ML19339B564
| ML19339B564 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 09/21/1977 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8011070254 | |
| Download: ML19339B564 (51) | |
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ATTACitMENT I INSTRUCTIONS FUR IMPLEMENTING 1
i' SUPPLEMENT NO. 7 TO PROPOSED CilANGE No.145
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SUPPLEMENT NO. 7 TO PROPOSED CHANGE No.145 IMPLEMENTATION INSTRUCTIONS s
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Replace Section 4.0 "Small Break Analysis" of Appendix A to " Yankee t
Nuclear Power Station Core X111 Performance Analysis" with the j
corresponding Section 4.0 attached in its entirety (including all tables 1
and figures).
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4.0 SMALL BREAK ANALYSIS 4.1 In t roduct ion i
j Modifications to the ECCS (Section 2.0) and to the low flow film boiling correlation used by Yankee (Section 3.2) were made to assure that I
the consequences of the previously identified most limiting small break LOCA (2.25 in I.D. thermal sleeve) are within the limits specified in 10CFR50.46. As previously discussed, the unique break postulated to occur is a complete severance in a small length of ECCS piping (1 to 2 feet) immediately downstream of the check valve which is nearest to one of the RCS cold leg injection points. This bre location results in reactor coolant (RCS) blowdown through a 2.25 a.
I.D. thermal sleeve and ECCS spillage through a 3.438 in. 1.D. ECCS line to containment.
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The revised ECCS assures that the consequences of this most l
improbable event and other small breaks are acceptable by providing separate i
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llPSI through each of the ECCS injection trains to the four RCS cold legs.
4 Sufficient injection to the three intact cold legs is sssured by supplying
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resistance in each of the separate llPSI lines to yield a HPSI header pressure in excess of 1000 psia when spilling through the ruptured ECCS line prior to injection to the intact loops. This is accomplished using throttle valves in each of the four lines connecting the HPSI header to the LPSI injection lines (refer to Section 2.0).
To confirm that the combination of modified low flow film boiling heat transfer and the revised ECCS provide the margin necessary to limit the 4
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i consequences of small break LOCA events within 10CFR50.46 criteria, a complete small br. ak LOCA spectrum analysis was performed.
Sections 4.2 through 4.4 provide the analysis of the 2.25 inch ID thermal sleeve, 4.0 inch ID, 5.0 inch lu, 7.5 inch ID and 10 inch ID small breaks.
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4.2 Method of Analysis
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S The trans ient depressurization of the RCS was calculated using the NHC approved RELAP4-EM digital computer code as modified by the Exxon Nuclear 1
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Company (RELAP4-EM/00311/07/75 95 ENC 20) and Yankee (refer to Section 3.2).
For rod heatup calculations, the approved TOODEE2 Versio2 ENC 13 as modified by Yankee (refer to Section 3.2) digital computer code was used.
The reactor coolant system was nodalized into control volumes interconnected by flowpaths as shown in Figure 4-1.
The broken loop was modeled explicitly while the intact loops were lumped together. This model is identical to the nodalization used in the Core XII analysis excepting:
(1) changes in the ECCS portion of the model necessary to represent the modified ECCS and i
1j (2) changes in the ECCS model necessary to accurately reproduce both the
- i results of the most recent LPSI and HPSI pump tests and the 1972 accumulator flow tests.
Figure 4-2 provides the Core XII model for comparison.
i (3) changes to the input HPSI fill curves.o reflect recent HPSI pump tests.
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4 For conservatism The results of these tests are provided in Figure 4.3.
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and to allow for future potential measurement uncertainties, the minimum
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HPSI perf ormance curve (i.e., Pump No. 3 f rom 790 paid to 160 psid, and Pump No. 1 f rom 160 psid to runout conditions) was reduced by 7.5%
both in flow and head. This revised performance curve is also included i
in Figure 4-3.
1 As stated above, the only actual change in the model is in the ECCS modeling. However, a minor change in secondary system modeling was required due to RELAP restrictions. The change is in the andeling of the feedwater flow to the steam generators.
In the Core X11 analysis, feedwater flow to the steam generators was modeled as a time dependent fill junction.
It was found in preparing the current small break model that due to separation of the HPSI pumps from the LPSI pumps (Core X11 LPSI and HPSI pumps were additive and treated as a single fill junction) that a total of six (6) fill junctions would be required if the primary system model I
were to be retained intact.
