ML19210D299

Forwards Info Re Power Operated Relief Valve & Safety Valve Lift Frequency & Mechanical Reliability,In Response to NRC 790928 Ltr

ML19210D299
Person / Time
Site:	Davis Besse
Issue date:	11/20/1979
From:	Roe L TOLEDO EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
NUDOCS 7911260275
Download: ML19210D299 (9)

Text

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TOLEDO

%ms EDISON LOWELL E. ROE Docket No. 50-346 v,c. m..a.es 5.c...c... c..... .n .

License No. NPF-3 (d'S' 25S-52'2 Serial No. 558 November 20, 1979 Director of Nuclear Reactor Regulation Attention: Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Reid:

This letter is in response to your letter to the B&W operating plants on "PORV and Safety Valve Lif t Frequency and Mechanical Reliability" dated September 28, 1979. Attachments 1, 2 and 3 provide our responses to the subject letter. The attachment number corresponds to the item number provided in your letter.

Attachment I also provides those documented events at B&W operating plants where the PORV stuck open. The data provided in Attachment 3 is current for Davis-Besse Unit I as of October 12, 1979.

Should you have any questions, please advise.

Very truly yours, 1 /

a e LER:SCJ cc: ~

h R. A. Capra Project Management Group Bulletins and Orders Task Force U. S. Nuclear Regulatory Cor: mission Washington, D. C. 20555 Y

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THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISCN AVENUE TOLEDO. OHIO 43652 7011260 N _

1-1 A T T A C; M I: M T l' Request i According to statements made by B&W, there are approximately 146 documented occasions where PORV actuation occurred at B&W facilities prior to the accident at Three Mile Island, Unit 2 (TMI-2). For each of these events which have occurred at your facility (ies), provide the following information: ,

a. The cause of the event;

b. the initial power level prior to the transient;

c. Indicate which of these trarsients caused the reactor to

, trip on hiah RCS pressure and/or caused the safety valves to lift; and,

d. If you assume that the present setpoints for hign RCS pressure trip and PORV actuation were in effect at the time of each of these transients, estimate whether either of the followina would have taken place:

(1) PORV actuation, and (2) lifting of the safety valves.

(For this item assume no credit for the anticipatory control-grade reactor trip on loss of feedwater or turbine trip.)

,Respcnse The requested information has been compiled in the followina plant specific tables (Reactor trips with a PORV actuation).

In addition, there have been seven (7) instances when the PORV stuck open - three (3) when the plant was at power and four (4) when it was not producinq power.

a.

At power:

(1) Oconee-3, June 13,1975 (Feedwater oscillations while shutting down)

(2) Davis-Besse 1, Sept. 24, 1977 (Loss of Feedwater)

(3). THI-2, March 28, 1979 (Loss of Feedwater) 159d145 e

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• I-2 ATTACHMEtlT I (Cont'd)

b. tiot producinq power:

(1) Oconee-2, August 15,1973 (Pre-op Testing)

(2) Oconee-2, tiovember 6, 1973 (RCS Heatup)

(3) Davis-Besse 1, October 13. 1977 (Hot Standby)

(4) TMI-2, March 29, 1978 (Zero Power Physics Testing)

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PCRV ACTUATIONS - DAVIS-BESSE-l I-3 REACTOR TRIPS WITH A PORV ACTUATION INITIAL IT PRESENT SETPOINTS POWER P:R. HAD SEEN "SE3 LEVEL SAFETT "RANSID47 *R!? (I R"L VAL *.T.S P0RV LIFT SAFEU 2ATI CU S!!TICA*'04 3!T:AL CAUSE OF -'tA':S!r:? FCL'ER1 L!?H3? AC*"A!!ON' " AI."*' S ?

