ML20073R508
| ML20073R508 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 06/29/1982 |
| From: | Firebaugh L, Reeside B DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20073R489 | List: |
| References | |
| AP-1-A-5500-01, AP-1-A-5500-1, NUDOCS 8305040260 | |
| Download: ML20073R508 (8) | |
Text
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Form SPD-1002-1 DUKE POWER COMPANY '(1) ID No: AP/1/A/5500/01 ,
, PROCEDURE PREPARATION Change (s) O to
- PROCESS RECORD 0 Incorporated (2) STATION
- McGuire Nuclear (3) PROCEDURE TITLE: Reactor Trip i
(4) PREPARED BY: Bill Reeside DATE: 6-28-82 (5) REVIEWED BY: u ,T DATE: 6 / 2. P/ #' 1
- Cross-Disciplinary Review By
- N/R: /M (6) TEMPORARY APPROVAL (IF NECESSARY):
By: (SRO) Date:
By: Date:
(7) AFFROVED BY: 6 Date: b-M~b 4
(8) MISCELLANEOUS:
- i Reviewed / Approved By
- Date:
Reviewed / Approved By: Date:
Date/ Initial Verified with Control Copy /
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8305040260 830502
- j, PDR ADOCK 05000369
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I AP/1/A/ 10/01
, PAGE 2 0F 4 DUKE POWER COMPANY McGUIRE NUCLEAR STATION REACTOR TRIP .
1.0 Sy=ptoms 1.1 Any Alar = on Reactor Trip First Out Panel.,
1.2 All rod botto: lights are ille=inated.
1.3 Nuclear Instrumentation indicating a rapid decrease in Neutron flux.
2.0 I==ediate Action 2.1 Auto =atic 2.1.1 All rods drop into core.
2.1.2 Turbine-Generator-trip.
- 2. .3 Feedvater Isolation when Tavg decreases to 564*F.
2.1.4 Stea= Du=ps Ar=-Actuate and/or Main Stea= PORV lift.
2.1.5 CA Pu=ps start and feed all S/G's if 2/4 Lo-Lo level exist in 1/4 S/G's.
2.2 Manual NOTE If Reactor Power does not decrease rapidly and control rods are not inserted, or turbine fails to trip, this is an " Anticipated Transient With-out Scra=" event.
i j 2.2.1 If the reactor or turbine fails to trip when required, proceed to O AP/0/A/5500/54 ( Actioru Requited for. c.n Articipated T.t:auiert Witheat Sc, tam Evcitt) and perfor= applicable steps concurrent i
j with subsequent. steps in this procedure.
l 2.2.2 Verify Feedwater Isolation when Tavg decreases to 3,564*F.
(_ 2.2.3 Verify Stea: Du=ps arc-actuate and/or Main Stea= PORV's operate
>1125 psig.
t 2.2.4 Secure all boron dilution operations.
- 2.2.5 Verify CA Pu=p start and feed all Stea= Generators if 2/4 Lo-Lo level exists in 1/4 S/G's.
3.0 subsecuent Action l 3.1 Verify all required I:=ediate Acticas have occurred.
AP/1/A/5500/01 PAGE 2 0F 4 NOTE AP/1/A/5500/02 (Turbine Generator Trip) should be run concurrent with this procedure if applicable.
3.2 If any pressurizer PORV's open on high pressurizer pressure, ensure re-seating at 2315 psig decreasing.
NOTE If PORV f ails to close and pressure is less than 2315 psig, close the associated FORV isolation valves.
3.3 Ensure the CA System flow to Steam Generators. If not, =anually start the
=otor driven CA Pumps.
NOTE If the CA Pu=ps receive an auto-start signal depress the Train "A" and Train "B" CA Modu-lating Valves Resets in order to regulate flow to the steam generators.
3.4 Select " Reset" on the Moisture Separator Reheater Panel and prepare the MSR's for a hot start or shutdown per OP/1/3/6250/11 (Moisture Separator Reheater).
3.5 Verify no-load PZR level (25%) and Pressure (2235 psig) are restored and
=aintained. Verify charging and letdown flow normal.
3.6 Verify Tave is maintained >557'F and <562*F.
3.7 Verify Stea= Generator levelt are in the narrow range and S/G pressures are
'1090 psig.
3.8 Announce occurrence over plant paging systen.
3.9 Note the cause of the trip on the first out panel before resetting the alar .-
3.10 If all rods are not fully inserted, borate 150 pp: for each rod not inserted per OP/1/A/6150/09 (Boron Concentration Control).
