ML20154S762
| ML20154S762 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/25/1986 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8604080085 | |
| Download: ML20154S762 (9) | |
Text
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Northern States Power Company 414 Nicollet Mati Minneapolis, Minnesota 55401 Telephone (612) 330-5500 March 25, 1986 Director Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PIANT Docket Nos 50-282 License Nos DPR-42 50-306 DPR-60 Additional Information to Support the License Amendment Request dated January'13, 1986 The attached information is being provided in response to NRC Staff questions concerning the January 13, 1986 License Amendment Request.
Please contact us if you have further questions concerning this submittal.
b Ym David Musolf Manager - Nuclear Support Services DMM/TMP/tp Attachment c:
Regicnal Administrator-III, NRC NRR Project Managcr, NRC Resident Inspector, NRC MPCA Attn: J W Ferman G Charnoff B604080005 060325 i
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a Prairie Island Nuclear Generating Plant Northern States Power Company Additional Information to Support the License Amendment Request dated January 13, 1986 Applicability of this License haendment Kaquest The changes proposed in this amendment are applicable to the operation of Unit 1 Cycle 11 and Unit 2 Cycle 10 and are a result of the changeover from Exxon fuel to Westinghouse fuel occurring on Cycle 11 of both units.
Althou,gh Westinghouse OFA fuel is not being utilized in Unit 2 Cycle 10, the proposed changes are applicable to the operation of Unit 2 Cycle 10 as discussed in the Exhibit A.
Unit 2 Cycle 11, and future cycles, will be evaluated per 10 CFR 50.59.
If changes to the Technical Specifications are required, a license amendment will be submitted.
K(z) Curve Changes This license amendment proposes the elimination of the third line segment of the K(z) curve, Figure TS.3.10-5.
K(z) is a normalized factor used to limit the power in the upper half of the coca. The existing curve consists of three segments. The first segment is flat and has a value of one.
The second segment has a slight negative slope.
The third line segment has a large negative slope. The third line segment is based on the small break LOCA analyses.
In the past, it was felt tbr.t power muet be restricted below the third line segment to ensure that 2200 F cladding temperature was not exceeded during a small break LOCA.
The analyses submitted in Exhibit F and G clearly demonstrate that operation with power distributions up to the itss restrictive second line segment are not close to the 2200 F limit.
The power shape used for the small break LOCA analyses is shown as the solid line in the attached Figure 1.
The K(z) restrictions were also plotted on this curve.
Since the K(z) factor is a normalized function, a value of Fq must be used in order to plot this curve on a plot of power density versus core ht4gnt.
The K(z) restrictions assume an Fq of 2.5.
Currently, an Fq of 2.3 is being requested.
The analyses assumed a value of 2.5 since future submittals are planned to be made which will support an r'q of 2.5.
The power shape used in the analyses summarized in Exhibits F and G is conservative for the value of F proposed in this amendment. It should be noted that no q
operating modes have been postulated which would produce power shapes as extreme as the one used for these umall break LOCA analysus. Therefore, these 2.nalyses justify the deletion of the third line segment of the d(z) curve.
Large Break LOCA The large break LOCA analysis was performed with the 81 Model approved by a Safety Evaluation Report attached to a letter transmitted 12/1/31 J R Miller (NRC) to E P Rahe (Manager of Nuclear Safety, Westinghouse).
The LOCA 1
analysis summarized in Exhibit E used a chopped cosine power shape. The chopped cosine power shape has been found generically to be limitirg for power shapes bounded by the K(g) curve (See the attachment to the letter transmitting the above referenced SER page 12, Section 2.6.3 paragraph 1).
The chopped cosine power distribution has also been demonstrated to bound other power shapes for Prairie Island in the power shape sensitivity study performed with the 81 Model on Exxon fuel (See Letter dated 11/4/85, D M Musolf to the Director of NRR).
The LOCA analysis accounted for hydraulic mismatches between the Westinghouse and Er.on assemblies with a 10 F PCI penalty. The hydraulic mismatch between the two assembly types was small enough that only the crossflows due to rod size and grid designs needed to be evaluated.
Differences in the grid design between the two assembly types can produce local flow ma1 distributions of up to 2.5% in the Westinghouse assemolies.
