ML19282C738
| ML19282C738 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/27/1979 |
| From: | Roe L TOLEDO EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 497, NUDOCS 7904300444 | |
| Download: ML19282C738 (2) | |
Text
.
APPENDIX H .
TOLECO TELECOPIED LOWELL E. ROE April 27, 1979 7,*LZ"a"1. .
(419) 2S9-5242 Docket No. 50-346
= License No. NPF-3 Serial No. 497
~
= Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Denton:
In your meeting of April 24, 1979 with representatives of Babcock & Wilcox and four licensees, including Toledo Edison, who have B&W nuclear steam supply systems, a number of concerns were expressed by you and your staff regarding certain features of the B&W plants. These concerns were further detailed in your NRR Status Report on Feedwater Transients in B&W Plants of April 25, 1979. In this report, on page 1-7, certain suggested steps were outlined which, if taken, would provide assurance to you that the B&W plants could continue to operate without undue risk.
While we feel that a number of design features already incorporated in the Davis-Besse Unit i fully meet or exceed the criteria you are requesting and that Davis-Besse can be operated without undue risk, we are proposing the following actions:
. t A. Auxiliarv Feedvater System Reliability and Performance The auxiliary feedvater system for the Davis-Besse Unit 1 is a
- reliable full safety grade system with redundancy for meeting the single failure criteria. The principal features are detailed in Table. 2.1 of your report.
We, however, will continue to review all aspects of this system to further upgrade components for added reliability and performance.
One such ites is an installation of dynamic braking on the auxiliary feed pu=p turbine speed changer to further minimize level fluctuation in the steam generator when on auxiliary feed.
B. Integrated Control System (ICS) Influence on Auxiliary Feedwater Control The Davis-Besse auxiliary feedwater control system is a full safety grade system completely independent of ICS. The auxiliary feedwater caster control is capable of being switched to ICS for a backup means of control, but this option is to be re=oved iz=ediately by administra-tive procedures. i 7904300 @
THE TCLECO EDISON CCMPANY ECISCN PLAZA 3C0 MACISCN AVENUE TCLECC. CMO 10552
Mr. Harold R. Denton, Director Page 2 ;
April 27, 1979 .
i C. Anticinatory Scram of Reactor Addition of a hard wired control grade reactor trip on loss of main feedwater or turbine trip.
D. Small Break Analysis Work with B&W to complete the analyses for potential small breaks and
- develop and implement any necessary operating procedures to define .
operator action.
E. Operating Procedures and Ocerator Training ,.
All procedures needed to be developed or modified by actions A thru D will be completed and training of the operators in the procedures will be done. All licensed shift operators will have received B&W simulator training on the IMI-2 incident.
All of the proposed actions outlined in A thru D above would be taken prior to start-up from the current =aintenance outage. l Toledo Edison will continue efforts to provide additional improvements related to A thru D as follows:
A. Continue to review performance of the system for assurance of reliability and performances B. The failure mode and effects analysis of ICS is under way with priority by B&W and will be submitted as soon as possible.
C. The reactor trips will be revised to safety grade as far as possible.
D. Continuing attention will be given to transient analysis and procedures for management of small breaks.
1 -
E. Continue operator training and retraining as a part of our ongoing program to continue to assure the high state of readiness of our operating staff. ,
We are confident that these actions on our part will satisfy your concerns and provide additional and full assurance for public safety.
Yours very truly, t .- r -
pg / G Lowell E. Roe Vice President Facilities Development The Toledo Edison Company LER.r
APPENDIX I Metropolitan Edison Company y
,gj
,e K r'f ( J l
Post Office Box 480 Middletown, Pennsylvania 17057 717 944-4041 April 16,1979 GQL 0527 On -
-lear Reactor Regulation IL . . aarold Denton, Director
. U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Docket No. 50-320 License No. DPR-73 Attached is a Preliminary Sequence of Events spanning the first approximately twenty hours following the TMI-2 accident which was initiated at 4:00 a.m. on March 28, 1979.
For this chronology of events, a reference clock was established with the time of the turbine trip, 0400:37, defined as time zero. The time of each event in the sequence is given as the number of hours, minutes and seconds relative to 0400:37, followed in parenthesis by the real time using a 24-hour clock. For example, 1:52:43 p.m. on March 28 would be written "9:52:06 (1352:43)". Depending upon the accurecy of the source of data for .each event, the times appear alone or with the notation " approximate".
The sequence has been reconstructed from various infomation and data sources, including control room logs, strip chart recorders, alam printouts and reactimeter printouts. Please note, however, that the alarm printer was out of service from 01:13:27 (0513:59) to 02:47:31 (0648:08) and during the course of the accident
. was running well behind the actual time of events. Efforts to annotate this chrono-logy and to develop graphs of various plant parameters as a function of time are
, unde rway. This additional infomation will be provided as soon as it is available and we will keep you infomed of our progress.
Sincerely, 1
1 1
. G. Herbeln ice President-Generation JGH:RAL:djh Enclosure Metronniitan R$enn rnmnany .e a va or nf tra nana ai p. +ne i r. >+c.c cetom -
,' PRELIMINARY SEQUENCE OF EVENTS TMI 2 ACCIDENT OF MARCH 28, 1979 Issued April 16, 1979
-00:05:00 Three Mile Island Unit Two was at 97% power with the Integrated (0355:36)
Control System in full automatic. Rod groups one thru five were fully withdrawn, rod groups six and seven were 95% withdrawn and rod group eight was 27% withdrawn. Reactor Coolant System total flow was approximately 107.5% of design flow and the Reactor Coolant Syster. pressure was 2155 psig. Reactor Coolant Makeup Pump B (MU-P-1B) was in service supplying makeup and Reactor Coolant .-
Pump Seal injection flow. The Reactor Coolant System soluble boron concentration was approximately 1030 parts per million. Pressurizer Spray Valve (RC-V1) and the pressurizer heaters were in manual ,
centrol while spraying the pressurizer to equalize boron concen-trations between the pressurizer and the remainder ,of the Reactor Coolant System. Normal Reactor Coolant System letdown flow was established.
Steam Generator parameters were as shown in the following table:
Steam Generator A Steam Generator B -
Loop Feedwater 5.7459 MPPH* 5.7003 MPPH*
- Operating Level 56% 57.4%
- Startup Level 158.8 inches 163.4 inches Steam Pressure 910 psig 889.6 psig Feedwater Temperature 462.7F 462.7F
- MPPH is Million Pounds Per Hour
Steam Generator Feedwater .Penps (FW-P-1 A and FW-P-1B) were in service, Condensate Booster Pemps (CO-P-2A, CO-P-2B and CO-P-2C) were in service, and Condensate Pamps (CO-P-1A and CO-P-1B) were in service. An attempt was being made to clear a clogged resin transfer line in the standby demineralizer.
-00:00:01 Condensate Pump A (CO-P-1A) stopped.
,- (0400:36)
-00:00:01 Feedwater Pumps (FW-P-1A and PW-P-1B) stopped at essentially the
~
(0400:36) same time resulting in a loss of feedwater flow to both steam generctors.
00:00:00 Main Generator was tripped followed by a turbine trip.
(0400:37) 00:00:00 Three Emergency Feedwater Penps (IF-P-1, 2A, 2B) started .
(0400:37) 00:00:03 The Electromatic Relief Valve (RC-RV2) opened at the setpoint (0400:40)
Approximate of 2255 psig.
00:00:08 Reactor tripped on high pressure at 2345 psi. Setpoint is 2355 psi.
(0400:45) 00:00:08 The operator placed the Pressurized Spray Valve (RC-V1) and pres-(0400:45)
Approximate surizer heaters under automatic centrol.
00:00:13 The operator started the Reactor Coolant Makeup Pump A (MU-P-1A),
opened High Pressure Injection Is:lation Valve A (MU-V16A) and isolated letdown flow in anticipa: ion of the expected pressurizer level decrease.
.. p, 00:00:13 The Electromatic Relief (RC-RV2) solenoid de-energized giving (0400:50)
Approximate a non-open indication to the control room operators. The Elec-trematic Relief Valve (RC-RV2) should have reserted at about this time (closure setpoint of 2205 psig).
00:00:14 The Emergency Feed Pumps (EF-P1, 2A and 2B) achieved normal dis- ,
(0400:51) '
charge pressure.
00:00:15 Water hammer in the condensate piping occurred.
(0400:52)
Approximate 00:00:30 Pressurizer Safety Valve (RC-RV13) and Electromatic Relief Valve (0401:07)
(RC-RV2) discharge line temperature alarms printed out.
00:00:38 Steam Generator A level reached the 30-inch setpoint where the (0401:15)
Approximate Emergency Feedwater Valves (EF-VilA and EF-VilB) open. Feedwater was not admitted because Emergency Feedwater Block Valves (EF-V12A and EF-V12B) were shut.
00:00:39 Reactor Coolant Makeup Pump A (MU-P-1A) was stopped. .
(0401:16) 00:00:40 Steam Generator B level reached the 30-inch setpoint where the +
(0401:17)
Approximate Emergency Feedwater Valves (EF-VilA and EF-V11B) open. Feedwater
- was not admitted because Emergency Feedwater Block Valves (EF-V12A and EF-V12B) were shut.
00:00:41 Reactor Coolant Makeup Pump A (MU-P-1A) was restarted. With (0401:18)
Reactor Coolant hakeup Pumps A and B (MU-P-1A and MU-P-1B) oper-ating, pressurizer level rate of decrease slowed.
..e
{x 00:01:00 Pressurizer level started ~ increasing. Reactor Coolant System hot (0401:37)
Approximate leg and cold leg temperatures reached 575F. Reactor Coolant Drain
~
Tank pressure was increasing.
00:01:00 The Pressurizer Safety Valve (RC-RVlA) high discharge line temper-(0401:37) ature alarm was received.
00:01:26 Reactor Coolant Drain Tank temperature normal alarm printed out.
(0402:03) s 00:01:45 Steam Generators A and B have boiled dry at this time.
(0402:22)
Approximate 00:02:01 Reactor Coolant Makeup Pump B (MU-P-1B) was stopped due to (0402:38)
Engineered Safeguards actuation.
00:02:04 High Pressure Injection Pump C (MU-P-lC) started automatically.
(0402:41) 00:03:12 Reactor Coolant Drain Tank Relief Valve (WDL-RI) lifted at 120 psig.
(0403:49)
Approximate 00:03:14 High Pressure Injection portion of Engineered Safeguards was manually
, (0403:51) bypassed. Both Reactor Coolant Makeup Pumps A and C (MU-lP-1A and MU-P-lC) were operating.
00:03:26 Reactor Coolant Drain Tank high temperature alarm received at 127.2F.
(0404:03) 00:04:38 Reactor Coolant Makeup Pump C (MU-P-lC) was stopped.
(0405:15) 00:04:38 The operator throttled the High Pressure Injection Isolation Valves (0405:15)
Approximate (HU-Vl6's).
p, 00:04:52 Intermediate Closed Cooling Pump (IC-P-1A) started.
(0405:29) 00:04:58 First alar = indication received that letdown had been secured.
(0405:35) 00:05:06 Presurizer level stopped its sharp increase at 376 inches and (0405:43) began to turn down. It reached a minimum of 372 inches and then started back up at 5 minutes, 21 seconds into the transient.
00:05:15 Condensate Booster Pump B (CO-P-2B) tripped.
(0405:52) 00:05:50 Reactor Coolant System pressure stopped its sharp decrease and began (0406:27)
Approximate to turn up. Minimum value reached was approximately 1350 psig.
00:05:54 Pressurizer level increased beyond the range of the ins:rument (0406:31) indication.
00:06:58 Letdown flow of 71.4 gallons per minute was re-established.
(0407:35) 00:07:31 Reactor Building Sump Pump A (WDL-P-2A) started.
(0408:06) .
00:08:00 Emergency Feedwater Block Valves (EF-V12A and EF-V12B) were opened.
(0408:37) .
Approximate 00:08:15 Reactor Coolant System hot leg and cold leg temperatures began to (0408:52) decrease.
00:08:30 Reactor Coolant Systen pressure began to decrease.
(0409:07) 00:10:00 Pressurizer level came on scale.
(0410:37) g 00:10:19 Reactor Building Swsp Pusp B (WDL-P-2B) started.
(0410:56) 00:10:24 Reactor Coolant Makeup Pusp 1 (MU-P-1A) tripped.
(0411:01) '
00:10:27 Reactor Coolant Makeup P=sp 1 (MU-P-1A) was started.
(0411:04) 00:10:28 Reactor Coolant Makeup Pesp i (MU-P-1A) tripped.
(0411:05) s 00:10:40 Reactor Building Sump high level alarm received. Setpoint is (0411:25) 4.650 feet.
00:11:40 Reactor Coolant Makeup Pe=p A (MU-P-1A) was started.
(0412:17) 00:14:50 The Reactor Coolant Drain Tani rupture diaphragm (WDL-U26) failed.
(0415:27) 00:24:58 The operator requested cesputer printout of the Electromatic (0425:35 Relief Valve (RC-RV2) outlet :emperature. The reading was 285.4F.
00:25:00 Intermediate Cooling System high radiation alarm annunciator (0425:37)
Approximate received at the Radiation Monitor Panel.
00:36:08 L=ergency Feedwater Pump 23 (IF-P-2B) was stopped.
, (0436:45) 00:38:10 Reactor Building Sump Pu=p A ,7DL-P-2A) was stopped.
(0438:47)
-00:38:11 Reactor Building Sump Pe=p 3 .7DL-P-2B) was stopped.
(0438:48) 01:10:54 Reactor Building air cooling : oils emergency discharge alarm (0511:31) printed out.
. - ,.. . p w
01:13:29 2eactor Coolant Pump 2B (RC-P-2B) was stopped.
(0514:06) 01:13:42 Zeactor Coolant Pump 1B (RC-P-1B) was stopped.
(0514:19) 01:13:27 The alarm printer became unavailable at this time and remained (0513:59) out of service until 02:47:31 (0648:08).
01:20:31 Operator requested printout of the Electromatic Relief Valve (RC-RV2) outlet temperature. The reading was 283.0F.
- 01:40:37 Inactor Coolant Pump 2A (RC-P-2A) was stopped.
(0541:14) 01:40:45 Inactor Coolant Pump 1A (RC-P-1A) was stopped.
(0541:22) 01:42:00 Operator started raising Steam Generator A level from 30 inches (0542:37)
Approximate en the Startup Range to 50% on Operating Range. Reactor Coolant System Loops A and B cold leg temperatures both started decreasing.
Eaactor Coolant System pressure started decreasing.
01:54:00 Raaetor Coolant System Loop A hot leg temperature began increasing. '
(0554:37)
Approximate .
02:00:00 Steam Generator A level reached 50% on Operating Range. ,
(0600:37)
Approximate 02:00:00 Raactor Coolant System Loop B hot leg temperature began increasing.
(0600:37) 02:12:00 Raactor Coolant System Loop B hot leg temperature increased to (0612:37) offscale at 620F.
p, 02:17:53 Operator requested Electromatic Relief Valve (RC-R2) outlet (0618:30) temperat ure . The reading was 228.7F.
02:22:00 Ihe Electromatic Relief Block Valve (RC-V2) was shut.
'0622:37)
Approxisate 02:30:00 Operator started increasing Steam Generator B from 30 inches ==
, (0630:37)
Startup Range to 502 on Operating Range.
. 02:45:00 Several radiation alarms were received.
(0645:37)
Approximate 02:45:00 Reactor Coolant Makeup Pump C (MU-P-lC) was stopped.
(0645:37)
Approxi ate 02:45:00 Operator opened Main Steam Isolation valves (MS-V4B and MS-773?.
(0645:37)
Approxina te 02:50:C0 Site Emergency was declared. Notifications to offsite au-heri:ies (0650:37)
Approx i=a t e and organizations were initiated.
. 02:51:57 Operator attempted to start Reactor Coolant Pump 2A (RC-P-21).
(0652::4)
? cmp would not start.
02:53:19 Operator attempted to start Reactor Coolant Pump 1B (RC-P- 33.
(0653:53)
?c=p wouid not start.
02:54:C9 Operator started Reactor Coolant Pump 2B (RC-P-2B).
(0654:45) 02:54:49 High Pressure Injection Engineered Safeguards actuation logic (0655:25) reset on increasing Re4ctor Coolant System pressure.
.. -8 -
(
02:56:19 Steam Generator B was isolated. Main Steam Isolation Valves (0656:56) (MS-V4B and MS-V7B) were shut. .
Approximate 03:00:00 Reactor Coolant System pressure increased to 2130 psig.
(0700:37)
Approximate 03:03:39 Steam Generator A pressure control was shifted from the Turbine Bypass (0704:16)
Approximate Valves (MSV-25A and B and MSV-26A and B) to the Power Operated
- Emergency Main Steam Dump Valves (MSV-3A and B).
03:10:27 Emergency Feedwater Pump 2A (EF-P-2A) was stopped.
(0711:04) 03:12:28 Electromatic Relief Block valve (RC-V2) was opened.
(0713:05)
Approximate 03:12:53 Reactor Coolant Pure 2" (RC-P-2B) was stopped.
(0712:53) 03:20:13 Reactor Coolant Mtkeup Pump C (MU-P-IC) was started. Reactor Coolant (0720:41)
Makeup Pumps C ane: A (MU-P-C and A) were operating.
03:23:13 Ceneral Emergency was declared. Notifications to offsite (0724:00)
- Approximate authorities and c rganizations were initiated.
03:30:00 Electromatic Relief Block Valse (RC-V2) was shut. *
(0730:37) ,
Approximate 03:35:08 Emergency Feedwiter Pump 2A (EF-P-2A) was started.
(0735:43) 03:37:00 Reactor Coolant Makeup Pump C (MU-P-lC) was stopped.
(0737:37) 03:51:00 Electromatic Relief Block Valve (RC-V2) was opened.
(0751:37)
Approximate e
Ep.
. 03:55:39 Engineered Safeguards actuated on low RCS pressure. Setpoint is (0756:16) 1640 psig.
03:55:39 The Reactor Building high pressure isolation signal actuated (0756:16) and isolated the Reactor Building. The Reactor Building isolation set point is 4 psig.
03:56:04 Reactor Coolant Makeup Pump C (MU-P-lC) was started.
(0756:41) s 03:59:23 Reactor Building Emergency Cooler B was shutdown.
(0800:00) 03:59:53 Reactor Building Emergency Cooler B was started. '
(0800:30) 04:06:00 Electromatic Relief Block Valve (RC-V2) was shut.
(0806:37) 04:08:37 Reactor Coolant Pump 1A (RC-P-1A) was started.
(0809:14) 04:09:14 Reactor Coolant Puwp 1A (RC-P-1A) was stopped.
(C809:51) 04:17:17 Reactor Coolant Makeup Pump A (MU-P-1A) was stopped.
(0817:54) 04:17:22 Reactor Coolant Makeup Pump C (MU-P-1C) was stopped. No makeup (0817:59) pumps operating.
04:18:17 Operator attempted to start Reactor Coolant Makeup Pump A (MU-P-1A) .
(0818:54)
The pump would not start.
04:18:30 Electromatic Relief Block Valve (RC-V2) was opened.
(08L9:07)
Approximate
(
04:21:53 Reactor Coolant Makeup Pump B (MU-P-1B) was started.
(0818:30) 04:26:59 F.eactor Coolant Makeup Pump C (MU-P-1C) was started, tripped, (0827:36)
Approximate and was restarted.
04:30:00 The Electrometic Relief Block Valve (RC-V2) was shut.
(0830:37) .
Approximate 04:30:45 Condenser Vacuum Pumps IA and 1C (VA-P-1A and VA-P-lC) were *
(0831:22) stopped and vacuum was broken.
04:30:45 Power Operated Emergency Main Steam Dump Valve (MS-V3A) was opened.
(0831:22)
Approximate 04:54:00 The Electrasatic Relief Block Valve (RC-V2) was opened.
(0854:37)
Approximate 05:18:00 The Electromatic Relief Block Valve (RC-V2) was shut.
(0918:37) 05:54:00 operator commenced filling Steam Generator A to 99% on the Operating (0954:37) -
Approximate Range instrumentation.
07:30:00 Electromatic Relief Block Valve (RC-V2) and the Pressurizer Spray *
(1130:37)
- Approximate Valve (RC-VI) were opened.
08:11:26 Core Flood Tank A high level alarm was received.
(1212:03) 08:30:00 Power Operated Emergency Main Steam Dump Valve (MS-V3A) was shut.
(1230:37)
. . . . p
__ _ - A 08:31:06 Decay Heat Removal Pumps lA and IB (DH-P-1A and DH-P-13) were (1231:43) started.
08:54:56 Core Flood Tank A alarm printed out at a level of 13.13 feet.
(1255:33) 09:04:18 Reactor Coolant Makeup Pump C (MU-P-lC) was stopped.
(1304:55) 09:49:44 Reactor Building Isolation and Containment Spray were actuated by
., (1350:21)
Engineered Safeguards. Engineered Safeguards actuation started Reactor Coolant Makeup Pump C (MU-P-lC) and Reactor Building Spray Pumps A and B (BS-P-1A and BS-P-1B). .
09:49:50 Reactor Building Spray Valves (BS-VlA and BS-VilB) opened.
(1350:27) 09:49:58 Reactor Coolant Pumps lA and IB (RC-P-1A and RC-P-1B) inlet air (1350:35) temperature high alarms annunciated and Pressurizer Safety Valves (RC-RIA and RC-RIB) discharge line temperature high alarms annun-ciated.
