ML17055B524
| ML17055B524 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/21/1985 |
| From: | Abbott R, Gayne R NIAGARA MOHAWK POWER CORP. |
| To: | |
| Shared Package | |
| ML17055B512 | List: |
| References | |
| N2-IOP-101A, NUDOCS 8604280274 | |
| Download: ML17055B524 (184) | |
Text
Enclosure 817 NINE MILE POINT NUCLEAR STATION UNIT 82 INTERIM OPERATING PROCEDURE PROCEDURE NO. N2-IOP-101A PLANT START-UP DATE AND INITIALS APPROVALS SIGNATURES REVISION 0 REVISION 1 REVISION 2 Supervisor Operations q/~ ss.
k Station Superintendent y,p~,]~
c(~v'ummer of Ps es Revision 2 (Effective 6/21/85 )
~Pa e Date 1-2 April, 1985 4-5 May, 1985 i, 3, 6"62 June, 1985 NIAGARA MOHAWK POWER CORPORATION THIS PROCEDURE NOT TO BE USED AFTER INITIAL FUEL LOAD.
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f TABLE OF CONTENTS T!TLE PAGE SECTIOII Technical Speci fications Requirements System Description Plant Operating Requirements Precautions Startup Procedure Normal Operation 17 Shutdown Off-Normal Procedures 18 Procedure for Correcting Alarm Conditions 18 19 Modified Master Startup Checklist. 28 Attachment II Attachment III Master Startup Checklist N2-IOP-1 01 A -i- June, 1985
N2-IOP-101A PLANT START-UP TECHNICAL SPECIFICATION REQUIREMENTS Oue to the fact that most, ff not all, of the Technical Specfffcatfons apply to plant operation, they are not lfsted individually fn this procedure.
SYSTEM OESCRIPTION As a descrf ptf ons 'f plant start-up fs affected 1 1 by all plant systems, not be df scussed f n thi s procedure.
individual system indfvfdual operating procedures for descriptions of systems that are of See the interest.
PLANT OPERATING REQUIREMENTS 1.0 All main plant systems and their auxflfarfes are required to be fn operation, or fn a condition of standby, where their operatfon could be accomplished fn direct accordance with their particular operating procedure, and the technfcal specifications which apply to them.
2.0 Portions of redundant mechanical pumping systems, or electrical distribution systems, are permitted to be inoperable for maintenance purposes under certain conditions, described as such, in technical specifications.
3.0 Duration of such inoperabflftfes must comply speciffcally with the systems related technical specifications.
- 4. 0 DI SCUSS ION It fs the intent of this procedure to outline the many steps requi r ed to achf eve saf e reactor start-up. The sequence suggested may require deviation by the CSO and SSS to allow for the many possible existing plant conditions.
4.1 General Functf ons aQ Plant Operating Orders:
The order to start the reactor and systems required for power generation or to shutdown the station is issued by the P1ant Superintendent or hfs designated alternate.
N2-IOP-101A Aprf 1, 1985
- b. Checkoff Lists:
+j't C ~ Log Book:
Dfrectfon for placing moor equfpaent fn and out of service will be given by the control room CSO who will record the time and status in the log books as the work i s accomplished.
PRECAUTIONS 1.0 Ensure that all control rod motion fs done fn accordance with an approved control rod sequence.
2.0 Extra caution should be used when pulling control rods fn the r egf on of crf tfcal i ty to avoid short per f ods. Cri tf cal predf ctfons should be used only as a gross estimate of the crf tfcal rod pattern sf nce there are many calcul atf onal uncertainties fn the prediction process. It should be noted that fn many previous short period incidents throughout the industry the operator thought that the reactor was substantf al ly sub-critical due to unexpectedly low SRM readings. Addftfonally, the "conti nuous wf thdrawal " mode shall not. be used when approaching critical ty. i The followf ng reactor conditions and characteristics fnfluence the point of criticality and the rate t
at whf ch f f s approached:
2.1 Xenon Concentration Xenon tends to suppress the flux fn previously hi.gh-powered.
regions of the core (generally. bottom and center). Since control rod worth fs a function of the flux to which it is exposed, rod worth fs diminished in high Xenon concentration regions and enhanced in other regions.
2.2 Moderator Tem erature At higher temperatures, neutrons travel further fn the slowing down process; and therefore, have a greater probability of reaching and being absorbed in a control rod. This results in increased control rod worths at higher temperatures.
2.3 Control Rod Position The zero worth of a control rod depends on fts axial position as follows:
N2-IOP-1 01 A Aprf 1, 198S
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Marth O 4 Low High 8-12 Highest 12-16 High 16-24 Low 24-48 Minimal 2.4 Order of Mithdrawal If the rods of an array are defined such that sufficient radial separation is maintained, the flux will be loosely coupled and the rods may be withdrawn in sequential order with decreasing rod worths (i.e., the first rod of a group is generally worth more than successive rods in that group).
3.0 The precautions in each of the specific procedures referenced to herein are to be adhered to.
E. START-UP PROCEDURE 1.0 PREREgUI SITES l,l One of the following three check lists complete.
- a. Master Startup Checklist completed and attached. This checklist applies to all circumstances, except those defined in b. and c. below.
- b. Modified Master Startup Checklist completed and attached.
This checklist is used in those circumstances greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, following an outage, where the work done involved only a few specific items.
C.. Short Form Startup Checklist completed and attached. This checklist is used for a startup that begins within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of shutdown, and involves no moor work which could affect the status of the plant systems.
1.2 Master Outage Check Sheet complete and atta-hed (not required for a Short Form Startup or a Modified Master Startup).
1.3 Condensate and Feedwater Systems flushed at least four hours prior to start-up per N2-IOP-3, N2-IOP-101A June, 1985
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1.4 Criticality prediction obtained from the Rx Physics Department.
