ML18087A722
| ML18087A722 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/08/1983 |
| From: | SCHAFFER R E, SWEENEY R Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18087A720 | List:
|
| References | |
| PROC-830308, NUDOCS 8303160473 | |
| Download: ML18087A722 (85) | |
Text
APPENDIX A
- *
- Public Service Electric and Gas Company 244 Chestnut Street Salem, N.J. 08079 Phone 609/935-8560 Nuclear Training Center March 11, 1983 TRAINING OBJECTIVES Following are the objectives for Anticipated Transient Without a Trip (ATWT) training conducted at Salem Generating station. Upon completion the student will be able to: 1. 2. 3. 4. 5. 6. Define Anticipated Transient Without a Trip (Condition II FSAR Event with a common-mode-failure).
Describe how the inherent stability of the reactor will reduce reactor power on a "Loss of Heat Sink" (ATWT type event) . Explain how "dead-heading" of the centrifugal charging pumps could occur during an ATWT type event. State which ATWT type event will cause the highest pressure transient on the reactor coolant system. Explain why pressure could increase to 2974 psia on an ATWT type event. Knowledgably discuss the ATWT event of February 25, 1983 and know it was classified as an Alert, in accordance with our Plan, classified a Significant Event, in accordance with 10 CFR 50.72, and required Immediate Notification (within one hour). 7. State which reactor trip breaker coils are operated when: a. An automatic trip signal is generated via the SSPS (Solid State Protection System) . b. The (handle-type) reactor trip switches are used. c. The "bezel" pushbuttons for the reactor 'trip breakers are used . The Energy People
- *
- 8. State which coil(s) energize to trip the reactor trip breakers.
9o State which coil(s) de-energize to trip the reactor trip breakers.
- 10. Identify which component within the reactor trip breaker prevented an automatic reactor trip on February 25, 1983 and probably prevented the automatic reactor trip on February 22, 1983. 11. List the Immediate Actions (both automatic and manual) for a Reactor Trip, EI-4.3. 12. State which reactor trip breaker and* bypass breaker is operated via SSPS Train "A" actuation.
- 13. State which reactor trip breaker and bypass breaker is operated via SSPS Train "B" actuation.
- 14. 15. Explain the difference between a demand and a confirmation trip signal . List the five "confirmation" trip signals displayed in the control room for a reactor trip. 16. Describe and explain the design and function of the safety-related systems and sub-systems at Salem Nuclear station for mitigation of ATWT Events. 17. Outline the steps that the Senior Shift Supervisor and Operations Manager must perform after a reactor trip in accordance with AD-16, Post Reactor Trip/Safety Injection Review and Unit Startup Approval Requirements.
- 18. Explain the approval process for reactor plant startup following:
- a. Planned unit outages. b. A reactor trip/safety injection actuation with the cause known. c. A reactor trip/safety injection actuation with the cause unknown . RES:jcp R. Pr Supervisor SALEM GENERATING STATION INSTRUCTOR LESSON PLAN TITLE: ATWT HANDOUT il LESSON NO: DURATION:
REVISION NO: DATE: 3/8/83 SUBMITTED:
R.
DATE: DATE: 3/8/83
- ANTICIPATED TRANSIENT WITHOUT A TRIP TRAINING ( SALEM UNITS I & II ) OJ:?JECTIVES
.... Upon completion the student will be able to 1. Define Anticipated Transient Without a Trip (Condition II FSAR Event with a common-mode-failure).-
- 2. Describe how the inherent stability of the reactor will reduce reactor power on a "Loss of Heat Sink" (ATWT type event). 3.. Explain how 11 dead-heading" of the centrifugal charging pumps could occur during an ATWT type event. 4. State which ATWT type event will cause the highest pressure transient on the reactor coolant system. 5. Explain why pressure could increase to 2974 psia on an ATWT type event. 6. Knowledgably discuss the ATWT event of February 25, 1983 and know it was classified as an Alert, in accordance with our Emergency Plan, classified a Significant Event, in accordance with 10 CFR 50.72, and required Immediate Notification in one hour). 7. State which reactor trip breaker coils are operated when: a) An automatic trip signal is generated via the SSPS (Solid State Protection System}. b) The W (handle-type) reactor trip switches are used. c) The "bezel" pushbuttons for the reactor trip breakers are used. 8. State which coil(s) energize to trip the reactor trip breakers.
- 9. State which coil(s) de-energize to trip the reactor trip breakers.
- 10. Identify which component within the reactor trip breaker prevented an automatic reactor trip on February 25, 1983 and probably prevented the automatic reactor trip on ary 22, 1983. 11. List the Inunediate Actions (both automatic and manual) for a Reactor Trip, EI-4.3 .
- 12. State which reactor trip breaker and bypass breaker is operated via SSPS Train "A" actuation.
- 13. State which reactor trip breaker and bypass breaker is operated via SSPS Train "B" actuation.
- 14. Explain the difference between a demand and a tion trip signal. 15. List the five "confirmation" trip signals displayed in -the control room for a reactor trip. 16. Describe and explain the design and function of the safety-related systems and sub-systems at Salem Nuclear station for mitigation of ATWT Events. 17. outline the steps that the senior Shift supervisor and operations Manager perform after a reactor trip in accordance with AD-16,Post Reactor Trip/Safety Injection Review and Unit Startup Approval Requirements.
- 18. Explain the approval process for reactor plant startup following:
a) Planned unit outages. b) A reactor trip/safety injection actuation with the cause known. cl A reactor trip/safety injection acutation with the cause unknown. SAOP8303:3 INSTRUCTOR
REFERENCES:
TRAINING MATERIAL REQUIRED:
Sc...r:.:
eerJ STUDENT HANDOUTS:
- 1 CLASSROOM REQUIREMENTS:
- NOTES Handout 4f 1 R-=1()1RAOP:6 I. Pref ace On the morning of February 25, 1983 a transient occurred on Unit #J during the start-up phase after the refueling.
This transient fell under the category known as "Anticipated Transient Without a Scram" or "Anticipated Transient Without a Trip". In 1969, a question was raised concerning the effects of anticipated transients without a reactor trip. Initially, it was believed by the AEC that this was a very low probability event; until the AEC staff took a closer look to find that there were,.....
lO/year/plant(co.i.Jd1T1ot1Jie<JC
.. Thus, they concluded that the combined probability of anticipated transients ana*a common mode failure could be such that a safety problem did really exist. On Aug. 19, 1981, a Mr. Steitler put together a brief overall package concerning this matter. Handout #1 contains a few pages of his discussion.
© TP-1 Condition II transients (15 listed in FSAR) in Chapter 15. The SSPS was evaluated for random conponent failures and the likelihood of no trip following of condition II events was on the order of magnitude of 2 x 10-7. G> TP-2, 3, 4, 5, 6 Page J 3/8/83 Rev. 0 ---
NITICIPATED TRANSIOOS WITHOITT SCRLII'1 A1WS ..
- Al.JGUST 19) 1981
- BASIC GROJNDRULES FOR A1WS TI-IAT HAVE NOT 01.6NGED Pl'ITICIPATED TR#JSIENTS
-THESE ARE THE CQ\IDITICN II TPJ\NSIBITS IN TI1E FSAR. TI-!ESE TRflNSIENTS SHDt/ TRIVIAL RESULTS BECAUSE A fV\CTOR SCRPM IS GENEPATED.
wrrnrur SCRDM -SO'f}!Or/
A a,f IN THE F£ACTOR " 4 ---. SYSID1 PREVENTS THE PDIB FRCM FALLING rrrro COff. ALL INPUTS AND OlJTFUT TO THE FEACTOR PROTECTICN SYS"f8'1 FAIL AT THE SM TIJVE. -. BECAUSE OF THE LOW PROBABILITY OF AntS EVENTS THE NRC HAS All.OtlED FOR TI1E USE OF BEST ESTirvlATE INITIAL A1WS LIMITS NO GROSS FUEL DPfv1AGE * .. NO GffiSS GI.ER PfBSURIZATICN NO COOAINBIT PRESSURE EXOJRSICN
... DIFFERENT TYPES OF ATh'S EVENTS REACT IV Ill' EXQJRS I ... LOSS OF HEAT SINK IfGPJillATION OF HEAT ltlTh'AL IN RCS
- LIBS OF HEAT SINK POrJER MI8"1ATG1 SU0-1 11-IAT PRIMl\RY HEATS UP FAS1ER TI-JPN SECCTIDARY CAN RB1JVt HEAT NET RESULT IS THAT PRIMC\RY CCOLPNr IDlFERATURES NID PRESSURE INCREASE J THIS IS SENSED BY mff DIFFEIDIT TRIP SIGNALS * . PND THE PRHV\RY IS VIA A SCPJJM WITHOJr SCPAVJ -TI-lE PRIMl\RY SIDE IB1JE:M1UPE mo PRESSURE RISE ARE LIMITED BY INHEPfNTLY NEGATIVE TEMPEMTURE CCEFfI CIENTS EVENTS CCNS IDERED EXCESS LOAD INO£PSE LCSS OF FEEUIATER LCSS OF LOAD CllJRBINE TRIP)
.. -!it .