9-:-77 Turoine *ria La RC?* OTSG Overfed ay 47.3 No No No 0perator

?-;.-7* Losa sf Fevcwater Manual " Half-! rip" of CFRC3 No No 7 No Isolated GTSCa 10-23-77 Lcss of Feecwater Lo RCP STRC3 Caused Isolation. 16 No No No of 1 JTSC. Later Sotn

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - P r io r t o Com:ne r c ial O pe ra t ion- - - - - - - - - - - - -

12-16-77 ICS in Manual to RC? Overfed "3" 0TSG. 11 No No No Operator had MFW Pu:np in Hand

' 12-30-77 Loss of Feedwater to RCP WP Tripped on 31gh 72 No D No Exhaust Casing Water

, Level  :

1 1-21-78 Loss of Feedwater Manual Sequential Loss of N 70 No No No

• 1-31-78 I Loss of Feedwater ili RCP Spurious STRCS Trip 67 No No No j Af ter Perfor:ning STRCS l Monthly Test l 3-1-78 Loss of Feedwater I R1 RCP STRCS Actuated on 49 No No No W/STM Pressure 4P; l Deserator Level Cont.  !

Valve Failed Shut 4- 2-78 Turbine Trip Lo RCP TT Test - During 75 No No No Runback, Rx Tripped, ,

Overfed OTSGs 4

6-2-78 Overcooling to RCP W Oscillations 40 No No No 9-10-78 Turbine Trip to RCP Tripped Turbine for *b75 No No No Test TP-800-14 l

9-28-78 Instrument Failure to RCP Loop 2 RCS Flow XMTR '

90 No No No Failed Low, Runback

@ 202/ Min Initiated. l-Operator Lost Control 10-3-78 Turbine Trip to RCP TT Caused by Starting 68 No No No 2nd EHC Pump. OTSGe  :

Overfed I i

10-29-78 loss of Feedwater to RCP EM Relief Cycled and 4 No No No Stuck Open Too Long l

1-12-79 Loss of Feedwater Hi RCP Technician Shorted 100 No No No Inverter Causing Loss  !

of Vital Bus T2; STRCS Trip l I

2-22-79 Loss of Feedwater Manual Malfunction in Turbine 87 No No No l

Speed Control System Led to STRCS Actuation

• RCP is Reactor Coolant Pressure D** lD lD U Ml -

.me n e MUE h@L .

!PO 147

. II-l ATTACHMENT II E

Reouest 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 followinq items:

a. The cause of each event;

b. The initial power level prior to the transient; c.

Indicate safety whichtoofopen; valves these and, transients caused the PORY and/or

d. If the old (pre-TMI-2) setpoints for high RCS pressure and

' PORV actuation were in effect at the time of these transients, estimate whether any or all of the following would have taken place:

(1) PORV actuation; (2) Reactor trip on high RCS pressure, and ,

(3) Lifting of the safety valves.

Response

This information has been compiled and is presented in the attached tables.

The hypothetical actuation of the PORV and/or safety valves and high pressure trip assuming the old s.etpoints had been in effect is not based on any analytical technique. Rather,'Engineerina judgement, coupled with past operating history led to these results.

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a REACTOR 1 RIPS S1.\CE BI!-2.DWtS-Bt..; .L -1 Il-1 If 01d SetpoIr,ty_Fid p g Uspd Date Transient Classificaticn Initial PCRV P2R PCRV Trip on Hisn Lift Saicty Cause of Trarsier.t Power Lifted? Safety Actuation? Pressure? Valves?

Level Valves e lifted *

. 9-18-79 Turbine Trip Porturbation In 99.3 No No Ves No No EIC Fluid Pressure 9-26-79 Turbine Trip Failure Of Power 100 No No Yes No No Supply For Turbine Throttle Pressure Limiter XMTR O

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ATTACHMENT III (3 PAGES) ,

Request 3:

Provide an estimate of the increase in reactor trip frequency since lowering the high pressure trip setpoint and adding the anticipatory reactor trip. Include a review of the design criteria for the number of reactor trips over the plant life and evalu' ate the ef fect 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).

Response

A. An increase in reactor trip frequency will result from a lowering of the high pressure trip setpoint and the addition of the

1. anticipatory reactor trip. To estimate the effect of these changes, total reactor trips were divided into two (2) categories:

1. Category 1 - Trips that should not be affected by the above changes (e.g., total loss of.feedwater, since this led almost invariably to a reactor trip with the old setpoints; power to flow trips; test trips; etc.)

2. Category 2 - Trips that are affected by the above changes (e.g., high pressure. trips, feedwater upsets, an'i turbine trips).

