ML19210D322
| ML19210D322 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/15/1979 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 1-119-10, NUDOCS 7911260300 | |
| Download: ML19210D322 (6) | |
Text
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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BCX 551 UTTLE ROCK. ARKANSAS 72203 (501) 371-4000 Nov ember 15, 1979 1-119-10 Director of Nuclear Reactor Regulation ATIN:
Mr. Robert W.
Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D.
C.
20555
Subject:
Arkansas Nuclear One-Unit 1 Docket No. 50-313 License No. DPR-51 PORV and Safety Valve Lif t Frequency and Mechanical Rel i a bil i ty (File:
1510.1)
Gen tl em en :
In repons e to your requ est for additional information con-cerning the above subject, the following is provided.
Request 1 According to statements made by B&W, there are approximately 146 documented occasions where PORV actuation occurred at B&W f acilities prior to the accident at Three liile Isl ad, Unit 2 (TMI-2).
For each of these events which have occurred at your f acili ty(ies), provid e the followi ng information:
a.
The cause of the event; b.
the initial power level prior to the transient; c.
indicate which of these transients caused the reactor to trip on high RCS pressure and/or caused the safety valves to lift; and, d.
if you assume that the present setpoints for high RCS pressure trip and PORV actuation were in effect at the time of each of these transients, estimate whether either of the following would have taken place:
791.1260 1390 2~8/
MEMBER MICOLE SOUTH UTILITIES SYSTEM
,)
1-119-10 Mr. Robert W. Reid Nov embe r 15, 1979 (1)
PORV actuation, and (2) lif ting of the safety valves.
(for this item assume no credit for the anticipatory con-trol-grade reactor trip on loss of feedwater or turbine trip).
Response
The requ ested information has been compiled in Attachment 1.
Request 2 Provide a complete listing of reactor trips for your facility (ies) which have occurred subsequent to the revised setpoints for PORV actuation and high RCS pressure trip.
This listing should include the following items:
a.
the cause of each event; b.
the initial power level prior to the transient; c.
indicate which of these transients caused the PORV and/or safety valves to open; and, d.
if the ol d (pre-TMI-2) setpoints for high RCS pres-sure and PORV actuation were in effect at the time of these transients, estima te whether any or all of the following would have taken place:
(1)
PORV actuation; (2) reactor trip on high RCS pressure, and (3) lif ting of the safety valves.
Response
This information has been compiled and is presented in Attach-ment 2.
Request 3 Provide an estimate of the increase in reactor trip frequency since lowering the high pressure trip setpoint and adding the anticipa tory reactor trip.
Include a review of the design 1390 279
1-119-10 Mr. Robert W.
Reid Novermber 15, 1979 criteria for the number of reactor trips over the plant life and evaluate the effect of the increase in trip frequency on these criteria.
Also, provide the basis for the acceptable number of reactor trips in terms of the limiting component (s).
Res po ns e A.
An increase in the reactor trip frequency can be expect-ed to occur as a result of lowering of the high pressure trip setpoint and addition of the anticipatory reactor trip.
In order to estimate the change in trip frequency due to these changes, specific pieces of information and assumptions are necessary.
For instance, the Rx trip frequencies before and af ter implementation of the design changes are needed.
Followi ng commerical operation through the time of the design changes in question, 24 reactor trips occurred in a period of 53 months for a frequency of 0.45 trips /
month.
Following the design changes, 2 reactor trips have occur-red in a period of approximately 5 months for a frequen-cy of 0.40 trips / month.
A simple comparison indicates a slight decrease in fre-quency since the design changes.
However, one should consider that as operating experience and history was gained, the frequency of trips on AN0-1 has decreased.
For instance, in 1977 and 1978 only 7 reactor trips were experienced yielding a trip frequency of 0.29 trips /
month.
Comparing this wi th the past design change frequency of 0.40 trips / month, indicates an increase of 0.11 trips / month.
Probably a more realistic measure of the increase in trip frequency due to the design changes is to evaluate past plant transients which did not result in reactor trips but can reasonably be expected to have caused trips if the PORV s etpoint, high RC pressure trip setpoint and anticipatory changes had been in effect. Review of pa st data shows that 4 such transients occurred during the 53 months following commercial operation and before the desig n changes.
This is an increased frequency of 0.08 trips / month.
In evaluating the consequences of the design changes the following must be considered:
1390 280
1-119-10 Mr. Robert W.