- i'ODERATOR IDPEPA TU ff CCEFFI CI B'ff IDFPLER IDfEMTUff CCEFFICIENT
' PEAK PRESSURE IS A STRCNG FUNCTICN OF FDRV 1 S Pl'ID SAFETY VN..VE v/ATER PELIEF PATES INITIAL FOYfR LP/EL . S/G HEAT TBPNSFER CAPPJ31U.ITIES HEAT SINK r1JST BE F£-ESTABLISHEn VIA AUXILIA.RY FEEIWATER
- Nff ATWS EVENT THAT IS A HIGH PPESSUFf CCNCERN ALSO RESULTS IN INCREASE IN COOLANT T81F£RATUFE ll-1E INCffASE IN COJUWT TBYPERATUFf RESULTS IN NEGATIVE REACTIVE INSERT I CN illE TO NEGATIVE CCEFf IC I 8'IT TI1E NEGATIVE REACTIVE IS BAlJiN(f]
BY IXlPPLER FEEDBACK SUGl THAT POtJER IS REDJCED -I.E. PLPNT STAYS CRITICA.L SALEM GENERATING STATION INSTRUCTOR LESSON PLAN TITLE: ATWT HANDOUT *2 LESSON NO: DURATION:
3/8/83 DATE: REVISION NO: DATE: SUBMITTED:
DATE: APPROVED:
- INSTRUCTOR
REFERENCES:
TRAINING MATERIAL REQUIRED:
oveRherx:i S.c,Re e. IJ STUDENT HANDOUTS:
- ;G CLASSROOM REQUIREMENTS:
- T HANDOUT 2 II. Westinghouse Owners Group/Emergency Response Guidelines (WOG/ERG)
From the WOG/ERG seminars held, guidelines were set forth to handle various types of transients.
One of which was the ATWS (Anticipated Transient Without scram). A. Purpose 1. Add -p when rods did not insert upon demand 2. Establish heat sink for primary 3. Prevent/minimize damage to fuel and release of radioactivity B. Symptoms c. 1. Rx trip bkrs. fail to open 2. Rod position indication
- 3. No rod bottom lites 4. @ level not decreasing Go thru 9 steps addressed in ECA-1 Step 1 The first action is. to mitigate the consequences as quickly as possible.
To do this, he must perform the 1st three steps. They are to be performed without delay. Rx trip -}:i switches Pushbuttons Drive rods in Trip Turbine -this is especially important if th ATWT was caused due to loss of F.W. Manual P.B. OST switch Turning off EH pumps Manually R.B. turbine Page 1 Date 3/8/83 Rev. 0 NOTES
- TITLE: ATWT HANDOUT #2 R301SAOP:6 0 steps #2 and 3 Must maintain a heat sink for the primary. If need be, manually start MDEFP's; Open steam valves for SDEFP. Ensure adequate inventory in AF storage tank. Ensure ll's and/or 21's are operating properly.
Steps i4 and 5 If attempts have failed via C.R. actions, then local operation must be accomplished.
Principle concern would be the Reactor followed by Turbine, feedwater.
Opening Rx trip bkrs., deenergizing MG sets, local trip lever on turbine front standard.
0 91 Caution" During an ATWS, RCS pressure can rise above pump shutoff head and, therefore, the pump mini-flow valves must remain open in order to avoid dead-heading the pumps. step #6 To get to Step *6, all attempts have failed to trip the reactor and neg. must be added to bring reactor P subcritical.
Emergency boration Injecting the BIT S.I. actuation in mind, however, s.I. actuation also trips MFW pumps.) Page 2 Date 3/8/83 Rev. 0 ---
.* NOTES
- If necessary to + RCS pressure below shut off head of charging pumps, the PZR. PORV's could be used, but ensure successful closure. Q Step #7 When using PZR. PORV, there is a chance that the PRT rupture disk will blow. Even though fuel failure is not expected, as a precautionary step, you should isolate cont. ventilation.
© step Boration must continue until an adequate SDM has been attained.
If the PZR. goes solid, a bleed and feed method thru the pressurizer PORV will be reguired . Pa e 3 Date 3/8/83 ---n
"
N OTES TP-4
- III. WOG/ERG Ao As stated previoudsly, Condition II events are classified under 15 types of events covering events such as: c Uncontrolled Rod group withdrawal from subcritical Rx 0 Uncontrolled Rod group withdrawal from power 0 start-up of inactive loop G} Blackout Loss of normal f eedwater Because of the numerous types of events, the response of the primary initially may be quite different; however, the operator response is the same once he identifies an ATWT has occurred.
Additionally, the response of primary would be different for the same event, depending on when in core life it occurs; especially since the value of MTC changes throughout the fuel cycle. -For the worst cases, addressed as long as a reactor trip is generated within 10 minutes, a turbine trip within 30 seconds (for loss of MFW); aux. F.W. established within 60 seconds, acceptable consequences will result. B. The one ATWT that causes the highest RCS pressure excursion is a main turbine trip from 100% due to loss of vacuum which additionally causes a loss of MFW. RCS Pressure Transient Pressurizer relief valve lifts in 5 sec.; S-G safeties lift in 11 sec., which corresponds to the 1st peak in pressurizer pressure.
Water expanding causes pressurizer to fill solid and ? ae 4 Date 3/8/83
- TITLE: ATWT RANDO . 2 NOTES TP-5/6 TP-7 .... -,...,....
r c. safety valves lift at 100 sec. Then peak pressure reached at 120 sec. of 2974 psia. Nuclear Power Transient By overlaying the graph of power upon RCS Tave, you will see that it is the inverse. Thus, as Tave t power +. Power initially
+ to and continues to + at sec. o At minutes, AFW is established.
There are 3 main results that need to occur for an ATWT: o Trip reactor o Trip turbine 0 Establish AFW (Review Figure #6) Page 5 Date 1L.§_l_83 Rev. O
- {
i 1 So pt. 19'21 ACTION/EXPEcn.D RESPONSE RESPONSE NOT OBTAINED The purpose of this guideline to add negative reactivity to the core when the control/shutdown banks are not irumed upon demn.nd, to establish and tain a heat sink for conditions amenable to long term cooling, and to prevent or minimize to the fuel and release of e."tc:essive rad.ioru:tivizy.
are symptoms of an anrici'pared tran..c;ierit without scram condition:
- l. Reactor trip bres!<.en fail to open l. Rod position indicators 5how failure of CRDMs to insert J. Rod bonom lights not lit 4. Neutron le'Yel not decreasing r..q:iidly corresponding to large negative reaaivtcy i.ns<mion 1cl3
!. -
N<>../l>a'l"f>
jgCAml ANilCl?ATEO WITHOUT (C©n1o) l Sop'i. 1931 -0
...
...... mt e°'7'T$' .........
mcmc=t ..........
==s mmrm.cte ... ,. "' ....... ,.
- o Circled numbers show immedicae actions ne-ps. o If at (JJ'ly time a reactor trip occurs 9 immediately go to E-0, REACTOR TRJP OR SAFETY INJECTION, STEP lo _ (0 P<3lrform rlctioo::J t'I. iry to mp iS"at:ror
- b. Try to trip turbinra morruol!y
© Orz-c1 bif\v f!!Jw.Wg:
- a.
pump indicator lights -LIT b. Tuibinet-driven pump stoom supply valves -OPEN Omdr M"1':J '! filwo ri. AFVJ
-PROPER EN\ERGENCY ALIGNMENT
(.!J 0 If TiKl Nisr-J0 Ct:airro.cl:
- a. Rsoctcr trip -b. Turbin\! trip --fl) Emttr plmlr
/i.rz. ci. iry to manually insert corrrro! rods. b. Try to runbod< ruri:lin0.
ci. Manually start pumps. b. Mornially open vclves. a. Monuolly or dose valves as
- a. If r.ot, try ttl trip r20ctor lc:allly.
1 )
plant means.] b. If not, try to trip turbin@ lcailly. 1 ) ['Emilf plant sp;a-cific:
meons.]
( J ;r:..rm, g\ci;;, i c: l Sept.
ANT1ClPAiED iRANSlENT WliHOUi SCRAJV1
...........
p\CTJON/'EXPt;Cit:D RE..s?ONSE V@ri-fy AfW j:L;: fl. AFW flow im;licotors CHKX FOR A.OW !'<EPONSE NOT Al NED o. Pmorrn actions of 2 and 3 lccally. Charging pW7ip mirdflow valves must rrmwin open whm RCS presswe is fP'ea.ur rha:n pu:mp shutoff h!Zculo lrfri"imo fI©lpicl B-0.'V?irm Of hlei 1i:J a. Start chorgiruJ pumps b. Align borotion flO\"i/ pm+t fjJ e. RCS
-Lm iHAN E! FSlG Vo-rifv Cc!:T'roi.m:il-Qm Adsqoota S1u.rtdot7n t'k.iw-gm G@ t@
ITl!lP S.AfITT SiU' e. Open PORVs, as l'lei:esscry, until RCS pressure is !!.! µsig. isolation hos NOT OCUJrred, lliQi monuolly isolate CDnttiinmerrt Vl;}rrtilcrtion.
