ML20147B303
| ML20147B303 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/27/1997 |
| From: | Beard P FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, NUDOCS 9701300210 | |
| Download: ML20147B303 (7) | |
Text
.
Florida Power CORPORATION EM January 27, 1997 3F0197-05 U. S. Nuclear Regulatory Commission Attn: Document Control Desk l
Washington, D.C.
20555-0001
Subject:
120 Day Response to NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions"
References:
A. FPC to NRC letter, 3F1096-19, dated October 18, 1996 B. FPC to NRC letter, 3F1296-05, dated December 13, 1996
Dear Sir:
Florida Power Corporation (FPC) is submitting this response in accordance with Generic Letter (GL) 96-06 and Reference A.
FPC has evaluated the conditions identified in the GL as they affect systems and components at Crystal River Unit 3 (CR-3). A summary of our evaluation is reported in the following paragraphs.
Waterhammer and Two-Phase Flow in Containment Air Coolers The containment air cooling heat exchangers (AHHE-31A, B, and C) are supplied cooling water from the Nuclear Services Closed Cycle Cooling System (SW). As the name implies this is a closed loop system. The SW system utilizes a surge tank for volume control which is pressurized to a minimum of 71.7 psig via nitrogen overpressure. Surge tank pressure is alarmed in the control room on either high or low pressure via a computer annunciator point.
Calculations were performed to evaluate the effects of post accident containment heat up on cooling water in the SW system following either a loss of coolant accident with loss of offsite power, or a main steam line break and loss of offsite power. Calculation results demonstrate that flashing will not occur in any portions of the SW system. This is a result of the increased saturation temperature of the SW system due to the nitrogen overpressure on the SW surge tank relative to predicted containment conditions. Since flashing cannot occur, neither waterhammer nor two-phase flow will occur. Therefore, the containment air cooling units will remain operable, and no corrective actions are required for waterhammer or two-phase flow %
concerns.
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V. S. Nuclear Regulatory Commission 3F0197-05 Page 2 of 6 Overoressurization of Isolated Pioina Sections
. FPC evaluated the effects of post accident containment heat up on isolated sections of piping penetrating the containment building.
Mechanical piping I
penetrations were screened to determine susceptiblity to overpressurization that could threaten containment integrity. Penetrations that had one of the following attributes were considered to be not susceptible:
- 1) containing a compressible fluid,
- 2) being open to the containment atmosphere,
- 3) being open to a surge volume, or
- 4) having an installed relief valve.
j.
Table 1 lists the penetrations screened out, and the reason for determining each to be not susceptible to overpressurization.
4 i
The penetrations not screened out were determined to be potentially subject to overpressurization, and were evaluated to determine their operability under post accident conditions.
Additionally, since the SW system may perform an active safety function following an accident, SW system penetrations that included installed relief valves were further evaluated to determine if the relief valve supplied sufficient relieving capacity. The penetrations considered susceptible, including SW, which were evaluated in detail are listed in Table 2.
i The majority of the existing SW system relief valves provide sufficient relieving capacity for fluid expansion due to post accident heat up; however, the relief valves associated with the containment air cooling units are not sufficient and must be replaced.
These are indicated by a "Y" in the last column of Table 2 1
denoting that a modification (Mod) will be made. The relief pressure and reseat pressure for relief valves on the containment air cooling units will be i
sufficient to prevent system pressure from dropping to saturation pressure in operating cooling units, and will be sufficient to protect an isolated unit from rupture during accident conditions.
)
During the evaluation of SW piping loops inside containment, FPC found that there are SW piping fittings and connections for the reactor control rod drive mechanisms that are not qualified for the post accident containment temperature conditions. These components could fail due to high temperature and cause leaks in this SW loop inside containment. This loop serves only the reactor control rod drive mechanisms. FPC is continuing to evaluate corrective actions for these components, and will notify the NRC of the corrective actions chosen by April 25, 4
1997 The susceptible penetrations, other than SW and Penetrations 314 and 318, will be modified by installation of expansion chambers to allow for expansion of the 1
process fluid. The expansion chambers are connected directly to the penetration piping, and an air volume is maintained inside the expansion chamber by isolation of the process fluid with a rupture disc. The expansion chamber is designed and considered to be part of the containment pressure boundary. This design allows for expansion of the fluid in the isolated process piping without violating containment integrity.
The expansion chambers will be located outside of the i
containment.
