ML19270G445
| ML19270G445 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 05/31/1979 |
| From: | Groce R YANKEE ATOMIC ELECTRIC CO. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-05.A, TASK-3-5.A, TASK-RR WYR-79-64, NUDOCS 7906080271 | |
| Download: ML19270G445 (6) | |
Text
I Telephone 617 366-90ll i
TwX F M, 3 9 0-0 7 3 9 YAflKEE ATOMIC ELECTRIC COMPAllY u.a.2.1 WYR 79-64
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%!"!O 20 Turnpiko Road West!arough, Massachusotn Oi581 May 31, 1979 e
United States Nuclear Regulatory Commission Washington, DC-20555 Attention: Office of Nuclear Reactor Regulation Mr. Dennis L. Ziemann Operating Reactors Branch #2 Division of Operating Reactors
Reference:
(a) License No. DPR-3 (Docket No. 50-29)
Dear Sir:
Subject' SEP Topic III-5. A, High Energy Line Breaks Inside Containment in response to your request at our meeting on March 21, 1979, we are submitting further details for the development of input for the review of SEP Topic III-5.A.
We have completed the definition of the systems which are classified as high energy systems in the vapor container and have developed a list of the lines in those Lystems.
These are included as Attachment A.
Our program for evaluating the effects of a break in one of these high energy lines is fi rs t to idratify the lines and equipment in the containment which are required to safely shutdown the plant.
This piping and equipment is that which is required to insert negative reactivicy into the reactor core, to maintain main coolant inventory and secondary water inventory, to control primary plant overpressure, or to control cooldown of the primary plant.
Then, using an effects-oriented approach, the safe shutdown piping and equipment are surveyed to determine if the br eaking of any of the high energy pipes listad in Attachment A wil; have any affect on. safe shutdown of tte plant.
The following are the conditions which are postulated in determining the piping and equipment which must remain available af ter a high energy line break:
1.
A simultaneous, unrelated, single f ailure is not postulated with the high energy pipe break.
2.
The ropture of any system, except the main coolant system, shall not cause a LOCA.
3.
A simultaneous, unrelated, LOCA is not postulated with the high energy pipe break.
4.
The rupture of any system (not a LOCA) shall not prevent the plant from safely shutting down.
5.
The rupture of any system shall not create an accident of a different type which has not been m,alyzed.
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United States Nuclear Regulatory Commission Page 2 May 31, 1979 6.
The effects of pipe whip will not rupture equal diameter or larger piping which has equal or greater wall thickness.
7.
The piping boundary is taken as the first normally closed valve, checx valve, relief / safety valve or first valve capable of remote or automatic closure.
8.
Effects of pipe whip and jet impingement f rom rupture of piping 1" nominal size and smaller are not required to be analyzed.
9.
Containment integrity required only for pipe breaks which result in a loss of main coolant.
We have performed a preliminary survey of the piping and equipment in the containment according to the criteria given above, and have determined that there are not very many high energy pipes in containment that are not part of the main coolant system boundary.
This is especially true in the four main coolant loop compartments because the boundary valves are in close proximity to the main coolant piping.
The physical layout of the piping also contributes to minimizing problem areas because the piping penetrations are at the lower quandrant of the containment and are well-shielded from the loop The penet rations for the main steam and feedwater pipes are at the areas.
bottom of the elevated containment and the pipes are routed outside the loop shield wall directly to their respective steam generator which minimizes their exposure.
The electrical penetrations are up at t h e"e q ua to r" level and are thus well isolated f rom most high energy pipes.
An overview of our preliminary results indicates that there should be a minimum of problem areas with most of the high energy piping and that analyses will be required for the effects of the main steam and feedwater line breaks; for several break locations in smalle r lines (e.g., steam generator blowdown lines); and for the effects of breaks on penetrations, conduits, and cables.
We plan to identify to the extent possible the pipe breaks which will require no further analysis and identify those problem areas where only further analysis will resolve the problems.
When the work is completed to this point, wh ich we
.pect to be mid-Augus t, 1979, we will be prepared to discuss the status and results of our work with your staff, including the numb r of breaks that requir.e further analysis and the estimated time.and cost for doing these analyses.
at this point we will expect to receive from NRC further direction on the extent of work required to complete these analyses.
Our ability to expedite this work for SEP has been severely disrupted by the additional effort required to supply information for NRC in the 'ollow-up to the Three Mile Island incident.
Specifically, the same personnel who have been working on the SEP project for Yankee Rowe have had to be rensaigned to provide answe rs to NRC bulletins which address numerous areas of concern, provide data for the many informal requests for additional information, and prepare for meetings with NRC to address these concerns.
Because of this serious depletion of manp wer and in some cases, duplication of the SEP e f f o~ r t, it is requiring more time than we had anticipated to complete tasks for SEP.
Howe ve r, we believe we can meet our existing TMI-related commitments and the program and schedule identified above for this SEP review.