09:50:24 Reactor Coolant Makeup Pump C (MU-P-1C) was stopped.
. (1351:01) 09:55:30 Reactor Building Spray Pumps A and B (BS-P-1A and BS-P-1B) were
. (1356:07) stopped.
09:56:58 Decay Heat Removal Pumps A and B (DH-P-1A and DH-P-1B) were (1357:35) stopped.
10:24:00 Reactor Coolant System hot leg Loop A terperature decreased to (1424:37)
Approximate within the instrumentation range.
,3 10:31:25 Reactor Coolant Makeup Pump C (MU-P-lC) was started. Reac tor (1432:02)
Coolant pressure was approximately 440 psig.
10:35:55 Reactor Coolant Makeup Pump C (MU-P-LC) was stopped.
(1436:32) 11:06:00 Pressurizer level started decreasing.
(1406:37) .
Approximate
- 11:12:00 Reactor Coolant System cold leg Loop A temperature started to .
(1512:37)
Approximate increase from 200F to 400F. Reactor Conlant System hot leg Loop A temperature decreased from above the instrument range to 560F.
11:18:34 Reactor Coolant Mckeup Pump C (HU-P-lC) was started.
(1519:11) 11:24:00 Pressurizer level stopped decreasing at 180 inches and started (1524:37)
Approximate increasing, going off scale during the next cour.
11:28:12 Reactor Coolant Makeup Pump C (MU-P-1C) was stopped.
(1528:49) 11:32:37 Reactor Coolant Makeup Pump C (MU-P-lC) was started.
(1533:14) -
11:35:48 Reactor Coolant Makeup Pump C (HU-P-1C) was stopped.
(1536:25)
- 11:36:00 Operator commenced filling Steam Generator B to 972 on the Operating (1536:37)
Ap pr oxi=a t e Range instrumentation.
12:00:00 Steam Generator A level was 97% on the Operating Range.
(1600:37)
Approximate 12:48:00 Pressurizer level came on scale.
(1644:00)
Approxima te
g
/
13:02:23 Concenser Vacuum Pump 1C (VA-P-lC) was started.
(1703:00) 13:08:22 Normal steam generator feedwater supply was put in service.
(1708:59) <
Approximate 13:13:10 Condenser Vacuum Pump 1A (VA-P-1A) was started.
(1713:47) 2 13:23:04 Reactor Coolant Makeup Pump C (MU-P-1C) was started.
(1723:41) s 14:43:15 Reactor Coolant Makeup Pump C (MU-P-lC) was stopped.
(1843:52) 14:54:00 RCS pressure reached 2350 psig. -
(1854:37)
Approximate -
15:24:00 Reactor Coolant Pump 1A (RC-P-1A) was started.
(1924:37) 15:24:10 Reactor Coolant Pump 1A (RC-P-1A) was stopped.
(1924:47)
Approximate 16:04:00 Reactor Coolant Pump 1A (RC-P-1A) was started.
, (2008:37) 22:15:00 Reactor Coolant System and Steam Generator conditions were:
. (0215:37)
Approximat e Reactor Coolant System pressure = 1065 psig.
Pressurizer Temperature = 551F (pressurizer heaters maintaining temperature).
Pressurizer Level = 397 inches.
Reactor Coolant System cold leg Loop A temperature = 288F Stehm Generator A steaming to the Main Condenser.
Steam Generator B isolated.
Reactor Coolant Makaup Pump B (MU-P-1B) operating to supply Reactor Coolant ? cap seal injection flow.
Reactor Coolant System cold leg Loop A temperature = 256.4F.
Reactor Coolant System cold leg Loop B temperature = 252.4F.
Reactor Coolant System hot leg Loop A temperature = off scale low, '
i.e. ,less than 520.0F.
Reactor Coolant System hot leg Loop B temperature = off scale low, i.e. ,less than 520.0F.
s s
e D
e t
i
?
APPENDIX J COMMENTS ON CRYSTAL RIVER FEEDWATER SYSTEMS Introduction These comme'nts were compiled during the. week,of April 15, 1979 from information in the FSAR and from telecons with the licensee.
Summary Concerns about the design of Crystal River's auxiliary feedwater system Tre:
(1) Seismic event could cause loss of all AFW pump suction sources.
. (2) The AFW pumps may not self-vent because of the s'ystem geometry.
(3) The AFW oump auto start logic is not single failure-proof. .
. (4) Vacuum breaker valves on main condenser can cause loss of suction to -.
both AFW pumps from hotwell. ,
(5) For several scenarios with single failures, operator action would be required to get auxiliarv feedwater to the steam generators.
Normal Feedwater_
Two turbine driven pumps with steam sources from (1) reheat steam; (2) main steam; and (3) auxiliary steam. Shutoff head = 2550 ft. No s' trainers in feedwater system. Condensate demineralizers are automatically bypassed by an air-operated valve on high differential pressure across the demineralizers.
(This valve could fail in ~any position on loss of air because it uses air as ,
its motive force in both directions.) There are no automatic bypasses around FW heaters. Feedwater;is shut off to the faulted steam generator when its steam pressure < 600 osig. ,
Auxiliary Feedwater (AFW) Sources Mrmal supply: condensate storage tank; first backup supply: condenser hotwell;
, se ond backup: demineralized water from the fossil units. Switchover from the nomal supply to the first backup can be performed from the control room in
- approximately 1 minute. Concern: There is no seismic category 1 source of auxiliary feedwater. .
Auxiliary Feedwater Pumps Two pumps,1 motor driven 'and 1 turbine driven, 740 gpm each. Shutoff head:
3400 ft (motor) and 3500 ft (steam). Concern: The pumps are not the low point in the system and may not be self-venting.
Auxiliary Feedwater Pump Drives The motor is on a Class lE power supply. Steam is supolied to the turbine driver from two main steam lines upstream of the main steam isolation valves.
The turbine driven pump is operable with steam pressure at least as low as 200 psig.
AFW Pumps Auto Start The motor driven pump has no auto start signals. It must always be started by the operator. Turbine driven pump auto starts on loss of both main FW pumps (as sensed by low oil pressure on both pumps). (Does not start on safety features actuation signal ) Auto start signals are not redundant or Class 1E. .
Automatic Trips of AFW Pumps Turbine driven: overspeed Motor driven: motor protective trips , , _
closed suction valve Concern: If taking suction on the hotwell (first backup), the suction valves are interlocked with the condenser vacuum breaker valves. If they are closed, the suction valves close and you lose suction to the AFW pumps. Only the motor-driven pump trips on suction valve position. The turbine-driven pump could be damaged before it trips on overspeed.
AFW Indication .
Turbine driven: steam stop valve position .
Motor driven: motor on-off lights, ameter Common: flow in startup FW line (would reoutre valve realignment to use). - '
Level Control -
On loss of main FW pumps, the ICS controls level at 30" after the operator closes a valve that bypasses the FCV. If all 4 reactor coolant pumps are lost, the ICS controls level at 250"after operator closes FCV bypass.
Operating procedures and practice require the operator to maintain these levels if the ICS fails to do sn.
Independence of AFW Trains Appear to be independent with the following exceptions: '-
(1) common, non-seismic suction source (condensate , storage tank). ,
(2) ICS inputs to flow control valves of both trains.
(2) suction valves from main condenser to AFW pumos are closed by common signal (see concern under " Automatic Trips of AFW Pumps").
Effect of Surveillance Test on System '
To test one pump, its motor-operated discharge valve is closed and recirculated to condensate storage tank through the mini-flow line. Operator action would ,
be required to open the discharge valve before that cump would deliver water to the steam generators. (Note that only 1 AFtl pump starts on auto.) -
Comt.on Mode Failures That Would Cause Loss of Main FW and Auxiliary FW None identified (except seismic event).
Seistric Event (1) There is no seismic source of suction to the auxiliary feedwater pumps. Therefore, a seismic event could cause total loss of feedwater.
(2) The initiating logic for AFW pump is non-seismic. Therefore, the pump may not auto-start even if suction source is available.
Loss of Offsite Power Would cause almost immediate loss of both main FW pumps. Only the turbine- ,
driven AFW pump would auto start. Operator action would be required to start the motor-driven AFW pump on the diesel generator. Several emergency loads may have to be stripped to allow starting of this pump on the diesel generator.
Loss of Offsite Power with Single Failure
. (1) Worst identified single failure is loss of turb'ine-driven AFW pump.
This would require operator action to start motor-driven pump. Other emergency loads may have to be stripped before starting AFW pump on ~
diesel generator.
s (2) ICS - Have not investigated whether single failure can cause both AFW flow control valves to close after the operator has closed the FCV beypass valves.
System Design Response to LOCA None.
Alternate Cooling Mode Without Main FW cr AFW None identified.
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APPENDIX K COMMENTS ON RANCHO SEC0 FEEDUATER SYSTEMS Introduction These comments were compiled during the week of April 15,1979 from informaticn in the FSAR and from telecons with the licensee, Summary 1
(1) Loss of offsite power with a single failure in the turbine driven pump train requires operator action to provide water to the steam genera tors. .
(2) For a loss of main feedwater (or loss of all reactor coolant pumps) while performing surveillance test on one train, it'would require operator action to realign the train being tested to provide flow to the steam senerators. ,
(3) We' don't know if there are single failures in the ICS that could cause loss of both main and auxiliary feedwater or both trains of auxiliary feedwater.
(4) We are not certain tha' each train of auxiliary feedwater has the capacity assumed in th generic LOCA analysis. R. C. Jones of B&W informed us on April 2T , 1979 that the analysis assumes 500 gpm per steam generator (1000 pm total) at 1050 psig. We need the pump head curves to evaluate this.
Normal Feedwater 2 turbine driven pumps with steam sources from (1) reheat steam; (2) main steam; (3) auxiliary steam. Shutoff head = 2750 ft.
No strainer s in feedwater (FW) system. No automatic bypasses for condensate demineralizers or feedwater heaters. Feedwater is shut off to the faulted steam generator when its steam pressure <435 psig.
Auxiliary Feedwater Sources -
Normal supply: condensate storage tank (seismic category 1); first backup supply: canal (non-seismic); second backup: reservoir (non-seismic). There is a manual switchover from normal to backup that takes apprcximately 5 minutes.
Auxiliary Feedwater (AFW) Pumps Two pumps - 1 motor driven and 1 with both a motor and a turbine driver, S40 gpm each. Shutoff head, with steam 3050 feet, with motor 3100 ft.
AFW Pump Drives Steam supplied from main steam lines. The motors are on Class lE power supplies. Steam driven pump has been demonstrated operable with steam pressure to the turbine drive as low as 213 psig.
AFW Pumps Auto Start Both pumps start on either of the following:
(1) 1.oss of both main feedwater pumps as sensed by discharge pressure; each main FW pump (850 psig.
(2) All reactor coolant pumps off as sensed by the power monitor.
The turbine driven pump only also starts on Safety Features Actuation .
Signal (SFAS). Electrical power to all initiating signals is from -
Class lE sources. .
Automatic AFW Pump Trips
- Motors: electrical faults (breaker). Steam turbine: overspeed.
AFW Indication Motors:, on-off li'ghts, ammeters.
Steam supply valve position. (3 at separate control room locations)
Level Control On loss of main FW pump's, ICS controls at 30 inches. On loss of all reactor coolant pumps, ICS controls at 318 inches. Operating procedures and practice require operator to maintain these levels using manual control if ICS fails to do so.~ On SFAS, the AFW flow control valves are bypassed, delivering -
full flow from AFW pump (s) to the steam generators. (Only the turbine driven pump starts automatically on SFAS.)
Independence of AFW Trains '
Appear to be indep'endent with three exceptions:
(1) They have a common suction source (seismic category 1 condensate storage tank). ,
(2) Cross-tie valves between the discharges are normally open (remote.
manual MOV's from Class lE power supply). -
(3) ICS inputs to both flow control valves. '
Effect of Surveillance Test on System To test one system, the discharge cross-tie valve and flow control valve are closed from the control room. Operator action would be required to get AFW to that steam generator if auto demand signal was received.
Common Mode Failures That Would Cause Loss of Main FW and AFW The ICS is not Class lE. There may be single failures that would cause the main FW flow control valves and the AFW flow control valves to close and remain closed. (ICS does not inhibit SFAS controls.)
3-Seismic Event only effect on AFW system would be that if offsite power were lost as a result of the earthquake, only one AFW pump would auto start on demand.
The operator would have to manually start the motor driven pump on the diesel generator.
L oss of Offsite Power Would cause almost immediate loss of both main FW pumps. Only the steam e driven AFW pump would start automatically. Operator action would be required to start the motor driven pump on the diesel generator.
Loss of Offsite Power with Single Failure _
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Worst identified single failure would be loss of steam driven AFP.. This would require operator action ~to restore AFW by starting motor driven pump. Questions on ICS - Have not investigated whether single failure can cause both AFW flow control valves to close.
AFW System Design Response to LOCA When SFAS is initiated, the turbine driven AFW pump is started (regardless of whether main FW pumps or reactor coolant pumps are tripped). SFAS also opens bypass valves around the AFW flow control valves, thereby allowing the AFW pump to putx420 GPM into each steam generator. When steam generator level exceeds 30" (or 318" if operator has tripped RCP's), the auxiliary FW flow control valves are closed by the ICS. -
Alternate Cooling Mode Without 11ain FW or AFW Nuclear Service Cooling Water System. ,
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APPENDIX L COMMENTS ON OCONEE FEEDWATER SYSTEMS Introduction
- These comments were compiled during the week of April 15,1979 from information in the FSAR and from telecons with the licensee.
- Summary (1) Seismic event could cause loss of all 3 units' emergency feedwater pumps. .
(2) Several scenarios could result in feedwater not being supplied to the -
steam generators for 10 minutes or longer while operator manually realigns systems.from other units or the auxiliary service water pump. g (3) Auto start signal is not single failure proof or seismic Category 1.
(4) There is only one EFW train per unit. R. C. Jones of B&W informed us on April 22, 1979 that the generic LOCA analysis assumes 500 gpm per steam generator (1000 gpm total) at 1050 psig. Apparently this one train does have that capacity; however, there is no redundancy.
(5) EFW injection valves are powered by non-Class lE batteries.
(6) Technical specifications don't have operability requirements for other units' EFW systems.
Normal Feedwate~r Two turbine driven' main FW pumps with steam' sources;_from (1) extraction steam; (2) main steam; (3) auxiliary steam. Shutoff head = 1253 psia. There are
, suction strainers for the hotwell pumps. The condensate demineralizers are automatically bypassed by air-operated valves (fail open) on high differential pressure across the demineralizers (40 psi). There is no automatic action to isolate a staem generator. on break of main steam or feedwater lines.
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Emergency Feedwater Sources .
Normal supply: upper surge tank; first backup: main condenser hotwell; second backup: other units' upper sruge tanks (all sources non-seismic category 1).
Switchover from normal to first backup is remote manual and requires approximately 1 minute.
- Emergency Feedwater (EFW) Pumps One pump per unit - turbine driven. Capacity = 1080 gpm at 1050 psig.
Shutoff head = 1465 psia.
EFW Pump Drive Steam supply to turbine driver is from main steam. Pump will operate with steam / pressure to drive at least as low as 300 psig.
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EFW Pump Auto Start (1) Loss of both main FW pumps as detected by low header discharge pressure <750 psig or by main FW pump turbine stop valve position
. on both pumps.
(2) EFW pump does not start on SFAS.
(3) Auto-start signals are from non-Class 1E sources.
Automatic EFW Pump T' rips "
(1) Overspeed t (2) Low hydraulic pressure EFW Indication (1) Pump discharge pressure n
(2) Flow -
Level Control On loss of main FW pumps, the ICS controls steam generator level at 25". On loss of all reactor coolant pumps, the ICS controls level at approximately 260". Operating procedures and practice require operator to maintain these levels using manual control if the ICS. fails to do so.
Independence of EFW Trains Not applicable: only I train. Time required to align EFW from another unit '
is 10 minutes or longer.
Effect of Surveillance Test on System '
Close manual block, valves. Would require op,erator to reopen manual valves and close recirc valve to get EFW to steam generators if demanded during surveillancetes). '
Common Mode Failures That Would Cause loss of Main FW and EFW ICS is not Class 1E. There may be single failures that would cause the -
control valves that normally would feed both main FW and EFW to the steam generators to close. Operator action would be required to open an. air
- operated valve in a line which bypassed the flow control valves. ,
Seismic Event It appears that a seismic event could fail all three units' EFW pumps because of their location in a non-seismic Category 1 building, and fail all sources of suction for the EFW pumps.
Loss of Offsite Power -
Could cause almost immediate loss of both main FW pu-ps. The EFW pump would start automatically if main FW pumps tripped.
Loss of Offsite Power with Single Failure _
Assume single failure is the EFW pump for that unit. It would require operator at least 10 minutes to get water to the steam generators by manually realigning part of the EFW flow from the other units or to manually start the auxiliary service water pump.
AFW System Design Response to LOCA None.
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Alternate Cooling Mode Without Main FW or EFW
.' Auxiliary service water pump. One pump for the site. 3000 gpm.~ at 75 psig.
Shutoff head = 100 psig. Must be started manually. Takes suction from circulating water inlet line. Located in seismically-designed auxiliary
., building. Powered from Class lE source. __
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APPENDIX M COMMErlTS ON DAVIS-BESSE UtlIT 1 FEEDWATER SYSTEMS I ntroduction These comments were compiled during the week of April 15, 1979 from information in the FSAR and from telecons with the licensee.
S ummary (1) A single failure in an AFW train would require operator action to
- provide water to both steam generators. ~
(2) Apparently, each train of the AFW system has less capacity than assumed in the LOCA analysis. (R. C. Jones of B&W informed us on April 22,1973 that the analysis assumes 500 gpm per steam generator (1000 gpm total) at 1050 psig.)
(3) Suction strainers on both AFW pumps could possibly be blocked following seismic event by debris from the common, non-seismic category l- suction source.
Two turbine , driven pumps with steam supplies from (1) reheat steam, (2) main steam, (3) auxiliary steam. Shutoff head = 2560 ft. Condensate pumps have suction strainers. Main FW is isolated from both steam generators when one is faulted (steam and feedwater rupture control system). This system also starts the auxiliary FW pumps and aligns both to the good steam generator.
Auxiliary Feedwater Sources Normal supply: condensate storage tank (non-seismic category 1): fdrst backup supply: deaerator (non-seismic category 1); second backup: fire water system (non-seismic category 1); seismic category 1 supply: service water pump discharge. Auto transfer of either pump's suction to the seismic category 1
- source when on any of the other sources and get low suction pressure.
(Redundant Class lE pressure switches.)
A uxiliary Feedwater (AFW) Pumps Two pumps, both turbine driven. Each 1050 gpm at 1050 psig (250 gpm of this is recirc flow each pump). Shutoff head = 3150 ft.
Auxiliary Feedwater Pump Drives Steam supplied from the main steam lines upstream of MSIV's. Pumps demonstrated operable down to T = 280 F (Psat ' 50 psia).
AFW Pumps Auto Start Both AFW pumps start on any of the following signals:
(1) Steam pressure greater than feedwater pressure by 170 psi (for feedwater -
break or loss of FW pumps).
(2) Steam generator low level.
(3) Loss of all reactor coolant pumps (sensed by RPS. power monitor).
(4) Low main steam line pressure (600 psig).
AFW Pump Auto Trips ,.
Either pump trips on:
(1) overspeed (2) low suction p'ressure (3) low steam (to turbine drive) after 25 seconds AFW Indication (1) Discharge pressure each pump.
(2) Speed indication each pump. -
Level Control ,
Auto essential level control system controls at 120". However, operating instructions require operator to control at 35 inches-if there is no SFAS (until dual level setpoint is installed).
Independence of AFW Trains Appear to be independent with the exception of:
(1) The suction source (non-seismic condensate storage tank). However, each pump auto transfers to seismic category 1 redundant sources on ,
low suction pressure.
Effect of Surveillance Test on System No effect on normal valve lineups. A pump is started and the mini-flow "recirc" is aligned to the sump as usual. If a demand signal is received during the surveillance test, the injection valves open and normal emergency injection.
begins.
Seismic Event (1) Only effect would be loss of the normal suction source to both pumps.
The suction for each pump automatically transfers to the seismic category 1 source on low suction pressure.
(2) The seismic event could possibly damage the condensate storage tank in a manner that could cause blockage of the suction strainers on both AFW pumps.
Loss of Offsite Power
- Could cause almost immediate loss of both main FW pumps. Both AFW pumps would start automatically after diesel generators are started.
. Loss of Offsite Power with Single Failure -
Worst identified single failure is loss of one train of AFW. Opera tor action is required to deliver water to both steam generators from one AFW pump (open cross-tie M0V's from control room). .
AFW Desi~gn Re'sponse to LOCA Does not start directly from SFAS. LOCA analysis assumes AFW flow. The AFW pumps would be started by different signals on many accidents which initia te SFAS.
Alternate Cooling Mode Without Main FW or AFW Startup FW pump 250 gpm at 1050 psig. '
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APPENDIX N COMMENTS Off ARKANSAS UNIT 1 FEE 0 WATER SYSTEMS Introduction These comments were compiled during the week of April 15, 1979 from information in the FSAR and from telecons with the licensee.