1.5 A properly approved .Rod Sequence Package has been obtained from the Rx Physics Department.
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1.7 Verify the following surveillances are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of start-up unless completed within the previous seven days:
- a. Source Range Monitors (N2-IOP-92):
- 1. SRM Channel Functional Test N2-ISP-NS-M8010 (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to moving the mode SM from shutdown).
- 2. SRM Channel Check N2-OSP-LOG-S001 (every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />).
- b. Intermediate Range Monitors (N2-IOP-92):
- l. Verify N2-OSP-LOG-S001 (IRM channel check) has been performed within the last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />).
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- 2. Perfom NZ-ISP-NMS-'0001 (IRM channel functional test) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to start-up, i f not performed within the previous seven days.
c~ Average Power Range Monitors (N2-IOP-92):
- 1. Check APRM bypass switches and record any bypassed APRM' in the CSO log along wi th the reason for bypassing.
- 2. Verify NZ-OSP-LOG-S001 (APRM channel check) has been performed.
N2-I OP-1 01 A May, 1985
- 3. Verify N2-ISO-K-W9003 (APRM neutron full set down function) has been performed.
- d. Rod Worth Minimizer:
- 1. Perform N2-XXX-XXX-XXXX (RWM operability) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to comnencing control rod withdrawal for the purpose of making the reactor critical).
NOTE: lhe Reactor mode switch may be placed in "START-UP/HOT STANDBY" for the purpose of completing this test. See T.S. 4.1.4.1.
- 2. Verify RN is not bypassed.
NOTE: If the 884 is inoperable, comply with Action a of T.S. 3.1.4.1 and docunent in CSO log including name of rod pattern verifier.
- e. Rod Sequence Control System
- 1. Perform NZ-OSP-RMC-8004 (RSCS operability) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to reactor start-up.
1.8 Verify all IRM's full inserted.
1.9 Position all IRM/APRM and IRM/RBM recorder select switches to the "IRM" position, 1.10 Yerify all IRM range switches on position l.
1.11 Check IRM bypass switches and record any bypassed IRM's in the CSO log along with the reason for bypassing.
1.12 Verify all SRM's fully inserted.
1.13 Position SRM recorder channel select switches to record the two highest channels.
NZ-IOP-1 01 A May, 1985
1.14 Verify a reading of at least 3 cps, on at least 3 SRM channels.
1.15 Check that Rx water level is in the normal operating range and WCS system is operating to control water level per N2-IOP-37, H.11.0 through H.11.7.
1.16 Verify Rx vessel head vents are open and lined up to the suppression pool (MSS-MOV118, MOV119 open and MSS-MOY108 closed).
1.17 Verify the shutdown cooling mode of RHS is secured and that RHS is lined up in the standby mode (N2-IOP-31, H.3.3).
1.18 Verify the main turbine is on the turning gear (panel 851 "Turning Gear Engaged" and either "Motor Fast" or "Motor Slow" red lights lit).
1.19 Verify Rx recirc. system is in normal operation on the LFMG's (N2-IOP-29, E.3.0 through E.3.7).
1.20 On panel 824 verify group I, II and III steam drain valves are open.
1.21 Verify authorization received for start-up from Station Superintendent or alternate.
1.22 Announce "Primary and 'Secondary Containment now in Effect".
1.23 Notify dispatcher of impending start-up, 1.24 Announce reactor start-up is about to cogence.
1.25 Place the condenser Low Vacuum Bypass Switches in the Bypass Position (2 switches: Div. I, panel 2CEC-PNL609; Div. II, panel 2CEC-PNL611).
N2- I OP-101A June, 1985
2.0 APPROACH TO CRITICALITY, HEATUP AND VESSEL PRESSURIZATION 2.1 Mithin 15 minutes prior to control rod withdrawal, verify reactor coolant pressure and temperature to the right of the critical line per N2-OSP-RCS-8001 (TS Fig. 3.4.6.1-1C).
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2.2 Place the Reactor Node Switch in START HOT STBY position.
2.3 Verify all Rod Blocks Clear.
2.4 Place SRM and at least one IRM recorder on fast speed.
NOTE: The following rules apply to control rod withdrawal:
Mhen 25% of the control rods have been pulled to position 48, DISCONTINUE use of the "Continuous Mithdraw" switch for control rod withdrawals between positions 0 and 12 where notch worths are high.
Mhen 50% of the control rods have been pulled to position 48, DISCONTINUE use of the "Continuous Mithdraw" switch between rod positions 00 and 24 while start-up and reactor heatup are in progress (less than one bypass valve open or until the generator is on line). See Precaution Numbers 1.0 through 2.4 (above) and Section D of N2-IOP-96.
as Commence Control Rod withdrawal (N2-IOP-46, E, F) in accordance with the specified sequence as follows:
- a. Select the appropriate control rod to be withdrawn by depressing its respective select pushbutton.
- b. Mi thdraw the selected control rod by momentarily depressing the "withdraw" pushbutton for notch withdrawal or by ho1ding down the "Continuous Mithdraw" pushbutton and then the "Mithdraw" pushbutton for continuous withdrawal as applicable (see preceding NOTE).
c~ Continue to withd;aw control rods as above while monitoring neutron flux, reactor period and actual control rod movement.
N2- IOP-101A June, 1985
2.6 Perform Control Rod coupl fng integri ty for each rod as f t fs pulled to Position 48 and fnftfal the correspondfng rod pull sheets.
2,7 When reactor reaches criticality (neutron count rate fncreases at a .logrfthmfc rate without control rod movement), establish a positive stable period greater than 60 seconds and record the following information fn the CSO Log:
- a. Time b, Rod Number ci Rod Group
- d. Rod Posftion
- e. Reactor Period (1.44 X Doublfng Time)
- f. Reactor Water Temperature go Person Pulling Crftfcal 2.8 Announce over the paging system, "The Reactor fs Critical".
2.9 Yerf fy the SRM/IRM Overl ap of one-half a decade by vf sual ly observing that all IRM's are above downscale before any SRM Count Rate i s above 105 CPS wf th the SRM' fully inserted, per N2-0SP-LTR, TS Table 4.3.l.l-l, Note (b).