~ U. S. Nuclear Regulatory Commission 3F0197.
Page 3 of 6 Penetrations 314 and 318 for secondary side steam generator drains will be fitted with rupture discs on the penetration boundary inside the containment, without expansion chambers.
This is considered to be an acceptable and conservative alternative in this special case because these lines are normally isolated and drained during operation. Should some leakage past the steam generator isolation valves occur, the rupture disc will provide overpressure protection.
No significant steam generator water inventory loss will occur since the isolation valves are closed during operation, and loss will be limited.to leakage past the isolation valves.
CR-3 is currently in cold shutdown and no actions are immediately necessary to assure continued operability of affected systems. In this condition containment air cooling units are r.ot required to be operable, and'there is insufficient thermal energy to threaten functionality of isol ated penetrations.
The replacement of SW relief valves with larger valves, corrective actions for low temperature components in the control rod drive cooling loop, and the installation of expansion chambers and rupture discs on the other susceptible systems will be completed prior to start-up from the current CR-3 outage. This item is identified as Restart Issue D-8 in Reference B.
Sincercly, m
es P. M. Beard, Jr.
Senior Vice President Nuclear Operations PMB/SCP:ff Attachment Regional Administrator, Region II xc:
Senior Resident Inspector NRR Project Manager
U. S. Nuclear Regulatory Commission 3F0197-05 a
Page 4 of 6
{
Table 1 Reactor Buildina Mechanical Penetrations Screened Out Penetration System or Service Line Size, Relief Mechanism No.
In.
105 Main Steam 24 Compressible fluid 106 Main Steam 24 Compressible fluid 107 Main Steam 24 Compressible fluid 108 Main Feedwater 18 Open to the OTSG 109 Emergency Feedwater 6
Open to the OTSG 110 Station Air 3
Compressible fluid j
111 Instrument Air 2
Compressible fluid i
112 Instrument Air 2
Compressible fluid 113 Reactor Building Purge 48 Compressible fluid 116 Leak Rate Test 2
Compressible fluid 117 Demineralized Water 3
open to the RCP standpipe 119 Spare 20 Compressible fluid 120 Spare 20 Compressible fluid 121 Leak Rate Test &
3 Compressible fluid H Recombiner 3
122 RB Leak Rate Test &
3 Compressible fluid H Recombiner 2
123 hitrogen 1
Compressible fluid 124 Nitrogen 1
Compressible fluid j
125 H Recombiner 3
Compressible fluid g
201 Main Steam 24 Compressible fluid 202 RB Leak Rate Test 2
Compressible fluid 206 Industrial cooling 2.5 Relief valve 207 Industrial cooling 2.5 Relief valve 216 Spare 12 Compressible fluid 217 Spare 12 Compressible fluid 305 Post Accident Venting 6
Compressible fluid 306 Post Accident Venting 6
Compressible fluid 306 Containment Monitoring
.5 Compressible fluid 306A Contalnment Monitering
.5 Compressible fIuid 315 RB Air Sample 1
Compressible fluid 316 Secondary Vent from Steam Generator 1.5 Compressible fluid 317 Nitrogen 1.5 Compressible fluid 319 RB Pressure Sensing 1
Open to containment 320 Secondary Vent from Steam Generator 1.5 Compressible fluid 332 Reactor Butiding Air Sample 1
Compressible fluid 336 High Pressure Injection 3
Pressure relief to RPV via check valves 337 High Pressure Injection 3
Pressure relief to RPV via check valves 338 Reactor Coolant Pump Seals 4
Pressure relief to RPV via check valves 340 Reactor Building Spray 8
Open to containment 341 Reactor Building Spray 8
Open to containme..t 342 Low Pressure injection 10 Pressure relief to RPV via check valves 343 Low Pressure injection 10 Pressure rotief to RPV via check valves 344 Decay Heat 12 Relief valve 345 R8 Sw p Recirculation 14 Compressible fluid 346 RB Sump Recirculation 14 Conpressible fluid 348 Fuel Transfer Tube 30 Compressible fluid 349 Waste Disposal 2
Compressible fluid
U. S. Nuclear Regulatory Commission 3F0197-05 Page 5 of 6 4
Penetration System or Service Lir.a Size, Relief Mechanism No.
