December 24, 1964, authorization to load up to 104 Type III-F fuel assemblies in the Dresden reactor.
Supplementary information was submitted by addendum dated March 2, 1965.
Discussion The proposed Type III-F fuel is quite similar to the Type III fuel presently in the reactor. Table I presents a comparison of the characteristics of these two types of elements.
The fuel loading proposed is a scattered configuration similar to that used in loading the Type III fuel elements.
Sketches of possible loading con-figurations are included in the application. The calculated lattice reactivity,
,both uncontrolled and controlled, for Type III-F fuel is higher than for any of the fuel elements previously authorized.
However, calculations performed using the maximum reactivity Type III-F and Type III loading configuration indicate that the resulting reactivity level is expected to be lower than that of the initial Dresden core.
In addition, the control rod worth for Type III-F fuel is greater in the cold condition than for any of the other Dresden fuel elements.
Type III-F fuel utilizes gadolinia as the burnable poison. Calculatior.s indicate tha t the reactivity increase due to gadolinia depletion is less than the reactivity lost due to burnup of the fuel at all times in core life.
Thus, maximum lattice reactivity occurs at the beginning of life. The cold shutdown margin at this time with one control rod s tuck full-out is 1.77. Zi K.
If the gadolinia were depleted instantaneously, the reactor would be 0.5% tiK shutdown under these conditions.
Critical proof tes t measurements performed at Vallecitos confirm the ability to predict the initial reactivity ef fects of gadolinia.
l.
The temperature and void coef ficients of reactivity for the full core mixed 0
lattice have been calculated to be negative at operating temperatures (546 F),
as was the case with previous cores.
The temperature coefficient, which is positive at room temperature, is expected to become negative at about 420 F.
80090'O h 6 5
i TABLE 1.
CG(PARISON OF THE CHARACTERISTICS OF DRESDEN TYPE III AND TYPE III-F FUEL ASSEMBLIES Type III Type III-F Type III-F (pellet)
(powder)
Clodding Material Zr-2 Zr-2 Zr-2 0.D., inches 0.555 0.5625 0.5625 Wall thickness, inches 0.035*
0.035 0.035 Configuration 6x6 6x6 6x6 Regular Rods Number Required
'31 29 28
- 7. Fuel Composition 99.857. UO2 1007. UO2 1007. UO2 0.157. Er2 03
- 7. UO2 Enrichment 1.83 2.34 2.34 Pellet Diameter, inches 0.478 0.482 0.4925 Special Corner Rods Number Required 5
6 6
- 7. Fuel Compos ition 99.857. UO2 1007. UO 1007. UO 2
2 0.157. Er2 03
- 7. UO2 Enrichment 1.83 1.77 1.77 Pellet Diameter, inches 0.438 0.482 0.4925 Spacer Rod - Pellet Number Required 1
same as Type III-F regular pellet fuel rod Poison Rod - Pellet Number required 1
1
- 7. Compos ition
- 95. 37. A12 03
- 95. 37. A12 03 4.77. Gd2 03 4.77. Gd2 03 Pellet Diameter, inches 0.465 0.465 OWall thickness for the corner rods is 0.055 inches
. Typical power distributions relating to Type III-F fuel are essentially the same as those for the Type III ' fuel. Therefore, operation of the reactor with Type III-F fuel will not appreciably change the thermal performance of the reactor and will permit the reactor to operate below.the maximum heat flux and above minimum burr out ratio limits.
Evaluation of the type III-F pellet fuel heat flux limit at 125% of rated power, 19.4 kw/ft, yields a center fuel temperature just below the melting temperature of UO2 using the thermal conductivity correlation of Lyons et al, ANS Transactions, June 1963.
This is equivalent to a value of 85.5 watts /cm for the integral of k dt from 0 F to the center fuel temperature.
Thus, we believe the criteria of no center fuel melting will be satisfied.
Evaluation of the Type III-F powder fuel heat flux limit at 125% of rated power, 19.4 kw/ft, also yields a center fuel temperature jus t below the melting temperature of UO2 using the thermal conductivity curve presented in the addendum dated March 2,.1965.