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United States Nuclear Regulatory Commission Page 3 May 31, 1979 We t rust this information is acceptable to you; however, should you have any ques tions, please contact Mr. James W. Stacey of this of fice.
Ve ry t ruly you rs,
YANKEE ATOMIC ELECTRIC COMPANY Robert 11. Croce Licensing Engineer DAll/jgh Att.?chment 2337 199
,.- - m
ATTACliMENT A CLASSIFICATION OF IIIGH ENERGY 3YSTEMS The following systems have been classified as high energy based upon a temperature 12000F and/or a pressure 1275 psig for a period < 2% of the time that plant is at normal temperature and pressure.
1.
Main Steam 2.
Feedwa te r 3.
Steam Gene rator Blowdown 4.
Charging 5.
Steam Generator Instrumentation 6.
Main Coolant (includes following systems up to first normally closed valve, check valve, safety valve, relief valve, and remote or automatic isolation valve) a.
Main Coolant Vents and Drains b.
Safety Injection d.
Charging e.
Letdown f.
Pressure Control and Relief IDENTIFICATION OF llIGH ENERGY LINES The lines considered to be high energy are listed on the following table:
SECONDARY Line Size Designation Remarks 14" Loop 1 Main Steam 14" Loop 2 Main Steam 14" Loop 3 Main S team 14" Loop 4 Main Steam 8"
Loop 1 Fecdwater 8"
Loop 2 reedwater 777 qqq 8"
Loop 3 Feedwater i')
LUU 8"
Loop 4 Feedwater 2"
Loop 1 S/G Blowdown 2"
Loop 2 S/G Blowdown 2"
Loop 3 S/G Blowdown 2"
Loop 4 S/G Blowdown 1"
Loop 1 S/G Level (1) 1" Loop 2 S/G Level (1) 1" Loop 3 S/G Level (1) 1" Loop 4 S/G Level (1) 4" VC IIcater Steam Line 1 (2) 4" VC licater S team Line 2 (2)
'4" VC IIeater Steam Line 3 (2) 4" VC lleater S team Line 4 (2) 3" VC lleater Condensate Return Line 1 (2) 3" VC 11 eater Condensate Return Line 2 (2)___
3" VC IIcater Condensate Return Line 3
' '(2) 3" VC lleater Condensate Return Line 4 (2)
PRIMARY Line Size Designation Remarks 20" Loop 1 Hot Leg 20" Loop 2 Hot Leg 20" Loop 3 Hot Leg 20" Loop 4 Hot Leg 24" Loop 1 Pump Suction 24" Loop 2 Pump Suction 24" Loop 3 Pump Suction 24" Loop 4 Pump Suction 20" Loop 1 Cold Leg 20" Loop 2 Cold Leg 20" Loop 3 Cold Leg 20" Loop 4 Cold Leg 2"
Normal Charging Loop 4 TH 2"
Alternate Charging 2"
Letdown 4"
Pressurizer Surge Line 5"
Loop 1 Crossover 5"
Loop 2 Crossover 5"
Loop 3 Crossover 5"
Loop 4 Crossover 2"
Pressurizer Solenoid Relief Inlet 3"
Pressurizer Safety Valve Inlet 3"
Pressurizer Safety Valve Inlet 3"
Solenoid Relief Discharge (2) 4" Safety Valve Discharge (2) 4" Safety Valve Discharge (2) 6" Combined Relief 6 Safety Valve Discharge (2)
< 1" Primary System Instrument Transmitter Lines (1) 1" Loop 1 Safety Valve (1) 1" Loop 2 Safety Valve (1) 1" Loop 3 Safety Valve (1) 1" Loop 4 Safety Valve (1) 11/4" Pressurizer Spray 3/4" Feed & Bleed Heat Exchanger Vent (1) 3/4" Pressurizer Vent (1) 3/4" Loop 1 Vent (1) 3/4" Loop 2 Vent (1) 3/4" Loop 3 Vent (1) 3/4" Loop 4 Vent (1) 1 1/2" Loop 1 Hot Leg Drain 1 1/2" Loop 2 Hot Leg Drain 1 1/2" Loop 3 Hot Leg Drain 1 1/2" Loop 4 Hot Leg Drain 2[ })) 1 1/2" Loop 1 Cold Leg Drain 1 1/2" Loop 2 Cold Leg Drain 1 1/2" Loop 3 Cold Leg Drain 1 1/2" Loop 4 Cold Leg Drain 1" Pressurizer Drain (1) 6" Shutdown Cooling Suction (2) 6" Shutdown Cooling Discharge (2)
} Line Size Designation Remarks 2" Safety Injection at Loop 1 2" Safety Injection at Loop 2 2" Safety Injection at Loop 3 2" Safety Injection at Loop 4 NOTES: (1) Eliminate because of size (2) Eliminate, used less than 2% of the time 9,, w .) e /}}