ML18348A756
| ML18348A756 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/14/1977 |
| From: | Bixel D Consumers Power Co |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18348A756 (4) | |
Text
consumers Power company
~~guf atory Docket
- Fila __
General Offices: Z12 West Michigan Avenue, Jackson, Michigan 49201
- Area Code 517 788-0550 February 14, 1977 Director of Nuclear Reactor Regulati 'fl Att:
Mr.Albert Schwencer, Chief Operating Reactor Branch No 1 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255, LICENSE DPR PALISADES PLANT - PRIMARY SYSTEM PRESST.JRE ~NCREASE TO 2,100 PSIA Ir. response to your requests in our meeting of Janua;ry 19, 1977 and in subsequent telephone conversations, the following additional information is being proVided ir. support of the ~echnical Specifications Change to increase the Palisades primary system pressure to 2,10.0 psia (submitted by lett~r dated Janu~ry 3, 1977).
. A question was raised :concerning the adequacy of our fracture toughness reqi'.ure-ment and associated curves.
The analysis an.cl associated curves' provided,by letter dated January 6, 1977 did assu:zne an operating pressure greater than 2,100 psia (see. Page 3-10 of reference letter).
- In answer to your qu~stions concerning. st~am generator tube stresses and steam generator tube plugging criteria, please refer to the report entitled Analysis To Determine Allowab1:.e Tub~ Wall Degradation for Palisades Steam Generators,"
which was transmitteq ~der cover letter dated February 12, 1976 and subse~uently amended in letters dated March 22, 1976 and.April 6, 1976.
Review of this report indicates that:
- 1.
All transients analyzed in the report, ~ncluding LOCA, steam line break, and normal operation, assumed a primary system operating pressure of 2,100 psia.
- 2.
The question.of the e'ffect of different LOCA break sizes and locations on s'team *generator tube stress was addressed in the report via a sensitivity
,stuaY.. Refer to Figures B. 5-and B. 6 in the report~
- You als~ requested pertinent heatup code input. at end-of-bypass and beginning-of-core-recovery for the worst break as reported in XN-76-52,,"Operation of the
- P,alisades Reactor at 2100 Psia:
A LOCA, Thermal-Hyaraulic, Neutronics, and 1_665
~i.
2 foechanfoal Design Assessment," and for the same break at 1,800 psia as reported in X:N-76-4, Supplement 2, "Palisades LOCA Analysis Using.the ENC wREM-Based PWR*
ECCS Evaluation Model."
The worst break is the l.O*DES/PD b:r:eak in Loop 1 for Type D fuel.
This information is provided in Tables 1 through 3 (attached).
David A Bixel Nuclear Licensing Administrator cc :
JGKeppler, usrrnc
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~
TABLE 1. Selected Hot Rod Temperatures at EOBY*
Axial Node Elevation From Core Bottom, In Fuel Average Temp, °F 2,100 Psia 1,800 Psia Clad Outside Surface Temp, °F 2,100 Psia 1,800 Psia 6
11 15 19**
59.4 74.4 86.4 98.4 1,213 1,322 1,351 1,435 1,264 1,377 1,409 1,471 1,060 1,144 1,166 1,231
- EOBY = 16.82 s for 2,100 psia case; = 21.85 s for 1,800 psia case.
- Peak clad temperature node.
TABLE 2.
Selected Hot Rod ~emperatures at BOCREC*
1,214 1,315 1,343 1,383
- BOCREC = 35.34 s for 2,100 psia case; however, the temperatures given correspond to 34.32 s because a data edit was not available for the exact time of BOCREC.
BOCREC = 40.54 s for. 1,800 psia case.
- Peak clad temperature node *.1
TABLE 3 ~. Reflood Input to Heat up Code at BOCREC Coolant Saturation Temperature
' 2,100 Psia 1,800 Psia Time Tsat2 °F Time Tsatz °F 16.82 33.20 53~20 83.20 120.10 160.10 200.10..
- 250.10 Core Inlet Subcooling 2,100 Psia Case - 178.6°F 1,800 Psia Case - 169.0°F Core Flooding Rate 2,100 Psia Case - See Figure 275.0 2*69.0 265.0 262.0 257.6 254.o 250.3 246.o 21.85 35.2 45.2 65.2 100.1 150.1.
250.1 3.23 of Reference 1 1,800 Psi a Case - see Figure* 2.22 of Reference 2 Inlet Core Flow
. 2,100 Psi a Case - See Figure 3.27 of Reference 1 1,800 Psia Case - See Figure 2.26 of Reference 2 References 272.3 268.1 266.2
. 262.4 259.3 254.4 246.1
- 1.
XN-76-52, Operation of the Palisades Reactor at 2100 Psia:
A LOCA, Thermal-
. Hydraulics, Neutronics, and Mechanical Design Assessment
- 2.
XN-76-4, Supplement 2, Palisades LOCA Analyses Using the ENC WREM-Based PWR ECCS Evaluation Model,