Reid No v em be r 15, 1979 1.
There has been a relatively short period of opera-tion since the changes, and 2.
As operators become f amiliar with the revised set-points and operating conditions, it is reasonable to assume the trip frequencies may decrease.
B.
The structural design criterion for the number of reac-tor trips over the life of the plant is to keep the f atig ue usage f actors of all RCS components below 1.0 as supported by the component stress analysis.
In general, this usage factor is made up of contribution due to all specified transients.
Since the largest contribution to the fatigue usage factor is attributable to heatup and cooldown transients, with reactor trips producing only a small effect, the increase in tr'ip frequency should only have a small effect relative to plant life.
As a part of the total allowabl e trans i ent pic tu re, 400 reactor trips are specified.
Assuming a 40-year life, this translates into 10 trips per year or 0.83 per month.
Since the overall frequency for ANO-1 commercial opera-tion is 0.45 per month there does not appear to be a reason for concern.
Howev er, for the reasons discuss-ed in A above, it is premature to draw any conclusions over the life of the plant based on the little data available wi th these setpoints.
C.
To determine the acceptable number of reactor trips in terns of the limiting component (s), it is necessary to review the stress report for each component and plant and evaluate the fatigue usage factor.
If the number of trips were to exceed 400 on any plant, that plant would have to be re-analyzed based on actual transients and the limiting component would be a func-tion of these actual transients plus those that would be expected throughout the remainder of the plant's life.
Very truly yours, MMd 5
David C.
Trimble ManLger, Licensing DCT/ ERG /ew 1390 28i
e a
9 ATTACHMENT 1 ARKANSAS NUCLEAR ONE - UNIT 1 REACTOR TRIPS d!TH A PORV ACTUATION If present Se roints P2R Had Seen used Initial Safety Transient Trip Power Valves PORY Lif t Sa fety Date Classification Signal Cause of Transient Level Lifted?
Ac tu a t t o r.?
halves?
10-15-74 Loss of Feedwater Hi RCP
'A' FWP Tripped on High 45 No No
%o Vibration 12-6-74 Loss of Feedwater Pressure /
Loss of Vacuum Due to "B" 60 No No 40 T emp Main Chiller Getting het and Shorting
..... -. -. -... -. -. -... -... - -.... P r i o r t o C oe r e r c i a l O p e r a t i o n. -.. - -.. -... -.
1-6-75 Load Rejection Hi RCP Generator Tripped on 98 No No ho Of fferential Current Due to Loss of Buss Cooling 6-6-75 Instrument Failure PWR/
Loose Connection on Loop 100 No No Mo Imbalan e
'B' T, Signal Flow 7-23-75 Loss of Feedwater Pressurt/
Adjustment on FW Heater 94 No No ho T em p Level caused Trip of Mtr Drain Pump d> FWP Trip 7-8-76 Loss of Fe ed wa t e r/
Hi RCP Inst. Techs Shorted hN!-
94 No No ho Power Supply Power Supply Fa il u r e 9-23-76 Turbine Trip Pressure /
Turbine Tripped When Vibra-100 No No ha
~~
Temp tion Trip Module was Reinserted by Technician 12-20-76 Rod Drop / Manual Hi RCP Rod 8 in Group 4 Dropped.
100 No No no Runback Manual Runback Too Rapid 6-19 78 Turbine Trip Hi RCP Technician or Operator 100 No No h3 Error in Opening Wrong Feeder Breaker 10-13-78 Ins. ument failure Pressure /
RPS Channel
'B' RC 100 No No No T emp Flow Signal Fa il ed 12-20-78 Instrument Failure Pressure /
Low Steam Pressure 99 No No ho T emp Caused by LVOT Ltnkage Breaking 1390 282
ATTACHMENT ?
ARKANSAS NUCLEAR ONE-UNIT 1 REACTOR TRIPS SINCE TMI-2 DATE 7-8-79 8-13-79 Transient Classification Rx Trip on Turbine Trip High RCS Press.
Cause of Transient Governor Val ve Switchyard Control Fa il u r e Relay Failure Initial Power Level 76 75 PORV Lifted?
No No PZR Safety Val ves Li f ted?
No No If Old Setpoints Had Been Used PORY Actuation?
Yes Yes Safety Valves Lif ted?
No No Trip on High Pressure?
Unabl e to De'.ermi ne - Too Close 1390 283