3.11 Transfer NR-45 to one source range channel and one intermediate range channe;
- or indication. Ensure a negative period and decaying count rate.
CAUTION Ensure Pri=ary and Secondcry Syste=s have stabilized before transferring
, Stea= Du=p Controller to Pressure Mode.
I 3.12 Place or verify the Stea= Du=p M/A Station in AUTO. Transfer Stea: Du=p Controller to PRESSURE MODE. Verify stea dumps operate to control steam i pressure at approximately 1092 psig.
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AP/1/A/5500/01
. PACE 3 0F 4 3.13 Verify Volu=e Control Tank level is being =aintained.
3.14 Reset Hi Flux at' Shutdown Alar = when neutron flux decreases below setpoint.
3.15 Notify Che=istry to obtain a NC Syste= boron sc ple. Perfor= a reactivity balance calculation and maintain a shutdown cargin ecual to or greater than 1.6% ak/k-per OP/0/A/6100/06 (Reactivity Salance Calculation).
3.16 Notify Plant Manager or Superintendent of Operations per Station Direttive 3.1.6(Notifying Managenent of Operating Conditions).
3.17 Notify NRC Operation Center by ENS phone within one hour as described in Station Directive 3.1.4 (Conduct of Operations).
3.18 Place boilers in operation per OP/1/3/6250/073 (Electric Boilers) as neces-sary.
3.18.1 Close 1AS-9 ( C Etr. Eleed to AS). As lAS-120 (Aux. ilec. Bir. A
& B to AS Isol.) is opened slowly, throttle close 1AS-12 (SM to AS).
3.19 Shutdown the MG sets per OP/1/A/6150/OS (Rod Control).
3.20 Reset the " Negative Rate Trip" bistables on the power range drawers.
3.21 Close the reactor trip breakers.
3.21 Take manual control and close the following valves:
1CF-104 (A S/G CF Cntr1. Viv. Eypass) 1CF-105 (B S/G CF Cntrl. Viv. Eypass) 1CF-106 (C S/G CF Cntrl. Viv. Bypass) 1CF-107 (D S/G CF Cntrl. Viv. Bypass) 3.23 Reset Train A & B CF Isolation.
3.24 Open the following valves:
ICF-126-3 ( A S/G CF to CA No::le Isol.)
1CF-127-B (B S/G CF to CA No::le Isol.)
ICF-128-B (C S/G CF to CA No::le Isol.)
1CF-129-B (D S/G CF to CA No::le Isel.)
3.25 start a Feedwater Pu=p per OP/1/A/6250/01 (Condensate and Feedwater System) and secure Auxiliary Feedwater per OP/1/A/6250/02 (Auxiliary Feedwater Syster when desired and =aintained stea: generator levels at ne load value.
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AP/1/A/5500/01 PAGE 4 0F 4 3.26 If thermal power output was >15% at ti=a of reactor trip, notify primary che=1stry to perfor= isotopic analysis for iodine in accordance
. with Tech Spec 4.4.9. Note that this sa=ple =ust be taken and analyzed no sooner than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the trip and no later than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the trip.
3.27 Notify EP to perfor= required radioactive gaseous vaste sampling in accordance with Tech. Spec. 4.11.2.1.2.
3.28 Notify Projects and Licensing Engineer to contact site NRC inspector and infor= of trip. If P&L engineer cannot be reached then notify site NRC inspector of trip.
3.29 Deter =ine the cause of the reactor trip and cor:ect the proble=. If restart is' desired, proceed to OP/1/A/6100/05 (Reactor Trip Recovery).
NOTE If Reactor Trip occurred during startup and <15%
power, restart may co==ence per OP/1/A/6100/01 (Controlling Procedure for Unit Startup) .
3.30 If shutdown is necessary, proceed to OP/1/A/6100/02 (Controlling Procedure for Unit Shutdown).
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. : flevision 0 DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECK LIST (1) STATION: /NS-2.-O//9 UNIT: 1 2 ' 3
, OTHER:
i (2) CHECK LIST APPLICABLE TO: (NSM) #1/r - 2. - O / / 7 [ 6-~< hl/r - /_ /MM O C/E Div. O M&N Div. ETElect. Div. Q%il Divs.
(3) SAFETY EVALUATION -PART A
. The item to which this evaluation is applicable represents:
(RE)* Yes V No A change to the station or procedures as described in the FSAR?
(SRAL)Yes No / A test or experiment not described in the FSAR?
If the answer to the above is "Yes", attach a detailed description of the item being evaluated and an identifi-cation of the affected section(s) of the FSAR. sere uspi gc,t ceTauco parce pneA .
