ML20024G531

Safety Evaluation Accepting Sys Concept for Facility Prompt Relief Trip Which Will Provide Core Protection

ML20024G531
Person / Time
Site:	Monticello
Issue date:	04/08/1974
From:	James Shea, Ziemann D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20024G510	List: ML20024G509 ML20024G531
References
NUDOCS 9102120609
Download: ML20024G531 (7)

Text

A UNITED STATES ATOMIC ENERCY C010!ISSIOU,

_S.A.FLTY EVALUATION BY THE DIRECTORATE OF LICDiSINO NORTHERN STATES POWER COMPANY

}DNTICELLO NUCLEAR GENERATI!U_PIRg 8, x 8 FUEL ASSI:MBLIES I

IliTRODUCTIOg i

horthern Staten Power Company (NSF) has submitted (12) a description and safety analysis of 8 x 8 (64 rods) roloed fuel assemblica and a segmented test rod (STR) assembly (3.4) which will replace 116 7 xx7 (49 rods) fuel assemblies (full core contains 484 assemblies) during the spring 1974 refueling outage. All of the remaining 44 tenporary control curtains are to be removed during the refueling outage. Fuel assemblies containing gadolinia havo been authorized (8) by the Directorate of Licensing for use in 20 reload 1 fuel assemblica that were inserted (6) into the Monticello core during the spring 1973 refueling outage. Therefore, the use of gadolinia in the 8 x 8 fuel assemblies is not a new feature. The outside dimensions of the fuel assembly l

remain unchanged.

t ne principal differences between the 7 x 7 and 8 x 8 fuel assemblies, in addition to the greater number of fuel rods in the 8 x 8 fuel assemblias, are 1.

The average uranium enrichment which is 2.62% is higher for the 8 x 8 fuol assemblies.

j 2.

The 8 x 8 fuel rods are smaller in diameter than the 7 x 7, but the clad thickness is greater; i.e., 0.034 inch va 0.032 inch clad thickness.

l 3.

The 8 x 8 assembly containe an internal asymmetric water-filled i

spacer capture rod.

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The 8 x 8 fuel rod array has 137. nore heat transfer surface than fuel assemblica previously used in Monticollo.

5.

The unter to fuel volume ratio is higher for the 8 x 8 reload i

fuel assechlics than the 7 x 7 fuel assemblies.

6.

The B x 8 fuel assemb.

,tains only 94': of the total uranium in the original 7 x 7.

'semblies, but. 10.5% raore U 235.

t:SP also subnitted(3.4) proprietary information related to the design features of an 8 x 8 segraented test rod (STR) assembly for the Monticello Le fuel nosembly, according to the ::SP cubmittal, satisfica thu core.

saw design and dauage criteria as thu fuel assemblies that have been used in Monticello and the reload 2 fuel 2ssemblies that are to be inserted in the Mouticello core during April 1974.

[yALUATION Lince there are 63 UO2 rods in the 8 x 8 fuel assembly compared with 49 in the 7 x 7 fuel assecLly, the avera;;e fuel rod linear heat generation rate in an average 8 x 8 fuel assembly at rated core power level (1670 Wt) vill be reduced from 5.8 kW/ft to 4.57 kW/ft. We combination of reduced power per rod and increased heat transfer surface will i

result in lower fuel rod temperatures which may result in some improve-ment in fuel rod integrity. Peak fuel clad temperatures following the design basic loss-of-coolant accident (LOCA) for the entire spectrum of breaks calculated in accordance witg AEC Interim Acceptance Criteria I

(LAC) with allowance for densification are acceptably below the 2300*F limit.

Metal-water reactions following a lhCA are similarly calculated to bo less than 0.21, well within the AEC-IAC limits. Although the average fuel temperatures vill be lower because of more surface and more fuel rods, the capacity to etere heat is reduced about 6% due to the 6% reduction in the amount of fuel. Eis effect, however, is i

partially offset by the 17% increase in the acotat of zircaloy clad.

I The combined effect on fuel heatup during adiabatic conditions has been considered in the calculations. As noted, peak clad temperatures and metal-water reaction following the postulated design basia LOCA remain within the LAC 11:.its with Isrger margins than calculated using the 4

same methods for the 7 x 7 fuel assemblies, f

With regard to the postulated design basin refueling accident, the

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fission product inventory of interest is about the same for the D x 8 fuel as the 7 x 7 fuel.

The lower 8 8 fuel temperatures could reduce the amount of noble gas and halogen activity that collects in the fuel I

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rod plenur.m and is available for release if the cladding is damaged, but the amunt would be negligibly ar.all.