{}) Emar pltmr spmfic f2) Emu ptmi1 sp;r:;:fic
- JW'711J sfruio ff lta
- a!i..
fJJ £nuu 200 btziow pl.Mr sp..."Otic pumrwnuoff hmJd.
I \, ECA-1 4256B:l BACKGROUND INFORMATION FOR WESTINGHOUSE EMERGENCY RESPONSE GUIDELINES ECA-1 ANTICIPATED TRANSIENT WITHOUT SCRAM BASIC REVISION SEPTEMBER 1, 1981 *. !
- . ( ( I o INTRODUCTION The guideline ECA-1, "Anticipateo Transient Without SCRAM 11 ifies the mitigating actions required ATWS events -a family of Condition II accidents requiring a reactor trip but, through some failure in the protection system, the trip is not obtainedo The reactor coolant system conditions at the time the operator an ATWS event can be very different depending on the initiating event. Loss of main feedwater, control bank withdrawal at power, and a spurious opening of a pressurizer PORV are examples of the differing nature of ATWS events. The required operator actions following identification of an ATWS event are the same but the reactor coolant system conaitions may be very different; pressure can exceed 2800 psia following a loss of main feedwater ATWS but will never exceed the nominal operating pressure following the spurious opening of a pressurizer PORV. Opera-. tors must be aware of such system responses ana not rely on any signals or indications other than those for reactor trip. Transient response is also highly dependent on the time in fuel cycle at which an ATWS occurs. Response is much more severe very early in cycle lifetime than later in the cycle. For primary pressure may exceed 3000 psia for a turbine trip ATWS at beginning of cycle life but not exceea 2575 psia (the pressurizer safety valve setpoint + tion) near the end of cycle life. This guideline only addresses the short-term operator actions. The operator must attempt alternate means of reactor trip and maintain a secondary-side heat sink via turbine trip {for a total loss of main feedwater) and auxiliary feedwater actuation.
Analyses have shown that if a reactor trip is generatea 10 a turbine trip within 30 seconds (for a loss of main feedwater), ana auxiliary feedwater ated w_ith in 60 seconds 1 acceptable consequences result. These times are for the limiting ATWS transients.
Tur*bi,1e trip and actuation of liary feedwater would normally be 1enerJteo by the reactor protection system b.ut it is assumed that the :.ame fault that prevents a reactor . trip also prevents these functions.
the guiaeline proviaes ECA-1 42568:1
- for these functions to be performed without delay by the operator in the control room. Long-term coolaown to c61d shutdown conditions is not adaressed.
Plant coo1down in a controlled manner is to be performed utilizing normal procedures as much as possible.
II. DESCRIPTION OF EVENT TRANSIENT The ATWS transient analyses were performea using composite plant ters to bound as many Westinghouse plants as possible, rather than using parameters for any specific plant. Sensitivity stuaies were performeo for the limiting cases to demonstrate that the conclusions are valio for all plants covered by the generic approach.
The analyses consioereo 2-, 3-, and 4-loop plant configurations with 51 and 44 Series and Moael D and F steam generators.
The reference plant was defined to be a 4-loop, 51 Series steam generator plant. Though numerous ATWS events have been analyzed, this section will be devotee to a description of the loss of load transient only, since this is a limiting ATWS transient with respect to peak pressure.
For a more detailed treatment of this transient and descriptions of other ATWS transients, refer to the two ATWS reports listed in the Reference sec-tion (Section V) of this document.
For. loss of load without reactor* trip: A major loss of load could result from either a loss of external trica) load or from a turbine/generator trip. In either case, unless a loss of ac power to the station auxiliaries also occurs. off-site power would be available for the comninea operation of plant components, such as the reactor coolant pumps. In this case. the loss of loaa acciaent was analyzea assuming that the control roes fail to drop into the core following .a turbine trip from full power, which would prpduce the mum possible load loss. The most severe plant conuit"ions that coula result a loss of load occur following a turbine trip from full )
- ( ' ( ( *
- l *power when the turbine trip is causeo by a loss of condenser vacuum. Since the main feedwater pumps may be turbine driven viith steam exhaust to the main condenser, loss of feedwater may also result from a loss of condenser vacuum. Plant behavior was evaluateo for a turbine trip ana loss of main occurring from full power with the assumption that the control rods fail to drop into the core following generation of a reactor trip signal. This evaluation showed the effectiveness of RCS pressure-relief devices and the extent of any approach to core safety limits. The results are presented below for the reference plant only (4-loop, 51 Series steam generator plant). Figures l through 5 show the plant transient response for a loss of load without reactor trip for a 4-ldop plant with a 51 Series steam generator and a moderator temperature coefficient valid for 95 percent of core life (95 percent MIC). Sequence of events for this transient are shown in Table 1. The first peak in pressurizer pressure occurs when the steam generator safety valves lift, and the second, higher peak (maximum system pressure*
of 2974 psia) occurs after the pressurizer is filled with water oue to a coolant volume surge resuJting from a rapia reouc-t ion of steam generator heat transfer.
Nuclear power decreases to a value of 68 percent due to negative reactivity feeaback causea by ator (coolant) heating. Further coolant heatup, causeo by loss of steam generator heat transfer, decreases nuclear power further, starting at about 110 seconds. The DNB ratio does not drop below its initial value during the transient.
"* It should be noted that there is a difference between "pressurizer pressure" and "system pressure" as used here. When pressurizer pressure is given, it refers to the pressure in the pressurizer, whereas the system pressure is defined to be the pressure taken at the discharge of the reactor coolant pump, the maximum pressure in the coolant system. The system pressure oefinition inc*uaes pump head and elevation heao and be higher than pressure by as much as 100 psi. ECA-1 42568: 1 At ten minutes into the transient, conditions are stablized, with liary feedwater prbviding heat removal capability and with an intact Reactor Coolant System and core. Thus, the operator could begin down operations through roe insertion, actuation uf the safety injection system, or through the BORATE or EMERGENCY BORATE modes of the Chemical and Volume Control system. Transient results for 3-loop and 2-loop plants with 51 Series steam generators are similar to those presented for the 4-loop case. A peak reactor coolant system pressure of 2861 psia results for a 3-loop plant, and a peak pressure of 2753 psia results for a 2-loop plant conf igura-tion. The following conclusions have been drawn from this particular event -loss of load: During a loss of load with failure of roa insertion after a reactor trip signal generation, core safety limits are not exceeded since the DNB ratio does not go below its initial value and the peak reactor coolant pressure is limited to 2974 psia for the 4-loop, 51 Series reference case. Furthermore.
plant conditions are stabilizea at 10 minutes such that the operator can begin shutdown operations.
ECA-1
- ( i \. TABLE 1 SEQUENCE OF EVENTS FOR LOSS OF LOAD A REACTOR TRIP FOR THE REFERENCE CASE* Time (seconds)
Event Turbine trips Reactor trip signal generated on turbine trip 0 Pressurizer relief valves lift 5 High pressurizer pressure reactor trip setpoint reached 6.4 Overtemperature t.T reactor trip setpoint reached 8 .4 Steam generator safety valves lift 11.
feed pumps begin delivering flow 60 Pressurizer safety valves lift and pressurizer fills with* water Maximum reactor coolant pressure (2974 psia) reachea 99 120
- Reference case: 4-loop plant with a 51 Series steam generator.
95 percent M1C ECA-l 4256B: l
- ECA-1 42568:1 LOSS OF LOAD ATiiS REFERENCE CASE 95:!. ITTC Fllil.JRE l \ II
- (
- l -0 u.. l.!J 0:: ::i I-:§ I.I.I _g_ ECA-1 42568:1 :;:.: ..... I-I-....J 0 Cl <.,) "" t!l I.I.I ::;;. <:!: 700.00 650.CO 600.00 550.00 500.00 450.00 400.00 0 o* g 0 0 8 0 0 N 0 0 0 0 ""' TIME (SEC} LOSS OF LOAD REFERENCE CASE 95,; MTC FIGURE 2 0 0 8 ..,,.