Summary (1) There is no auxiliary steam supply to the main feedwater pump turbines.
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If both emergency feedwater pumps inoperable, must rely on auxiliary
- feedwater pump to get to cold shutdown. Auxiliary feedwater pump is not seismic category 1 and is not Class lE. -
.. ( 2) Several components in the emergency feedwater system are not powered -
by a Class lE source. The pump motor is not normally powered by a Class lE sourc~e but can be manually aligned to a Class lE source. Additionally, some of the system instrumentation is not Class 1E.
(3) For any, dema6d sequence, a single failure of the turbine driven EFW pump would require operator action to get emergency feedwater into the steam generators because the motor driven pump does not auto-start by design.
(4) It is questionable whether each train of emergency feedwater has the capacity assumed in the generic B&W LOCA analysis. (R. C. Jones of B&W informed us on April 22, 1979 that the analysis assumes 500 gpm per steam generator (1000 gpm total) at 1050 psig.) -
(5) We don't know if there are single failures of the ICS that could cause loss of both emergency FW trains or simultaneous loss of -
main and emergency feedwater. ,
(6) The emergency feedwater pumps do not directly tart on ECCS initiation
. signal. (However, the turbine driven EFW pump would be started by other signals for many of the accidents which initiate ECCS signal.)
(7) Portions of the auto start instrumentation are not redundant. Concern is for single failures.
(8) The pressure switch on EFW pump suction that alerts operator to switch to backup suction source (of water) is not redundant.
Two turbine driven pumps with steam sources from (1) reheat steam, (2) main steam. Shutoff head = 1090 psig. The condensate pumps have suction strainers. No automatic bypasses around demineralizers or FW heaters.
The Steam Line Break Instrumentation and Control (SLBIC) system isolates main FW to both steam generators if the pressure in either is less than 600 psig.
Emergency Feedwater (EFW) Sources Normal supply: condensate storage tank (non-seismic Category 1); backup ,
source: service water pump discharge (either of two) (seismic Category 1).
Switchover to backup source is by remote manual M0V's and can be done in seconds from control room. However, the pressure switch which gives low suction pressure alarm is not redundant or Class lE. The valves which ~*
must be realigned are Class lE.
EFW Pumos Two pumps, one tuibine driven and one motor driven. Each pump 780 gpm at 1112 psig. Shutoff head unknown.
EFW Puma Drives (1) Motor - not normally powered from a Class lE source but the licensee is currently evaluating this possibility.
(2) Turbine - supplied by main steam upstream of MSIV's. -
EFW Pumos Auto Start (1) Motor - no auto starts ,
(2) Turbine - auto starts on any of the following signals: (a) SLBIC (steam generator pressure less than 600 psig). This is a Class lE .
signal; (b) loss of FW (as sensed by governor latch on both main FW pumps) coincident with low discharge pressure of the " auxiliary" -
FW pump (signal from ICS); and (c) loss of all reactor coolant pumps (sensed by breaker position). Note: does not start on SFAS. -
(2) With exception of SLBIC, the start signals are not Class lE.
Automatic EFW Pumo Trios Motor - electrical faults Turbine - overspeed EFW Indication Discharge pressure each pump.
Level Co'ntrol On loss of main FW pumps, ICS controls level at 20 inches in one steam generator; 24 inches in the other. On loss of all reactor coolant pumps, the ICS controls level at 50% on the operating range (approximately 300 inches). Operating procedures call for the operators to manually control level to control reactor coolant system temperature.
Independence of EFW Trains Appear to be independent with the exception of the following:
(1) Normally open cross-ties valves between discharge lines.
(2) Common normal suction source (non-seismic Category 1 CST) and suctica line.
(3) Common suction line from backup source (service water).
(4) Ncn-redundant pressure switch that alerts operator to switch suctions on loss of normal source.
(5) ICS inputs to flow control valves of both trains. There may be single failures of the ICS that would cause valves in both trains to close. -
Effect of Surveillance Test on EFW System Operator opens manual recirc valve on train being tested. Injection valves are normally closed. If EFW is needed, operator must close recirc valve to align full EFW flow to the steam generators. "'
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Common Mode Failures that Would Cause loss of Main FW and Emeraency FW
_ The ICS is not Class IE. Neither is some of the EFW system instrumentation.
Failure modes may exist which would cause the main FW valves to close and
, prevent the EFW injection valves from opening. Operator action would be required to open the EFW injection valve bypass valves (MOV's).
Seismic Event A seismic event could cause loss of the normal suction source. Debris from the non-seismic category 1 condensate storage tank could cause damage to both emergency feedwater pumps if it entered both pumps. (This is not unique to this facility.)
Loss of Offsite Power Could cause almost immediate loss of both main FW pumps. Only the turbine driven EFW pump starts automatically. Operator action would be required to start the motor driven pump on the diesel generator. (Operator action required to start motor driven pump with offsite power available also.)
Loss of Offsite Power with Single Failure Worst identified single failure is loss of the turbine driven EFP train.
This would require operator action to restore EFW by starting motor driven pump. Questions on ICS: We have not investigated whether single failure can cause both EFW flow control valves to close.
EFW System Design Response to LOCA System will not start automatically on ECCS initiation. LOCA analysis assumes flow. (Mowever, the turbine driven EFW pump would be started by other signals for many of the accidents which initiate ECCS signal.)
- Alternate Cooling Mode Without Main FW Pumps or EFW Pumps Auxiliary feedwater pump. 1150 gpm at 1000 ft. _,.
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AFPENDIX 0 CRYSTAL RIVER, UNIT 1 FLORIDA POWER CORPORATION Response to Item 2 of I&E Bulletin 79-05A Eacn Licensee for a B&W operating plant was requested to respond to
, Item 2 of IE Bulletin 79-05A. Item 2 v:as stated as follows:
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" Review any transients similar to the Davis-Besse event
- . (Enclosure 2 of IE Bulletin 79-05) and any others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (f es).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a de-scription of any corrective actions taken. Reference may be made to previous information provided to the NRC, if appropriate, in responding to this item "
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i Trip: 79-1 Da t e : Janua ry 6, 1979 Event: Excessive Cooldown Rate Due to Stuck FW Block Valve initial Condit ions: 71% RL'T, 59 5 t1We DESCRIPTION At 0242 on January 6, 1979, the turbine tripped and feedwater block valve
,' FWV-30 stuck in an open or partially open position. Cont rol room operators -
took action to shut the main feedwa te r cross-connect valve, FWV-28, nd ,
t rip feedwa ter pump 'A". At this point feed flow was stopped to " A" steam -
genera tor and "B" feedwater pump was supplying "B" steam genera tor. Whe n steam supplying the "B" feedwater pump turbine automatically shifted from reheat steam to main steam (a normal occurrence due to loss of reheat s t.* am ,,
pressure when the main turbine tripped), the feedflow to "B" steam genera tor decreased and Tave increased to 600*F. Reactor coolant pressure peaked at 2235 psi when the electromatic pressurizer relief valve opened momentarily. Pressurizer icvel peaked at 307 inches. The reac tor was manually tripped by the control room operator and the turbine driven eme rgency feedwater pump was star *.cd to res tore feedwater flow. FWV-3 0 wa s found to be stuck on its backseat and was taken of f the backseat manually and closed. FWV-28 was reopened. The cooldown transient , which resulted when the reactor was tripped and emergency feed initiated, resulted in a loss of pressurizer level indication (low) for approximatal; 2 minutes.
Reactor coolant system pressure decreased to 1600 psi during the same time frame and T ave decreased tu 521*F. At 0615, FWV-30 was proven operable by surveillance procedures and auxililary steam had been brough,t in to restart no rmal feedwa ter pumps. Plant parameters of interest are shown in the attachments.
SlCNIFICAST DEVIATIONS There were no deviations from expected performance except fo r the failure of FWV-30 to shut. This failure had no impact on sa fe shutdown of the plant since feedflow could be stopped by shutting FWV-28 and tripping "A"
- feedwater pump. Redundant emergency feedwater systems were available and
- o pe rabl e .
Simpson (NRC10)
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Trip: 77-33 Da t e : April 16-23, 1977 Event: Shutdown From Outside Control Room Test Initial CondLt tons: 157. RTP DESCRIPTION The shutdown f ron outside the control room test s imula t ed an eme rgenc y situation requiring evacuation of the control room. All plant controls l were left in automatic unless remote indication required taking them into manual modes of ope ration. The purpose of the test was to demonstrate that the unit could saf ely be brought to hot standby conditions from outside the Control room.
The test was started from 151 power. Le tdown flow was s topped and the control room evacua ted by the normal shif t com plemen t of ope rato rs. The opera tors manned their remote shutdown stations as shown in the attached table. A complete second set of operators was lef t in the control room to assume plant control if the test fa il ed. The reac tor was tripped remotely and the plant allowed to come to hot standby automatically. The operators outside the control room were to take control of various equipment if it.
was not pe rfo rming adequa tely in automatic.
On Aprili6, 1977, the first run at shutdown from outside the control room was attempted and aborted af ter approximately 18 minutes due to feedpump a P oscillations. Main feedwater pump speed control had been shif ted to hand by the operators in the control room early- in the test, After the reactor trip, high leakage through the startup valves resulted in overfeeding both steam generators. The test was terminated due to loss of steam generator level control and greater than desired cooldown of the primary plant. As a result of this expe rience, the plant emerg'ency procedure for shutdown from outside remotely.
the control room was changed to require tripping the main feedpump
. start This would allow the steam driven emergency feedwater pump to and take over steam genera tor feed requirements.
' On April 22, 1977, the test was repeated with the above modifications.
- This run was stopped af ter approximately 9 minutes due to low level.; in both steam generators. Subsequent investigation revealed that Initial conditions
' for steam pressure to the steam driven emergency feedwater pump were not met.
This resulted in both steam generators being dry
- until flow was escablished by the electrical driven emergency feedpump. The plant emergency procedure was modified to requ l ce the ope ra to r to c h e r- k the steam driven pump and, if that is not opera t I ng prope rly, sta I tho o lec t r t e cil driven pump.
- The steam generators were designed for 20 allowable thermal cycles equivalent to being boiled dry.
Simpson (NRC10)
D63
Trip: 77-31 Da t e : April 21, 1977 Event: Partial Loss of Power to the ICS Initial Conditions: 46% RTP, 323 MWe DESCRIPTION At 0430 on April 21, 1977, the "X" power supply to the integra t ed cont rol system (ICS) was los t , resulting in. the following events:
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A. The ICS saw an erroneous :ero reactor coolant flow condit;on (no reactor coolant pumps running) and signaled the feedsater system to maintain steam generator level at 50% in the operating range.
B. An ensuing increase in turbine header pressure caused all turbine bypass valves and atmospheric dump valves to o pen.
C. The main feedwater block valves went shut and emergency feedwater b.ock valves opened in response to ICS demand.
D. As a result of decreased steam flow to the main turbine. Megawatt electric output decreased to the point where the control room -
operato rs openea the generator output breake rs and tripped the turbine. ,
E. The control room operator manually opened the startup block valves and maintained minimum required feedwater flow.
At this point , power was restored to the ICS when the electricians replaced a blown fuse. A no rmal plant recovery followed. Minimum and maxinum pressuri:er levels attained during this transient were 40 inches and 270 inches respectively. Other plant parameters of interest are shown on the attachments.
SIGNIFICANT DEVIATIONS There were no deviations f rom expected pe rformance.
OTHER COMMENTS An earlier transient of this type was experienced during reactor trip 77-13 on March 2, 1977, when Inverter *3" tripped, causing a similar loss of ICS po we r . There wa s a concur rent loss of a main feedwater pemp necessitating use of the e=ergency feed system for steam denera tor level con:rol.
Simpson (NRC10) 063
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On April 23, 1977, the test was run successfully. The control room was evacuated with the plant at 15 po we r . The running main feedpump was tripped remotely (which trips the main turbine). The reac to r , however , was not tripped unt il a t least one minute later, resulting in low steam generator levels. The operators started the electric-driven feedwater pump, took manual control of both feedwater startup valves and restored level in both steam generators. Twenty minutes into the test, an operator remotely added wator en the makenp tank, othersise the plant rem ined in a fully automatic mode of operation and came to a hot standby cond Lton. The test was allowed to run for thirty ininutes to verify that the opera tors
, outside the control room had complete control of the plant. At this time, plant parameters were at or near their final steady state values and the test wa s e nd ed .
- . Although level and feedflow indication did not show zero, post test analysis indicated that the steam generators were dry in about seven minutes. This occurred because of a combination of problems with reference legs, flows, and/or calibration errors. This could be ver ified by not ing that during the dry period, main steam pressure was below the saturation pressure and recovered as soon as feed flow was re-established.
Charts of significant plant parameters are shown in the attachments. 'Jo rs e case transients were experienced during the aborted tests on April 16 and April 22. Test results on April 23 were acceptable.
SIGNIFICANT DEVIATIONS Deviations fron expected perfo rmance were
- experienced on Apcil 16 and April 22 as discussed above. Corrections were made to the emergency procedure governing this evolution and the test was completed satisfactorily.
The test proved that the reactor can be brought to and maintained in a safe hot standby condition from locations outside the control room by the normal shift com plemen t of ope rato rs.
Simpson (NRC10) 063
Y. ]
/ -
/
Trip: 77-35 Da t e : April 23, 1977 Event: Loss of Of fsite Power Test Initial Conditions :
The Loss of Offsite Power Test consisted of two (2) parts. The first pa c t approximated a totel plant blackout f rom 15% reactor power; the second part was performed from a shutdown condition and verificd a diesel generator's ability to start and pick up certain vital loads. .
DESCRIPTION With the plant at 157. power , the reactor and startup trans fo nner were .'
simultaneously tripped. This immediately reduced total plant power to the emergency batteries and the above mentioned diesel genera tor. The 3A diesel generator was timed as it started , came up to speed and picked up certain pre-detennined loads on its ES Bus. After allowing the plant to o pe ra te in this condition fo r fif teen minutes, the s ta rtup trans fo rmer was r e-en e rg i zed. Loads considered necessary to allow plant equipment to survive the test were shifted from the 3B to 3A diesel generator. The 33 diesel was then s topped and the startup trans fo rmer again tr tpped. This allowed the timing of the 33 diesel as it came up to speed and picked up
- its pre-selected loads.
At that point, the test wa s c om ple t e. However, it was observed that there was a large imbalance in feedflow. Subsequent investigation and evaluation revealed the following sequence of events :
EVENT Tripped power, both feedwater pumps stopped, and all feedwater flow was lost.
Steam driven emergency feedwater (EF ) pump automatically up to speed and feeding both steam generators.
Operator started electric-driven emergency feedwater pump. "A" steam water.
generator is being preferentially fed, but botn cre getting Operator stopped steam-d riven EFW pump. "A steam generator is ,
being filled (startup level indication), "B" s team genera tur startup and operating range level indicators are both apparently at the bottom of their range. This can be verified by noting that loop "B" hot and cold leg temperatures indicate that little or no heat t rans fer is occurring t hrough "B" Ic ?.
(Figure 4.15-1)
"3" loop startup feedwater flow indica t io n is at the bot tom of its range. (This was confirmed by a zero calibration check of the instrument made on 5/11/77. This check showed the zero had shifted to 1.5 x 105 Lbm/Hr. A similar check of "3" startup flow made on 5/11/77 showed its zero shif ted to .95 x 105 Ldn/qc.)
Simpson (SRC10)
D63
/
Trip: 77-48 Da te : October 26, 1977 Event: Loss of Inverter "A"/ Loss of Vital Bus ~~ A "
Initial Conditions: 100% RTP, 838 f1We DESCRIPTION At 0427 on October 26~ 1977, Inverter "A" tripped causint; a loss of powe r
. to vital Bus 'A". As a subsequent result, the main turbine tripped.
Fe edwa te r Pum p " A" tripped and Feedwa te r Pump " B" ran back to nintmum s pe ed . Er. cess heat production resulted in hip,h reactor coolant uvstem pressure and a reactor trip. The cont rol room opera tur s ta rt ed both
'. emergency feedwater pumps. Atmospheric dump valves , which opened to relieve excess steam from the steam generators f ollowi ni; turbine trip, f ailed to close at the expected design setpoint but did close at a lower pressure. With the high steam generator feedrate and the late closing of the atmospheric steam dumps, an excessive rate of cooldown was experienced.
Pressurizer level, decreasing due to reactor coolant system cooldown. was maintained in the ind ic a t ing range using manual cont rol of the h I e,h pressure injection system. !!inimum and maximum pressurf.er levels achieved during the transient were 35 inches and 245 inches r e s pe c t i v ely. Other ,
plant parameters of interest are shown on the attachments.
SIGNIFICANT DEVIATIONS -
The only deviation from expected performance was the fa ilure of t'te atmospheric dump valves to close at the pr escribed design se,t po i n t s .
Although this failure occurred and contributed to an excessive reactor coolant system cooldown rate, it was not considered to be a critical failure since the dump valves are only sized to pass 7. 5% of ra t ed s t e am flow. Even if the valves stuck open for the entire transient, the cooldown rate, which would have been experienced, would be insignificant when compared to a main stea::t line break accident. Each atmospheric dump valve can be manually isolated by an associated upstream rect ealve.
Simpson (NRC10)
D63 ,
/ -- 1
/
/
Operator restarted steam driven EFW pump, started feeding "B" steam genera tor again.
Operator opened parallel valve (EFV-162) In the feedwa ter train to " A" s team genera tor by mis take.
Ope rator shut EFV-162 and opened a paralle.1 valve (EFV-161) in the feedwater train to "B" steam genera tor.
"B" .
steam generator filling , on its way to recovery SIGNIFICANT DEVIATIONS The plant should have responded automatically by starting the steam-driven emergency feedwater pump within I minute and filling each steam generatar to 50% on the ope rating range. However, with one pump feeding both steam genera tors any imbalance in steam pressure will result in one gouera tor getting more feedwater than the other.
(approximately two minutes), the cold feedwater coolsAfterthe primary flow has coasted down primarv water in the steam generator. This results in a continuing lowe ring of t he pressure in the steam genera tor already being f ed, thus increasing its feed flow- ,
This feedback ef fect allows one steam generator to be underfed untti the other one reaches a level of 50% at which point its feed valve will shut.
The emergency procedure for loss of of fsite power has been' changed to require the opera tor steam generato rs. to monitor levels and keep the feedflow shared between s
~
Simson (NRC10) 063
Trip: 79-2 January 17, 1979 Da t e :
Turbine Building Flooding /1.oss of Feed 9ater Event: 100% RTP, 848 MWe initial Conditions:
DESCRIPTION d on CWV-2 a solenoid failure was experienceexchanger "A')
At 1010seawater on January 17, 1979, block valve to secondary services heat exchanger (inlet causing it to fail The associated secondary services heatWhen CWV-2 thisopen.
time fo r cleaning.
- was opened at exchanger onto the 95 f t. elevation of the Turbine from of the open heat out Building.
Control room operators Attempts were alerted by flood.ng were made reportsbut to close C'WV-2 scene.
maintenance personnel At on the1015, the cLeculat'ng water pump, water which was the
,* were unsuccessfui. Input of seawater stopped but At 1018, both source of flooding, was secured. flow across the floor. local control already in the building continued to At 1021, feed-condensate pumps tripped due to wcter contacting and shorting At 1020, the main turoine was manually tripped.to tripped due a low deaerator switches. .
water booster pumps and mainsystem feedwater pressure.
pumps 1022, the reactor wa level. At pressurizer level and reactor coolant The tur-was maintained below 227f) psi.1030, the control Maximum reactor coolant pressure At bine-driven emergency feedpump started automatically.feedwater pump and sec room operator started the motor-driven At 1040, the emergencybegan plant to recover 1760 psi from the and was the turbine-driven pump. pressure reached a minimum of1110, electrical transient. Reactor coolant 0 At restored to a stable reading of 2150 psi by 110 .During the transient, Plant pres-the condensate pumps was restored. ~
300 inches.
power to surizer level wasare eaintained shown on the between attachments. 80 inches and parameters of interest -
51CNIFICAST DEVIATICNS ystem within the This transient was controlled by the emergency feedwater sThere wer intended design envelope. seawater contacting andting shorting out the local
~
ance.
The initiating event (i.e., l local condensate pump controllers) was corrected b floor level.
by re oca ccadensate pump controllers to higher elevations a ove Simpson (NRC10)
D62 1
Trip: 79-3 Date: January 30, 1979 Event: Lnss of Feedwater Flow to the "E" OSTG Initial Conditions: 1007. RTP, 84 5 StWe , Full ICS Auto.
DESCRIPTION At 0515 on January 30, 1979, a reactor trip occurred due to a loss of feed f rom the "B" main feedwater pump. The feedwater pump did not ac .W.
tually trip flow had reduced significantly in the ~S" loop. The but F.W.
crossover valve, FWV-28, did not open since FW P-2 B d id no t trip. This ,
caused a loss of feedwater to the "b" steam genera tor which resulted in excessively high reactor coolant pressure and a degradation o' a header pressure. The turbine reacted to the reduced header pressure by rapidly ,e
- reducing 51We in an attempt to regain plant stability. The control system rea c t ed to reduction in 11We and comnenced runni ng the plant hace to a lower powe r level. A sho rt time luto the runback the operator took action to res tore F.W. flow to the "3" OTSC by opening FWV-28. This resulted in overfeeding the OTSC's for the imnediate power level which induced a rapid RCS cooldown and outsurge of the pressurizer. The resultant reduction of RCS pressure tripped the reactor on low R. C . S . pressure. The excessive .
feed rate an:i subsequent cooldo-7 was te rminated by the ICS f une:Lon of RCS pressure closing all F.W. control valves following the reactor tri;3 degraded to slightly less than 17C0 psig but high press ure' injec t ion wa s not ac t ua t ed . Pressuri:cr level maximum and minimum achieved curing this transient were 290 in. and 35 in. r es pe ct ively.