2.10 Withdraw SRM's as required to maintain a count rate between 102 and 105 CPS.
2.11 Range up on the IRM range switches,'ndividually, as required to mafntafn a reading between approximately 25 and 75 on the 0-125 scale.
NOTE: Perform N2-OSP-RCS-M01 (Reactor Heatup/Cooldown) while heating up the reactor until the reactor is fully pressurized with pressure being controlled by the tur bine bypass valves.
2.12 When the heating range is reached, as evidenced by a .downturn in reactor power, withdraw control rods, as necessary, to maintain a heatup rate of less than 100oF in any one hour period.
N2- IOP-101A June, 1985
2.13 Fully withdraw the SRM's when the IRM's are on Range 3 and on scale (N2-IOP-92, E.2.12 through E.2.18).
2.14 Using the WCS system, reject water as necessary to maintain reactor water level in the normal control band (N2-IOP-37, H.ll.0 through H.ll.7).
NOTE: Condensate .demfneralfzers are placed in service and removed from service as required by condensate system .demand.
Demfneralfzers are placed in service prior to any
.demfneralfzer exceeding 3000 gpm or a system delta p of 55 psfd. However, .do not add so many demfneralfzers that the flow/demfneralfzer is less than 500 gpm.
2.15 Prior to reactor water temperature reaching 212%, verifyN at least one condensate pump and (as required) one condensate booster pump (N2-IOP-3, E.l through E.13) and 2 to 4 demfnei alfzers (N2-IOP-5,E.l) are operating normally through Lv137 to ensure water makeup capacity to the reactor vessel (N2-IOP-6, E.4.0 through E.4.4). Operate the condensate/feedwater system fn accordance wfth N2-IOP's 3, 5, 6, and 7.
NOTE: If any steam line paths (f.e., even drains) are open before the condensate and condensate booster pumps are on lfne, reactor water level will have to be monitored closely as the temperature reaches 212oF. A point will be reached where the steam flow will exceed RDS cooling flow, and this loss fn inventory could result in a low water level scram.
2.16 Throughout the remafnder of the heat up, monitor the drywell temperature and maintain its average less than 150oF (T.S.
3/4.6.1.7). Start additional drywell cooling units (N2-IOP-60) as required.
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2.17 Start up the main steam system fn accordance with N2-IOP-1.
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CAUTION On-off cold f.edwater flow cycling at low power operation fs one of the major contributors to feedwater nozzle thermal duty and shou1d be avoided.
N2- I OP-101A June, 1985
2.18 Throughout the remainder of the heat up, monitor the drywell pressure and maintain it between -0.5 and +.75 psig, drywell equipment hatches and air locks closed, the pressure wil',
With the
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increase with the increasing temperature until a drywell equilibrium temperature is achieved (average less than 150oF).
The pressure increase is kept within limits by opening the containment purge (CPS) 2" outlet line to atmosphere th~ough the standby gas threatment bed (A or B).
If this does not adequately reduce pressure, the respective f'r Standby Gas Treatment System (GTS) fan fs started and run brief periods of time.
Closely monitor drywell pressure during this operation.
aO Open DW Purge INBO 2CEC*PNL875 Outlet Isol Vlv 2CPS*AOV108
- b. Open DW Purge OTBO 2CEC~PNL873 Outlet Isol Ylv 2CPS~AOV110
- c. Open PNL Cont Purge 2CEC*PNL873 Exhst to SGTS 2GTP'SOV102
- d. Open SBGTS Train A 2CEC*PNL870 Inlet Valve (*PNL871) 2GTS*MOV2A (2B)
- e. Veri fy SBGTS Train A (B) 2CEC*PNL870 Closed Decay Heat Cooling (*PNL871)
- f. Close SBGTS Train A (B) 2CEC*PNL870 CROSS BLEED PIPE (*PNL871)
Val ve
- g. Open SBGTS DISCH FAN 1A (1B) 2CEC*PNL870 DISCH Valve (*PNL871) 2GTS*MOV3A (3B)
- h. If the drywell pressure can be maintained and reduced by this method, skip to step k. If it cannot be maintained or reduced, do steps i,,i, and k (below).
Start 5BGTS DISCH FAN 2CEC*PNL870
- FNlA (*FNlB) (*PNL871)
Adjust Reactor Bldg 2CEC*PNL870 as In/OUT Diff Press (*PNL871)
Required N2- IOP-101A June, 1985
This Controller adjusts the recirculation valve *PV5A
(*PY58) between FN1A (18) inlet and outlet. A fully closed
- PV5A (58) would allow a more rapid depressurfzatfon of the drywell.
- k. At the desired minimum drywell pressure (approximate1y 0 psfg), stop'SBGTS "DISCH FAN, *FN1A (18)"; close the valves a through d and h above; open the valve in f, reset controller *PY5A (58) to its original setting.
2.19 Start-up the clean steam reboiler and establish main turbine steam seals in accordance with N2-IOP-25, E.l through E.29.
2.20 Shutdown the MSR steam blanketing system in accordance with N2-IOP-2, E.2.0 through E.2.3.
2.21 Draw condenser vacuum using the condenser air removal pumps in accordance with N2-IOP-9, E.1.0 through E,l.9.
2.22 Preheat the off-gas system fn accordance with N2-IOP-42, section E.
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2.23 Mhen reactor pressure reaches 5 psig, close reactor vessel head vent valves (MSS-MOY118, -MOY119). Open head vent to 'A'ain steam line (HSS-MOY108).
2.24 If MSIV's were not open prior to reactor start-up, open the MSIV's fn accordance with N2-IOP-1 (less than. 5 psig, Section E, greater than 50 psig use H.l see third note below).
2.25 Mhen reactor vessel pressure reaches 75 psfg, reset the ICS low pressure isolation and commence warming and aligning the ICS steam line in accordance with N2-IOP-35, E.1.0.