- '2 -"' "-..... c:;:: ECA-1 11 :::i "' Vl ..... c:. Q,,. a:. ..... N -°" ::::l ..,., Vl L>.l Q,, 3000.0 2750.0 2.500.0 2250.0 2000.0 -1750.0 1500.0 12.50.0 l 000.0 900.0 0 0 0 0 g g 0 0 ..... 0 0 g ..., TiflE (SEC) 0 0 0 a q LOSS OF LOAD A"l)lS REFERENCE CASE. MTC FlGlJRE 3
\. ( ( ...... :::i ..SI ..... ::1:: :::i c5 ::>-" ..... N -:::i "' "' w ci:; Cl-ECA-1 42568:1 2000.0 1750.0 1500.0 1250.0 iCOO.D 750.0 500.0 250.0 100.0 c 8 8 g 0 c 0 0. 0 0 c:::i 8 0 g 0 8 0 0 0 N .... ""' ...... \g T!f1E (SEC) LOSS OF LOAD AT\:!S REFERE!lC.E CASE 95% ITTC *FIGURE 4
..... c:t: .... V'I c.. ..... c:: :::::i "" "" ..... 0:: c.. ..... I'-V'I
- ECA-1 1000.00 > 750.00'"" 500.00 250.00? 0.0 0 0 0 c: 0 0 g 0 0 N 0 c: 0 0 ..., TrnE {SEC) LOSS OF LO.a.D AT'dS REFERENCE CASE 95% l'ITt: FIGURE 5 0 0 0 0 0 0 0 '<tr o.n . 0 c: 0 0 '°
- ( (
- I ' * \ Ill. RECOVERY DESCRIPTION The recovery technique employed in guideline ECA-1. "Anticipatea sient Without SCRAM." is composea of three main functions
-reactor trip, turbine trip, and auxiliary feedwater actuation
-that must be performed without delay from the control room, org if this proves cessful, operators must be dispatchea to perform these actions locally. If the reactor is tripped, the operator is to leave this guideline and proceed to E-0, "Reactor Trip or Safety InJection," step 2. However, if the reactor cannot be tripped, a rapia boration of the RCS must be initiatea and continueo in oraer to establish adequate down margin. As a precautionary measure, containment ventilation must be isolated at this time. The final step of the recovery involves taining adequate shutdown margin once it is achieved.
A block aiagram description of steps in guideline ECA-1 is given in Figure 6 . ECA-1 42568:1
( (
- FIGURE *6. ANTICIPATED TRANSIENT WITHOUT SCRAM (ECA-1) TRIP REACTOR Ml-'\11UALLY TRIP TURBINE ACTUATE AUXILIARY FEEDWATIR 42568: l GO TO E-0 STEP 2 RAPlDL Y !!.ORATE RCS ISOLATE COITTAHrrlENT VENTILATION MAINTAIN SHUTDOWN GO TO E-0 STEP Z TRIP TURBINE LOCALLY _ ACTUATE AF\-1 LOCALLY
- ( ( IV. DISCUSSION OF SPECIFIC GUlUELlNE AND NOTES Step 1 Once an operator has diagnoseo an ATWS event using the ;noications 1isteo in the section, his first actions must be an to got* th* consequences in the quickest means possible, i.e., from the main control board. lhe note before this step alerts the operator that the first three steps are immediate actions to be performed without delay. In the control room, the op er a tor must f,i rs t attempt to trip the reactor by use of the manual trip buttons or, if necessary, by manually insert-; ng the c ont,ro l roos. If the trip is s ucces sf u l l y obtained, the A 11,s portion of the transient is terminated with the consequences then being no more severe than those of the initiating transient.
1he note instructs the operator to immediately proceea to guideline E-0, "Reactor . Trip or 5-af ety Inject ion," step 2 if, at any time during the conouct of the ATWS procedure, a reactor trip occurs. In guideline E-0, diagnosis of the initiating event is performed.
The operator should next attempt a turbine trip, if one has not tically occurred to maintain steam generator inventory.
A turbine trip is required for the loss of main feedwater ATWS. For the remaining ATWS events, with the exception of the case when a turbine trip is the ating event, manual tripping of the turbine will create an unanalyzed situation, superimposing a loss of load tyµe of event on the initiating ATWS transient.
This unanalyzed case may yield a somewhat higher system pressure depending on the initiating event and time in core life, The turbine can be tripped in the control room by the manual trip buttons, use of test s*itch, and turning off EH control oil *pumps.: If these method* fa i 1 , the opera tor shou 1 a at tempt to manua 11 Y run back the turbine. ECA-1 42568:1 Steps 2 and 3 The second component in maintaining a secondary side heat sink is the actuation of auxiliary feedwater.
lf the auxiliary feedwater system doesn't start automatically, the following actions should be performed from the control room: a. Manually start motor-driven AFW pumps; b. Manually open turbine-driven AFW pump steam supply valves; c. Manually open or close AFW valves as appropriate for proper valve alignment (observe valve status lights) d. Verify that the water supply to the suction of the pumps is available by observing CST level indication, absence of CST low level alarms and AFW suction pressure low alarm, ano aaequate pres-sure at the suction of the AFW pumps.
- 4. If any of the required functions in Step 1 have not been successfully achieved when attempted from the control room, an operator should be dispatched to perform the actions locally. The local actions are done after those from the control room since they are more time consuming; the control room actions can all be completea quickly without cant impact on the time of 16cal attempts should they be necessary.
Local reactor trip actions are perforrneo first since the sooner a trip is obtained tl1e less severe the AHJS transient wi 11 be r The reactor can be trtpped locally by opening all reactor trip breakers, de-energizing the MG .sets providing power to the rods, or other plant specific means of cutting power to the rods. Plant specific means may gizing buses that supply power to the MG sets, opening switches '.that supply power to the power cabinets, etc.
42568:1 i I
( ( \ ( 1. Local attempts for turbine trip and auxiliary feedwater actuation are needed to mainta'ln a secondary side heat sink. Analyses have -tal<en credit for il tllrbfoe trip sooner than auxi1 *i ary f.2ed*;1ater actuation so the 1oc'11 actfons ¢lre to be performed in that order. P1 ant specific &rveans for tripp'irig the turbine may include using the 1oca.1 trip lever, using local ovei"speed test equipment, locally securing EH pumps. venting EH system, closing the main steam1ine isolation valvesp etc. Step 5 Total auxiliary feedv1ater flow is required following
- rr,1s events. Therefore, any of the actfons of steps 2 and 3 that cou1d not be form2d *in the control room shou1 d be done 1oca11y **io obtain fu11 ,!l.FW f1 (;';{. Step 6 The operator has progressed to this step only if a11 means of obtaining a reactor trip have failed. Negative reactivity must be added to the core through lxlrati on of the reactor cool ant system to bring the reactor subcritical.
The caution precedfr1g this step has been added for the protection of the charging pumps that are required for boration.
normal conditions, the charging pump shutoff head is greater than the RCS d;es *i gn pressure.
Howe'ler during an AHJS event RCS pressure can rise above the pump shutoff head and, therefore, the pLllTlp rniniflow valves must remain open in order to avoid dead-heading the pumps. ihe method of boration is a function of plant configuration and, hence, the boration flew path ttnst be aligned according to plant specific Methods of boration include injecting the BITP and injection actuation.
It should be that SI 4ctuation trip the main feedwateiA pLimps. If th*is is undesirab1eD the operator can manua11y align the system for safety injection.
However, the R\'IST valves to the of che Sl pumps shou1 d be opened first before up the BIT '\Jal ves. ECA-1 42568: 1
- 1f RCS pressure is above the shutoff head of the charging or SI pumps, borati on wil 1 be impeded o Therefore, pressurizer PORV s must be opened, as necessary, to reduce RCS pressure and obtain injection fl c-w. The PORVs sh ou1 d be closed when primary pressure drops 200 ps *i bel v;;J the operating boron injection pump shutoff head. The operator must verify successful closure of the PORVs, closing the backup isolation valves, if necessary o SteE.. 1 The pressurizer relief tank rupture disk may have burst during the event. _If containment ventilation has not been automatically isolated the operator should do so manually.
This is ;;i precautionary measure; sis has sho\'m that fuel failure is not expected following an Anis event. Step a Boration must continue to establish the required Technical Specification shutdo1tm margin and cool the RCS to no-1 oad T avg* A l'lb1 eed and feed" method of boration through a pressurizer PORV be required if the pressurizer goes water solido In this case a PORV should be opened as needed to permit injection of borated µ1ater through reduction of RCS pressure and rsr:oval of 1o>v1-concentrate liquid volume from the RCS. The operator must verify successful ci osure of the PORV. closing the. bad up isolation valve if necessary.
Once adequate shutdown margin is established, the operator is instructed to maintain this condit*fon.
It should be noted that shutdoWf'l margin cal cu1 ati ons shou1 d compensate for any rods that wi11 not insert.
Once reactor coolant system conditions have stabilized and the reactor is subcritical the operator should proceed to E-0. "ReactorJrip or ECA-1 42568:1
- \ ' Safety Injection" step 2 to establish conditions allo-wing the ope,rator to continue to a cold shutdoi;m condition; Technical Specifications require this if two or more control rods are Mt fully inserted upon tiEw.and.
ECA-1 4256B: 1
,-1 *
- V. REFERENCES
- 1. Anticipated Transients Without SCRAM for Westinghouse Plants, NS-TMA-2182, Letter Anaerson to Hanover, December 30, 1979. 2. Westinghouse Anticipated Transients Without Trip Analysis, WCAP-8330, August, 1974. ECA-1 42568: l
- \_. ( ( J / l ** >
- WESTINGHOUSE OWNERS GROUP EMERGENCY RESPONSE GUIDELINES CONFIGURATION CONTROL SHEET GUIDELINE DESIGNArOR:.
ECA-1 GUIDELINE TITLE: Ant'icipated Transients SCAAM REVISION:
Basic DATE: September 1 s 1981 The guideline described above has been revi e'r1ed and approved for implementation by the Ovmer 1 s Group Procedures S_ubcomrnittee and the \*lestinghouse Nuclear Technology Division.