SIGNIFICANT OEVlArloNS There were not significant d evia t io ns fro >n, expected system pe r f o r.7a nc e during this transient.
Simpson (NRCIC) i 003
APPE ! DIX P RANCHO SEC0, UNIT 1 SACRAMENTO MUNICIPAL UTILITY DEPARTMENT Response to Item 2 of I&E Bulletin 79-05A Each Licensee for a B&W operating plant was requested to respond to
. Item 2 of IE Bulletin 79-05A. Item 2 was stated as follows:
" Review any transients similar to the Davis-Besse event N (Enclosure 2 of IE Bulletin 79-05) and any others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a de-scription of any corrective actions taken. Reference may be made to previous information provided to the NRC, if appropriate, in responding to this item."
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its 2 .
Paview any transients sicilar to the Davis-Besse Event (Enclosure 2 cf IE Bulletin 79-05) and any others which contain similar t!erents fran '
the enclosed chrer. ology (Er. closure 1) which have occurred at your facility.
If cny significant devictions free expected ;:erfor:2nce are idantified in yo:;r re.f es, provide details and an analysis of the safety significan:c tuetter with a descricticn cf any corretti ve actices taken. Reference
- '-y be cade to previo;s infora- tion provided to the N?.C. if a::prcprie:e, in responding cc if:is i*ar.
Resconse to Item 2 The District has reviewed transients at Rancho Seco Unit No. T in order to determine any having similar elefrents to the chr:r. ology of evert: Si Three Mile Island Unit 2 and D3YiS-Oe$se Unit 1. VG hSVC DGt e
R. H. Engelken April 11. 1979 fcur.d any transients which are siciilar, however, we have reviewed one transient with a cooldown which resulted in operation outside the Technical Specification pressure-tecperature licits. This has been reported previously as a reportable occurence as follows:
90 78-01 March 30, 1978 and
.- March 31. 1978 This event was analyzed by B&W, the NSSS vendor. The analysis concluded no dange occurred which would affect further operatice of Rancho Seco. The District's Management Safety P.eview Cemittee evaluated this event and has approved the folicwing correcthe actions to be irclecentEd at FOncn? Seco Unit No.1 by the end of the next refueling outage:
- 1. A nonconducting foan rieber plug has been developed to insert in the back lighted push button codule wher.c.er the lamp bulb section of the redule is lifted out.
- 2. Testing has been perfcrred cc the existing NNI-Y p&er supply systera to detercine the trip point of the pxr supply conitcrs, the time delay cf the trip circuit breakers, the current li=iting point of the pcwer succlies, the transfer voltage point of the AC automatic tracsfer switch, the perfor:rance characteristics of the pcwer su; ply fuses, and verification of original trip conditions en the pcwer supplies.
- 3. Lower rated fuses will be installed in each group of redules where analysis has shcwn this can be dcne safely.
4
- 4. The pcwer supplies will be improved to ninimize the ned:er of c:cponents affected by a power failure.
- 5. Pmcedures have been changed and instrtrentaticn will be ins.talled to provide control rece indicat co of the critical N.U-X cr NNI-Y signals.
The safety significance of this transient is due only to the structural integrity of the reactor coolant systee and possible equiprent da= age. Neither c::cdition posed any threat to the fuel. Voids =cre not for:cd. Pressurizer level did decrease below visible indication uptn safety features actuation but was restond with high pressure injection.
Subcooling in the reactor coolant systec (excluding the pressuri:cr) was enintained at a cinicum of 35*F. Reactor coolant flew was raintained with a cinimu:n of one p :p per loop at all tices during the transient.
APPENDIX Q OCONFE, UNITS 1, 2, AND 3 DUKE POWER COMPANY Response to Item 2 of I&E Bulletin 79-05A Each Licensee for a B&W operating plant was requested to respond to Item 2 of IE Bulletin 79-05A. Item 2 was stated as follows:
" Review any transients similar to the Davis-Besse event (Enclosure 2 of IE Bulletin 79 05) and any others which
. contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a de-scription of any corrective act1ons taken. Reference may be made to previous information provided to the NRC, if appropriate, in responding to this item."
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Response
Based on initial infor=ation relative to the recent Three Mile Island Unit 2 occurrence, Duke Power Company initiated on March 29, 1979 a review regarding similar transients at Oconee Nuclear Station. On March 30, 1979, a summary of this early review was provided verbally to NRC/0IE, Region II. Subse-quently, the review of Oconee transients was continued, particularly to address additional TMI-2 information as such became avai: sble. At the present time, Oconee transients considered applicable for purpose of the subject review are categorized as follows:
~
(a) Feedwater Transients Resulting in Reactor Trip -
(b) Pressurizer Relief Valve Stuck Open (c) Loss of Offsite Power With regard to feedwater transients resulting in reactor trip, Oconee has ex-perienced approxi=ately 42 such incidents as tabulated below:
UNIT YEAR 1973 1974 1975 1976 1977 1978 1979 1 11 1 3 3 2 2 0 2 4 1 4 1 0 3 0 3 N/A 2 1 1 0 2 1 As can be seen, the greatest number of these transients (per un; ., per year) occurred during the initial operation of Unit 1 in 1973. Subse 2ent ex-perience is consistent with classification of this event as one of moderate frequency.
Eleven of the above 42 incidents occurred at or near full power (Unit 1-7, Unit 2-2, Unit 3-2) and demonstrate the ability of the Oconee units to safely respond to such events. Several feedwater transients which resulced in reactor trip have been identified however as involving deviations from -
expected perfor=ance. These and transients in categories (b) and (c) are summnrized, in chronological order of occurrence, below:
(1) On January 4, 1974 while operacd.ng at 75% full power, Unit 2 tripped due to a loss of off-site electrical power. The reactor coolcat pumps (RCP) ,
tripped, and natural circulation cooling was established. RCP seal injec-tion and component cooling were lost for approximately 31 seconds at the
ITEM 2 (Continued) 4 time of the trip. Both were again lost for 15 seconds about 25 minutes after the trip. Subsequently, because pressuri=er level was increasing due to excessive =akeup flow, an attempt was made to initiate letdown flow, but no flow was indicated. High Pressure Injection (HPI) was secured, and all seal return valves closed in order to reduce =akeup volume. A leak was discovered which was the result of a blown gasket on the upstream side of the letdown flow indicator. The letdown line was isolated to control leakage, and the emergency makeup valves were closed. During this time, HPI was turned on again for approximately a =inute, then secured again. When the seal return valves were closed, RCP seal cavity pressure went to system pressure. Seal injection flow resumed about 20 minutes later when EPI was once again started.
No design or Technical Specification limits were exceeded during this
. P.ransient , and the event was not considered to have any safety signifi-cance. Hardware and procedural changes were made, however, to provide better monitoring and control during future similar incidents.
(2) On June 13, 1975 while Unit 3 was operating at 15% full power, a feedwater transient resulted in an RCS pressure transient which resulted in the pressurizer power operated relief valve (PORV) opening. The PORV failed to close when pressure decressed and the subsequent RCS depressurization was terminated by closure of the PORV block valve by operator action.
Additional information regarding this incident is provided in Mr. William O. Parker's letters of June 27, 1975 and August 8, 1975 to Mr . No r=an C.
Moseley, Director, NRC/0IE, Region II - see Enclosure 2-1.
(3) on July 12, 1976 while Unit 2 was being shut down in order to repair a main turbine steam leak, the ICS induced an oscillation in feedoater parameters. The feedwater pu=ps tripped on low feedwater pressure, causing a turbine trip. The trubine trip caused RCS pressure to rise sufficiently to open the pressurizer PORV, relieving pressure to the quench tank. The quench tank rupture disc burst. The PORV reclosed properly when RCS pressure decreased. This RCS transient was of short duration and not observed by the operators whc were responding to the turbine trip. The alarm typer, another source of plant equipment status, was out of service. The unit was shut down and turbine rupairs were effected, but the quench tank rupture disc was not replaced since its rupture had not been noted. The unit was operated until July 27, 1976,
- when it was shut down to repair a reactor coolant pump. At that time the burst rupture disc was discovered and replaced.
No design or Technical Specification limits were exceeded during this transient, and the event was not considered to have any safety significante. Operations personnel were subsequently instructed, how-ever, to observe quench tank instrumentation more closely following transients in order to note indications of high quench tank pressure or a burst rupture disc.
ITEM 2 (Continued)
(4) On December 14, 1978, an electrical short in the Unit 1 :.CS RCS average temperature (Tave) recorder caused the temperature indication to be approximately 130F low. To compensate for the low Tave indication, the ICS initiated an increase in power (from approximately 98% full power),
but operations personnel had been instructed not to allow power to increase above 99% full power until an earlier problem had been resolved.
Therefore, manual control of the reactor was assumed, causing the ICS to switch Tave control from the reactor = aster to the feedwater master.
Feedwater flow decreased to compensate for the -13 F error in the ICS.
Upon observing increasing hotleg temperature, decreast'g reactor power, ;
and decreasing feedwater flow, operations personnel placed the feedwater master in =anual and began increasing feedwater flow. However, before the increasing RCS temperature could be corrected, the reactor tripped on
- high temperature. Feedwater flow was decreased as rapidly as possible, and the resulting high discharge pressure caused both feedwater por.ps to trip. The emergency feedwater pu=p was started and ran until the feed-water pumps were reset and started. However, the levels in the two steam generators continued to decrease; level in the 1A steam generator reached a low of six inches, while steam generator 1B went dry. Opera-tions personnel opened the feedwater valves and the emergency header block valves in order to feed the steam generators through the emergency feed header. Level was partially restored, although steam generator 1B level re=ained significantly lower than that of steam generator lA. This was probably due to the failure of the 13 emergency header block valve to open fully. In order to increase the IB steam generator level, the emergency feedwater pump was restarted and fed through the emergency header. RCS pressure dropped rapidly due to the quick cooldown of steam generator 1B, causing the feeuaater pumps to trip on low suction pressure, and removing feedwater flow from steam generator lA. Flow was re-established to that steam generator by lining up the emergency feedwater pump to feed it. RPI was initiated when an Engineered Safeguards actuation signal was received due to low RCS pressure. All ES components operated properly.
Additional information regarding this incident is provided in Mr. William O. Parker's letter of January 15, 1979 to Mr. James P. O'Reilly, D':ector, .
NRC/0IE, Region II - see Enclosure 2-2.
(5) On December 25, 1978, Unit I was at approxi=ately 10% full power and
- increasing in power following a reactor trip when power to the ICS was lost as a result of blown fuses. When ICS power was lost, both feed-water pumps tripped. The emergency feedwater pump was started, but Control Room instrumentation indicated a discharge pressure of less than 100 PSIG. Personnel were dispatched to increase the discharge pressure to its normal range of 950 to 1000 PSIG. The pump indicated a discharge pressure of 600 PSIG, and it was later determined that the control rcom instrumentation required approximat21y five =inutes to provide an accurate indication.
ITEM 2 (Continued)
Approximately one sinute after the feedwater pumps tripped, the reactor tripped on high RCS pressure. When ICS power was lost, the normal feed-water startup header valves began to close and the emergency heade. block valves opened. Level in steam generator lA was restored, but IB went dry.
it appears that the block valve failed to open fully. The feedwater pumps were reset and restarted, and flow to the 1B steam generator resumed through the nor=al feedwater header. The steam generator was dry for approxi=ately 15 minutes.
l The reason the emergency header block valve failed to open fully has not been deter =ined. The governor control valve on the emergency feedwater pump has been checked to assure that it is properly set. Operations person-nel have been instructed as to actions to take to supply flow to the affected steam generator if flow cannot be established through the startup feed valve and auxiliary feedline i= mediately af ter the loss of main feedwater pumps.
A procedural change, applicable for all units, has been made requiring operators to bypass the block valve in the event the block valve fails to open. The operator can, from the Control Rcom, operate one valve to provide emergency feed flow bypassing the block valve to the affected steam generator. The emergency feedwater pu=p discharge pressure instrument has been adjusted to decrease its response time. This event was not considered to have any safety significance.
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APPENDIX R DAVIS-BESSE, UNIT 1 TOLED0 EDISON ELECTRIC COMPANY e
Response to Item 2 of I&E Bulletin 79-05A Each Licensee for a B&W operating plant was requested to respond to
- Item 2 of IE Bulletin 79-05A. Item 2 was stated as follows:
" Review any transients similar to the Davis-Besse event (Enclosure 2 of IE Bulletin 79-05) and any others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a de-scription of any corrective actions taken. Reference may Le made to previous information provided to the NRC, if appropriate, in responding to this item."
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Davis-Besse, Unit 1 Response to Item 2 All transients that have occurred at DB-1 that have been initiated by either a loss of feedwater flow or excessive feedwater flow hcve been reviewed to determine if any significant deviations from expected per-formance occurred. During this review the following information became evident regarding the five simi'ar transients discussed below:
a) Out of the five similar transients; found the first four occurred during the first year of operation prior to the time that the final tuning of the Integrated Control System (ICS) was completed. ICS .
controls the main feedpump turbine speed and the main feedwater '
control valves.
b) No offsite radiation raleases resulted from any of these events. -
The Davis-Besse Unit 1 event referenced in Enclosure 2 of IE Bulletin 79-05 occu red on November 29, 1977. This event was addressed in previous information provided to the NRC, reference Reportable Occurrence NP 77-20 on the Davis-Besse Unit 1 docket, dated December 12, 1977. At the time of the occurrence the Unit was in Mode 3. The loss of power aspect of this event is discussed in Reportable Occurrence NP-33-77-98 dated December 16, 1977. The corrective action modified the emergency procedure to preclude manual tripping of the generator main breakers on a turbine trip.
With respect to Item 3 on page 2 of Enclosure 2 to IE Bulletin 79-05, reference is made to "A special analysis has been performed concerning this event. This analysis is attached as Enclosure 1." The Enclosure 1 referred to is a letter from L. E. Roe to R. W. Reid dated December 22, 1978, Serial No. 475. This letter analyzed costulated Davis-Besse Unit 1 transients resulting from the operator not controlling steam generator level at 35 inches in accordance with current operating pro-cedures. The two overcoolina transients examined are a loss of offsite power and a loss of feedwater. The loss of feedwater transient results .
in the greater volumetric contraction of the reactor coolant system because the forced coolant flow with reactor coolant pumps operating -
causes a faster rate of heat rejection to the steam generators.
On September 24, 1977 a depressurization of the Davis-Besse Unit I "
reactor coolant system occurred that contained some similar elemen.s to the chronology of Enclosure 1 to IE Bulletin 79-05A. At the time, the Unit was in Mode 1 with power at 268 MWT with the tu-bine off the line.
The details of this event are included in Supplement to Reportable Occurrence NP-32-77-16 dated November 14, 1977. System and equipment modification and testing actions are included in that report.
2-1
On December 11, 1977 Davis-Besse Unit 1 was tripped for the 40% reactor trip test. During recovery from the trip with the Unit in Mode 3 (power at 0 MWT), control of both auxiliary feed pumps was lost. Modifications were made to the controls and surveillance testing was modified to demonstrate operability. See Reportable Occurrence NP-33-77-110 dated January 3, 1978.
. On April 29, 1978 Davis-Besse Unit I had one high pressure injection pump inject water for two minutes while the RCS pressure was below 1700 psig. The Unit was in the process of shutting down from 420 MWT for maintenance. The cause of the occurrence was the sensitivity of the
- feedwater controls while on three reactor coolant pump operation, and improper operator action in taking manual control of the feedwater.
This resulted in overcooling the reactor coolant system. HPI actuation was according to design. Corrective action was completed per Reportable Occurrence NP-30-78-01, Letter No. 1-23, dated July 28, 1978.
On January 12, 1979 an accidental ground caused the loss of a 120 VAC essential bus due to an improper fuse in the 120 VAC switchgear. The loss of this 120 VAC essential bus caused a loss of level indication on steam generator (SG) 2. After the reactor tripped, the level in SG 2 fell low'enough to cause a full Steam and Feedwater Rupture Control System trip and isolation of both steam generators. Steam generator 2 level was restored in about 5 minutes by operation of auxiliary feed pump 2, which had been out of service for surveillance testing as required by the DB-1 Technical Specifications. Auxiliary feedwater was supplied to steam generator 1 normally. The improper switchgear fuse was replaced.
See Reportable Occ4rrence NP-33-79-13 dated February 9,1979. At the time of the occurrence, the Unit was in Mode 1 at 2772 MWT.
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APPENDIX S ARKANSAS NUCLEAR ONE, Ut'IT 1 ARKANSAS POWER & LIGHT COMPANY Response to Item 2 of I&E Bulletin 79-05A Each Licensee for a B&W operating plant was requested to respond to
,. Item 2 of IE Bulletin 79-05A. Item 2 was stated as follows:
" Review any transients similar to the Davis-Besse event
- . (Enclosure 2 of IE Bulletin 79-05) and any others wnich contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a de-scription of any corrective actions taken. Reference may be made to previous information provided to the NRC, if appropriate, in responding to this item."
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Arkansas Nuclear One - Unit 1 Response to Item 2 We have reviewed similar transients at ANO-1 inclusive of Loss of Offsite Power, Loss of Feedwater, Turbine Trip, Load Rejection, and Reactor Trip. For all transients, ANO-1 performed as expected with no significant deviations with the following exception.
Following a reactor trip from 100% power in December,1974, (start-up testing) and again following a reactor trip from 100% power in May,1975, ANO-1 experienced a momentary loss of pressurizer level .
indication. The loss of indication ranged from approximately 20 to -
40 seconds. Following these occurrences the Plant Safety Committee (PSC), the Safety Review Connittee (SRC), and B&W thoroughly analyzed the situation. .-
The results of the analyses indicated that pressurizer level had dropped only approximately 8 inches below 0a indicated.
Approximately 96" of actual pressurizer water level remained in the pressurizer.
It was further determined that the level drop was due to RCS shrinkage from cooling.
Following investigation, we determined that RCS Tave following a reactor trip was slightly lower than design. We further determined that by fine tuning the Integrated Control System (ICS) runback of feedwater and setpoints of steam relief and bypass valves we could maintain an approximately 2F higher Tave which would reduce shrinkage such that pressurizer level indication would no longer be lost following a reactor trip.
ICS runback of feedwater and setpoint adjustment of the steam bypass valves were subsequently adjusted in early 1976 and in early 1977 setpoint adjustment of the steam relief valves was .
subsequently adjusted to increase Tave. As a result, level indication has not been lost on any subsequent transients. -
The results of the PSC, SRC, and B&W reviews indicated that the -
momentary loss of pressurizer level indication was not a safety issue. The loss of indication was not an anomaly of the system,
- but was due to a lack of fine tuning of the system. The two occurrences compared favorably, that is, pressurizer level re-sponded approximately the same in both instances. Further, should pressurizer level have decreased further, Safety Injection would have been automatically initiated at approximately 1500 psig. 1500 psig in the RCS would have been reached considerably before pres-surizer level dropped out of the pressurizer. Therefore, level indication would have been restored by HPI and, as desired, the steam bubble would have remained in the pressurizer.
Further, the NRC recently raised this issue on another B&W unit.
In a February 14, 1979, meeting in Lynchburg, Virginia, with the NRC Special Investigative Team, B&W owners and B&W, we presented information on +5e ANO-1 occurrences and analyses.
We have reviewed our analyses of 1975, and maintain that our con-clusions at that time were and are still valid.
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APPENDIX T
. EXCERPT FROM TMI-2 FSAR
['
\ "
15 1.8 LOSS OF NORMAL FEEDWATER 15 1.8.1 Identification of Causes A loss of feedvater accident results from either a reduction in or the complete loss of nor=al feedvater flow to the steam generators. With loss or reduction of feedvater to the steam generators, the capability of the secondary system to remove the heat generated in the reactor coolant system is impaired. Reactor trip, however, occurs before the steam generator heat transfer capability is significantly reduced. Since the emergency feedvater system is also available to remove the decay heat generated following reactor trip, fuel and reactor
, coolant system boundary system damage vill not occur. Loss of feedvater may re-
~
sult from abnor=al closure of a feedvater valve, pump failure, or a feedvater line break.
. 15.1.8.2 Analysis of Effects and Consequences 15.1.8.2.1 Safety Evaluation Criteria The safety evaluation criteria for this accident are:
- a. The cora ther=al power shall not exceed 112% of rated power.
- b. Reactor coolant system pressure shall not exceed code pressure limits.
15.1.8.2.2 Methods of Analysis
(
A B&W digital computer code (14) was used to determine the characteristics of this accident. Included were e complete kinetics model, pressure model, aver-age fuel rod model, steam demand model with secondary coastdown to decay heat level, coolant transport model, and a si=ulation of the instrumentation for pressure and flux trip. The initial conditions were normal rated power opera-tion without automatic control. Only the Doppler and moderator coefficients of reactivity were used as feedback. The nominal values used for the main pa-rameters in evaluating this accident are given in Table 15.1.8-1. For trip, a
the minimum control rod worth that satisfies the criterion for a shutdown mar-gin of 1% ak/k at the hot standby condition is used through the analysis.