Q ff A(TER 7 cF FSAsP /5 f,pp ecr3D ,
(4) SAFETY EVALUATION -PART B (SRAL)
Yes No / Will this item require a change to the station Technical Specifications?
, if the answer to the above is "Yes", identify the specification (s) affected and/or attach the applicable page(s) with the change (s) indicated.
(5) SAFETY EVALUATION - PART C (SRAL)
As a result of the item to which this evaluation is applicable:
l Yes No # Will the probability of an accident previously evaluated in the FSAR be increased?
Yes No / Will the consequences of an accident previously evaluated in the FSAR be increased?
, Yes No / May the possibility of an accident which is different than any already evaluated in the FSAR be created?
Yes No / Will the robabilit of a malfunction of equipment important to safety previously evaluated in the FdAR be increased? ~
Yes No / Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
, Yes No May the possibility of malfunction of equipment important to safety different than any already evaluated m the FSAR be created?
i Yes No / Will the margin of safety as defined in the bases to any technical Soecification be re-duced?
- If the answer to any of the preceding is "Yes", an unreviewed safety question is involved. Justify the conclu-sion that an unreviewed safety question is or is not involved. Attach additional pages as necessary.
3[29h3 (6) PREPARED BY (SRAL): - . DATE:
-(7) REVIEWED BY (SRAL): -
DATE:
(8) REVIEWED BY (RE): b- DATE: 3- M-A3 (9) Page 1 o
'RE = Responsible Engineer SR AL = Safety Review, Analysis & Licensing Division l
MG-1-ll61o Rev. O MG-2-0119, Rev. O ATTACHMENT In the case of the modification to add the capacity for automatic actuation of the reactor trip breaker by means of the shunt trip mechanism, the following logic was applied to each of the items in section (5) of the Nuclear Safety Evaluation Check List (Attached).
- 1. Will the probability of an accident previously evaluated in the FSAR be increased?
No. The shunt trip mechanism was already in place in i.he manual trip circuit and malfunction is no more probable now than previously. Addi-tion of the automatic trip by means of the shunt trip will noi. inter-fere with, degrade, or replace automatic reactor trip by means of the undervoltage trip mechanism. For the above reasons, no previously evaluated accident is more probable.
- 2. Will the consequences of an accident previously evaluated in the FSAR be increased?
No. There will be no degradation of the reactor trip mechanism or its associated systems as a result of this modification. Implementation of this modification will enhance the reliability of the automatic reactor trip function.
- 3. May the possibility of an accident which is different than any already evaluated in the FSAR be created?
No. This modification enables automatic actuation of the shunt trip.
The shunt trip was already in place and in use as a manual trip for the reactor trip breakers. Enabling automatic initiation of both the undervoltage and shunt trips will increase the reliability of the reactor trip function and will not degrade the reactor trip system.
- 4. Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
No. As stated previously, the shunt trip was already in place and is no more likely to fail now than before. Interference of the shunt trip mechanism with the undervoltage trip mechanism is not considered credible.
The circuitry enabling receipt of the automatic trip signal by the shunt trip mechanism was designed to the high standards of the ESF system,
- 5. Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR by increased?
No. As stated previously, there will be no degradation of the reactor trip mechanism or its associated systems as a result of this modification.
Implementation of this modification will enhance the reliability of the automatic reactor trip function.
- 6. May the possibility of malfunction of equipment important to safety differ-ent than any already evaluated in the FSAR be created?
No. As stated previously, the only new malfunction mechanism would be
interference of the shunt trip mechanism with the undervoltage trip mechanism. This is not considered credible.
- 7. Will the margin of safety as defined in the based to any Technical Specification be reduced?
1 No. The reactor will trip at least as reliably and quickly as with-out this modification.
The addition of the capability for automatic actuation of the reactor trip breaker by means of the shunt trip mechanism will enhance the reliability of the reactor trip function. All trip mechanism problems to date have occurred in the under-voltage trip mechanism. Although enabling automatic initiation of the shunt trip is not an alternative to the undervoltage trip or a fix for the undervoltage trip problems, implementation of this modification provides additional assurance that the reactor will trip when an automatic trip signal is generated. The shunt trip mechanism is highly reliable and implementation of this modification has increased Duke Power's confidence in the ability of the reactor trip system to trip the reactor on demand.
NOTE: This package was developed to provide documentation of the logic applied in determining that an unreviewed safety question does not exist due to the modification enabling automatic trip of the reactor trip breaker by means of the shunt trip mechanism.
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