- NOTICE: THIS GUIDELINE REVISION IS TrlE ORIGINAL ISSUE OF GENERIC GUIDANCE ON ITS SUBJECT MATTER FOR THE rnERGENCY RESPONSE GUIDELINE SET *. ANY GENE.RIC GUIDANCE ON THIS SUBJECT BEARING AN ISSUE DATE EARLIER THATJ SEPTEMBER
- 1 , 1981 IS SUPERSEDED BY THE EMERGENCY RESPONSE GUIDELINE SET. *':',: File this sheet \>-'Jith the approved version of this guideline in your Emergency Response Gtlidel *!ne Set. /?
Manager, Standard Plant Engineering Westinghouse Nuclear Technology
. '.
- SALEM GENERATING STATION INSTRUCTOR LESSON PLAN TITLE: ATWT LESSON NO.: DURATION:
REVISION NO.: 0 DATE: 3-08-83 DATE: 3-08-83 DATE:
- NOTES TP-1 TITLE: ATWT II. Feb. 22, 1983 Transient At 2155 with reactor @ 20% Group Bus 'F' was in the process of being transferred from the SPT to the APT. Due to a faulty limit switch on the APT bkr it prevented the closing coil from energizing, resulting in a de-energized bus. Most important load dropped was #13 RCP. Being < 36% power no auto Rx/Turbine occurred as is designed.
Another load lost was 14 MAC panel which dropped control power to #12 SGFP. Supervisor noticing #13 SG level dropping rapidly ordered that the unit be manually tripped. Other events took place during this transient; the one of concern at this particular point is when did the reactor actually trip? 000 => 21:56:35 1163 > 21:56:54.4 1163/60 = 19.4 sec. Rx trip should have occurred due to Lo-Lo level in #13 S/G 1381 => 21:56:58 1381/60 = 23 sec. Operator went to trip with W switch 1385 => 21:56:58.1 sec. 1385/60 = 23.08 Rx trip bks open This was initially not picked up due to other numerous events that had occurred due to dropping of a bus; swapover of the lB vital bus to the #12 SPT, acutation of S.I.; lifting of the PZR PORV 1 s; no spray flow, etc . '---------'-------------=----=--------------------
- -----* Fage 1 Date 3-8-83 8305-SAOP:l Rev. 0 I 11;Ja;:
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. \ OPERATIONS DEPARTMENT REPORT ON REACTOR THIP/SAFETY INJECTION February_
- 22. 1983 *, -I \ At February 229 Salem Unit 1 was synchronized and the.
loading in accordance with established Operations Department procedures.
At 20 3*reactcr powerv epprczimately 200 Mwe electrical load, procdure 5.34 of Hot Standby to.Minimum Load, required that the 4Kv Group Bus power supplies be transferred from the Station Power Tran sf armers to the a:ry Poo-1 er Transf orme:r. The 1G 4Kv.Group Bus was successfully transfered and at 2155 hours0.0249 days <br />0.599 hours <br />0.00356 weeks <br />8.199775e-4 months <br /> the operators attempted to transfer the 1F 4kv Group Bu5 to the Auxiliary Power Transformer.
Upon depressing the close push button for the Auxiliary Power Transformer infeed breaker. the Station Power Transformer infeed breaker openedv however, the infeed breaker from the Auiiliary Power Transformer failed to close. The failure of the Power Transformer to close cau.sed the 1F l\Kv Group Bus to deenergize.
The opening of the Station Power Transformer Breaker prior to the closure of the Auxiliary Power Transformer Breaker is in accordance with the power transfer design scheme.
- Upon the less of pouer to the 1F 4Kv Group Bus all equipment powered from the bus was deenergized.
The most important equipment deenergized Ba5 13 Reactor Coolant Pump and 14 MAC 115 volt distribution panel. The control room lighting was also lost. However 9 this. did not present a problem as the emergency lighting provided sufficent light. There was no immediate Turbine/Reactor trip on loss of reactor coolant flow to the 13 reactor coolant loop because the Reactor Protection System allows single loop los5 of flow below 36 $ power. The loss of 14 MAC however, causect*12 Steam Generator Feed Pump to trip due to the loss of feed pump control The less of the feed pump interrupted the feedwater flow to all steam generators and 2team generat6r levels started to immediately decrease.
The operators noted that 13 Steam Generator level was decreasing rapidly due to the combined effects of the cf feedwater flow and the shrink in the steam generator due * -tO-.lo.&S" *or :reactor' cooL-:mt flO\tJ. In addi t.ion to the loss of the pUIDp, a significant amount of in:strument.ation lost po';Jer when 14 MAC panel*was de-energized.
This caused the instruments to fail to the mid-soale The supervisor in the control room at the time of incident ,..ealized that with the f'ate of clecrea_s_e-in 13 Steam Gen-erator level not be possible to bring the reactor to less than 10 power in a manner so he ordered the Unit to be manually tripped. At 2157 hours0.025 days <br />0.599 hours <br />0.00357 weeks <br />8.207385e-4 months <br />, simultaneously, with the j issuance of the order, the reactor was autcmaticly tripped by the ( eactor Protection on 13 Steam Generator LO-LO level. equipment functianad aa designed called upon by the Reactor Trip. The Reactor Trip/Turbine Trip set up the automatic transfer circuit for the buses to be tranafered from the Auxiliary PoBer Transformer to the *station Pohler Transformer.
The and 1E 4Kv Group were already 5upplied from the Station RoBnr Transformers as swapped over to the i.i.uxiJ.iary T:ransfarrner:.
1G Group Bus swapped over ES per design. The transfer scheme also provided a close signal to the Station PoBer infeed to the 1F 4Kv Group Bus, causing the breaker tc close and the bus to become re-energized.
The simultaneous etarting of all the connected loads on the 1F 4Kv Group Bus caused the 11 Station Power Transformer voltage to decrease.
The decrease in the transformer voltage sensed by the second level of undervoltage protection en the 1B 4Kv Vital Bus and the bus automatically sBapped over to 12 Station Power The locked rotor relay protection for 13 Reactor Coolant Pump operated shortly after the bus was reenergized due to the inability of the pump to start without the aid of the lift pump and the voltage condition on the At 2204 Bn automatic Safety Injection occured because the steam in *13 Stea!r.:l Generator
'l.:1as 100 psi less then the steam pressure in any other 2 generators.
The cause of the steam pressure decreasing 100 psi was due to the combined effects cf the addition cf cold feedwater by the Auxiliary Feedwater Pumps which started automatically on response to the LO-LO level in 13 Steam Generator?
the cooling effect cause by the off of steam from the 11 and 13 Steam Generator by the turbine driven auxiliary
_feedwater the reduced circulation in the Reactor Coolant System due to the loss of 13 Reactor Coolant Pump and the fact that the MSIV's had been closed to control the decrease in Tave. The closure cf the NSIV 1 5 to limit the cooldown following a reactor trip is 5tandard practice early in core life when there is little or no decay heat available to maintain RCS temperature.
All required Safeguards Equipment functioned as designed.
causing the pressurizer level to increase during the Safety Injection.
As pressurizer level increased the bubble Has compressed and pressurizer pre5sure increased cau5ing the Power Operated -Relief Valve5 to relieve to the Pressurizer Relief Tank. The in pres.su:rizer p1essur"2 Hould normally be held below the point cif the Power Operated Reliefs by the pressurizer valves. HoBevar 9 in this case no spray flow hlas since 11 end 13 Reactor Coolant Pumps were cut of No.11 Reactor Cool&int .P1J.mp ha¢ tripped zt 2206 for no apparent c 2 The operators responded to the Safety Injection in accordance with the Emergency Instructions for Safety Injection.