2.26 At 90 psig verify ADS operability. Perform N2-ISP-ADS-M005 (ADS chan TST - comp gas lo press alarm); N2-ISP-ADS-M006 (ADS auto
'act funct TST); N2-OSP-ADS-R002 (ADS accum leak rate test);
N2-ISP-ADS-R105 (ADS chan calib comp gas low press alarm).
N2- IOP-101A June, 1985
i NOTE: AOS system must be verified operable prior to reactor pressure exceeding 100 psig. See also Step E.2.26 above and E.3.12 below.
ICS must, be verified operable prior to reactor pressure exceeding 150 psig (see E.2.25 above and E.3.11 below).
Condenser vacuum must be established prior to opening a bypass valve with Uie vacuum being maintained by the SJAE's and not the condenser air removal pumps (N2-IOP-9, E.2.0 through E.2.15).
NOTE: While shell warming avoid a sudden spike in shell pressure as it could result in a reactor scram.
could make it if (A pressure spike the thermal power is greater appear as than 30%. The turbine stop valves are less than 95% open.)
2.27- At 100 psig, coranence turbine shell warming (chest also warms in the process) and warm as reactor pressurization continues in accordance with N2-IOP-21, E.2.0.
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2.28 Establish main condenser vacuum using the SJAE (N2-IOP-9, E.2.0 through E.2.15).
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2.29 Concurrent wi th step 2.28, start-up the . off-gas system in accordance with N2-IOP-42, E.19.0 through E.23.0.
2.30 At approximately 150 psig, the EHC system will open a turbine bypass valve, verifying bypass system response to the pressure regulator. As soon as a bypass valve opens raise the pressure regulator setpoint 50 psig above system pressure (this keeps the bypass valves closed) and maintain the setpoint 30 .to 50 psig greater than reactor pressure until 940 psig is reached.
2.31 Shut or verify shut: 2MCS-MOV 106 (WCS ORN to Maste Collector Tank) and 2MCS-MOY 107 (MCS ORN to Hain Condensor) and open or verify open their respective motor breakers and mark up and tag out all valves and breakers.
2.32 At 200 psig, start-up the auxiliary steam system in accordance with N2-IOP-I, E.25, E.26, and change the reboiler and SJAE supplies over to auxiliary steam.
N2-IOP-101A June, 1985
2.33 Shut or verify shut 2MSSWOVll2 (Mafnsteam line Drain out"=:srd fsol.ation valve) and open or verffy open fts Motor Breaker> and mark them both up with hold out, tags.
2.34 When the condensate booster pump dfscharge pressure is still about 200 psig'bove reactor pressure, start a motor driven feed pump and control water level fn accordance with N2-IOP-3.
CAUTION As reactor power increases, do not allow any unbypassed APRM channel to exceed 12% while in the start-up mode. (A control rod block occurs at 12% and a reactor scram occurs at 15% any time the mode switch fs not in RUN.)
2.35 Verify APRM comes on scale during power fncrease.
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2.36 Verify main steam line low pressure annunciators c ear a grea er than 765 psfg.
2.37 Place the low condenser vacuum bypass switches fn normal after main condenser vacuum fs establfshed (two switches, Dfv. I, II; 2CEC*PNL609 and *PNLP611).
2.38 Perform the Anal Drywell Inspection ff required, and verify that the Primary Containment AC Circuit in TS 3/4.8.4 are de-energized.
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3.0 TRANSFER OF MODE SMITCH TO "RUN" 3.1 Add condensate demfneralfzers as required to maintain between 500 gpm and 3000 gpm per on line demfneralfzer and a system delta p less than 55 psfd (N2-IOP-5, E.l).
3.2 Stop pressure increase when the steamline pressure reaches 940 psfg and verify that the bypass valves begin to open to control pressure.
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N2-IOP-101A June, 1985
3.3 Continue to pull rods until APRM downscale lights have cleared.
- 3. 4 Yerk fy the AP ' RM are reading greater than 5% by changing alternate recorder select switches from IRM to APRM leaving one recoeder in each channel in the APRM position.
3.5 Verify
- a. Steamline press greater than 765 psig. /
- b. APRM on scale between 5%-12%.
- c. APRM downscale alarm has cleared.
- d. MSIY's are greater than 94'X open.
- e. Condenser vacuum greater than 25" Hg.
f, Low vac alarm has cleared Div. I/II.
3.6 Shift the reactor mode switch to "RUN".
3.7 Select the APRM or RBM position on all IRM/APRM or IRM/RBM recorders.
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3.8 Continue to increase power by wi thdrawing control r ods in sequence. (The recirculation system remains on 15 Hz with the flow control valves wide open.)
3.9 Fully withdraw the IRM's from the core. /
3.10 Mith data from th'e reactor arialyst, verify the thermal limits:
APLHGR (N2-XXX-XXX-XXX, TS 3/4.2.1); bCPR (N2>>0SP-XXX-XXXX, TS 3/4.2.3); and LHGR (N2-0SP-XXX-XXXX, TS 3/4.2.4) are equal to or less than their respective limits within 12 hours after completion of a THERMAL POMER increase of at least 15% of RATED THERMAL POMER.
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3.11 Perform ICS 1000 psi g operabil i ty technical speci fication surveillance N2-0SP-ICS-(002,(TS 4.7.4.b) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.
3.12 Perform ADS val ve operabil i ty technical fi speci cation sur veil 1 ance N2-OSP-ICS-R001 (TS 4.5.1.e.2. b) wi thin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.
N2- IOP-101A June, 1985
4.0 TURBINE STARTUP 4.1 Begin chest warming if shell 500oF warming is complete ~nd the chest temperature is less than (N2-IOP-E21, i.~.0. through E.3.4).
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4.2 Continue to increase power until 2 to 3 1/2 turbine bypass valves are open.
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4.3 Notify Power Control that the turbine is ready to roll.