If the noble gas and halogens activity in the fuel plonuca is conservatively assumed to be the saue for the 8 x 8 and 7 x 7 fuel casamblies, the fission products released i

to the water during the design basis refueling accident are unchanged l

l and the fission product release to the environs, therefore, remains j

unchanged. Since the S x 8 fuel rod cladding is thickar and because there are more rods to absorb the fall shock, a lower percentage of the

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6 x 8 fuel assembly rods may be damaged and the release would be i

correspondingly lower than that assumed. We have concluded that the l

consequences of the postulated design basis refueling accident will not change significantly because of the change to 8 x 8 fuel assemblies.

We rolcase of radioactive materials to the environment as a result of

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a postulated main steam line break outcide of containment is governed by the mass flow rate of steam from both open ends of the break.

The i

steam flow rates are limited by critical flow conditions at steam line flow restrictors. Since the source of radioactivity, steam flow rates following the assumed pipo rupture, and valve closure tinea are unaffected by the change to 8 x 8 fuel assemblies, we have coacluded that the i

consequences of the postulated desi n basis steam line break accident F

are unchanged from those previously evaluat ed.

J The postulated design basis control rod drop accident is b sed upon a maximum rod worth of 1.3% delta k/k causing fuel damage to the j

extant that (1) peak fuel enthalpy during the accident is no mre than 280 cals/gm, the conservativeJy assumed value for prompt energy deposition s

i into water, and (2) all fuel abova 170 cals/gm causes clad failure. For these asstased accident conditions, the release of fission p:oducts into the primary coolant would be about the same. We have concluded that the release of radioactive material to the primary coolant due to a i

postulated rod drop accident is about the same for 8 x 8 fuel as for l

7 x 7 fuel assemblies although comparison shows that the maxinum rod j

worth is further below 1.3% delta k/k limiting condition and the con-i sequences of an assuned rod drop accident with 8 x 8 fuel assemblies j

instead of 7 x 7 fuel assemblies, tharefore, would be reduced.

l We increased enrichment of the 8 x 8 fuel assembly is not a result of changing fuel. Such an increase would have been required if 7 x 7 fuel assemblies were used as replacesient fuel. We increase in enricisment 1

is required and had been pisaned to compensate for the change in core characteristics to an equilibrium core rather than a startup core; i.e., core constituted with partially depleted fuc1 rather than 484 new fuel assemblies and poison curtains. We have reviewed (5) the nuclear design of the 8 x 8 fuel assemblies, including the increased enrich-ment which is not unique for the 8 x 8 fuel assemblies, and have found the design to be consistent with previously approved fuel and, there-fui a En.my L 5IJ185 <

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- 4 The increased clad thicknano with smaller rod diamoter and reduced power per rod vill provide greater resistance to clad failure. Never-theless, :iSP has agreed (7) to participate with General Electric in a surveillance program to monitor the performance of a precharacterized 8 x 8 fuol bundle in the Monticello reactor beginning with cycle 3.

We have concluded (5) that the cladding integrity of 8 x 8 fucl is acceptable and the effect of the unheated rod near the center of the l

bundle will not be significant.

The beneficial effects of additional heat transfer curface as well as the increased coolant flow resistance due to increased surface drag havo been included in the thermal hydraulic evaluation. We have notod(5) that in general the 8 x 8 fuel has greatur thermal margins to design limite than the 7 x 7 fuel, indicating that improved heat transfer off sets the slight decrease in 8 x 8 anscubly flow.

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ne increase in water to fuel volu:ao poses no particular problem.

Increancs of this magnitude can be and have been accomuodated. For examplo, the water to fuel ratio for reload 1 fuel inserted during

.t e spring 1973 plant outage (l) te 2.53 in contrast to the initial core f tw1 ratio of 2.47 - approximately the same magnitude of the change f rom 2.53 to 2.60 for the 8 x 8 fuel water to fuel volume. Uo have i

concluded (5) that the nuclear design of the 8 x 8 assembly is acceptabic, i

We STR asseably(3,4), in addition to satisfying existing design criteria and qt.ality assurance requirencnts and operating at a more restrictive

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linear heat generation level, will be removed from the core during each refueling outage to remove selected rods for destructive testing.

Care-ful inspection before returning the bundle to the core vill provide added assurance that the fuel perforcance is in accordance with design expectations.

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The proposed operating characteristics for the 8 x 8 fuel assemblier; i.e.,

• 13.4 kW/ft maximum L11GR (operating Limit), 45,000 mwd /Te maximin local exposure, and 4 6 years incore residence time.

are consistent with the design changes that have been identified and are indicative of continued improvement in quality assurance during fuel f abrica tion.

The changes in the 8 x 8 fuel assembly design, including f nel pellet preparation, are relatively small and can be acconcaodated by existing calculational methods.