Previous analyses produced sensitivities showing a 5% flow maldistribution could result in a 19 F peak cladding temperature rise.
iherefore, the 2.5%
flow maldistribution could cause a 9.5 F peak cladding temperature penalty.
8 A 16 F peak cladding temperature penalty was added to the Westinghouse assemblies.
Offsite Doses from Analyzed Lockes Rotor Events Fuel failures project 2d in the Locked Rotor analyses (8%) will be bounded by the 100% fuel failures assumed in the LOCA Offsite Dose Analysis.
ENC (U-3) vs OFA DNB Values For both nominal and transient conditions, the MDNBR has been calculated for both the ENC and the Westinghouse fuel explicitly. Rod bow penalties are then applied to the calculated MDNBR values for each particular fuel type before comparison to the de sign limit. With this method, the design MDNBR limits ar6 a constant 1.30 and 1.17 for the W-3 and WRB-1 correlations respectively and the calculated values are reduced to account for the effect of rod bow.
Exhibir H (NSPNAD 8600) reported the MDNBR results for the ENC fuel using tha W-3 correlatior., since the ENC fuel was more limiting with respect to MDNBR than the Eastinghouse fuel. this is true even after the rod bow penalties are accounted for (the rod bow 7enalty for Westinghouse fuel is 5% and less than 3% for ENC fuel). Table 1 shows a comparison of nominal MDNBR conditions for the two fuel types.
Fuel Mechan gal Design Comparison The fuel mechanical design comparison is shown in Table 2.
The major differences are in fuel rod outside diameter and grid design.
The compatibility of the two fuel types is discussed in Reference 5 of Exhibit H.
A detailed discussion of the effect of fuel rod bowing is contained in Section 3.4 of Exhibit H.
2
Use of the W-3 CHF Correlation Below 1000 psia The use of the W-3 CHF correlation was justified previously in the Prairie Island FSAR. The discussion is contained on pages 14.2-30.
RCS Flow Rates The MDNBR methodology used in Exhibit H assumed t RCS total flow rate ot 178,000 gpm, which is the minimum allowed by Prairie Island Tech. Spec.
3.10.J.
When Prairie Island measures the RCS flo'r rate, instrument uncertainties (2.3%) are applied to the measurer.ent before comparing it to the minimum allowable value. This method is ased to ensure that the actual RCS flow rate is always abova the Tech. Spec.
(and thereby transient analysis) value.
ENC Grid Composition The ENC grids do include spacer springs made of Inconel-718 Core Composition There are 40 Westinghouse OFA design assemblies and 81 Exxon TOPROD design assemblies in the Prairie Island 1 Cycle 11 core. Unit 2 Cycle 11 will also add 40 Westinghouse OFA design assemblies.
The OFA design assemblies to be inserted in Cycle 11 of both units will nave axial natural Uranium blankets and Gadolinia bearing fuel pins.
This will be the first time Cadolinia will be used in Westinghouse designed fuel. However, these features have been used in the similar Exxon TOPROD fuel for the last 6 Prairie Island cycles (three on each unit).
The similarity of Unit 1 Cycle 10 (with all ENC TOPROD fuel) and Unit 1, Cycle 11* is shown in Table 3.
The parameter changing the most is the moderator coefficient. This is an effect of the smaller diameter Westinghouse fuel.
Even though there will be less boron in the core at beginning of life for Cycla 11 (which by itself would tend to make the moderator coefficient less positive), the moderator coefficient is more positive due to additional coolant in the core with the smaller diameter Westirghouss fuel rods.
Average Flow Velocity in the Coge The average flow velocity in the core is as follows'.
All ENC TOPROD Core 14.149 ft/see All Westinghouse OFA Core 13.267 ft/see Prairie Island 1 Cycle 11*
13.844 ft/see Moderator Temperatures at Full Power Tin, n minal 530.5 F T
nominal 560 F ay,,
- 40 0FA assemblies, 81 ENC TOPROD assemblies used in Exhibit H 3
Grid Height ENC TOPROD Westinghouse OFA Middle 5 2.25" 2.25" Top & Bottae 2.25" 1.50" The higher oft 6 loss coefficients are a result of using thicker pieces for the grid construction.