15 1.8.2.3 Results of Analysis For a loss of feedvater accident due to a feedvater valve failure, feedvater pu=p failure, or feedvater line break upstream of the first feedvater line up-stream check valve, the complete loss of normal feedvater has been analyzed as this is the most conservative case. The sequence of events (see Table 15 1.8-2) and the evaluation of consequences are as follows:
- a. Termination of all feedvater results in a reduction in secondary sys-tem heat re= oval capability.
[
15.1.8-1 Am. 43 (T/15/76)
- b. Increased reactor coolant system pressure results in a reactor trip -
which causes the turbine to trip, s .
- c. The turbine trip closes the turbine stop valves.
- d. The turbine driven and the electric dr ?.n emergency feedvater pumps are started on loss of main feedvater pumps, loss of all 4 RCP's or lov feed-line/steamline dp,
- e. Following closure of the turbirs step valves, secondary system steam is relieved through the turbine bypass and steam safety valves. ,
- f. Steam vill be vented to the atmosphere until the turbine bypass valves can handle all excess steam generated.
- g. Eventually, ther=al equilibrium is reestablished; i.e. , the heat re-moval rate (steam flow through the turbine bypass valve) is equal to the heat input (core decay heat).
- h. Decay heat re=cval and cooldown of the reactor coolant system is then provided by steam relief to the condenser through the turbine bypass valves with the feedvater being supplied by the e=ergency feedvater system.
- 1. Figure 151.8-1 shows neutron power, ther=al power, reactor coolant .
system pressure, and core average =oderator te=perature for the tran-sient. (
Since the core ther=al power does not exceed 112% and the reactor coolant sys-tem pressure does not exceed design limits, the safety evaluation criteria are met.
15.1.8.2.h Eaviron= ental Consequences ,
The loss of normal feedvater due to a feedvater line break between the first .
feedvater line upstream check valve and the steam generator results in doses no vorse than those reported for the steam line break accident, Table 15.1.15-h. -
The loss of feedvater due to equip =ent =alfunction or feedvater line break up-stream of the first feedvater line upstream check valves results in doses no -
vorse than those reported for the loss of AC power accident, Table 15.1.9-1.
For either situation the resultant doses are vell within the guidelines of 10 ~
CFR 100.
15 1.8.2.5 Reactor Building Pressure For the reactor building pressure evaluation, the worst conditions following a feedvater line break occur as a result of a break in the =ain feedvater header to a cteam generator. This break location results in the fastest steam genera-tor blovdown and thus the fastest high enthalpy = ass release to the reactor building.
L 15 1.8-2 Am.~$0 (12-8-76)
The flow fran the feedvater system side of the break was computed using the RELAP 3 computer code (USAEC Report IN 1321). All main feedvater flow vac assumed to bypass the steam generators and exit through the break to the reactor building.
A digital computer program ( vas used to determine the affected steam genera-tor blowdown characteristics. This multinode model permitted the detailed pro-gramming of the steam generators and their interconnecting piping and valves within the main steam system. The following assumptions were made:
,. a. The main steam isolation valves and turbine stop valves were left open.
- b. Flov to the turbine is cut off as soon as the secondary pressure drops be-lov the turbine steady-state value (this forces the mass / energy that would have gone to the turbine to go out the break).
- c. Provisions were made to allow the inventory in the unaffected steam genera-tor feedvater line to boil off and pass through the steam generators and out the break (this effect begins when the pressure drops below the sat-uration pressure of the feedvater).
After the blowdown, the building's cooling capability is adequate to handle the residual heat removal frc= the RC system by auxiliary feedvater flow to the affected steam generator.
The mass and energy released to the reactor building are given in Table 15 1.8-3
( .
Reactor building pressure calculations vere made using the CONTEMPT code des-cribed in 6.2.1 3.2.
Using these methods, the peak containment pressure is 35 psig, assuming that passive heat sinks and two emergency fan coolers are available. Thus, the results of the containment pressure analysis for the feedvater line break accidant are within those predicted by the DEA (see 6.2.1.3.2) and the reactor building design pressure of 60 psig.
15.1.8-3 Am 43 (7/15/76f
TABLE 15 1.8-1 -
LOSS OF NORMAL FEEDWATER ACCIDENT PARAMETERS
\
Doppler coefficient at rated power (BOL),
-1.22 10 ak/k/F Moderator coefficient at rated power (BOL),
-N +0.9 10 ak/k/F Trip parameters Delay for high-pressure trip, s 0.5 Delay for high-flux trip, s 0.3 Control rod travel time to 2/3 insertion, s 1.4 ,
TABLE 15.1.8-2
SUMMARY
OF LOSS OF NORMAL FEEINATER ANALYSIS Reactor trip, s 13.4 Emergency feedvater initiation, s 40 Maxi =us reactor coolant system pressure, psia 2515 M=vi e core ther=al power, % 100
/
15.1.8-h Am. 20 (9-27-74)
Table 15.1.8-3. Feedvater Line Break Transient Mass and Enersv Realease Rates Time after -
rupture during S0 Mass release Energy release blowdown e rate. Ib/s rete. Stu/s 0 0 0 0.1 1.01+k 5.67+6 0.2 1.0k+k 5.71+6 0.3 1.0k+b 5.69+6 0.4 1.03+k 5.65+6 0.5 1.01+4 5.60+6 0.6 9.88+3 5.5k+6 0.7 9.65+3 5.L8+6 0.8 9.69+3 5.L8+6
" 09 9.53+3 5.k3+6 1.0 9 10+3 5.32+6 1.2 9.37+3 5.ho+6 1.k 9.17+3 5.35+6 1.6 9 10+3 5.3k+6 e 1.8 9 2?+3 5 38+6 2.0 8.C3+3 5.29+6 2.2 9.20+3 5.3?+6 2.h 9.21+3 5.39+6 2.6 8.98+3 5.32+6 2.8 9.03+3 5.33+6 3.0 9.19+3 5 3S+6 3.2 8.80+3 5.27 6 3.h 8.96+3 S.31+6 3.6 8.80+3 5.27+6 3.8 8.86+3 5.30+6 k.0 8.65+3 5.22+6 h.6 8.3k+3 5.15+6 5.0 8.03+3 5.ek+6 5.2 7.73+3 k.9h+6 5.h 7.28+3 k.76+6 5.6 6.67+3 h.56+6 5.8 6.05+3 k.3?+6
(' 6.0 6.2 5.56+3 5 13+3 b.1k+6 h.0 +6 6.h k.81+3 3.9 +6 6.6 k.60+3 3.8k+6 6.8 h.ho+3 3.77+6 7.0 k.19+3 3 69+6 7.2 3.96+3 3.60+6 7.h 3. Th+ 3 3.52+6 7.6 3.57+3 3.k5+6 7.8 3.k +3 3.39+6
' 8.0 3.26+3 3 33+6 8.2 3.12+3 3.26+6 8.h 3 00+3 3.?!+6
- 8.6 2.89+3 3.18+6 8.8 2.81+3 3.1 k + 6 9.0 2.7k+3 3.12+6 9.2 2.69+3 3.c9+6 9.h 2.63+3 3. c7 +6 9.6 2.59+3 3.35+6
. 9.8 2.58+3 3.ek+6 10.0 2 5k+3 3.06+6 11.0 2 51+3 3.1k+6 17.0 2.<1+3 3.15 +6 13.0 2.k9+3 3.1k+6 lb.0 2.k7+3 3.11+6 15.0 2.k3+3 3.07+6 16.0 2.38+3 3.01+6 18.0 2.P9 63 2.90+6 20.0 2.18+3 2.77+6 22.0 2.07+3 2.63+6 2k.0 1 9h+3 2.k6+6 26.0 1.83+3 2.33+6 29.0 1.69+3 2.1566 30.0 1.60+3 2.03+6 32.0 1.k9+3 1.90+6 g 52.0 1.19+ 1.k2+6 Th.1 1.19+ "I 1.k2+6I *I TOTAIS 163. 7+ 3 167.2+6
- Extrapelsted.
15 1.8-5 Am. 43 (T/15/76)
TABLE 15.1.8-3 [
FEEDWATER LINE BREAK TRANSIENT MASS AND ENERGY RELEASE RATES (CONT'D)
Time after Mass release Energy release rupture, s rate, lb/s rate, Btu /s Feedvater Picing Release 0.0 - 0.1 5.1700 (3) 2.0739 (6) 0.1 - 0.2 5 7439 (3) 2.3007 (6) -
0.2 - 0 3 5.7097 (3) 2.2862 (6)
O.3 - 0.4 5.6962 (3) 2.2801 (6) 0.4 - 0.5 5.6935 (3) 2.2782 (6) 0.5 - 0.6 5.6933 (3) 2.277h (6) -
O.6 - 0.7 5.6835 (3) 2.2727 (6) 0.7 - 0.8 5.6714 (3) 2.2670 (6) 0.8 - 0.9 5.6592 (3) 2.261h (6) 0.9 - 1.0 5.6h72 (3) 2.255L ')
1.0 - 1.1 5.6352 (3) 2.2503 (6) 1.1 - 1.2 5.6235 (3) 2.2448 (6) 1.2 - 1 3 5.6125 '-) 2.2396 (6) 1.3 - 1.4 5.5996 (3) 2.2337 (6) 1.h - 1.5 5.59h6 (3) 2.2309 (6) 1.5 - 1.6 5.5862 (3) 2.2267 (6) (
1.6 - 1.7 5.57T6 (3) 2.2225 (6) (
1 7 - 1.8 5.5692 (3) 2.2184 (6) 1.8 - 1.9 5.5606 (3) 2.2141 (6) 1.9 - 2.0 5.5521 (3) 2.2100 (6) 2.0 - 2 5 5.5264 (3) 2.197h (6) 2.5 - 3.0 5.4826 (3) 2.1759 (6) 3.0 - 3.5 5.h372 (3) 2.1538 (6) 3 5 - k.0 5.3903 (3) 2.1311 (6) 4.0 - 4.5 5.3410 (3) 2.1070 (6) 4.5 - 5.0 5.2913 (3) 2.0835 (6) '
5.0 - 5 5 5.2373 (3) 2.0578 (6) 5 5 - 6.0 5.1833 (3) 2.0323 (6) 6.0 - 7 0 5.0931 (3) 1 9901 (6) 7.0 - 8.0 h.96h5 (3) 1 9265 (6) 8.0 - 9.0 h.8190 (3) 1.86h8 (6) .
9.0 - 10.0 h.6545 (3) 1.7916 (6) 10.0 - 12.0 4.3704 (3) 1.6675 (6) 12.0 - 1h.0 3.9281 (3) 1.h773 (6) 14.0 - 16.0 3.4667 (3) 1.2868 (6) 16.0 - 18.0 3.1987 (3) 1.1680 (6) 18.0 - 20.0 2.977h (3) 1.0680 (6) 20.0 - 22.0 2.8682 (3) 1.0087 (6) 22.0 - 24.0 2.6894 (3) 0 9255 (6) 24.0 - 27.0 2.2193 (3) 0 7418 (6) 27 0 - 30.0 1.8091 (3) 0.58509 (6) 31.0 - h0.0 1.4706 (3) 0.45196 (6)
(.
15.1.8-6 Am. 20 (9-27-74)
TABLE 15.1.8-3 FEEDWATER LINE BREAK TRANSIENT MASS AND ENERGY RELEASE RATES (C0h'r'D)
Time after Mass release Energy release rupture, s rate, lb/s rate, Btu /s Feedvater Piping Release 41.0 - 50.0 1 3460 (3) 0.39547 (6)
J 51.0 - 60.0 1.2969 (3) 0.36483 (6) 61.0 - 70.0 1.2765 (3) 0.34301 (6) 71.0 - 80.0 1.2478 (3) 0.31969 (6) 81.0 - 90.0 1.3528 (3) 0.33303 (6) 91.0 - 100.0 1.4351 (3) 0.33252 (6) 101.0 - 110.0 1.3108 (3) 0.28811 (6) 111.0 - 120.0 1.1257 (3) 0.23583 (6) 121.0 - 130.0 0.9215 (3) 0.185215 (6) 131.0 - 1h0.0 1.0393 (3) 0.20153 (6) 1h1.0 - 150.0 0.8766 (3) 0.16h58 (6)
Total releases 2.1693 (5) 8.29007 (T)
L 15 1.8-7 Am. 20 (9-27-Th)
4 4
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TEST TARGET (MT-3) 1.0 lg a su lE I.I b'l;?Il!E El=4 1.8
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< 6" MICROCOPY RESOLUTION TEST CHART
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---. . -l ~=--..._a .
120 Heu tron 80 -
Power,7 60 -
40
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20 120 100 Th ermal 80 -
Power,f. 60 -
40 -
I i i e i i i i i g
+6 Average +6 -
Core +4 -
Mode r ato r +2 -
Tempe ratu re 'O Ch ange, F -2 -
_q t i i i i i i i I 2600 Reactor 2500 -
System Pressure, 2400 -
psi a ,
2300 -
2200 O 2 4 6 8 10 12- 14 16 18 20 Time, s LOSS OF NORMAL FEEDWATER FROM RATED POWER T11REE MILE ISLAND NUCt EAR STATION UNIT 2 6
~ , -
FIGURE 15.1.t-1
e
- i i APPENDIX U LETTER FROM BABC0CK & WILCOX APRIL 30,1979 e
e e
S B s4 u C C u"w' .'."u vOn" pcm co-,m ~. ce:o P.C. Scr 1250. L, : : rg, V:. 2 C-Tele::h re: 7,5;a' 224.nI1 April 50, 1979 Dr. R. J. '!a t ts c n Director, Division of Systems Safety .*
Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Subject:
Babcock 5 Wilcox Company's Commitments
Dear Dr. Mattson:
Attached is a summary of the Rabcock 5 Wilcox Company's commitments that have resulted from various meetings and correspondence over the past few weeks; this is a complete list of our commitments as I see them. I have indicated the date by which B5W intends to complete each commitment: however, no allowance has been made for prior review of these submittals by the licensees. Should they require prior review, submittals to the Nuclear Regulatory Commission could be delayed slithtly.
It should be noted that some of the dates have been extended beyond those originally discussed with the staff because of the very signific. ant work effort required in connection with the small break guidelines and procedures. *
~
Also attached is a copy of the meeting minutes from the April 26, 1979 meeting with the NRC staff. ,
If you have any questions, please call me (Ext. 2817). .
Very truly yours, s
{0/pQ}'(O'
' James H. Taylor
'j'
- MU l'
f Manager, Licensing JHT:nw Attachments bec: E A_Wenack (f,.F. Wa_iqjQ_'r '
D.W. LaBelle B.M. Dunn D.H. Roy
+ - -
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I. Conmitments f rcr 2-':/:;?C Meeting on . Acril 17, 1979 (Reference _1 ,..
A. Perform calculations, worst-case break without AF':
for 30 minutes.
Due: April 21, 1979 Submitted: April 21, 1979
- Outstanding: Detailed results discussed with staff on April 26, 1979. Detailed report to be submitted May 21,197 9. (See Reference 2, Item 5.)
B. Document natural circulation tests conducted at Davis Besse and Oconee.
s Due: May 7, 1979 C. Document all occurrences of natural circulation which happened inadvertently- include a description of unexpected behavior.
Due: May 7, 1979 D. Document natural circulation analv.tical methods.
Uce. May 16, 1979 E. Summarice and document sensitivity in key paraneters (not to be started until release of R. Tedesco report) .
Due: Eight weeks following receipt of Tedesco report.
i 1 -
d
% e F. Deleto'I G. Define and document thernal shock criteria fcr cpara:ic.-
at lo .- terrpora:ure tith HP purps running.and no natural circulat;cn.
Due: ?.-co wee?s follc.;ing rcccipt of Tedas o report.
H. Assessment of the safety concerns raised in the report of Dr. Michelse..
Due: May 7, 1979 Outstanding: Basis for concluding that Michelson concerns do not invalidate ICCFR50.46 analyses for small breaks were discussed with staff en April 17, 1979 (See Reference 1, page 7) and on April 25, 1979.
II. Commitments in Taylor to Mattson Letter of April 25, 1979 (Reference 2)
A. CRAFT Analyses
- 1. Item 1 (Reference 2)
Due: May 4, 1979 Status: Some detailed results submitted herewith Ficures 23-30.
Remainder to be submitted May 21, 1979.
- 2. Item 2 Due: May 4, 1979 Status: Details discussed with and some results submitted to NRC staff on April 16, 1979. .
Figures 17-21 submitted herewith.
Remainder to be submitted May 21, 1979.
- 3. Item 3 Due: May 4, 1979 To be cc=bined with Item 4.
- 4. Item 4 Due: May 4, 1979 Some detailed results discussed with staff on April 26, 1979. Finures 6-11 nulmittel herewith. RemainI 5er to be submitted M.r. Jt, 1".
2 -
- , . . . . ~ ~ - . .-
- 5. Itc= 5 Due: :.2y 4, 1979 Submitted: See Item I..'.. above
- 6. Item 6 Due: "ay 4, 1979
,. To be submitted ':ay 21, 1979.
- 7. Item 7 Due: May 4, 1979 Detail results discussed with staff on April 26, 1979. Figures 22-27 submitted herewith.
Remainder to be submitted May 21, 1979.
- 8. Item S Due: May 4, 1979 To be submitted May 21, 1979. .
B. CADDS Analyses
- 1. Item 1 (Reference 2)
Due: May 4, 1979 Detail results discussed with staff on
- April 26, 1979. Figures 1-3 submitted herewith.
Remainder to be submitted May 9, 1979.
- 2. Item 2.a.
Due: May 4, 1979 Detail results discussed with staff on April 26, 1979. Ficures 4-5 submitted herewith.
Remainder to be subditted May 9, 1979.
- 3. Item 2.b.
Due: May 4, 1979 To be submitted May 9, 1979.
3 -
_ _ _ _ . . _ . _ _ . . _ . . . _ _ . ~ . _ . . . .- __ . _ . _ . . . _ . . _ _ _ . . . . . . . _ _ . . _ . . _ . - . _ _ . . . . .
- 4. Item 2.c.
Due: May 4, 1979 To be subnitted May 9, 1979.
- 5. Item 2.d.
Due: May 4, 1979 To be submitted May 9, 1979.
III. Commitments in Roy to Mattson Letter of April 26, 1979 (Reference 3) .
A. Details of results of the analyses described in Reference 2.
Due:
Submitted: See II Above Outstanding:
B. Details of B&W's evaluation of the Michelson report Due:
Submitted:
Outstanding: See I.H. Above C. System response to total loss of steam generator heat sink To be submitted May 25, 1979.
D. Sensitivity study of system respon?a to auxiliary feedwater flow rate .
To be submitted May 25, 1979.
E. Effect of anticipatory trip on loss of main feedwater See II.B.2 above IV. Staff Requests for Additional Analyses at B&W/NRC Meetinc of April 26, 1979 (Reference 4)
A. Benchmark analysis of sequential auxiliary feedwater flow to OTSG's for LOMFM.
To be submitted May 21, 1979.
_ a _
B. System response to PORV and code safety val /c actuation.
To be submitted June 1, 1979.
C. Ideas on benchmarking of natural circulation modes of cooling CRAFT II.
To be submitted July 2, 1979.
I D. Evaluation of Michelson report concerns and outline of operating criteria for'small breaks.
, Due: See I.H. above and V.A. below. I
, E. Worst case small break with no auxiliary feedwater flow and single ECCS failure To be submitted July 2, 1979.
V. Analysis Commitments in MacMillan to Denton Letter of April 26, 1979 (Reference 5)--Reliab111ty Analysis of ICS Due:
, Submitted:
Outstanding: Scope and schedule were submitted on April 28, 1979, by letter J.H. Taylor to H.R. Denton.
VI..
Analysis Commitments in W. S. Lee to E. R. Denton Letter of April 26, 1979 (Reference 6)
A. Operating instruction for management of small breaks
. Due: May 15, 1979 (procedures in control room)
Submitted:
Outstanding: B&W to submit guidelines for developing procedure, approved by,NRC, to Duke Power Co. on or before May 12, 1979.
B. ICS FMEA Due: See V above 5 -
n..A B&W E:!GI::EERI'.:.0 (CTHER u, ,2 . _c.,,
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. r .' .' .= m TF.ANSIE::TS I:I B&W PLA:;TS" OF APRIL 25, 1979 I. Commit:'ents in MacMillan to Centon Letter of Acril 25, 1973 (Reference 1)--Caveico .'eans for Cecoupilng A'-.i d ar7 F2:L- '
wate r Centrol f ro-' ICS Submitted: April 23, 1979
- II. Commitments in MacMillan to Denton Letter of April 26, 1979 (Reference 2)
A. More hositive indication of the position of the pilot-operated relief valve Due: May 28, 1979 Completion on this date consists of transmittal of technical hardware description to operating plant owners.
B. Saturated condition indicator for reactor coolant Due: May 30, 1979 ,
Completion on this date consists of transmittal of technical hardware description to operating plant '
owners.
D 9
e 5
0
1725 K St eet N.7,'.. b.::.m-: nr., a C. M .; . -
s-t CCC.'sb',.r,}CCX Te!e::hene: (202) 230 0: 90 b E. .