This new thermal e onductivity curve was used since Type III-F powder fuel is vibratory compacted to only 85% of the theoretical density of UO.
In our opinion, the ten Type III-F powder 2
fuel elements authorized by this change should be considered developmental until more experimental data is obtained regarding the thermal conductivity and gap heat transfer coefficient for powder fuel.
Hazards Evaluation Type III-F fuel is sufficiently similar to Type III fuel, that no significant differences will result in the potential consequences of the accidents pre-viously considered for the Type III fuel.
One exception to the above statement is the refueling accident.
The reacti-vity worth of a Type III-F fuel element is greater than previous Dresden fuel elements. An analysis of lowering a fuel element worth of 1.9% pg K at the maximum design hoist rate into a near-critical core (10-8 times rated power) has been performed.
Commonwealth a.-;umed the following violations of pro-cedures or malfunctions:
1.
Two control rods, adjacent to the vacant fuel element position, were fully withdrawn.
2.
The operator failed to observe instrumentation.
3.
The period scram failed.
Results indicate the maximum fuel temperature would be 5080 F, the melting temperature of UO2 and a small amount of burnout might occur.
The excursion would be turned around by the Doppler ef fect and terminated by the high flux (125% rated power) scram.
Since the containment sphere would be in a con-fined status during refueling operations, the consequenaes of this accident are within acceptable limits.
. All potential criticality problems were also reviewed by the applicant in the addendum dated March 2, 1965. The cases involving criticality con-siderations of the fuel handling baskets, the fuel storage racks (flooded and dry), spent fuel storage, and dropping of a fuel element between s torage racks were found to yield a Kef f _ below 0.90.
Based on this evaluation, we believe the present fuel handling and storage facilities are adequate.
Technical Specifications f
To provide authorization of Proposed Change No. 10, the Technical Specifica-i tions of License No. DPR-2 should be amended as follows:
1.
Section B.2, page 1, in its entirety as follows:
Maximum core diameter (circumscribed circle) 129 in.
Maximum active fuel length - cold 112 in.
Maximum number of fuel assemblies by types:
Type I 263 Type II 102 Type III 192 Type III-F (s tandard) 88 Type III-F (removable experimental segmented poison rod) 6 Type III-F (powder) 10 Type PF-10 and PF-11 (one each) 2 Type SA-1 1
Maximum total number of fuel assemblies 488 The fuel assemblies may be located in any position of the reactor, provided that fuel assemblies Type PF-10 and PF-11 are each separated from the other by at least four Type I, T;pe II, Type III, or Type III-F fuel assemblies.
I The reactor may be operated at any power up to and including rated power with any configuration of the various types of fuel assemblies ins talled, provided the maximum number -and location are within the limits specified above.
2.
Section B.3, page 2, second paragraph, as folloss:
The nominal fuel pellet density averaged over a fuel segment is 94.3% of theoretical for all fuel assemblies except ten Type III-F assemblies which contain vibratory compacted UO2 powder with a nominal density of 85% of theoretical.
I 3.
The tabulation in Section D.3, page 12, is amended to. reed as follows:
Fuel Type 1 350,000 Fuel Type II 410,000 Fuel Type III 360,000
. Fuel Type III-F 360,000 Fuel Type PF-10 and PF-11 510,000 Fuel Type SA-1 425,000 4.
Table II (revised February 19, 1964) should be replaced by Table II (revised November 13, 1964) as set forth in Commonwealth Edison's application dated December 24, 1964.
Conclusion Based upon our review of the information submitted, we have concluded that operation of the reactor in accordance with the proposed change does not involve significant hazards considerations not described or implicit in the hazards summary report and that there is reasonable assurance that the health and safety of the public will not be endangered.
Accordingly, we believe that the Technical Specifications of License No.
DPR-2 should be revised as ' licated above.
J
/gl h kb )/lfic) l * /
.,.s,,
Roger S. Boyd, Chief
[
Research & Power Reactor Safety Eranch Division of Reactor Licensing Date:
MAR 2 3 555 i
.