4.4 Br ing the main turbine on line in accor dance wi th N2-IOP-21 (E.4.0 through 4.20),
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4.5 Synchronize the Main - erator to the grid per N2-IOP-68 (E.1.0 through 1.23) and then intaediately increase the load set on EHC to pick up 50-70 mwe load (this prevents motoring the generator).
4.6 Slowly increase the load set on EHC (keeping the rate of change of the first-stage shell metal temperature at or below >50F/hour) until all turbine bypass valves are shut, then increase the load set to 110% above turbine MWe (1188).
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4.7 Vent the feed water heater tubes as required, N2-IOP-B, E.l.5.
4.8 Transfer house service from Reserve to Normal power in accordance with N2-IOP-71, .F.2..
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4.9 Transfer Start-Up Feedwater Level Control Valve to Normal Level Control per N2-IOP-7 (E.6.0, E.7.0, E.8.0).
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4.10 On panel 824 take the Group I steam drain valve switch to close.
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5.0 INCREASE OF POMER TO RATED NOTE: The Reacto~ Scrams.. and the Containment Isolates at 1.68 psig'in the Drywell.
5.1 Commence inerting of Drywell and Suppression Chamber in accordance with N2-IOP-61A, E.2.0 through E.3.8.
N2- IOP-101 A June, 1985
NOTE: Drywell and Suppression Chamber atmosphere oxygen concentration must, be less than 4% by volume, based on noncondensible gases within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after thermal power exceeds 15% of rated (TS 3/4.6.6.4, N2-0SP-LOG-M001).
5.2 Add additional condensate demineralizers as required to maintain between 500 gpm and 3000 gpm per on line demineralizer and a system delta p of less than 55 psid. Seven demineralizers are required for 100% power (N2-IOP-5, E.l).
5.3 Increase reactor power with Control rod withdrawal in'ccordance with Reactor Analyst instructions.
5.4 Hetween 10-15% power close all feedwater heater shell side startup vents, N2-IOP-8; E.1.6.
5.5 At approximately 15% power, place the feedwater control system in "3 element" control in accordance with NZ-IOP-7, E.9.0 through E.9.2 (Feedwater Control System).
5.6 At approximately 15% power take the Group II steam drain valve switch (Panel 824) to .close.
5.7 At approximately 20% power add a second condensate, condensate booster pump N2-IOP-3.
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5.8 At approximately 30% power take the group III steam drain valve
,switch (Panel 824) to close.
5.9 At approximately 35% power add a second feedwater pump (N2-IOP-6,
-7).
5.10 Transfer reactor recirculation pumps to high speed at approximately 35% power per N2-IOP-29, Section E.5.0 through E.5.6. Leave the flow control valve on minimum flow.
N2- I OP-101A June, 1985
5.11 At approximately 45% power, start the fp>rth point heater drain pumps on minimum flow to clean the>r water volume prior to valving them into the feedwater (N2-IOP-8, E.2.0 through E.2.6).
5,12 Corrtinue power increase by control rod withdrawal until the 100%
rod. pattern line is intersected (figure 1, N2-IOP-29).
\
NOTE: Do not exceed the following parameters:
- a. Reactor Thermal Power 3323 MMt
- b. Total Core Flow 108.5 M lb./hr.
5.13 Continue (reactor analyst approved) power increase by increasing recirculation flow per N2-IOP-29 (E.7.7 and F.8.0) until 100%
power is achieved.
5.14 At approximately 65'X power place. the fourth point heater drain pumps in operation per N2-IOP-8 (E.2.0 through E.2.10).
NORNL OPERATION NOTE: During the late fall, winter, and early spring, Cooling Tower/C ir cul ating Mater temperature wil 1 have to be monitored closely to prevent freezing of the PVC piping in the cooling tower, see N2-IOP-10A, F.2 thru F.6.
1.0 LIMITATIONS 1..1 Core thermal power shall not exceed the limit. of 3323 MMT.
1.2 All modes of plant operation shall be within the limits of the technical specification.
1 3 All systems shall be operated within the guidelines of the respective operating procedure.
2.0 POMER INCREASES 2.1 Power increases can be make by control rod withdrawal or flow increases. Power increases shall not be made by control rod withdrawal and flow increases simultaneously.
2.2 All control rod movement shall be in accordance wi th the rod sequency that is in effect as .determined by the Reactor Analyst Department.
N2-IOP-101A June, 1985
2.3 During power'ascension,,'.for each control rod sequence, fuel pre-conditioning rates shall not be exceeded. Preconditioning will be in accordance with the reactor analyst procedures, 3.0 POWER DECREASE There are'o restrictions on power decreases.
3.1 Decrease recirculation flow, as required.
3.2 Insert control rods in proper sequence per Reactor Analyst's instructions.
4.0 POWER ASCENTS FOLLOWING PRECONDITION ING 4.1 Providing the fuel has been preconditioned, power ascents after power .decrease shall be made at the rate determined by the Reactor Analyst Department.
5.0 CONTROL ROD INTERCHANGE 5.1 Control rod interchange will be made using approved Reactor Analyst Procedures.
G. SHUTDOMN Shutdown procedure will be covered in N2-IOP-100B, Plant Shutdown.
H. OFF NORMAL PROCEDURES N/A
I. PROCEDURE
S FOR CORRECTIVE ALARM CONDITION N/A N2-IOP-101A June, 1985
ATTACHMENT I N2-IOP-101A NINE MILE POINT - UNIT 2 SHORT FORM" STA~LtET Date/Time Completed Reviewed By Approved By Items that cannot be satisfied shall be noted in Section G.l 0. Entry into Operational Condition Two (Startup) shall not be made unless the surveillance requirements associated with the Limiting Conditions for Operation have been performed'ithin the required surveillance interval (T.S.4.0.4).
For a system/component to be considered as operable, the following must have been accomplished:
A. The required mechanical/electrical checklists have been performed on that applicable system/subsystem/component.
B. All periodic survefllances satisfactorily completed on that system/subsystem/component.
The SHORT FORM STARTUP CHECKLIST shall be conducted by a licensed Reactor Operator and Reviewed by the Chief Shift Operator. The Shift Supervisor shall approve the checklist after he has verified that items in G.10 are not required for startup.