April 26, 1979 Mr. Harold R. Denton, Director Office of t'uclear Reactor Regulation Nuclear Regulatory Commission
. 7920 ::orfolk Avenue Bethesda, Maryland 20555
Dear Mr. Denton:
Subject:
Integrated Control System This letter documents the commitment of Babcock & Wilcox to undertake a reliability analysis of the Integrated Control System (ICS) which will include a failure mode and effects analysis.
This analysis will identify sources of transients, if any, initiated by the ICS and develop recommended design improvements which may be necessary to reduce the frequency of these transienrs.
In addition, means will be developed for decoupling of the auxiliary feedwater control of steam generator water level from the ICS. This modification will provide control of feedwater under emergency conditions independent of the ICS.
The scope of the reliability analysis and schedule for both the analysis and development of independent feedwater control will be provided within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- Sincerely,
"*\
Jo H. MacAil an
'N Vic President Nucl r Power Generation Division cc: W. S. Lee Duke Power Company John Mattimoe Sacramento Municipal Utility District William Cavanaugh Arkansas Powcr & Light
! William Griffin Florida Power Corporation John Herbein l Metropolitan Edison Cciapany Lowell Roc T.oledo Edison Company
. . _ . . _ . . . . . . . - . . . . - - . . . - - . . . . - - . . - . - - - . . - . . _ - - . . . . ~ . . . _. . . _ _ .
1725 K D e * * " -r- . O C 2^:25 Ea.c c c c,d &. ,,....ncox Teienomea2:23 2es.cm April 26, 1979 Mr. Harold R. Denton, Director Office of Nuclear Eeactor Reculation
- Nuclear Regulatory Ccamission 7920 Norfolk Avenue Bethesda, Maryland 20555
Dear Mr. Denton:
Subject:
Near-Term Design Improvements
. In the April 16, 1979 meeting with the ACRS, I identified several near-term actions which Babcock & Wilcox was cormitred to undertake. Two near-term design improvements which evolved from our evaluation of the TMI-2 accident are a more positive indication of the position of the power operated relief valve and a saturated temperature condition indicator for the reactor coolant systen. These instruments will provide additional information to the operators which i= prove their ability to identify an open relief valve and maintain subcooled temperatures in the reactor coolant system to provide adequate core cooling.
These improvements are curren..ly in the design and development phases. The schedule for completion is consistent with the six week commitment indicated at the ACRS meeting.
Sincerely, ,
A
- s Johb H. MacMillan .
Vice resident Nuclear Power Generation
- Division cc: W. S. Lee Duke Power Ccmpany John Mattimoe Sacramento Municipal Utility District William Cavanaugh Arkansas Power & Light William Griffin Florida Power Corporation John Herbein Metropolitan Edison Company Lowell Roe Toledo Edison Company
t MI::UTES OF THE MEETI::G OF n.n . . . ,..-..,..,--,......,,I.*, ,
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. J DISCCSSID:' C:: S:'.ALL LCCA A::AL' ISIS April 26, 1979 Short Term Transient 2nalysis l
CADD-S ,
See CADDS stufics list (P.4), Acril 25, 1979 ~
B&W letterl General Mcdel Characteristics o Used for analyses up until time system becomes 2-phase o Suitable for sensitivity analyses for delays of auxiliary feedwater, variation in feedwater flow, variations in reactor trip mode or se: point.
o System actions which will be investigated (cut to about 8-10 minutes) are:
Reactor trip time
- Peak pressures achieved in initial pressurization
- Time and valve of repressurization Time to fill pressuriner e Code Limitations
- Not valid for 2-p'mse, saturated conditions
. (use CRAFT)
- HPI not precisely modeled (use CRAFT)
One loop TMI-2 Benchmark (Case #1 of Ref. A CADDS study)
Curves of pressure, pressuriner level comparison to TMI-2 transient of March 28, 1979, are in Attachment 1, Figures 1-3.
Sensitivity of Reoressurization to Auxiliary Feedwater Initiation Time (Case #2, CADDS study)
Curves of pressure and pressuriner level for three representative cases are in Attachment 2, Figures 4-5.
Staff Recuest 1: Show benchmark to plant data (e.g.,
Davis J2:s cr equ valent) for case 'ehera generators fill in scquance.
Staff Recuest 2: Examine parametric behavior ~ of PORV 's and Safety valves on pressurizer.
What is cycrating experience with safety valves opening and cicsing?
Why not consider failure of the safety valve as a single failure? i What are che consequences and expected behavior of a stuck cpen pressuriner safety valv=?
Consider 3:eca and 2-phase f1c? discharge. ,
- Define basis of and treatment justification for fic'.-
model through the valves. Include cuench tank back pressure effect assessment.
Long Tern Transient Analysis CRAFT-2 See CRAFT analyses, April 25, 1979, E&W letter 1 Model Used e Noding as described in Figure 12 o Model handles three modes of natural circulaticn Solid water 2-phase mass movement Boiling / recondensation e Natural circulation model of B&W is believed to account correctly for these effects, and is similar to Comnis- .
sion audit models. Data for benchmark to~ actual system conditions is available .only for solid water mode. -
Staff Request #3: Further discussions, with the aim of developing benchmarks, are needed.
Michelson Repo_rt B&W considers interrupted natural circulation as an acceptable cooling mode.
Staff Recuest d4: Provide a descriptica of this cooling mode and outline of emergency operating criteria for the operator to handle it.
2 -
. . . . . .. . ... ~.- -. . . - . . . . . .
TMI-2 Banc5 trP (Case #4 cf Ref. 1)
See attachment, Figures 6-11.
Conclusion--Existinc ccdes are cacable of handline nhenenena seen n T:C _ case anc s;m iar transients.
Loss of Feedwater ir Coniunction rith 0.01 sc. ft. break (Cace 45 of Ref.1) e Break size selected to be the largest which would not automaticallj initiate ECCS high pressure injection in the initial depressurization transient.
o Shows that core damage will not occur in the first 20 minutes of operation in the following mode
- No auxiliary feedwater
- No ECCS injection 0.01 ft.' break Action within 20 minutes to establish either auxiliary feedwater* or HPI will avoid core damage for all plants except Davis Besse. At Davis Besse, feedwater would have to be restored, but time available to accomplish this without core uncovery will be somewhat longer due to loop configuration. (* Initiation of AFW will result in ECCS HPI initiation automatically.)
e For breaks larger than 0.01 ft.2, for which ECCS will automatically initiate, there is no need for auxiliary feedwater so long as ECCS function is unimpaired. See
- attachment, Figures 12-16.
- Staff Request SS: Analyze worst case small break assuming a single failure in the ECCS and no AFW. (B&W noted that this would be a very low probability event.)
Presented Loss of Feedwater with Stuck Open PORV Case with RCS pumps running with AFW (Case #2 of Ref. 1)
- AFW on in 40 seconds.
- PORV stuck open on initial pressure transient.
Date in attachment, Figures 17-21.
Results: No core damage Presented Stuck Open PORV as the Initiatinc Event (Case 47 of Ref. 1) -
3 -
S a. a -"**- a c . .." n. .". *. , r. i g " ". e s .7 2 - 2 7 .
Results: ':0 core damage Not Presented:
1 Case -1 of Ref. 1a see CRAFT anal. eses' See attac'r.ent, Ficures 23-30. -
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1 Case No. 2 Results e
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Reference:
Letter from J. H. Taylor to R. J, Mattson, April 25, 1979.
~
Case No. 7 Results b
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Reference:
Letter from J. H. Taylor to R. J.'Mattson, April 25, 1979.
t e
Case No. 1 Results O
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D Tabic 1.
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. 1. Main feedwater pumps trip at ti=e = 0.0 seconds. *
- Note: All RC pumps remain powered throughout the transient.
- 2. Pressuri:cr EMOV valve open at 6.0 seconds and re=ains open (stuck).
- 3. Reactor scracs at 12.0 seconds.
- 4. Auxiliary feodwater starts at 40.0 seconds after loss of cain feedwater (auxiliary feedwater level is set at 30 inches).
1
- 6. Long term cooling established at 155 seconds. .
- 7. Pressurizer goes solid at approximately 425 seconds.
4 b
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APPENDIX V k.5 COLD SHU'QCWU CAPA3ILI?f EXCERPT FROM TMI-1 FSAR .
'"he adequacy of the borated water storage tank as an interi= heat si for the Three Mile Island Nuclear Statio.1, Unit 1, reactor ecolant syste= has been evaluated for the folleving set of assu=ptions:
- a. Stes= line break occurs inside the inter =ediate or turbine building during rated power operation
- c. Loss of all feedvater to both stea= generators occurs
- d. Loss of off-site power occurs i In addition to this set of assu=ptions, tais evaluation is valid for any situation where reactor coolant syste= energy removal thrcugn the stea= generators is no longer available.
~.
There are three primary areas of concern for this condition. These areas are prevention against core uncovering, protection against excessive reactor building pressure, a=d the ability to achieve cold shutdevn conditiccs .
"he 3W digital ec=puter ecde C2AFI (10) vas used to deter =ine the characteristics of this accident with regard to core uncovering an:1 nass energy releases to the con-tain=ent. The = ass and energy release data frc= CRAFI was used in the digital cc=puter code CONIDIPT (11) for reactor building pressure calculations. Ihe assu=ptions and results of the analysis are s-wited in Table 6. A single stes= generator blev-devn was considered as the =ost censervative case since for a double blevdevn the HPI pu=p .rculd be started *ost instantaneously on icv reactor ecolant syste= pressure actuation (1500 psig) =ecning a lever probability of core uncovering.
Core uncovering is prevented by pu= ping vater frc= the borated water storage tank via the =akeup and purificatic syate= (EPI) into the reactor ecolant syste=. With one
akeup and purificatics (EPI) pu=p started 15 =inutes after the break, the =ini==
coolant level in the reacter vessel cecurs at approxi=ately lk0 =inutes and at no time falls below the top of the core. Operator action is assu=ed to occur 15 ninutes after the break in starting the =akeup and purification pu=p (high pressure injection).
Ihe building pressure ' increases during "the transient as boiloff occurs through the
. pressurizer safety valves (2515 psia). Assu=ing the bciloff goes directly to the building atmosphere with no credit for the f quench tank, the building pressure reaches the reactor building cooler and high pressure injection setpcint (k psig) ',8 ninutes after the break. With cne building cooler cperative at this time, the building pressure reaches a =axi=== value of 2k psig and never exceeds the design pressure li=it.
Further= ore, the reactor spray actuation setpoint (30 psig) is not reached and a single building cooler provides adequate protection throughout the transient against excessive reactor building pressure.
High pressure injection of SWST vater continues until the 3WST is depleted (approxi-
=ately 2k hours assu=ing one EPI pu=p is operating.) At this ti=e further cecidevn is achieved by using the decay heat (low pressure injection) pu=ps draving fren the reactor tuilding su.p to supply suction to the =akeup and purification (HPI) pu=ps.
Ihe su=p recirculation ecutinues until the decay heat re= oval syste= (LPI) can be actuated to reduce the syste= to cold shutdown. Cold shutdcvn is then achieved by venting the syste= pressure and actuating the decsy heat re=cval syste= to recirculate the reactor coolant through the decay heat ecclers.
Supple =ent 2, Part IX A=. h3 (11-1-73)
5.0 E!ERGENCY PRCCEDUPIS The e=ergency precedures belev are genert in nature since it is deened appropriate to allev for assessment of the incident prior to ulti=ately bringing the reactor to cold shutdevn.
51 STGTCMS (STEAM LINE BREAK)
- a. Rapid decrease of secondary steam pressure.
- b. A steas line break detection system actuated alar =.
- c. Megavatts generated reducing rapidly.
- d. Decrease in pressuricer level, reactor ecolant pressure, and cold leg te=perature.
- e. For a rupture cutside the reactor building noise vill be heard in the control rcen or a report =ade frc= personnel outside the control rocc.
5.2 n! MEDIATE ACTION
- a. Autenatic Action
- 1. Steas line bresk feedvater shut-off sfste= actuates (< 600 psi) and the lov 1 cad centrol valves 3"J-V-16A and 163, =ain feidvater valves P4-V-17A and 173, and emergency feedvater valves IF-V-30A and 303 close.
- 2. Reactor trips ,
3 ?rbine trips ,
- k. High pressure injection initiates if Icv reactor coolant pressure of .
1500 psig or reactor building pressure of k psig is reached.
5 Reactor building cooler actuation due to h psig in the reactor building,
- b. Manual Actics
- 1. Veriff that the reactor has tripped; if not, trip it.
/
- 2. VeriP/ the turbine has tripped (=ain stop valves closed); if not, trip it.
s 3 Notiff shirt fer-- - that the reactor has tripped.
Supple =ent 2, Part II A=. h3 (11-1-73)
L. Determine which steam generator has suffered the rupture frem the .
steam line break detection system in the ccatrol recs.
5 Verify that lov lead centrol valves FV-V-16A er 16B, = sin feedvater valves FW-V-17A or iTB, and emergency feedvater valves F.F-V-30A cr 20B en the affected steam generator are in the closed position.
- 6. Initiate e=ergency feedvater supply to the unaffected stes= generater.
7 Letermine if the mkeup and purificatien syste= (high pressure injection) has started. Manually initiate it if both steam generators are inoperative and pressure setpoints have been exceeded.
i 5.3 LCNG TEPR ACTICN (F.=ergency Feedvater)
If there are indications that the e=ergency feedvater system is not verking properly, enter the inter =ediate building as scen as possible. Inspect the e=ergency feehater syste= to deter =ine if it has experienced any an age.
Line up the'unda= aged e=ergency systems to supply water to the unaffected steam generator. Open stes= db=p valves on the unaffected stea= generator.
The valves nay have to be operated using handvheels if the cabling has been
- n= aged by the break. When the e=ergency feedvater valves have been . ;ned up, start the e=ergency feedvater pu=ps. Throttle the feedvater contivl valves to -Mntain high level in w ? -as generator.
- inen the reacter coolant system pressure has decreased sufficiently initiate the decay heat re=cval syste=.
5.k LCNG TIPR ACTICH (Feed and 31ead)
When the centents of the berated water storage tank are depleted as deter =ined from the borated water storage tank low-lev level ala:-s in the centrol recs, shift suction of high pressure i=jection frem the berated water storage tank to the reactor building su=p by cpening valves EH-V-TA and T3, EH-V-6A and 6B, and closing valves EE-V-5A and 53 (all re=otely centrolled frc= the control buil:iing) .
When reactor coolant system ta=perature is belev kk0 F, clcse the core flecd line discharge valves CF-1A and 13, secure the nakeup and purificatien pu=p (EPI) ,
and depressurice the reactor coolant syste= by cpening the pressu:1:er electro-
atic relief valve er pressurizer sample line (both re=ctely centrolled frc
the centrol building). To initiate deesy heat re=cval, cpen valves EE-1, OH-2,
- and EE-3 (nor=al decay heat let-down line) .
6.0 SIMARY AND CCNCLUSICNS
'"he results of this design review are st==arized as fc11cvs:
- a. A rupture of the high energy piping systems is considered highly unlikely.
The syste=s have been conservatively designed in accordance vith the criteria in the B31.1.0 Code for ?over Piping. Materials , fabrication, and quality assurance require =ents of the code have been utilized. In additica, the nain steam piping has been subject to 100 percent radicgraphy ,
of velds frem the steam generators to the turbine step valves , and the Supple =ent 2, Part IX Am. L1 (7-16-73)
TABLE 6 CHRONOLOGY OF ETCTIS FOR HIGH EERGY pI?E 3REAK Time hecends) Event 0 Double-endrd break of a 2k inch diameter stes=
e line en the secondary side 1 Reactor trip en variable icv pressure; turbine step valves close isolating t?.e uaaffected stes= i generator kT Da:.sged steam generator blevs dry ,
h50 Unaffected stes: generater provides no =cre heat sink; =ini=u= syste= pressure of about 1550 psia is reached 900 Operater action starts ene HPI pu=p 1200 Pri=ary loop becc=es solid with subecoled vater; pressuri::er code relief valve opens at setpoint
,- of 2515 psia
(
2300 Reacter building cceler actuation setpcint of k psig is reached 5700 Stea= first appears in the core 3500 Mini =u= coolant level in reacter vessel is
, reached; core re=ains cevered 8800 Contain=ent building pressure reaches the maxi =u=
value of 2h psig -
a o
N Supple =ent 2. Part IX A=. h3 (11-1-73)
APPEtiDIX W PORTLAfl0 GEf1ERhi. ELECTRIC COMPAtlY Responses to ACRS Questions on Pebble Springs A preliminary assess =ent has indicated that the double-ended rupture of up to .3 t'ubes during a LCCA would not seriously i= pair the capability to reflood and cool the core in accordance with the conservative requirenents of Appendix K to 10 CTR Part 50.
CUESTION 5 What is the maximum secondary system pressure developed af ter J curbine trip with first subsequent randem failure be.ing loss of nain feedvater flow control leading. to flooding of super-
~m beat section of steam generators. Assu=e turbine trip vichout bypass (loss of condenser vacuum).
Resconse to Ouestion 5 The maximum secondary side pressure developed, assuming turbine trip without bypass and a subsequent loss of sai= fa adwater flev control, is equal to the setpoint of the main stez: safety valves.
There are evo b2nks of safety valves. The "high" b. k setpoint is about 1315 psia which includes 3~. accumulation. The maxi =tr allowable steam generator pressure is 1375 psia.
CUESTION 6 Does applicant know that time-dependent levels vill occur in pressuriaer, steam generator and reactor vessel after a rela-tively small primary coolant break vnich causes coolant to approach or even partly uncover fuel pins? What does operator do in respect to interpreting level in pressuriaer?
During primary "sys tem refill frem high pressure injection pu=ps there is some period vben ::ither condensation nor natural convection is present to effect beat trans po rt to secondary side. How is transition to natural convection without assistance frem ori=arv coolant ou=es obtained.
Resoonse to Question 6 There are two overriding concerns with any LCCA:
(1) Initial removal of fuel-scored heat. .
(2) Continuous removal of core fission product decay heat.
For small breaks, fuel-stored heat is removed during the first few I
seconds of blowdovu. The B&W ECCS system, using internal vent valves, precludes the interruption of decay heat recoval for all accidents within the range of relatively small breaks (break size ,.
<0.01 ft 2). 3reak location, ECCS injection, coolant phase separation, Reactor Coolant System (RCS) mixture levels and sca.s generator conden-sation have been considered in arriving at this conclusion.
As we understand the question, the concern is related to possible interruption of steam condensation within a steam generator due to refilling of the primary sys tem. In general, such a situation can occur only at extended times during the final recovery stage of a LCCA when steam condensation is no longer required. However, even if this situation occurred earlier in time, the perfor=ance of the vent ralves vould be to equalize water levels between the hot and cold regions of the primary sys tes, thereby assuring continuous fluid coverage of the core with no adverse consequences.
This is substantiated by a sore detailed exznination of the fluid conditions during a relatively stall LCCA. Such an accident can be viewed as a very slow transient during which, at ary particular -
time, the sys tem is not meaningfully different frem steady-state .
cond'itio ns. The ECS can then be properly described as a sealed nano-anter. For the B&W system, because of the vent valves, this =anometer is double looped as illustrated in Figure 6-1 with i=portant volu=es identified by letters.
Many experi=ents have been run which show that as long as a fluid (quality less than, say, 70%) covers the core, no adverse core temperature excursion can occur at decay heat power levels. Thus, the design problem associated with s=all LCCAs is to achieve steady = ass and energy balances which assure that the core re=ains -
covered. This neans that = ass injec: ion equal to = ass loss, and energy removal equal to decay heat is achieved. For a spectrum of bresk sines appropriate for relatively small LCCAs, conservative analysis assures that no uncovering of the core occurs prior to achieving excess mass injec: ion. Thus, any concerns with very small-break LOCAs deal with the energy balance once excess
- injec: ion has been achieved.
For certain small breaks, the stea= generator would ac: as an energy re=oving device. Enargy re= oval occurs through a three-step sequence: initially, a solid flow-forced convection process would control heat re= oval, later a two phase natural circulacon process involving both convection rad condensation heat transfer would control, and finally a pure condensation node would result. In this latter mode, fluid has fallen to approximately' level 3 on Figure 5-1. As steza is produced in the core through boiling, it travels through D, F, and G and is condensed in the lower regions of E. Concerns over the i= pact of concendensible gases have been examined for this phase and the folleving points apply:
. (1) Insufficient noncondensibles are available in the
, initial RCS fluid to block the flew of stea= a: G (this is a 3-ft dia=eter pipe).
(2) Heat transfer coefficients with noncondensibles present are sufficiently large to condense stems in the lower regions of E. Even if the heat transfer were accentarily inadequa te, this would nerely cause a pressure increase and resultant temperature increase en:il the temperature difference ec=pensated for the lower heat transfar coefficient.
(3) The open sanometer paths D, F, G, H, and 3 assure that hydrostatic balances exist between regions H and A, and betvuen regions K and A. If these ,
balances do not exist, fluid novement will occur to produce dies.
After excess mass injection is achieved, the RCS starts to refill.
During refill, a rising water level in region H may eliminate g condensing heat tr ans f er. Note that a rise of level in H also ceans a rising level in I and A. Thus, no E==ediate core concern
'~
exists. Steam pockets vill be for=ed at J and C. If the level continues to rise, a two phase mixture vill be forced into D and F. This will occur through the necessity of =aintaining a hydro-static balance with H. However, if condensation ceases, the energy balance is no longer maintained. As energy is not being adequately removed from the system, the system =ust repressurize.