In.
j 350 Core Flood 1
Pressure relief to core flood tank 351 Core Flood Tank Vent 1.5 Compressible fluid l
354 Waste Disposal 1.5 Compressible fluid l
355 Nitrogen 1.5 Compressible fluid i
356 Reactor Building Air Sample
.5 Compressible fluid 357 Reactor Building Purge 48 Compressible fluid 366 Industrial Cooling 2.5 Relief valve 1
367 Industrial cooling 2.5 Relief valve 372 Nitrogen 1
Compressible fluid 373 Core Flood 1
Pressure relief to core flood tank 376 Reactor Building Air Sample
.5 Compressible fluid 423 Main Feedwater 18 Open to the OTSG 424 Emergency Feedwater 6
Open to the OTSG 426 RB Pressure Sensing 1
Open to containment 427 Secondary Drain from Steam Generator 3
Inside valves MSV-142 through -145 normally 4
open to OTSG j
428 Secondary Drain from Steam Generator 3
Inside valves MSV-124 through -127 normally open to OTSG j
l 429 RB Pressure Sensing 2
Open to containment i
430 Fire Service Water 4
Compressible fluid 434 High Pressure Injection 3
Pressure relief to RPV j
435 Makeup & High Pressure Injection 3
Pressure relief to RPV d
436 Fuel Transfer Tube 30 Compressible fluid 442 RB Pressure Sensing 2
Open to containment i
Abbreviations:
{
OTSG - Once Through Steam Generator i
RB Reactor Building RCP Reactor Coolant Pump RPV - Reactor Pressure Vessel 1
(
(
b d
a 4
4 4
4
\\
U. S. Nuclear Regulatory Commission 3F0197-05 e
Page 6 of 6 Table 2 Reactor Buildina Penetrations Potentially Susceptible to Post Accident Overoressurization and the Relievina Device used for Mitiaation Penetration System Service Relieving Device Mod
- No.
314 Main Steam Steam Ganerator Drain Rupture disc Y
318 Main Steam Steam Generator Drain Rupture disc Y
321 SW Letdown Cooler Relief valve N
322 SW Letdown Cooler Relief valve N
323 SW Reactor Coolant Pump Relief valve N
324 SW Reactor Coolant Ptanp Relief valve N
325 SW Reactor Coolant Pump Relief valve N
326 SW Reactor Coolant Pump Relief valve N
329 Decay Heat Pressurizer Spray Rupture disc Y
330 SW Control Rod Drive Relief valve N(1) 331 SW control Rod Drive Relief valve N(1) 333 Make-up Letdown Purification Rupture disc Y
339 Liquid Waste Reactor Building Sump Drain Rupture disc Y
a 347 Spent Fuel cooling Transfer Canal Purification Rupture disc Y
352 Core Flood Sample and Bleed Rupture disc Y
358/359 SW Containment Air Cooler Relief valve Y
360 SW Letdown Cooler Relief valve N
361 SW Letdown Cooler Relief valve N
362 SW Reactor Coolant Pump Relief valve N
363 SW Reactor Coolant Pump Relief valve N
364 SW Reactor Coolant Ptano Relief valve N
365 SW Reactor Coolant Pump Relief valve N
368/3v9 SW Containment Air Cooler Relief valve Y
370/371 SW Containment Air Cooler Relief valve Y
374 Liquid Waste Reactor Coolant Drain Tank Rupture disc Y
377 Reactor Coolant Letdown Reactor Coolant Pump Seals Rupture disc Y
425 Post Accident Sampling Reactor Building Sunp Rupture disc Y
439 Liquid Sampling Reactor Coolant / Pressurizer Rupture disc Y
440 Liquid Sampling Steam Generator Ruoture disc Y
441 Liquid Sampling Steam Generator Rupture disc Y
- " Mod" stands for Modification (1)
Components in this piping loop may be modified to improve high temperature performance.
U. S. Nuclear Regulatory Commission 3F0197-05 STATE OF FLORIDA COUNTY OF CITRUS P. M. Beard, Jr. states that he is the Senior Vice President, Nuclear Operations for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
P. M. Beard, Jr.
Senior Vice President Nuclear Operations Subscribed and swotn to before me, a Notary Public in and for the State and County above named, this 27th day of January, 1997.
P. M. Beard, Jr.,
is personally known to me.
[VAll][ $. TfDfIk VJ Notary Public (print)
Notary Public (signature)
MY COMMISSION # CC 514300 EXPMS: December 18,1980 tended Thru Netsy Pete(Nonumers i
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