The conditions are satisfied as listed in A through G.
- The SHORT FORM STARTUP CHECKLIST is used for a startup that begins within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of shutdown and involves no major work which could affect the status of plant systems.
N2-IOP-1 Ol A -1 9- Dune, 1985
tnt ci a i s j Uaze A. SCRAM IDENTIFICATION
- 1. The cause of the scram has been identified and corrected.
B. REACTOR AND EMERGENCY SYSTEMS
- 1. Shutdown Margin has been determined to be adequate per T.S.3,1,1.
- 2. Reactor Vessel temperature greater'han the minimum shown on T.S.
Figur e 3.4.6-1C (T.S.4.4.6.1.1).
- 3. Estimated critical rod position has been received from the Reactor Analyst Department.
- 4. Low Pressure Core Spray system (CSL) operable-IOP-32.
- 5. Three independent Low Pressure Coolant Injection subsystems operable.
6, High Pressure Core Spray system (CSH) operable.
- 7. All Control Rod Drives are operable (T.S. 3/4.1.3.1).
- 8. Reactor Core Isolation Cooling system (ICS) available.
- 10. Control Rod Drive Hydraulic system in operation. CHECK for CRD Accumulator.. Inop lights.
ll. Reactor Coolant 5 ECCS Leak Detection System in operation.
N2- IOP-101A June, 1985
x<>i viui vi uawC
- 12. Reactor Recirculation Pumps fn operation fn slow speed in accordance wi th N2-IOP-29, ~
'3.
Reactor-..Mater Cleanup System in operation and capable of re)ecting water to Radwaste facilities or to the main condenser in accordance wi th N2-IOP-37.
- 14. Main Steam System avaflable.
- 15. Standby Lfgufd Control System operable.
4
- 16. Redundant Reactivity Control System operable.
- 17. Verffy that the followfng valves are shut and that their associated motor breakers are open and that both valves and breakers are marked up and tagged out. (This fs an NMP2 Appendix R Cmeftment.)
Valve 1 Descrf tion KRPRR32A . Cond. Mode to ICS 2RHS~V32B RHS B Stm. Cond. Mode to ICS 2RHSWOV37A Ht. Exch A Drain to Sup Pool 2RHSAOV37B Ht. Exch B Drain to Sup Pool 2DER&OV129 DER Dr afn Isol Vlv from Reactor btm hd 2RHS&OV22A RHS "A" Stm Line Isol 2RHS+MOV22B RHS "B" Stm Line Isol 2RHS&OVBOA Stm Line A Isol Vlv Bypass 2RHS*MOV80B. Stm Line B Isol Vly Bypass 2RHS*MOV67A RHS A Sht. On Clg Cv Bypass 2RHS&OV678 RHS 8 Sht. Dn Clg Cv Bypass C. ELECTRICAL
- 1. Two 'physically independent circuits between the Offsite Transmission Network and the Onsite Class 1E Distribution system operable.
- 2. VERIFY Main Transformer (2MTX-XM1A, 1B and 1C), Reserve Station Service Transformer (2RTX-SXR1A, 1B) and Auxiliary Boiler Transformer (2ABS-Xl) are operable.
N2- I OP-101A Dune, 1985
Ini ti al s/Date
- 3. A separate fuel storage tank for each Diesel Generator containing a minimum of:
- a. 50,148 gal. each of fuel for EDG-1 (Div. I) and EDG-3 (Div. II),
- b. 33,879 gal. of fuel for EDG-2 (Div. III).
(T.S.3.8.1.1.b) 4, The following Power Distribution System division shall be energized the breakers open between redundant busses within the unit fn accordance with Technical Specification 3.8.3.1,
- a. Standby & Emergency A.C. Power Ofstrfbutfon System.
- b. Emergency O.C. Power Distribution System.
- 5. The onsite normal electrical distribution system fs energized.
- 6. The Normal D.C. Distribution fs available.
- 7. The 24 volt O.C. Distribution System fs available.
- 8. The EDG-1 (Dfv. I) Diesel Generator fs operable.
- 9. The EDG-3 (Div. II) Diesel Generator is operable.
'I
- 10. The EDG-2 (Oiv. III) Diesel Generator is operable.
D. CONDENSER ANO FEEDWATER SYSTEM
- l. Circulating Water System and cooling tower are operating fn accordance with N2-IOP-10A, 10B, 10C wi th a minimum of one cfr culating wate~ pump per condenser section running, bypassing the cooling tower.
- 2. Main condenser vacuum breakers closed (2CEC*PNL851).
N2-IOP-101A tune, 1985
- 3. Mechanical vacuum pump available,
- 4. Off Gas System available.
- 5. Steam packing exhauster available.
- 6. The Condensate/Feedwater System in Long Cycle Cleanup, unless it is being used for make-up to the reactor.
- 7. Condenser hotwell level is approximately 256 feet, with hotwell level control in auto, and makeup and re$ ect stations in service.
- 8. Feedwater Heater Drain System available.
1 S
- 9. VERIFY that the heater drain tank level controllers on 2CEC*PNL851 are in MANUAL ana the level control valves (LV24A, B, C) are CLOSED.
E. TURBINE GENERATOR AND AUXILIARIES
- 2. Generator hydrogen coolers in operation.
- 3. Main Steam aux. supply steam stop valves are open 2MSS-MOV19A and MOV19B.
- 4. Moisture Separator Reheater System ready for operation.
5, Main Turbine Bypass Valves closed (2CEC*PNL851), unless they are being used for pressure control.
N2-IOP-101A June, 1985
Ini ti a l s/Date
- 6. Bypass opening Jack at 0% (closed) (2CEC*PNL851).
7, Chest Marming selector set in OFF position (2CEC*PNL851).
8, Main Stop Valves, Control Valves, and Combined Intermediate Valves closed (2CEC*PNL851).