The duration of Safety Injection 7 minutes. Upon establishment of the Injection reset criteria by the operators, the Safety Ir1j acti or2 i;J3:3* te:t'!ni 2 t 2211 hours0.0256 days <br />0.614 hours <br />0.00366 weeks <br />8.412855e-4 months <br /> in accord a.nee with
- procedures and tha Unit uas returned to a stable shutdown . t . c cmd:l."ti on .. An _investigation uas conducted by and relay department into th,a incident. "The i'1vest1gatj.o:n a faulty 52 I/S limit in the 1F 4Kv Group Bus =
Power Transformer infeed breaker. This switch is normally closed uhen the breaker is fully racked up and is wired in series with the closing coil cf the The faulty limit switch therefore prevented the closing coil from energizing when the transfer control scheme called fer the breaker to clc5e. The investigation did not reveal any apparent problem with the ralay protection fer the 11 Reactor Coolant Pump nor did it reveal any cthar reason for the tripping of the pump. The Engineering Department has been requested through* a Design Change Request to provide undervoltage lockout relay protection for all Reactor Coolant Pump This Hill provide a breaker trip/lockout cf the coolant pumps in the event cf a total loss of voltage on the supplying bus.
will prevent the uncbntrolled s ta:rting cf the Rea cto:r. Cool ant plirops upcn1 bus gi.z ati on. ' " 3
- r--*----------------
1 SALEM GENERATING STATION INSTRUCTOR LESSON PLAN *
\ I TITLE: ATVIT LESSON NO o: DURATION:
DATE: 3-08-83 I I REVISION NO 0: 0 -\ I DATE: 3-08-83 SUBMITTED:
Rick
- /15 :JJ4t -APPROVED :-Vt/ / I DATE: f 8 _l, _________________________
_ \ \ \ -l NOTES 0322 1820 TITLE: ATWT IV. Reactor Trip Feb. 25, 1983 0 At 1200 on Feb. 24 the turbine was taken off line to perform OST. test successfully.
to with power 0 At OQ12 the generator -sync grid @ 14%. o Probs. experience with #12 S/G level and the S/G reached lo-lo level rx trip setpoint of 18%. The logic for the rx trip being 2/3 detectors on 1/4 S/G should have caused the trip. e The first out OBA F-10 actuated "S/G 12 Lo-Lo Level R.T." However, the OHA F-46 never actuated "Rx trip Turbine Trip". This* annunciator is actuated by P-4 which is actuated by having 1 A 1 and 'A' bypass bkr or 'B' and 'B' bypass open. © In approx. 25 sec. the operator recognized Rx and turbine did not trip and performed a manual reactor trip. © The lo-lo level signal did leave the SSPS cabinets because the MDEFP's did start. o Alert was declared at 0130 and terminated at 0200. 10CFR50. 72 ->-significant event+ notification w/in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by phone. 0021:38 #12 S/G Lo-Lo level channel #4 0021:43 #12 S/G Lo-Lo level Rx. ttip (Ch. #1) 0021:43 The SSPS called for Rx trip
- 11/12 AFP start 0022:08 Operator went to trip with trip switch 0022:08.4 Turbine Trip signal 0022:08.4 Rx main trip bkr 'A' tripped 0022:08.4 Rx main trip bkr 'B' tripped From the time that the SSPS called for a reactor trip to tne time that the opetator'turned the trip switch was: 1820 -322 = 1498/60 = 24.96 sec. Date 3-8-83 Rev. 0 1 8305-SAOP:l "NOTES
- There was a time lag in which the event occurred and when the NRC was informed.
The s3 anticipating the NRC question on how do you know it was an actual ATWT had the I&C Dept. run various tests to ensure it was a breaker failure and not a SSPS failure . Date 3-8-83 ___ _ 8305-SAOP:l Rev. 0
- I ----------
REACTOR TRIP FEBRUARY 25,1983 On Februrary 24,1983 Unit 1 was in the startup phase following a refueling outage. At approximately 1200 the turbine was taken off the line in order to complete the overspeed test. Following the successful completion of the test the unit was returned to power. At 0012 on February 25,1983 the generator was synchronized to the grid. Reactor power was approximately 14%. Problems were experienced controlling the level in No.12 Steam Generator and as a result the level decreased to below the Lo-Lo Level Trip Setpoint of 18%. A reactor trip signal was generated by the Reactor Protection System, however, the Reactor Trip Breakers failed to open. The operator, upon surveying*the controi _room indications as directed by Reactor Trip, observed there were no Rod Bottom Lights illuminated, the Individual Rod Position Indications still showed the control rods to be withdrawn and the turbine did not indicate it was tripped. Upon recognition of these facts, the operator concluded that the Reactor Trip had failed to initiate and manually initiated a trip utilizing the trip handle on the console. This ocurred approximately 24.8 seconds after the Reactor Trip .signaJ..
was gernerated, (see the attached s,,equence of events printout attached) and resulted in a reactor trip. All automatic functions associated with the reactor trip were then verified to have ocurred. (All control rods were fully inserted and the turbine tripped.)
The operator had concluded that a trip was warranted because the actual level indicated in No.12 Stearn Generator was below the trip setpoint of 18$ and all of the alarms associated with the Steam Generator levels. both on the console and the Overhead Alarms were illumiated.
Also, the Motor Driven Auxiliary Feedwater Pumps had started automatically.
These pumps start when the level in any Steam Generator decreases to less than 18%. As directed by the Emergency Plan Procedure EP-I-0 Part 2 Item 16, an Alert Classification was declared at 0130. All notifications were made in accordance with EP-1-2 1 Alert. The event was terminated at 0200. Attached are copies of the recorder charts for all four Steam Generator Levels, both narrow and wide range . 1 :._ ....
(_ . * '* ( ./ * ( RJ;;AQT()!l_
'!'!ll!' __ n11sT OU'.\' __ ------c---------------+
- ----REACTOR sTM OEN 11 . sTM GEN rr isr:rM nri:P p=----* coNDENSER sTM GEN rr COQLANT LOW-LOW FEEDWATER . LOW Pl VACUUM HI-HI PR RC LO FLOW 1H RANGE OR RCP BKR HI PRESS LEVEL LO LVL & FLO SI LOW LEVEL REAC TRIP REAC TRIP rEAC TRIP f{EAC TRIP , TURB TRIP 1 TURB TRIP 3 11 . 6 '37 51 C1 ----1.-R_E_A_C_T_O_R
__ , STM OEN 12 i,STM GEN 12 STM pIFF P rrURBINE STM GEN 12 LOW-LOW 1 LOW P2 BEARING HI-HI LO PRESS LEVEL LVL & FLO SI LOW OIL LEVEL REAC TR+.P 'l REAC TRIP 1t; TRIP u HEAC TRIP 1 z_ TURB TRIP']'J ; TURB TRIP c(o -PRESSURIY'E"'i=C'-sfM OEN i°3 13 '*sTM DIFF P'd? -rr*URBINE l ST'M.GEN 13 HIGH LEVEL LOW-LOW fEEDWATER LOW P3 THRUST 1 HI-HI LEVEL lLO LVL & FLO SI BRO FAIL LEVEL 1 ! . t..f. I REA9 TRIP .REAC 'rRIP 16 rEAC TfUP 1'7 TRIP _'"15 ' TURB TRIP ) PZR LO LVL STM GEN GEN l I ISTM DIFF p 11 TURBINE STM GEN 14 & RC LO LOW-LOW LOW P4 . lovERSPEED HI-HI PRESS sr '?-1 LEVEL zz LVL & FL.O ;sr I LEVEL I REAC TRIP REAC. TRI)?
2-at TURB TRI*P !l? ; TURB 1:1RIP COHTAimlENT I STM HI FLO -'TURB TRIP E-H ! REACTOR i t NEUTRON PRESS HIGH & LO PRESS & P-7 DC PWR FAIL I TRIP I RATE -. .* ' ISOL SI EA'C TRIP "25 !REAC TRIP Zl 2 1 REAC 'l'RIP .7S. er* jREAC TRIP cio TURB TRIP ,,,-TURB TRIP u. , 1 f
- I :-::-.:i/E=*
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fRONT '/l[W Of STUD LC(A T IONS fOR 2 POL[ flREAllEft WITH 2 OVfRCURR[NT rn1rs AND REVERSE CURRENT TRIP. REVE115E cummn Tf11P -----------TH[ FOi LO'NIHG Aff,\CilfACIHS CM{ [I[ '.,\J?f\.i[O
\'!I-ii i(1UT INCH[1\SllJG OVE f',1\Ll OlMUl SIOl*I s: I f1UXILIARY SWITCHES 02 CIRCUIT 2 SHUtH 1RIP I 3 UND[IWO:_TAGE TRIP (WIT\\ OR WITHOUT rnA[ DEL/IY) 4 f-t_NlM S'IJITCH 5 [L EC Tf\IC LOCKOUT 6 OPERATION COUtHER Fig. IA -Tyr;11 DB-50 Three Position Drawout Outline Dimensions
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\ \ I \ \ I * ]:lOI. ES 1 Fig :\t-16 v .. \' \ I \ \ \ l DB-50 ckt. operation of undervoltage trip attachment.
When energized the moving core is held against the stationary core which allows the rod to keep the reset lever in the reset position.
(4BV DC coil) When voltage + spring overcomes the magnetic attraction of the 2 cores and rotates the reset level which releases a latch pin and releases the latch. Now trip spring rotates trip lever. Thru rnech. linl<:.2>.ge blzr .. opens. 0 () Operation of Shunt trip device (125 VDC). This device also uses the moving ana stationary core idea. And also uses the trip spring and trip lever. This device energizes to perform its function and is actuated thru the manual trip switches on the board as is the UV coil also. The reactor trip breaker pushbutton on console operates the shunt trip device. --------*---
page 1 Date 03/08/83 Rev * -----C-.1 ______ ... ------------***--
- .,.. .. , . .-' ' 'i .:( ' l k<.. **-L--* I Type 1iDB" aJ.:r r:;i:r-cuit bre?J<:er is design2d to give continuous and :reliable service ss the :protecti v2 li cl: between foe power source * ---------* ----*-----
and associated pTaductive equi.pmenL This breaker is built to oi:1erats with a minL:-num of mai:c-,tenance, *while at sarne time Us simplified construction perrnits maximum accessj_bility for inspection and ad)ustn:..ent V/hen requirecL The ease vrith which tachments may be added or :removed is an outstanding feature of the "DB" design. For the greatest :measure o-£ safety to operating personnel and also to minimize maintenance requirements, the brealzer should be mounted in an enclosure suitable to local operating conditions, A selection of standard enclosures is available for v;:i.rious
\J,\ P 0 R T AH T : To a s s u r e p r o p er fun c ti on i n g , inspect ead*, br,:oker at int:::rvols in ac-'
- I ,* * * \_ .1 l coroance wi'tn o systemonc mo1nrenonce scn2ou e. The und choroc'r2r of the w i i ! for th" rn o st pad he de t e :min *2 d by the s e -* r ' d r l -r\_ .. '1erd)' o-r tne
- uty pe:*rorrnec.
ne rn1n1mum re-quirements, howe'1er,.
should consis*i*
of a light mon'ihly inspecticn, with a *ihoroU*]h inspection serni-cmnuolly.