Category 2 trips are listed in the following table (Table 1) in the "A" columns while total trips (Category 1 plus Category 2) are listed) in the "B" columns. The number of trips and frequencies in Table 1 are based on commercial calendar time. The post-TMI-2 frequencies should be viewed with caution as they are based on only a short operating history following the setpoint changes.

Ideally (i.e., with a large enough data base), the Category 1 trip frequency ("B" column minus the "A" column) should be similar for the pre- and post-TMI-2 periods. The fact that this is

' not the case can be attributed to the difference in calendar time and also statistical variations of the pre- and post-TMI-2 sampies.'

• However, on the average, it can be observed tha t the trip frequency in Catrgory 2 (i.e., "A" column) has increased by approximately a factor of 3 (0.23 to 0.71); which would almost double the total trip f requency.

Thus, although the data indicate approximately a doubling of tSe average trip frequency, the following must be considered-

1. There has been a relatively short period of operation since the changes.

2. There have been many startups and shutdowns during the post-THI-2 period.

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III-2 ATTACHMENT III (Cont'd)

3. As operators become familiar with the revised setpoints and operating conditions, it is reasonable to assume the trip frequencies may' decrease.

B. The structural design criterion for the number of reactor trips over the life of the plant is to keep the f atigue 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 contributions 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 trip frequency (indicated by the average data to date) should only have a small effect relative to plant life.

5As a part of the total allowable transient picture, 400 reactor trips are specified. Assuming a 40-year life, this translates into 10 trips per year or .83 per month. With the pre-TMI-2 setpoints, only the most ecent plants to come on line (Davis Besse-1, Crystal River-3, and TMI-2) exceed this figure. As these plants accumulate operating experience, their trip frequencies would be expected to decline under the pre-TMI-2 setpoints.

With the new setpoints, three plants exceed 0.83; however, for the reasons discussed in A above, it is premature to draw any conclusions over the life of the plant based on the little data available with these setpoints.

C. To determine the acceptable number of reactor trips in terms 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 reanalyzed based on actual transients and the limiting component would be a function of these actual transients plus those that would be expected throughout the remainder of the plant's life.

It is important to recognize tha t usage f actors below 1.0 represent design margin in the plant design. Any change that

, increases the frequency of transients causes a decrease in this margin whether the actual limit is reached or not.

Therefore, steps (such as raising the high pressure trip setpoint, etc.) should be considered to reduce the trip frequency thereby improving design and safety margins. .

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I11-3 TABLE 1 - ITifC Di' I'lT!Si 3 SIT-lOl'TS "

AND ANTICIPATORY RIXTC4: T;tIP (N 1 RIP filalNY g - .

DATE OLO.11\R TRIP IMTL Ot[TICAL W1111 CAILWR PLANT 02 m iCIAL TRIP TDE (nWS) M). OF TRIPS FREQLEN.Y (TRIPSAD.) NIX SET-FO:NTS TD:E (nWS) NO. OF 11 TIPS FRfQA.NOf (IRIPS/D.)

PRE-BII-2 POST-Dt!-2 I A B A B A B A B GNIE- 1 7-15-73 2082 31 52 .45 .76 , 5-18 79 125 2 4 .49 .97 QN.E- 2 9-9-74 1661 9 28 .16 .51 6-3-79 109 2 3 .56 .84 4% G 3 12-16-74 1563 17 27 .33 .53 -- -- - - - -

W IS EESSE I 11-21 77 492 5 23 .31 1.42 7-11 79 93 2 2 .65 .65 Af5TAL Al\U 3 3 13-77 745 8 28 .33 1.14 7-29-79 57 5 5 2.67 2.67

. .Ti) St.C0 4-17-75 1441 4 16 .0S .34 4-22-79 185 3 5 .49 .82 L

e.4M 12-19-74 1560 6 24 .12 .47 6-20-79 120 2 2 .51 .51

.MI-I 9-2-74 166S 3 6 .05 .11 -- -- - - - -

t! 2 32-30-78 88 1 3 .35 1.04 -- -- - - - -

Ji/4 --

11,300 84 207 .23 .56 --

689 16 21 .71 .93 A - consiJers only high pressure, IV upset, arJ turbine trip events (pre-Till-2 trips exclude IDFW) 8 - consiJers all trips Trip ircquer.cy (trips /m.) = hd$( g,) X 3h' {L - ,

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