9, ALL VALVES CLOSED speed set selected (2CEC*PNL851).
- 10. Pressure Regulator setpoint at 150 psig (2CEC*PNL851 ), unless it is controlling reactor reassure through the bypass valves and the main condenser.
ll. Maximum Combined Flow limit potentiometer set at 105% (2CEC*PNL851).
- 12. Load Limit set at 110% (2CEC*PNL851).
- 13. Load Set selector at O'X (2CEC*PNL851).
- 14. Turbine Supervisory Instruments in service.
r
- 15. Isolated Phase Bus Duct Coolers operable (N2-IOP-24).
- 16. Steam Seal System operable (N2-IOP-25).
F. INSTRUMENTATION 5 CONTROL Verify the following unit surveillances are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of'tartup unless completed within the previous seven days (T.S.4.3.7.6, 4.3.1.1, 4.1.4.3 and 4.3.6):
- a. SRM channel functional test (N2-ISP-NMS-M%10).
N2- IOP-101A J une, 1985
Jnl I4I 5/Udl 0 L
- b. IRM flux high channel functional test (N2-IOP-NMS-MM01).
- c. APRM channel functional test (N2-ISP-NMS-W9003).
- d. RBM channel functional (TS 4.1.4.3).
- e. Rx recirc flow channel functional test (N2-ISP-NMS-W8006).
- f. SRM detector no full in rod block functional test (N2-ISP-NMS-WK09) .
- g. IRM detector not full in rod block functional test (N2-ISP-NMS>>WN09).
- 2. List any bypassed SRM channels, and enter these channels in the CSO log.
/
- 3. List any bypassed IRM channels, and enter these channels in the CSO log.
- 4. List any bypassed APRM or flow unit channels, and enter these channels in the CSO log.
- 5. Process Computer available.
- 6. Reactor Pressure Vessel metal temperature recorder on panel 2CKC-PNL614 in service and inking properly.
- 7. Recirculation Pump suction temperature recorder B33-R650 or control room panel 2CEC*PNL602 in service and inking properly.
- 8. The Nuclear Boiler Instrumentation System in operation.
N2-IOP-101A June, 1985
Ini ti al s/Oate 10, The Reactor Protection System (RPS) is operable.
ll, The Rod Block Monitoring System operable.
- 12. The Traversing In-Core Probe System operable.
- 13. Reactor Manual Control and Rod Position Indication System operable.
14, Feedwater Control System available.
G. FINAL CHECKS
- 1. CHECK the Equipment Status Log to make sure no equipment is inoperative which would prevent plant startup.
3, Locked valve and open breaker checklist verified current.
NOTE: Reactor Coolant System Chemistry analysis results must be 'from within the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (T,S.4.4.4 and T.S.4.4,5). Gross activity determination should be made from the most recent. data.
- 4. Obtain current rad/chem daily chemistry analysis results of the Reactor Coolant System. Verify the following:
- a. Conductivity less than or equal to 2 umhos (T.S.4.4.4).
- b. Chlorides less than or equal to O.l ppm (T.S.4.4.4).
- c. pH between 5.6 and 8.6 (T.S.4.4.4).
N2-IOP-101 A June, 1985
- d. Gross activity less than or equal to 0.2 uci/gram DOSE EQUIVALENT I-131 (T.S.4.4.5),
- e. Name of person/time contacted (for items a-d above):
- 5. Reactor Protection System reset.
Channel A Channel B
- 6. Check the SSS Jumper Log to verify that all unnecessary jumpers and blocks have been removed and required equipment is not out of service.
- 7. REVIEM the Mark Ups for entries which may adversely impact unit startup.
- 8. VERIFY technical specification surveillances are curr ent.
ion 1 Supervisor (SSS)
List all inoperable or out of service control rods and accumulators.
- 10. Note any items on this checklist that could not be satisfied:
N2-IOP-101A June, 1985
ATTACHMENT II N2-IOP-101A NINE MILE POINT - UNIT 2 MODIFIED MASTER SDNTUFl'HECKLIST Date/Time Completed Reviewed By Approved By Items that cannot be satisfied shall be noted in Section H.10. Entry into Operational Condition Two (Startup) shall not be made unless the surveillance requirements associated with the Limiting Conditions for Operation have been performed within the required surveillance interval (T.S.4.0.4).
For a system/component to be considered as operable, the following must have been accomplished:
A. The required mechanical/electrical checklists have been performed on that applicable system/subsystem/component.
l B. All peri odi c survei llances sati sfactorily completed on that system/subsystem/component.
The MODIFIED MASTER STARTUP CHECKLIST shall be conducted by a licensed Reactor Operator and Reviewed by the Chief Shift Operator. The Shift Supervisor shall approve the checklist after he has verified that items in H.10 are not required for startup.
The conditions are satisfied as listed in A through H.
- the MODIFIED MASTER STARTUP CHECKLIST i s used in those circumstances greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, following an outage, where the work done involved only a few specific items.
N2-IOP-1 01 A June, 1985
k/II \ IQI4/UQO'C
~ ~
~
~
A. CONTAINMENT
- 1. Primary Containment established and integrity maintained.
- 2. Reactoi Building established, integrity maintained (T.S.3,6,5.1).
Reactor Building inter nal pressure less than or equal to -0.25 inches or water gauge (T.S.4.6.5.1.a.).
- 3. Suppression Pool operable with water between 24'l 5/8" and 23' 3/8" and the temperature is less than 90 oF'T.S.3.6.2.1).
- 4. All Drywell-Suppression Pool Vacuum Breakers are fully closed as per position indication at 2CEC*PNL628 (T.S.3.6.4).
- 5. Drywell Cooling System is in operation.
- 6. Both Standby Gas Treatment Trains ready for operation.
- 7. Primary Containment average air temperature is not greater than 150oF (T.S.3.6.1.7)..
- 8. Primary Containment internal pressure is between -0.5 and +.75 psig (T.S.3.6.1.6, N2-OSP;LOG-S001).