Occasional d112cks on tion as well es en coordinotion and freedom of all rnoYing ports, must be includ2d in the nance schedule.
Consult Westinghouse ing end ser*1ic2 p2rsonnel for iecomrnendotions peri'aining
- to specie\ op<::ruting of maintenance condilions.
-!------*-----------
I -DB-50 C\RCUlT port' '<EC' I\ 1600 i'.\llPERES
-60_Q VOLTS A-C I I ! \ \ l L ... -----*--
-,_.. --.. ----------(RESET) rzg. 3 -Cross-Sectional v* -------------.w:.* of Tvoo DB --. -, i . -:;0 c . .rrcm I Breaker I l _____ . _____ _\ CLOSED TRIPPED L Page 12 **In -r 0-nl::ic'n*r clos'ncr co;l be s11r"' *t-o ra..-.l . ....., _t-' --1-., :.::-*,,.... .._ .l .... ., .J. * (_, ., c ... -*. _ce brP.ss tuoe (0) so tciat sts.t10nary cor:; (4) 2.nd I.noving core (3) a:re aligned in ths: on right hand side of solenoid yoke (1) and allow moving core (3) to drop into brass tube (5). Pick up closi!1g coil \vith brass tube ancj moving core and bring out through the U-shapecl foot oil breaJ.;:er.
- tube. Re-2.ssemble closing coil anJ details in reverse order from reE1ovaL li the circuit break.er is pen:nancmtly mounted ne2r the floor so that the closing coil cs.nnot be dropped far enough for moval then follovr these directions.
Tr-ip breaker and :remove breaker manual erating handle and breaker fac.:; plate, Dis** connect closing coil leads from control cuit wiring. Ta2.ze off bolts (9), w2,shers (12), 2*elay :release arm (8), bolts (lOi, washers (11) ancl plate (2). Drop closing coil {'7) with brass tube (5) so that pin (6) is exoosed. Push pin (6) to right into ho1e Re-a.ssemble closing coil and details in reverse order from removal. Take care to align stationary core ( 4) and moving core (3) irr brass tube (5). OVERCURRENT TRIPPING DEVICE The overCLffr:::nt trip is an air del2.yed device that can be supplied \vith various rating coils ranging from 200 to 1600 peres. The construction, except for the coils, is simila:r foT all ratings. --------------
SPRlt'G MECHANiSM CIRCU!T BREAKEF\ 1500 AMPERES, 500 VOLTS A-C '-------------*---
Fig. 3A -Type DB-50 Spring Closing ,Assembly
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I \ I 1* I \ I I \ I i TEST FOSIT\ON 1.1ARK r I i Fig. 3B -Type OIJ-50 Three Posirinn M<'<'linnism
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- 26 Ene:cgize relay operating coil. Slowly close foe brea .... 1.::er manually.
The :relay release arm should 01)erate the :celay trip assernbly and the relay trip s:';sembly should 0:9en the relay contacts just before the breaker lsi.tches.
This position can best L'B determined by wat.ching the pawl in foe brea1cer opera.ting
- \vhich should snap in place just after the relay contacts open. If this operation sequence is not correct, the relay release arm should be bed to sui.t. l\lake sure that the :relay lease arm does not rub on either side of the relay trip assembly lever aperture.
V{hen the b:reaker is latched, de*-energizing and then energizing the relay operating coil should not cause the relay contacts to move toward the closed position.
Tr*ip brea..1.zer.
Reconnect closing coil leads to the trol circuit wiring, Check electric cJ.osing of SHUNT TRIP ATTACffi1,IENT The shunt trip mounts on top oi the platform immediately to the right of the operating mechanism.
{See Fig" 16 ,) 1t is non-adjus'cable and is intended for intermittent duty only. The shunt trip cir-* cuit must alv.rays be opened by an auxiliary switch contact. Tripping currents are tabulated in Table No. 2, Page 7. Inspection With the breaker in the open position, manually push the moving core against the stationary core and rotate the breaker handle tci the closed position.
The breaker should be trip free. The trip le,1er of the shunt trip should ' .1' _, /0" t .L /8 . ' l t nave _,:rom .l ,),:, o . --1ncn c earance o the trip ba:r. ..)
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J Fig. 17 -Under-voltage Trip Attachment
-Constn1ction Details o:f the arm<1.ture, to trip the breaker. The armature:
si1ould move 'Nithout friction, and should have approximately t:cav-el after tripping, Final inspection should 0e made after thr:; circuit COw"1ections are complete as shown in }'ig. 2, Page 10. 1Vl R.emove all power fr01n the brea..1-:er and :repeat "the II1echanical hS})ection given above. Check :for J.oose bolts and open cuit b potential coil. FIELD DISCHARGE S\V1TCH. The DB:F-16 breaker is a two-pole DP,-50
- .. ,,,,,1*n-"'pec1'n]
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Ct '-fied arcing contacts pJ.us a field discharg2 S\vitch mounted on the center pole 20). The field discharge switch is shipped with the gap setting shown in Fig. 20, for generator field protection.
Hoyvever, the I J ., .. ./
- * *Chec:z for }oose bolts and faulty coil. UNDERVOLTAGE TRIP ATTACHMENT The lmdervoltc.ge trip nE1ti.nts on top of the platform, to t1:1e :right of tl1e shunt trip. (S2e Fig. 17). H.s function is to trip the bre2.ker 'Nhen *tlJ.e volb.ge falls to betv1e2n 30 to 60 pe:ccsnt of no:msl. The co:re is noi-inaUy held m2-gneticaHy against the sb.tionary core to hold t.'rie i\1icarL".
- rc-rJ nr1d consequ2ntly tbe :csset 1ever, i11 tI1e reset position.
\Vhen the coil voltage i0 :reducsd sufficiently, the res*2t leve:r spring overcomes the mc.gnet}c 2ttr2.ction of the cores a11cl rot2.tes the reset leve:* clock'\vi.se.
r\.s the reset le*v91*
it c2.n::ies vrith it the latch pin which rntates relative to *foe laf:ch until the latch is relezised.
When the 1.3.tch rel:;ases, tt1e trip spring rotates -Che trip l2ver countercloc:z*,Nise to hip th-2 breaker. The latch is ye-.;et by the cross b3.r moving the adjustable lever as the breaker oi:;ens. Fig. 17 shows the c:rnss beer in the open position of the bre&.ker.
The seH-loc1dng scre'N in the movir1g core is set at the factory and not requixe 2djustm2nt.
It is used to lat.ch release the r:1oviDg core is 7 /32 outside the :Cc*ame. Ahvays connect the coil to the li.ne side of the brea..1-<:er unless t11e att2c11ment is equipped Yrith a time deJ.ay device. In this c2.se, the tiine delay *wil1 delay the tripping of the breaker long enough to permit energiz.abon of t}1e unde1"'10ltage coil from the le.ad side. Do "1ot use an au:x..'cliary switch con-tE:ct in the undervoltage ci.J:c1Jj_t.
The t.rip lever of the undervoltage should hzwe approxirnat21y J/16 inch c1eatance to the trip ba:: TJNDER.VOLTAG:s TLvlE DEL.A"_{ ATTACH!vtE;i'iT
'T'he undervolt2.ge ai..r da_shpot time delay :me21t n10i.mt..s on tb2 front of the undervo.lb.ge Fag e 27 trip, replacing movi:lg core cover. (See Fig. 17 .) 1l1e needle ,:2 scre-?1 in th2 top regulr.t:;s opening through 'Y'ihich the ,?.iris forced 2.\. :.1 the tirr1e delay. (S22 18.) rfhe attachrr1e7it does not have a qui.cl<: reset fe2ture and therefore approximately one minute should be 2110,,ved.
between ope:r2tions to perrait complete resetting.
It is s"'t to trip *within 4 to 7 seconds. Hold the trip ba.:r dov,rn a:nd close the bre,?ke::
Tf\..BDUally.
H.elec.:;e the trip bar slmvly, allo 1..ving the undervolt<:cge trip spring to raise the trip bLix 2nd trip the breaker. rilaintenance Check for loose bolts <1.:nd frmlty coils. REVERSE CURR.ENT TRIP ATTACJ-E\'lEI\T
'This atfachmed I!'.ounts dii-ectly on the cent2r molded pole unit bise, iJ1 the space ordi.n2.rily occupied by the overcurrer:t attachment. (See Fig. 19.) It is used to trip foe breaker whe:i the di..rection of curTent ilov; in that pole iE reversed.