The changes result from extensive performance analysis, inspection, end destructive examination of fuel j

rods that have beca irradiated in nucicar power plants. We ote that

' ~ ~*STR maxEum IJIGR 10.5 kW/ft l

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1 the trend toward smaller diameter and more rods per btaidle to improve i

i fuel rod integrity and reliability is supported by the reactor operating i

experience at Big Rock Point. It should be noted, however, that if i

l contrary to expectations the fuel integrity in reduced resulting in i

fission product release into the cools.nt, the existing technical specifi-j i

cation limits on radioactive releases into the primary coolant and i

effluant releases to the environe will continue to determine the extent of allowable fuel degradation and provide the same level of protection to the health and safety of the public.

i Coolant flow characteristics during normal and accident conditions are f

nearly the same since the flow cross sectional area for the 8 x 8 and 7 x 7 fuel annenblies is identical.

Iho small decrease in flow due to the increased heat transfer surface in the 8 x 8 rods is less than i

8*. and 10 accounted for in the calculations that show an MCHFR of 2.3 for the 8 x 8 fuel assembly compared with 2.03 for the 7 x 7 fuel assembly; i.e., the 8 x 8 assembly has a larger thermal margin.

g)NCLUSION Based on our evaluation of the differences between the 8 x 8 fuel i

f assemblies to be inserted during the April 1974 refueling outage and 1

the 7 x 7 fuel assemblies, we have concluded that design margins for i

safety or fuel damage limits are larger for the'8 x 8 fuel assemblies than the 7 x 7 fuel assemblies used up to the present time.-

the fusi should be more resistant to failure during normal and accident conditions j

according to the design calculations.

Inspection during each refueling i

outaSe of a precharapterir.ed reload 2 fuel assembly and the STR assembly l

plus-destructive tests of selected STR rods will provide timely infor-i mation to confiina design performance expectations or identify deviations from predicted performance. Changes to the Technical Specifications to j

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permit reactor operation above 1% power level with reload 2 fuel inserted will be made to reflect limits on'the 8 x 8 linear heat power generation l

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(13.4 W/f t), maximum average planar LHCR as a ftmetion of fuel bundle exposure, and other changes related to control rod scram time (3.5 I

seconds instead of i seconds). Plant modifications that are in progress will be completed before the Monticello nuclear plant returus to power in May 1974.

However, based on our evaluation of the 8 x 8 fuel assembly performance characteristics and the ' Technical Report on the General 6

Riectric Company 8 x 8 Fuel Assembly" prepared by the Ragulatory atsff(5),

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1 we havu concluded that the health and safety of the public will not i

be endangered by reactor operation with 8 x S fuel assemblies replacing 7 x 7 fuel assenblies in the manner described by NSP, 1he staff is evaluating the acceptability of a new system concept for I

Monticello, the Prompt Relief Trip (PRI) that has been proposed by SSP to provide core protection. However, this system is not required to protect the reactor until the end of cycle and even then onif if it is desired to extend the fuel cycle to the "all rods out" condition l

without prograroing power downward during this period prior to shut-down for refueling.

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Jares 3. Shea Operating Reactors.3 ranch #2 9

Directorate of Licensing l

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Dennis L. ".iemann, Otief Operating Reactors Branch #2 Directorate of. Licensing Date:

OPP 8 1974 I

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_REFERC4CES i

t 1.

USP "Second Ecload Subtittal dated Noverlor 19, 1973.

2.

j NSP nubmittal, " Supplemental Informtion to th- %,ticello Second l

Reload Submittal", dated February 8,1974.

l 3.

NSP oubmittal of LEDE-20179 -- tbnticello Seguented Test Rod Bundle

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Subraittal (Proprietary Information) - dated December 14, 1973.

l 4.

NSP subruittal, ' Change to NEDE-20179 lbuticello Segmented Test Rod Etnidle Submittal" (Proprietary Information) dated Jr.aary 15, 1974.

5.

AEC - Directorate of LicensinE letter (February 11, 1974) to MSP 1

regarding 8 x 3 reload fuel with the following encloourcs.

i 1.

Federal Register Notica regardinF authorization to operate j

lbnticello with 8 x 8 reload fuel, including a segmented j

test rod assembly.

I 2.

Technical Report on the Gonaral Electric 8 x 8 Fuel Assembly i

by AEC Directorate of Licensing dated February 5,1974.

6.

HSP aubnittal, Supplementary Information Regarding the First i

Monticello Reload", dated April 13, 1973.

7.

USP submittal, Special Survoillanco Program for S x 8 Fuel",

dated February 14, 1974,

i 8.

Directorate of Licensing authorization to operate lkinticello Nucicar Power Plant with reload 1 fuel asserablies in core dated March 5, 1973.

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