Since RELAP is restricted to a maximum of five (5) fill junctions, it was decided to modify the feedwater modeling since this was the most readily accomplishable change. Thus, feedwater input to the steam generator is modeled as a time dependent volume (Volume 26) connected to er.ch of the two steam generator secondary nodes (Volumes 16 and 17) with time dependent valves in each junction. The change was developed to assure that the feedwater flow, which is minimal since it is ramped from full flow to zero flow in two seconds, remained consistent with the Core X11 analysis and subsequently was confirmed in the analysis.
The peak clad temperature analysis was performed using the T00DEE2 digital computer code as modified by Yankee (refer to Section 3.2).
Figure 4-4 provides the 17 axial node model employed to simulate the peak rod.
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i Initial f uel rod temperatures were obtained utilizing RELAP4-EM/ HOT Channel as modified by Yankee (refer to Section 3.2) with a corresponding axial nodalization. Time dependent fluid conditions required as T00DEE2 input i
were obtained from the blowdown results.
I 1
4.3 Results "t
The following figures illustrate the key parameters for the various small breaks analyzed for Core X111 at BOC Conditions:
Figures 4-5.1 through 4-5.8 2.25 inch ID 12.85 kw/ft Figures 4-6.1 through 4-6.8 4.00 inch ID 12.85 kw/ft i
Figures 4-7.1 through 4-7.8 5.00 inch ID 12.85 kw/ft t
Figures 4-8.1 through 4-8.8 7.50 inch ID 12.85 kw/ft j
Figures 4-9.1 through 4-9.8 10.00 inch ID 12.85 kw/ft Table 4-1 provides a summary of the sequence of events for the above breaks and Table 4-2 provides a summary of the results.
i 4.4 Conclusions i
The ef fects of the revised ECCS and modified low flow film boiling I'
correlation yield small break LOCA transients whose consequences are well i
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within the limits specified in 10CFR50.46.
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i TABLE 4-1 YANKEE ROWE CURE X111 SHEL BREAK ANALYSIS u
TIME SEQUENCE OF EVENTS l
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l Event Time, Seconds I
j Break Size Equivalent. ID:
2.25" 4.0" 5.0"
- 7. 5" 10.0" Event Pipe Rupture 0.0 0.0 0.0 0.0 0.0' Loss of offsite Power 0.0 0.0 0.0 0.0 0.0 l
Safety Injection Signal 10.5 6.2 6.0 5.6 4.9 PCT Occurs 12.6 209.8 117.6 72.8 5.4 HPSI and LPSL Flow Begins 20.0 20.0 20.0 20.0 20.0 EECS Flow to Intact Cold Legs Begins 48.0 24.8 20.0 20.0 20.0 j
Core kecovery Occurs 616.4 218.5 126.8 76.0 56.0 5
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TABLE 4-2 i
YANKEE ROWE CORE XIII SMALL BREAK LOCA ANALYSIS l
SUMMARY
OF RESULTS*
i Break Size Equivalent Internal Diameter. Inches a
Pa ramete r 2.25 4.0 5.0
- 7. 5 10.0 Peak Clad Temperature "F
1133.5 1793.4 1522.4 1398.4 1625.3 Peak Clad Temperature Location, ft.
4.29 3.79 3.79 3.79 3.79 Maximum Local Zr/H O 2
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.13 0.70
.24
.17
.20 i
Maximum Local Zr/H O 2
i Reaction Location ft.
4.29 4.04 3.79 3.79 3.79 Percent of Total Core Zr/H O Reaction, %
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<1
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- CALCULATIONS PERFORMED AT THE FULLOWING CONDITIONS:
Power Level, MWt 618 Peak Linear Heat Generator Rate 12.85 j
Teial Peaking Factor 2.76 3
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Accumulator Water, Fr 700 Cold Leg Temperature, F 519 i
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2 KEt4*e/G 3 1/13/77 03/32/77 8
YR CORE 13 SMALL BREAK LOCR
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Figure 4-9,7 Cold Leg -Break Flow vs. Time for 10.00 Inch ID Small Breal l
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Figure 4-9.8 Peak. Clad Temperature vs. Time for pg 10.00 Inch ID Small Break bP m.
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