When the series coil is flo'iv-ing in *::he :Eor'1va.rc1 direction, Inovernent is pT2-vented by a stop. When the series coil current i;:; reversed, the armah .. u*e rotates in the opposite direction to trip t.l-ie b:teaker.
Calibration ment covers 5 a.nd 25 percent reverse current, based on noTrnaJ cru.Tent rating. P..:fter tripping the rev21se cu.-crent armat1-tL2 is reset by opening the potential coil circuit. For *t.i.'-iis pm-pose 2n "a" . cont-J.ct of the brea}z2E auxi1iary switch should be connected in with the potentiaJ coil. Inspection Close the breaker rnanually, and push "\Vard on the sprinr-r, stud located on the bottom
.. --..
SALEH STATION I INSTRUCTOR LESSON PLAN
.. *------------------------------
1 TITLE: ATWT I DUR..:'.\
TI ON: REVISION NO o 0 DATE: 3--08-83 DATE; 3-08-83 D .. Z\TE: I I I I I 1 ___________ , ________ ;------* -:------*------------------
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--....... 1. '.1 INSTRUCTOR REFERENCES; I
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.-------------------------------------------------------------------11' VI. I.E.Bulletin 83-01 t I I I l I o Addresses the problem of failure of rx trip breakers due to a sticking of the under voltage trip attachment.
0 0 Additionally it sites similar failures that have occured at other *plants (H.B. Robinson, Conn Yankee, St. Lucie and Prarie Island). Due to these problems I.E. bulletins, I.E. circulars and Westinghouse T2chnical Bulletins were generated.
Up to these previous failures had prevented a warranted rx trip because involved one of the 2 series breakers.
not 1. -'-only o The required actions for this I.E. bulletin are: l. Perform surveillance test of UV trip coil 2. Ensure maint. program with the W directive or freq. and lubrication of lhe trip mechanism
- 3. Review Provide written follow-up reply concerning actions performed.
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SPi.OP8303;3 Page :!. Date 03/08/'lJ
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INSTRUCTOR LESSON PLAN ,-*-----*-------
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\ l -TITLE: A'I'W'I' LESSON NO" : DUR.c'\T I ON P.EVI SION 0 DATE: 3-08*-83 DATE: 3-08*-83 I l I
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_,-____ _ I CLASSROOM REQUI;:>,frIENTS:
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\ 1\TV?r rritle:
'J:P \ \ \ \ \v11. Modification to EI-4.3 Review Immediate nAutornatic a.nd "Manual" actions. I \ my ],_
- I lf .<;"i rev. 1s OD-15 use of Operations neot. procedures
,_.___ ______ w ____ __._,, __ ,P_ o The following procedures shall be performed in a step by step sequence as written in the body of the procedure unless the procedure specifically states otherwise:
Emergency Instructions overall operating rnstructions surveillance procedures Radioactive Waste procedures
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,._J ______________________
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I SAOP2i303:3 P 3.9 e 1 D ate ___ Qj __ /j) _ _Pu'. 8 J Rev. Q
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. SSINS No_: f22G ** O!S 315G-OGDi2
- f. :rn *{ i.'J ti C*tl ti: :
lE.B 83=Dl IE ND. E3-tn: f"fiiLURE OF 1fUP SHEAXERS TO O?Er: on l'RlP A\? i'Jf"e'5:Hi:
.. i1'20
\)!';l"-t2r hcld-!ng rm 1ice;"1;.2
{DL) 1or lH>i::1'2i1r hlci1lU<e".!.
f 1 J? in *fo \;,;,.; t. i O'!"J
- Thie vt.rrpc:;.-e tf t:>u1"ie't1n is to 1;fcn'E! CP J.bo-:Jt r>1:-:-er.t fai t:f 'r:i' D-B -:-.,,yp12 i:frc.w1t 1.D tr1p
- on of
- trip fro= the ree.ctttr sy:;t@i':':
.;;1r;d to 1e-a1.ifr't!
o::t i '.'.;{ <:i 1 I 00.zrJ.t 1t:t; Z'ed. "t1b:lC'l:.GY'*S tn
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- 25. 1983, st2rti.:J::
- r Uri1t 1 pl.Jiit, both i'.PJ--5
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h,.;,s. !::2er1 Hi.Cfdr:g of urirJcrve1ti:9e tr1p 2tt3th:::t:11't.
Tf.-t trir;µi:d l!'Ji'r.Litl°l\t frn;r: the car.trDi r.001r . .Ca);::it1!
30 a-fti:.o-: .gi.rto;;;iitii::
t.r"l;c> s19ir<l 0 l 'o'as: ;r,anua\1y 1r.iti{.lti:d
\;:l.<iS by tb: shur.i.
in"Site.11\!ri :;:-.ath br.2;j;;2r.
1i> so.::2 r2i0.t:tor G'125igr:'!-, odcr;.,;iti.!::
'5ign.;ds
.:.ire tiri1y ?;!'."! i..;r.ti-.;;rvol
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rif ttie riz"'ctoT i:.-:t,r;u.:d el"'-2 b-vth ::.c IJ"\' tr*;J:' <:.nd tc.;;, si-<ur;t tdp ccil of j., "J*";.l
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OP<<21'1 ,:iuto:?::1*U-ca"i1y 0;12 rd UV trip . * -Th?'Y:t h;:i.v-c
- ir: L£5"{ 1;,; E2-D!Z/C3>.::...(
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or, :i983 1 _Sii1-c::-:
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- 25. l9a3 2 of 3 the UV trip *ttachment co !he RPS'havt been to 'the. NRC .. Thest fa I 1vres oo1y one of the t"a they did not in a tc iutcmatita11y the Said failures have occurrt!d
&t M. s.* RonfnsonT Canrt.e,cticut l3-1and and St.
in addttiar.
to those at Silem. A5 a t.nf HRC issutd iE Builet\n 71-02 *r.d IE Cfrcular No. Bl-1Z. and issued Bulletin Ho. NSD-78-14-1 dated
- 11. 1974 ind HSD Data Letter-?4-Z dated Ftbrvary l4. 1974.
for All of 02erati"g Licenses for Wat.er Re.actors:
L1eensees W OB type brealers lttachcent fn ReJctor Protective Syste:
- re to: l.
test of undervoltage independe"t of the Shunt trip W'fth1n 24 ctf cf this unless equf va1ent testfn; beeo Within 5 days. These plant5 for en-line testab1l,ty not pravid-ed Illy this resur::.i"lg C'1etation or \f currtt>tly
- .at the nett plaf't shutdown_ .
fvr to retotmiended progra. {attacnmer-t}
in=ludin9 to Ver1fy actual Qf th!! program. lf including 1>>bricatior.
does not conf:nr:,
- such within S of this bulletin provide an '.!1ternate the testing requ1red in ite'l
- ': l to c1ecl4r1ng tbe breaker OPERABLE.
..-. J. Nottfy al1 11cerr;.ed of)erat.e>ri of whi::h occvrred Salte_
the for t.h.e e-,ent of fa p with each upar. hi-s arriv& l on-sM ft .. d. Prcvide written reply 7 days cf Teteipt of bu11!t\n *
- a.
of T'estsonse to l, .. , b.
confonr.ar.ce of p,.ogral"'\
- nd ci2scribin9 results of direct1y A
- f this item Zy c.
- d. ir.
1iceft$ed c.perators of the Salem to the{r attention appropriat!
fa ila:re .. 1.-:::-tTip opor. th-ei r arrri'141 on-sr.i ft. Vou of ""tQuirements for not1f1cat1on 1n aci:crdr.nce*w\tn 10 CFrt 50.:ZZ 1n thi-cf u1 1noperaol' RPS br-eaker. . ** *I I ' j J j j j j .. j .:* j j . j *j . I I I tEB 83-01 f 1983 .. Pa9e 3 of 3 Ucense!5 flOt using the $Ubjeet a*1'deryo\tag.e ind therefore by this shall submit*
declaratfon withfn 7 days of tM:
.. The writt!n sha11 be to the Regional under Oith or under provisfch$
Act cf 1954, Tht CDPY tn.e cover end i :opy of the shail be transm1ttea to U. S, Nutlttr Co::tr.i$s1or..
_D.c.* ZuSS5 fer
- .. This for inforaation h'JS approved by the Ofnce cf )',ana9tt.ent-&l'd Budget under a blanket clearance n(J13'1.ber which 3Q, igss. Cotmients on bul'"den and
- ay c.e to the Office and Sudget, Roum lZOB, New Office Buiiding, Washington, D.C. 20503. If re91Tding please RegiDnal of the NRC Office the contact be 1ow. Technical I. V411alv1.
-. .Original slg"ed*by
£, L. Jor-dan .;.
C.
Director cf irtd I .. !lii* l.
ltr. 74-Z ... 2.
of Recently !!sued IE Bullttins
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