Two mechanisms are now possible:
(1) The break flew increases until it removes enough energy, or the break allows removal of enough = ass to reestablish condensation, or (2) Repressurtsation continues until energy removal is brought ab out through the pressurizer relief valve path E.
Most likely, mechanism (1) vill repeat for several cycles prior to .
zechanism (2) occurring. In any case, uncovering of the core can-
~
not take place. Again, if the core fluid . level is lowered, then the fluid level in H must be lov and condensation is a credible phenemenon.
The flow pattern in D, the horizontal section of the hot leg, is of interest during repressuri:ation. This is illustrated in Figure 6-2 along with the pressures within the system. The following hierarchy of pressures exists:
Po<?3<Pi<?
_9
- mm
G(160'ELGCW) j F(vvoER Her G G)
's y s (xtss&azz)
/
- pC(UP95A.Db) i l Q(L,7. aER !
N( D " mr t.m) te.wmg /
/-
B(vEmvsLvu) 7(um_s ...
C N wCOM7.R)
. 1
.. - - A(coq t0 cwa. -
temuc:.n.cg
. $ ~~
~
I.(CcabLEG)
Iiaure 6-1 35'4 Sys:a: as a Sealed
'danc=ecer ic: a Re12:1,f e;f 5:all LCCA
O s G(ledE1.Gcw)
/
fe f
-} '
7 7 (uppen.
F uct tr.c,)
s J 4 i -
9 yJ spetsseuac
- g, e,
/
g o e
goa C,
e' s.e
- p. -
,,4 a'
.,,sf i '
q<
.c( : wasao o n ez n(mm o K N p
~; .o s
'fl k \ u s 1 j -j u ria p IT'*1 0
} # '
Oc%4 COmtg -
( o gg ,
- A(cent) t(M -
~
' te w e-e.g A
4
"[ (CC'_b LEs)
Fisure 6-2 7;cv Pa::er_. ;uring RCS
. epressuri:2:icn Fo;;:.-i ,;
a Rela:1 fe17 g ,,31; . ;
CUESTION 15_
Considering such satters as (1) off-site power failure, (2) con- .
denser vacuum failure, (3) spurious main feedwater valve closure (see item 21 preceding) and recent incidents of failures in auxiliary feedwater systems it appears that, single failure criteria notwiths tanding , at least short term failures of the auxiliary feedwater system =ust be considered to estimate the needed reliability of such system.
=
Wat , for instance, would be thu peak pri=ary system pressure, consequences to prbary coolant system safety and relief valves and rate of pri=ary coolant loss following failure of the Auxiliary Feedwater pu=ps when needed?
Resconse to Question 26 The feedwater sfste=s are designed to current NRC regulations.
Since these regulations include criteria for design and analysis assuming one single failure, and the safety grade Auxiliary Feedwater System contains cultiple redundant trains (four 50%-size capacity pu=ps are installed with independent power sources), the Pebble Springs design eceplies with the latest require =ents. .
. Postulation of an event whereby all feedwater is lost requires
=ultiple failures in the sain and auxiliary feedwater systems.
Nonetheless , a preli=inary analysis has been made to determine the event sequence, assuming that all feedwater is lost instantaneously without regard for a realis, . =echanism. The folleving is an estimate of the sequence of events expected:
Ti=e Event 0 see All feedvater is lost and the RCS begins to increase in pressure.
- 7 see Reactor trips on high RCS pressure. .
e10 see Pressuri:er begins to relieve decay heat via steam to the RC drain tank at the pressuri:er g safety valve setpoint of 2500 psig (RCS pressure -
about 2740 psig). r
< 2 sin Reactor coolant expansion causes the pressurizer to bece=e water solid, and water relief to the RC drain tank begins (RCS pressure about 2500 psig).
<10 min containment pressure increases to the ESTAS setpoint (a psig), and high pressure ECCS coolant injection to the core starts automatically.
<45 min High pressure ECCS injection flow heat re= oval race is about equal to the decay heat generation race. Prior to this ti=e , boiling has occurred in the core; and af ter this ti=e , it will diminish.
- A coolable gec=etry is maintained at all ti=es.
Long tern ECCS high pressure injection vill continue .
to provide coolant frem the borated water '
storage tank (3WST). When the 3WST low-level signal is reached, the operator can switch the ECCS high pressure coolant injection to the recirculation =ede, if auxiliary or sain feedvater has not been restored (see Pebble Springs Section 6.3.1.4.1 for a discussion t on this mode).
r s
APPENDIX X IE BULLETINS (79-5,79-05A,79-05B,79-06,79-06A,79-06A Rev. 1,79-06B, 79-08) e O
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 IE Bulletin No. 79-05 '
Date: April 1,1979 Page 1 of 3 NUCLEAR INCIDENT AT THREE MILE ISLAND 5
Description of Circunstances:
On March 28, 1979 the Three Mile Island Nuclear Power Plant, Unit 2 '
experienced core damage which resulted from a series of events which were initiated by a loss of feedwater transient. Several aspects of the incident may have general applicability in addition to apparent generic applicability at operating Babcock and Wilcox reactors. This bulletin is provided to inform you of the nuclear incident and to request certain actions.
Actions To Be Taken By Licensees:
(Although the specific causes have not been determined for individual sequences in the Three Mile Island event, some of the following may have contributec).
For Babcock and Wilcox pressurized water reactor facilities with an operating license:
- 1. Review the description (Enclosure 1) of the initiating events and subsequent course of the incident. Also review the evaluation by the NRC staff of a postulated severe feedwater transient related to Babcock and Wilcox PWRs as described in Enclosure 2. ,
These reviews should be directed at assessing the adequacy of your .
reactor systems to safely sustain cooldown transients such as these. -
- 2. Review any trarsients of a similar nature which have occurred at your facility and determine whether any significant deviations frem expected performance occurred. If any significant deviations are fcund, provide the details and an analysis of the significance and any corrective actions taken. This material may be identified by reference if previously submitted to the NRC.
IE Bulletin No. 79-05 Date: April 1, 1979 P. age 2 of 2
- 3. Review the actions required by your operating procedures for coping -
with transients. The items that should be addressed include:
- a. Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability.
r
- b. Operator action required to prevent the formation of such voids.
' c. Operator action required to ensure continued core cooling in the event that such voids are forned.
- 4. Review the actions requested by the operating procedures and the training instructions to assure that operators do not override automatic actions of engineered safety features without sufficient cause for doing so.
- 5. Review all safety related valve positions and positioning require-ments to assure that engineered safety features and related equip-ment such as the auxiliary feedwater system, can perform their intended functions. Also review related procedures, such as those for maintenance and testing, to assure that such valves are returned to their correct positions following necessary manipulations.
- 6. Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the containment to assure that undesired pumping of radioactive liquids and gases will not occur inadvertently.
In particular assure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all such systems and indicate:
- a. Whether interlocks exist to prevent transfer when high radiation indication exists and,
- b. Whether such systems are isolated by the containment isolation signal.
- 7. Review your prompt reporting procedures for MRC notification to assure very early notification of serious events.
IE Bulletin No. 79-05 Date: April 1, 1979 Page 3 of 3 The detailed results of these reviews shall be submitted within ten (10) days of the receipt of this Bulletin. .
Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Construction Inspection, Washington, D.C. 20555.
For all other operating reactors or reactors under construction, this Bulletin is for information purposes and no report is requested.
Approved by GAO, B180225 (R0072); clearance expires 7-31-80. Approval -
was given under a blanket clearance specifically for identified generic p roblems.
Enclosures:
- 1. Preliminary Notifications Three Mile Island -
PNO-67 and 67A, B, C, D, E,F,G
- 2. Evaluation of Feedwater Transients w/ attachment
- 3. List of IE Bulletins issued in last 12 months e
4 9
r '
. w . - IE Bulletin 79 C 05' '
Enclosure 1 PN No. 79-67 and Subsequent Revisions ,
PRELIMINARY NOTIFICATION _
March 28,1979 PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL CCCURRENCE--PNO-79-67 ,
This oreliminary notification constitutes EARLY notice of event of POSSIBLE safety or ouclic interest sicnificance. Tne infor nation oresentec 1s as initially rcceived witnout verification or evaluation anc 1s cas1cally all tnat is known oy IE staff on :nis cate.
Facility: Three Mile Island Unit 2 Middlet:wn, Pennsylvania 4 (Docket No. 50-320)
Subject:
RF. ACTOR SCRAM FOLLCWED SY A SAFETY INJECTICN AT THREE MI!.E ISLAND - UNIT 2 -
The licansee notified Regicn I at approximately 7:45 AM of an incident at Three Mile Island Unit 2 (TMI-2) which accur ed at approximately 4:00 AM at 98% pcwer wnen the secondary feed pumps tripped due to a feedwater polishing system prcblem. This resultad in a turbine trip and subse-quent reactor trip en High Reac,cr Ccolant Pressure. A c::mbinaticn of Feed Pump Operaticn and Pressurizer Relief - Steam Generator relief valve operation caused a Reactor Coolant System (RCS) c::aldewn. At 1600 psig, Emergency Safeguards Actuaticn occurred. All ECCS components started anc operated procerly. Water level increased in the Pressurizer and Safety Injection was secured manually acproximataly 5 minutas after actuation. It was subsequently resumed. The Reactor Ceclant Ptress were secund when 1cw net pcsitive sucticn head limits were approached.
Abcut 7:C0 AM, high activity was noted in the RCS c::alant Sample Lines (apcroximately 6CO mr/hr contact readings). A Sita Emergency was then declared. At approximately 7:30 AM, a General Energency was declared
. based on High Radiatien levels in the React::r Building. At 8:30 AM site boundary radiaticn levels were reported to not be significant (less than
. 1 mr/hr).. 'The source of activity was stated to be failed fuel as a result of the transient, and due to a known preyfoud"irimary :: sec ndary leak in Steam Generator B.
The Region I Incident Rescense Cantar was activated ad 8:10 AM and direct c:mrunications with the licensee and IE:Headcuarters wal estab-lished. The Respcns7 Team was dispatched at 8: 45 AM and arrived at the sita at 1G:05 AM.
At 10:45 AM the Reactor Ccolant System Pressure was being held at 1950 psig with ta=perature at 220cF in the cold leg. By 10:45 AN, ndiatien levels of 3 mr/hr had been detacted SCO yards offsita.
CONTINUED
Page 2 March 23,1979 Continued PNO-79-67 There is significant media interest at the present time because of concern about potential offsite radiation / contamination. The Cormxanwealth of Pennsylvania and EPA have been infor:ned. Press contacts are being mada by the 1icensee and NRC.
Centact: GKlingler, IE x23019 FNolan,,IE x23019 SE3ryan, IE x23019
.e' W DistM bution: Transmitted H St .f 'l .
Chaiman e.endrie Cemissioner Bradford S. J. Chilk, SECY '
Ccin::issioner Xennedy Comissioner Ahearne C. C. Xa:nnerer, CA Ccmissioner Gilinsky (For- Distribution)
Transmitted : MNBS 3 Tu P. Bldg E 46 J. G. Davis IE _,
L. V. Gossick, ECO H. R. Centon, NRR Region J ~1: S q H. L. Ornstein, ECO R. C. CeYoung, NRR J. J. Fouchard, PA R. J. Mattscn, NRR N. M. Haller, MFA V. Stello, NRR (MAIL)
R. G. Ryan, OSF R. S. Boyd, NRR d. J. Cumings , OIA H. X. Shapar, ELD SS Bldg S*. S 2 R. Minogue, SD W. J.. Direxs NMSS PRELIMINARY NOTIFICATICN e
d s
M
PRELIMINARY NOTIFICATION March 29, 1979 PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PHO-79-67A This oreliminary notification constitutes EARLY nctice of event of POSSIBLE safety or ouolic interest sicnificance. The infor nation presentec 1s as initially received w1:nout verification or evaluation and is bas 1cally all tnat is known oy IE staff on tnis cate.
..- Facility: Three Mile Island Unit 2 Middletcwn, Pennsylvania (DN 50-320)
Subject:
NUCLEAR INCIDENT AT TriREE MILE ISLAND - UNIT 2 This supplemnts PNO-79-67 dated March 28, 1979.
As of 3:30 p.m. , on March 28, 1979, the plant was being slowly cccled dcwn with Reactor Ccolant System (RCS) pressure at 4S0 psi, using normal letdown and makeup ficw paths. The bubble has been collapsed in the A Reactor Ccolant Loop hot leg, and scme natural circulation ecoling has been established. Pressurizer level has been decreased .'.o the high range of visible indication, and some heaters are in operation. The secondary plant was being aligned to draw a vacuum in the main ecndenser and use the A Steam Generator for heat removal. The facility plans to continue a sicw (3CF/hr) cocidown, until the Cecay Heat Removal System can be piaced in operation at 350 psi RCS pressure, 3500F RCS temperature in 15-18 hours.
As of 3:30 p.m. , a plume approximately !s mile wide and reading generally 1 mr/hr was moving to the north of the plant. The ARM's helicopter is being used to define the length of the plume. Airborne iodine levels of up to 1 x 10-8 uCi/mi have been detected in Middletown, Pennsylvania,
- which is located north of the site.
Media interest is continuing. The Cctrenwealth of Pennsylvania is being kept informed by plant personnel.
Centact: GX1ingler, IE x28019 FNolan, IE x28019 SE3ryan, IE x2E019 Distribution: Transmitted H St hb D b' Chairman Hendrie Comiss1oner Bradford S. J. Chilk, SECY Cc:missioner Kennedy Ccmissioner Ahearne C. C. Xaninerer, CA Ccmissioner Gilinsky {.Q; g (Fcr Distribution)
Transmitted: MNSB h b P. Bldg IE L. V. Gossick, ED0' H. R. Denton, NRR J. G. Ufyis,l Region _L o 3 c,.
H . L. O rnstein , ECO R. C. CeYoung, NRP.
J. J. Fouchard, PA R. J. Mattsen, NRR N. M. Haller, MPA V. Stello, NRR (MAIL)
R. G. Ryan, OSP R. S. Bogd NRR J . J . Curmin gs , CI A H. K. Shapar, ELD 55 Bldg L@G R. Minogue, 50 W. J. Dircks, hMSS PRELIMINARY H3TIFICATICN
...__.....~...,,,e- Y MM l$l$
f
[- .
PRELIMINARY NOTIFICATION
~
March 30, 1979 PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PNO-79-o73 This oreliminary notification constitutes EARLY notice of event of P05515LE sa fety cr auc t ic interes t sicnificance. Ine information presentec is as initialiv received witncut verification or evaluation and 15 basically all tna't is (nown bv IE staf f on tnis care.
Facility: Three Mile Island Unit 2 g *
,- s Mi ddl e town , Pennsylvania (ON 50-320) ,,,.-
Subject:
Nuclear Incident at Three Mile Island e Plant Status
' Three Mile Island Unit 2 is continuing to remove decay heat through A-loop steam generator using one reactor coolant pump in that icap for coolant circulation. The reactor coolant pressure and temperature were stable and under control throughout the night of March 29. There has been some difficulty in maintaining coolant letdown flew due to resistance in the purification filters. The licensee notified IE at about 11:00 p.m. en March 29 that they expected to remain in this cooling made for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The licensee's engineering staff was requested by NRR to chtain a better estimate of the volume of the noncondensible " bubbles" one in the in reactor the coolant system. There are apparently two such bubbles pressurizer _ that has been intentionally established for control of pres:ure and level, and one in the reactor vessel head caused by the accumulation of nonconcensible cases from _ failed fuel and radiolytic tscomposition or water. The estimate is to be obtained by correlating pressuri:cr pressure and level indications over the past hours of stable .
Operatien. ,The volume of the bubble in the reactor vessel is of interest. -
in as:uring that sufficient volume remains in the upper head for collecticn cf more noncondensible gases arising from continued operation in the -
- p. :sent ecoling mode as well as to assess the potential for movement of tne bubble during a switchover to decay heat removal operation. ,
9.e licensee believes it is prudent to remain in the present cooling
- da due to the potential fer leakage of highly radioactive coclant fr:=
t a ducsy heat remcyal system into the auxiliary building, movement of
":rc r.densible gases inte the reactor coolant loop, and boiling in :ne c:rc wn2n the reactor coolant pump is shut dcwn. , ,,
CCNT3fUED
y .._ .;
- k. .
{ . _
Page 2 March 30,1979 Continued PNO-79-67B Fuel namea ,
Preliminary assessment of the extent of fuel damage from a reactor coolant sample taken at approximately 5:00 p.m. on March 29 indicates significant releanes of iodine and noble gases from the fuel. A 100
- illiliter samole taken from the primary coolant system via a letdown line was measured at about 1,000 R/hr en cuntact (70-60 R/hr at one foot and 10-30 R/hr at three feet). Preliminary analysis of a diluted sample in the IE mobile laboratory indicated fissicn product c::ncentrations of about S x 105 microcuries per milliliter. The sample will be flown to 5ettis Laboratory for further analysis.
Thermoccuple readings of coolant temperature at the outlet of the instru anted fuel assemblies indicate potential local core damage, possibly in one quarter of the total of 177 fuel assemblies and generally in the center of the core. Of the 52 readings at 5:00 a.m. on March 30, one was above the coolant saturation temperature of about 5500F, 7 were above 3500F, and 2 were off-scale, indicating temperatures higher than 7000F. Upon request of NRR, Sabcock and Wilcox is developing a proce-dure for use by the licensee in taking direct potentiometer readings-from the off-scale thermocouples since the temperature scale limitation of 70CoF is controlled by the pmcass c::mputer, not the themoccuole i ts el f. .
i Reacter Coolant System (RCS) Parameters The RCS parameters have remained relatively stable during the period.
Gradual RCS cooldown continued to about 1:30 a.m. , March 30, when tempera-
. ture was slightly increased to allow. additional margin batwen RCS operating parareters and Technical Specification minimum pressurization limits. Following are the primary system parameters over this period:
- 10:00 a.m. 7:00 p.m.12:01 a.m. 3:C0 a.m. 5:00a.m.
3/29/79 3/29/79 3/ 30/79 3/30/79 3/30/79 3t8 321 325 342 354 Pressuri:er Level (inches) 1055 1053 Fressurizer Pressure (psi) 853 -
945 1023 557 542 551 556 Pressurizer Temperature (OF) 529 Loop A Core 278 274 Inlet Temperature (OF) 281 277 275 Lc:p B Core 278 27' Inlet Te=erature (CF) 2S1 277 275 CONTINU5D r-
.i g;-
..a . ,' :~.-
, *:~
Page 3 March 30, 1979 Continued PNO-79-67B Environmental Status Two aerial surveys were conducted during the evening of March 29. The first flight was made about 8:15 p.m. during which measurements were taken in a circle around the site with a radius of about eight miles. No defined plume of radioactivity was detected but residual pockets of radioactivity were identified at various pc nts where the measured -
levels ranged frem .025 to .050 millircente.ns per hours. (Natural background levels are about .005 to .015 millircentgens per hour.)
During the second flight, at about 10:30 p.m., a plume was detected <
- northwest of the plant with a width equal to and c.cnfined within the beundaries of the river. The plume was touching down about one mile from the plant at Hill Island and then splitting into two parts - one on each side of Hill Island. Measurements at the east shoreline of the river, opposite Hill Isaind indicated about four millircentgens per hour and at the shoreline on mile north of Hill Island near Olmstead Air Fcrce Base about one millircentgen per hour. Additicnal measurements at five miles from the plant were on the order of .010 millf roentgens per hour and are in agreement with the earlier fli,ght.
During the early mcrning hours of March 30, an NRC monitoring team took radiaticn measurerents frcm a vehicle traveling both sides of the Susquehanna River frem 10 miles south of Three Mile Island to 4 miles north. Radiation levels were highest near Cly, a community just south of the facility on the west side of the river. The level at Cly was 0.15 millircentgen per hour.
0.05 millircentgens per hour. All other locatiens had levels less than
_0ther Infer =stion At acproximately 4:00 p.m. on March 29 two empicyees of Metropolitan Edison Co. received radiation exposures, in excess of the quarterly limit .
of 3 rems. The employees, an operatcr and a chemist, entered the auxiliary building to collect a sample of primary coolant. Present ,
estimates are that the operator received 3.1 rams and the chemist 2.4 rens.
The licensee released less than 50,000 gallens of slightly centaminated industrial wastes en March 29, 1979. This release was terminated at NRC request at approxima tely 6:00 p.m. , March 29, 1979, because of concerns ex:ressed by state representatives. At about 12:15 a.m. on March 30, N7.C gave the licensee permission to resume releases of the slightly contaminated industrial wastes to the Susquehanna River. This acticn was coordinated with the office of the Governor of Pennsylvania and a cress clease wa: issued by the State. Representatives cf the news media excres:ed cencern that they were not informed of the planned resum: tion of the release price to permission having been grantec.
CCSTINUED
G .* \ m., ,5 g . .
Page 4 March 30, 1979 Continued PNO-79-678 At 8:40 a.m., en March 30 the licensee began venting from the Basecus weste tanks. The impact of this operation is not yet kncwn.