Effect of Guide Tube Diameter on RCCA Drop Times The new 0FA fuel has a reduced diameter guide tube. This will cause an increased resistance to RCCA insertion, thereby increasing the time it takes for the rods to fully insert following a SCRAM. However, the increase in rod drop time will not be large enough to necessitate a change in the Technical Specificat.on limit of 1.8 sec.
i Westinghouse Standard Fuel Assemblies The currently operating Unit 2 cycle (Cycle 10) contains 4 Standard fuel assemblies. Originally, no Standard Westinghouse fuel was planned to be used in Unit 2 Cycle 10. During refueling several assemblies were damaged, and the Standard Westinghouse fuel was used as replacements. These assemblies are currently covered by a past Westinghouse analysis limiting them to an Fq of 2.21.
These assemblics are not covered by the LOCA analyses (allowing an Fq of 2.30) submitted with this analysis.
Therefore, we will limit these assemblies to the current Technical Specification F limit of 2.21 and the existing K(z) curve until they are removed in October of this year. At this time thimble plugs will be removed, Westinghouso optimized Fuel will be added and the new Upper internals will be installed in Unit 2.
Changes to Exhibit D In Exhibit D, the response to Item 3 should read "... loop above 10%
power" rather than "... loop below 10% power."
Add the following words to Item 7:
The impact of the Westinghouse assemblies on the Exxon assemblies was also analyzed and found to be acceptable.
4
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1 TABLE 1 I
Initial Conditions for MDNBR Analysis
, ENC Westinghouse Core Power (MWth) 1683 1683 Total RCS Flow (M1ba/hr) 68.62 68.62 Active Core Flow (Mlba/hr) 64.50 64.50 Coolant Inlet Temperature ( F) 534.5 534.5 1
Pressure (psia) 2220 2220 Core Power - 102% rated Total RCS Flow - Miniwum Tech. Spec. Value (178,000 gpm)
Active Core Flow - 94% Total RCS Flow Coolant Inlet Temperature - Nominal +4 "F Pressure - Nominal -30 psia Margin (5) - (1 - design limit / calculated value)
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p TABLE 2 TOPROD
__OFA Fuel Assembly Length (in) 159.71 159.71 Fuel Rod Length (in)-
152.00 151.8S Fuel Rod Pitch (in) 0.556 0.556 ruel Rods / Assembly 179 179 Guide Tubes / Assembly 16 16 Instrument Tubea/ Assembly 1
1 Clad Material Zr-4 Zr-4
, Clad O.D. (in) 0.417 0.400 Clad Thickness (in) 0.0295 0.0243 Pael fallec 0.D. (in) 0.3505 0.3444 Cuide Tube 0.D. (in) 0.541 0.528 Nunbar of Crlds 7
7 Crid !!aterial Middle 5 Grids Zr 4 Zr 4 i
.2 End Crids Zr 4 Inc-718 2
Active Surface Araa/ Assembly (ft )
234.5 225.0 2
Flow Area / Assembly (in )
32.62 34.79 Average Heat Flur. (M3tu/hr-ft2) 0.1984 0.2068 Average Linear Power (Kw/ft) 6.35 6.35 FO 2.30 2.30 Peak Linear Power (Kw/ft) 14.60 14.60 f
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d TABLE 3 s
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i Cycle 10 Cycle 11 Rod isorth AAI 6625-6267
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.SDM.
30C (p:m) 1928 1946 p
Eoc (pcm) 344 252 l
4 e
Modar.stoe Coe.f. (per/*F)
-3,0
.8 ARO, HIP, BOC t
Doppler Coef.. (pep /*r)
-1.7
-1.E i
Boron Concentration ARO, HZP, BOG 1447 1331 ARO,'dFF, 100 hvD/HTU 975 893 Boron Worth (pcm/ ppm)
+
BOC, HZP 1203 ppe
-9.00
-9.18 ENO WOFA Pressure drop across core used in saalysist 2",.3 pst 24.0 psf.
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lst Line Segment 14 12 LOCA Analysis, Assumeri Power Shape C
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CORE HEIGHT (FT)
FICURE 1 Srna.l Break Pcwer Distribution Assumed for Loca Analysis y