Contact:
DThc=psen, IE x28111; EJordan, IE x 28111
., Di stri bution: Transmicted H St 9 '8 0 Cha1 man Hendrie Ccmissioner Bradford S. J. Chilk, SECY Camissioner Xennedy Ccmmissioner Ahearne C. C. Ka:n erer, CA Comaissicner Gilinsky (For Distribution)
Transmitted: MNB8 /olo 2 P Bldg /o //f J . G. Davi s , IE L. V. Gossick, EDO H. R. Denton, NRR Region H. L. Ornstein, EDO R. C. DeYoung, NRR J. J. Fcuchard, PA R. J. Mattson, NRR N. H. Haller, MPA Y. Stello, NRR (MAIL)
R. G. Ryan, CSP R. S. Ecyd, NRR J . J . Cermi ngs , DI A H. X. Shapar, ELD (55 51dg R. Minogue, SD W. J. Di rcu , tiMS5 Attachments (7):
Aerial Survey (6)
Ground-Level Survey (1)
PRELIMINARY NOTIFICATIN
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\\.];E-AERIAL SURVEY
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2: 1573 4:30 p.=.
PI - in a H to NE direction, about 30' sector.
Fr1:.2rily Is-133. A distance of about 15 miles, rtdiatien ceasurements in the plume were aheut 0.1 cr/hr.
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- PRELIMINARY NOTIFICATIO'! _
. m March 30, 1979 PREi.IMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PNO-79-67C This ov eliminary notification constitutes EARLY_ notice of event of PG.iSIBLE safety or oublic interest sienificance. ~The information presented is as_ in1tlally received witnout verification or evaiuation ,
ana' is easically an tnat is known oy IC staff on tnis care.
' Facility: Three Mile Island Unit 2 3
Middletown, Pennsylvania (DN 50-520)
'Sub e'et: NUCLEAR INCIDENT AT THREE MILE ISLAND ..,
~
. Plant Status There have been intermittent uncontrolled releases of radioactivity into the atmosphere frem the prirary ecolant system of Unit 2 of the Three Mile Island Nuclear Pcwer Plant near Harrisburg, Pennsylvania. The licensee is attempting to step the intermittent gaseous raleases by transferring the radicactive ccolant water into the pri: nary containment building. The levels of radioactivity being measured have been as high as:20 to 25 millirem per hour in the innediate vicinity of tne site at ground level . Off-site levels were a few millircentgen.
At!about 11:30 a.m. EST, the Chaim.an of the MRC has suggested to Governer Thornburg of the Ccatunwealth of Pennsylvania that pregnant wcmen and pre-school children in an area within five miles of the plant site be evacuated. Members of the NRC technical staff are at the site and efforts to reduce the tem;:eratures of the reacter fuel are continuing.
These temperatures have been coming down !,1cwly and the final depres-sui-i::ation of the reacter vessel has been delayed. There is eviden e of severe damage to the nuclear fuel . Samples of primary coolant centaining high-levels of radioicdine and instruments in the core indicate high fuel temperatures in some of the fuel bundles, and the presence of a
- large bubble of non-cendensible gases in the top of the reacter vessel.
Because of these non-condensible gases, the pcssiblity exists of interrupting coolant ficw within the reactor when its pressure Several is to optiens
- further decreased and the contained gases expand. In the reach a final safe state for the fuel are under consideratien.
meantime, the reacter is being maintained in a stable condition.
Contact:
SE3ryan, IE x28188 ELJordan, IE x28188 Distribution: TransmittedHStlllL5 S. J. Chilk, SECY Chainran Hendrie Comissioner Bradferd C. C. Xamnerer, CA Cdmissioner Xennedy Ccmissioner Ahearne Commissioner Gilinsky (For Distribution)
Transmitted: MNBBi P.Bldgl1'Il J. G. Davis IE Region ~ L~# so L. V. Gossick, EDO
~
H. R. Centon', 'NRR H. L. Ornstein, EDO R. C. CeYoung, NRR J. J. Feuchard, PA R. J. Mattson, NRR V. Stello, NRR (FAIL)
N. M. Haller, MPA J. J. Cumings , 01 A R. G. Ryan, CSP R. S. Boyd, NRn SS Bldg R. Minogue, 50 H. K. Shapc.e, ELD _
W. J. Di_rcks, NM.SS PRELIMINARY NOTIFICATION -
IMMEDIATE P?.ELIMINARY _NOTIFICATION
- - - - - - - ~.- -
_ - , - __ _ = _ _ - ~ .
,._-.,.;n_.- .
March 30,1979 .
PRELIMICARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PMO-79-670 This or21imin:ry notification constitutes EARLY notice of an event of
,POSSIBLE safety or ouolic interest sicnificance. The infor. nation ,
rresentcc is as initially received without verification or evalut tien
.:'d is bcsice l iv al l tnat is knotn by IE staff on this cate.
Facility: Three Mile Island Unit 2 Middletown, Pennsylvania (DN 50-320)
Subject:
NUCLEAR INCIDENT AT THREE MILE ISLAND Pl.?nt Status G.'secus radicactivity frcm the primary coolant system lendo'm has been enncaincd in ticste gas decay tanks since the last gasecus release at t.pproximttely 2:50 p.m. March 30, 1979. At the present reacter coolant lctdc'.In rate of approximately 20 gpm it may be necessary to make a planned release of radioactive gas tomorrow to prevent gas decay tank relief valve operation at its setpoint of 100 psi. The licensee has i;;st-lled a tc:.1porary line frc.n the gas decay system back to reactor ctattinaant vthich is under evaluation before being placed in operation.
Contair.me.it pressure is being maintained slightly negative (-l psi) as a r:sult of fan cooler operation.
R:setor coolant temperature measured at fif ty-two 1ccations at the outlet of the core have continued to come down sicwly. Three Outlet
- t perature instruments continue to indicate above saturation temperature.
The NRC staff was informed by the licensee on Friday corning that extminatian cf centai;r :nt pressure data for March 28 indicates a pressure spike up ,
to approx;cately 30 psi occurred at approximately 1:50 p.m. NRC parscenel a.c evalutting the possibility that a hydrogen explosion was the ccuse ci the conttfr.msnt internal pressure spike.
Th? r: actor coolant path is through cne reactor ccclant pump and cr.e st::m ccnarator. The stc:a generator is being fed by an auxiliary feed-r ...:p . 5:"cral options for depressurizirg the reactor and continuing ccolde: n via the residual heat removal system are under considaraticn.
]
= -
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~
CChTINUED e-
March 30,1979 Faga 2 PNO-79-670
.---u--- - = . = . . , . = =
Contin ad - -
The volu;.c of rion-condensible gases in the reactor vessel has been esticited to be approxirately 1000 to 1500 cubic feet at 1000 psi. -
This vole. ? e is estimated to result in a water level of several feet over the top of the fuel. The rate of growth of the bubble in the r: actor vessel is estimated to be less than 50 cubic feet per day at 1000 psi.
e Tha Cir:ctor of the Office of Nuclear Reactor Regulation, the Director of the Rcgica I Office of Inspection and Enforcen'ent and the Director of the Division of Opcrating Reactors arrived at the site at approy.imately 2 p.m. t:i'c.y to dir:ct NEC activities at the site and site vicinity.
P. ;re:cntatives of HEW and EPA are providing coordination and assistance to tha NRC at the Incident Easponse Center.
LC personnel assembled at the TMI site and vicinity in addition to the t? par manage:cnt personnel censist of the following:
RI RII RIII Hq 8 5 4 f.:7.ctor Inspectors (IE) 12 12 10 Mcsith Physicists (IE) 4
.:;-Ith Physicists (SP) 1 1 1 D.:blic Af 7 airs 13 F.. :ctor System Analysts (i:RR) 4 C.dition ':!aste Specialists (NRR) 6 H; lth Physicists (NRR) 2
. C, cr:tir.g Licer. sing (MTS) 80 Total Staff
.-- - - ~ - ~ ~
- .- -.-.--.. - -x.-..
L -.;T :.;UED
4 March 30, 1979 Page 3 PNO-79-67D Centine:d - - - - - =- - - -
The folicMing ecuipment has bien assembled at or near the site for support of MRC operations:
Equipment Location 1 RRC Inctrument Van with Observation Center ,
2 telepheno lines 1 .iRC Office Van 1 Office Trailer (Supplied by Licensee) 200 Esnd-Held Fortsbie Radios frem US Forest Service Portable Mealth Physics Instrumentation 3 P.alicopters from COE for survey and support 2 Laboratory Yans DOE /Settis A sophisticated communications pod from COE/ NEST will arrive tc:.r.crrc.t.
~l:','IRG.T~NTAL STATUS:
nt appro.ximately 3 P.M. on March 30, 1979, NRC analysis of eightAtvegetation 5.30 P.M.
samples frera the offsite areas sho':ed no detectable activity.
the Fennsylvania State Radiation health Department repor'ad that environ =cntal t atte and air sampics collected in the vicinity of the Three Mile Isl:nd ~
~1 ant sh2:1-d no detectable activity except for scme Xenon-133 and Xenon-135. .
i: ilk s:..ple analysis sho.ied no activity levels abcve background.
'iffsita crcund level gc.T.na surveys in the Middletown and Goldsboro areas
.troen 3:00 and 6:00 P.!'.. on March 30, ranged from .01 to 1 millircentgens
- r hour. An serial survey was made by heliccater from 4: 00 - 6:00 P.M.
i.a " arch 30, the site was surveyed in concentric circles at approximately one mile
- itcrvals cnd at a height of 300 to 1,000 feet. The highest radiation
' c; ding.; :2r: cver tho site and measured 8 to 10 millircent;cns per hour.
.t ?..'. - p l u:: : th: highest radiation readings were 6 to S nillircentgens
- r h:;r. Th; pluma follow 2d the river in a northviesterly direction and level Site greenc
. ! r.0 : c'2':tc:c >ie bayend five to six miles frcm the site.
- . y; tar. ucted bar cen 7:30 - 8:00 P.M. rang d from .01 to 1.S
.ill;re:nhens per nour.
- - - - - - - ~ ~ - - - -
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CONTI!.JED
I Page 4 March 30,1979
,. Ccr.;i
- _n u e, d,, ,,, _ -_ - -
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F"0-79-670 ._,--- --
At 4 P.M. March 30, upper level winds were from the southeast. Forecast irdicates precipitation in the for.n of thunderstorms r.oving in after 12 midnight, March 30. At 5:00 P.M. winds onsite at Three Mile Is1cnd
'rt:re repcrted at 2 to 3 miles per hour generally frca east to '.tast.
- Contsct: EG:c.tard, IE x28111; EJordan, IE x28111 Distri S tion: Transmi tted H St /.* /c = #3 '
u.:;rc.un . Mar.drie CeaTssioner Bradforc S. J. Chilk, SECY C:..aissiencr Kennedy Comissioner Ahes.rne C. C. K .r=:rer , CA Ct .nissicner Gilinsky (For Distribution)
Trtnsmitted: MNBS ///7 P Sidg / .' 2 S ' J. G. Davis, IE L. V. Gcs ick, EDO H. R. Centen, NRR Region H. L. Ornstein, EDO R. C. CeYoung, NRR J. J. Fcuchtrd, PA R. J. Mattson, NRR H. H. Haller, MPA V. Stello, NRR (l' AIL)
R. G. .~c an , OSP R. S. Boyd, NRR J. J. Cu- nings , OIA H. K. Sh:;:ar, ELD (SS Blc:9 // 3 3 R. Minogue, SD W. J. Di rcxs , NMSS l'hite House Situation Rcc.: /2fM %d5 EPA t u..,. / d. ,.
C.nElEOC 2 : c., n .h,
.2dc..::nt (1)
T.' diation Survey Mao
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f IMMEDIATE' PRELIMIN RY NOTIFICATICN -
March 31,1979 i
PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRE f
This inmudiate oreliminary notification constitutes an update of event of safety and occiic interest sicnificance. The infomation cresented -
1s as int:1atiy received witncut ver1fication or evaluation and is_
casically lill that is know11 cv NRC staff at tnis time.
I Facility: Three Mile Island Unit 2 Middletown, Pennsylvania (DN 50-320) e
Subject:
NUCLEAR INCIDENT AT THREE MILE ISLAND Plant Status _
Reactor cooling continues usirg! the 1A main reactor coclant Changes pump with to this steam generator A steaming to the main condenser. An operability status cooling method are not planned for the near term.of equipment is b of existing operating equip: rent.
The hydrogen recortiner is in an operable status; hcwever, shielding of its piping and ccmoonents Lead for is not fullyhas shielding installed and isand been located presently will be cen-sidered inadecuate. Calculations of Pydrogen in moved to the site on an expedited basis.
centainment show that the present concentration is less than 4%, the staff's limit en allowed enncentration to ensure an explosi not obtained. I sample. i The waste gas decay tank pressures were 80 psiThe at 10:15 tank isp.m.
set on March and had been relatively censtant for about five hours.The radiation field (50 R/hr at to relieve pressure at 100 - 110 psi.
centact) prevents resetting re. lief points, l
Reactor ccolant temceratures measured by incare the Temperatures cradually decreasing.
in the core as measured frca cutlet ther==ccupie I
Envinnmental Status _ l Three AFMS flights of one-hour length were conducted beginningAtat a 9:30 p.m. on March 30, and at; midnight and 3:00 a.m. en March 31.
' CONTINUED
~
7
. . Continued . th 31,1979 Page 2 PNO-79-67E distance of one mile frem the plant maximum readings ranged from 0.5 millircentgens per hour (mr/hr) to 1.5 mr/hr. At the 18 mile point, readings of 0.1 to 0.2 mr/hr were obtained during the two earlier surveys and 0.5 mr/hr during the latest. Flights are being made at approximately three hour intervals.
Offsite ground level gamma surveys in the Middletown area and north, -
between 9:30 p.m. on March 30 and 1:00 a.m. en March 31, indicated levels frem 0.2 to 0.5 mr/hr. lThese measurements were taken in the general direction of the plumet measured in aerial surveys.
I At 3:00 p.m. on March 29, (prior to the releases of March 30) the licensee pulled thermoluminescent dosimeters from 17 fixed positions located '
within a 15 mile radius of the. site. The dosimeters had been in place for three months and had been exposed for about 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> after the incident. Only two dosimeters showed elevated exposures abcss normal ..
levels. The highest reading observed was on Three Mile Island, 0.4 miles north of the reactor at the North Weather Station. At this location, the quarterly accumulated exposure was 81 mr, approximately 55
=r abo > the normal quarterly exposure rate. The other high exposure was observed at North E idge, 0.7 miles NNE of the reactor at the entrance to the site. At this location, the total quarterly accumulated exposure was 37 mr or approximately 22 me above the nor al quarterly exposure rate. !
\
During the evening milking hours on March 30, milk samples were collected by the Pennsylvania Department of Environmental Resources at the follcwing locations ,
I Harrisburg (2 sites)'.
York l Middletown l Bainbridge i Etters j i
Analyses showed no detectable radiciodine. The cows had been fed on ,
stored feed but had been outside for exericse.
i The Pennsylvania Depart:nent of Environmental Resources also collected water samples at filtration plants at Columbia, PA (for the City of -
Lancaster) and Wrightsville on March 30 in the morning and early afterncen.
Both sample points are downstream of Three Mile Island. No detectable activity was 'qund.
' CONTINUED l.
NW
~ '
Continued rtarch 31, 1979 Page 3 PHO-79-67E I
Contact:
DThcmpsen, IE x2 Bill NCMoseley, IE x28111 Distribution: Transmitted H St O: CL Chair:ran Hendrie Comissioner Bradford S. J. Chilk, SECY Comissioner Kennedy . Comissioner Ahearne C C. Karrrnerer, CA Comissioner Gilinsky (For Distribution) -
Transmitted: MNBBh: Of P. Bldg 9 l M J. G. Davis. IE L. V. Gossick, EDO , H . R. Denten , N RR Region 1 O ; S Lh H. L. Ornstein, ECO R. C. DeYoung, NRR J. J. Fouchard, PA R. J. Mattson, NRR N. H. Haller, MPA ' V. Stello, NRR (MAIL)
R. G. Ryan, OSP NRR J. J. Cumings OIA H. X. Shapar,. ELD :R. S. Boy,d, G O
- SSBldg(_.]. R. Minogue, 5']
~ W. J. Di rexs , NMSS .
l .
White House Situation Recm EPA FDA/SRh i DOE /EOC j l
Attachoant (1) I Radiation Survey Map '
, IPM DIATE i
PRELIMINARY NOTIFICATION i
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shown above.
March 31,1979 1 :00 a.m. All grcund level readings were less than 0.1 mr/hr.
measurements made in.venicle travelling ecute 441 frem about ten miles south o' plan to route 75 and south along reads on the west side of the river.
IMMEDIATE PRELIMINARY NOTIFICATION March 31, 1979 Pf:ELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PNO-79-67F e This oreliminary -notification constitutes summary information of an event of satetv or oublic interest sion1ficance. The informs:1on oresentec 1s a surmtry of information as of 5:30 am date 3/31/79.
Fccilitv: Three Mile Island Unit 2 Middletown, Pennsylvania (CN 50-320)
Subiect: NUCLEAR INCIDENT AT THREE MILE ISLAND Plent Status There has been no change in the method of cooling the reactor since the previous report (PKO-79-67E). Reactor coolant temperatures measured At by present incore thermocouples at 52 locations have continued to decrease.
none of the temperature readings is above saturation temperature for this There has been a slight pressure (554 F). System parameters remain stable.
o' rop in pressurizer level from 215 to 191 inches.
Efforts continue to complete installation of components and piping on the hydrcq2n recombiner. Approxicately 220 tons of lead shielding Lead in various shielding shapcs and forms has arrived, or is on the way, to the site.
is being installed around the recombiner. A dacision to use the recombiner has net yet been made. Two sampics of containment atmosphere have been antly.:ed which show hydrogen concentrations of 1.7 and 1.Q%.
Effort: continue to estimate the volume of the noncondensible gas bubble cbove the cort. Licensee calculctions At of the size of the bubble at 2:a0 p about 4: 20 pm this was recalculaiec by t' -
- 5::s E20 cubic feet at S75_psig. lnis is being further evaluated.
liceis se to oc 621 cunic feet at 875 psig.
Environcental St:tus 00 Three AR'S flights were conducted at cbout 6:00 a.m. , 9:00 a.m. , and 12: Maximum neo, ca Mcrch 31. All flights reflected a rather stable situatien.
rc-dings in the plume were from 1.5 to 2.5 millircentgens per hcur (mr/hr) at a d; stance of one mile f rom the plant, from 0.5 to 1.0 mr/hr out to 7 ciles, and 0.1 to 0.2 mr/hr beyond 10 miles. The plume width isOfabout fsi te 1-1/2 te 2 miles. No radiciodines have been detected in the plume.
ground icvel gamma surveys performed in the predoninsnt vind direction indic.ed raxicen 1CYOls Of about 2 ar/hr at a.kout l/2 mile f rom the site
- . c:f r?ction of th: pluca. The wind was f rom the SSW at the time of the CONTINUED rntu2ninnnt nuitricx Avn e-
Continued March 31,1979 Page 2 _
PNO-79-67F ARMS flights. At about 1 PM the wirids shifted and are now blowing in a south easterly direction.
International Contacts NRC's Office of International Programs (OIP) has prepared daily status reports, transmitted by Immediate Department of State telegracs to official NRC contacts in the 25 foreign countries with which NRC has regular official relations. OIP is also receiving many foreign telephone calls. ,
Two scaf or safety experts from the Federal Republic of Gor =any (FRG) arrived lcts M ach 30 znd were briefed by HRC experts et the Operations Center, '
1cte March 30 and during March 31. Two French experts will arrive April 1.
t.'cshington Representatives or senior visitors of Japan, FRG, and Sweden also have been briefod in the Operations Center. OIP also has been briefing the President of the AEC3 of Canada, who offered to send any AECL or AECS experts who could be of assistance.
Contr.ct with Licensee NOC Regional Offices are transmitting to the utilities with operating licenses sue:ary information (in the form of Preliminary Notifications) as th y are prepared.
Centact: DThompson, IE x28111 EMHoward, IE x28111 Distribution: Transmitted H St 7.'0Cc.
Chairraan Hendrie Commissioner Bradford S. J. Chilk, SECY Co:niscioner Kennedy Commicsioner Ahearne C. C. Xac:erer, CA Co==issioner Gilinsky (For Distribution)
Tr2ns-itted: MNBB 7.'/dn P. Bldg 7.'/ # c J. G. Davis. IE
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U. J. Dircks,'N:'SS J. J. Cunnings, CIA R. Minocuc, 50 Whitt licuce Situ . ion Room Ek '
EPs ~ <
FDA/32: w-D E/ECC Att c.~iant (1)
F.cdiation Survey M:p
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IG:EDIATE PRELIMINARY NOTIFICATICN
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IE Bulletin No. 79-05 Date: April 1, 1979 Encicsure 2 Page 1 of 3 EVALUATION OF FEEDWATER TRANSIENT A loss of offsite power occurred at Davis-Besse on November 29, 1977, which resulted in shrinkage of the primary coolant volume to the degree that pressurizer level indication was lost. A recommendation to convey -
this information to certain hearing boards resulted in the attached discussion and evaluation of the event. This discussion includes a review of a loss of feedwater safety analysis assuming forced flow, ,
which predicts dispersed primary system voiding, but no loss of core cooling. During the Three Mile Island event, however, the forced flow appears to have been terminated during the transient.
Attachment:
Di;cussion and Evaluation of Oavis-Besse Transients e
4
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IE Sulletin No. 79 05
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Date: April 1, 1979
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Enclosure 2, Attachment
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