- 9. Containment Atmosphere Monitoring System is in operation.
ll. Automatic Depressurization System and Drywell Pneumatic Systems lined up with Nitrogen.
- 12. Primary Containment Vent Purge and Nitrogen System operable.
N2-IOP-101A June, 1985
- 14. Drywell Equipment and Floor Drains System available.
- 15. Containment Leakage Monitoring System available.
B. REACTOR AND EMERGENCY SYSTEMS
- 1. Shutdown Margin has been determined to be adequate per T.S.3.1.1.
- 2. Reactor Vessel temperature greater than the minimum shown on T.S.
.Figure 3.4.6-1C (T.S.4.4.6.1.1).
- 3. Estimated critical rod position has been received from the Reactor Analyst Department.
- 4. Low Pressure Core Spray system (CSL) operable.
\
- 5. Three independent Low Pressure Coolant Injection subsystems op'erable.
- 6. High Pressure Core Spray system (CSH) operable.
- 7. All Control Rod Drives are operable (T.S. 3//4.1.3).
- 8. Reactor Core Isolation Cooling system (ICS) available.
9.
~ All 18 Safety/Relief Valves (SRV) operable.
~
N2- IOP-101A June, 1985
- 10. Contro1 Rod Drive Housing Support fs fn place (T.S.3.1.3.8).
- 11. Contr ol Rod Drf ve Hydraul fc System fn operatfon. CHECK for CRD Accumulator Inop 1 f ghts.
- 12. Reactor Coolant 4 ECCS Leak Detection System fn operation.
- 13. Reactor Recfrculatfon Pumps fn operation fn slow speed fn accordance wf th N2-IOP-29.
- 14. Reactor Mater Cleanup System fn operation and capable of regectfng water to Radwaste facflftfes or to the main condenser fn accordance
. wf th N2-IOP-37.
- 15. Nafn Steam System available.
- 16. Standby Lfqufd Control System operable.
- 17. Redundant Reactivity Control System operable.
- 18. Verify that the following valves are shut and that thefr associated motor breakers are open and that both valves and breakers are marked up and tagged out. (This fs an N+2 Appendix R Commftment.)
Valve 8 Description 868%%32A lOPfTYB. c d. M d tCE 2RHS*HOV32B RHS B Stm. Cond. Node to ICS 2RHSAOV37A Ht. Exch A Drain to Sup Pool 2RHS~Y37B Ht. Exch B Drain to Sup Pool 2DERAOV129 DER Drain Isol Vlv from Reacto~ btm Hd 2RHS~V22A RHS "A" Stm Line Isol 2RHS~V22B RHS "B" Stm Line Isol 2RHS~VBOA Stm Line A Isol Vlv Bypass 2RHSWOY808 Stm Line B Isol Vlv Bypass 2RHS*MOV67A RHS A Sht. Dn Clg Cv Bypass 2RHS*HOY67B RHS B Sht. Dn Clg Cv Bypass E
N2- I OP-101A June, 1985
Initials/Date C. ELECTRICAL
- l. Two physically independent circuits between the Offsfte Tr ansmfssfon Network and the Onsfte Class lE Distribution System operable
- 2. VERIFY Main Transformer (2MTX-XM1A, 1B and 1C), Reserve Station Service Transformers (2RTX-XSR1A, 1B), and Auxiliary Bofler Transformer (2ABS-Xl) are available.
- 3. A separate fuel storage tank for each Diesel Generator containing a minimum of:
- a. 50,148 gal. each of fuel for EDG-1 (Ofv. I) and EDG-3 (Ofv. II).
- b. 33,879 gal. of fuel for EDG-2 (Div, III),
(T.S.3.8.1.1.b)
- 4. The following Power Distribution System division shall be energized with the breakers open between redundant busses within the unit fn accordance wf th Technical Speci ffcatfons 3.8.3.1,
- a. Standby 8 Emergency A.C. Power Distribution System.
- b. Emergency O.C. Power Distribution System.
5, The onsfte normal electrical distribution system fs energized.
- 6. The Normal O.C. Distribution fs available.
- 7. The 24 volt D.C. Distribution System fs available.
- 8. The EDG-1 (Div. I) Diesel Generator fs operable.
- 9. The EOG-3 (Oiv. II) Diesel Generator operable.
- 10. The EDG-2 (Div. III) Diesel Generator fs operable.
N2- IOP-101A Qune, 1985
- 0. OUTAGE MORK COMPLETION System System Title
- a. Mork Request Number Complete
- b. System Mar'k Up Cleared.
- c. Valve and/or electrical ltne up complete per N2-lOP-
- d. System Surveillances Complete, if required.
- 2. System System Title
- a. Mork Request Number Complete
- b. System Mark Up Cleared.
- c. Valve and/or electrical line up complete per N2-lOP-
- d. System Surveillances Complete, if required.
A NOTE: Use additional sheets as required.
~ N2- I OP -101A June, 1985
E. CONDENSER AND FEEDMATER SYSTEM
- 1. Circulating Mater System and cooling tower are operating in accordance wi th N2-IOP-10A, -10B, -10C wi th a minimum of one circulating water pump per condenser section running, bypassing the cooliof, tower.
- 2. Main condenser vacuum breakers closed (2CEC*PNL851).
- 3. ,Mechanical vacuum pump available.
- 4. Off Gas System available.
- 5. Steam packing exhauster available.
- 6. Condensate Demineralizer System available.
- 7. Condenser hotwell level is approximately 256 feet, with the hotwell level control in auto, and makeup and re3ect stations in service.
- 8. Condensate/Feedwater Systems in long cycle clean up.
- 9. The extraction Steam/Heater Drain System in operation.
- 10. Feedwater Heater Drain System available.
- 11. YERIFY that the heater drain tank level controllers on 2CEC*PNL851 are in MANUAL and the level control valves (LY24A, B, C) are CLOSED.'.
TURBINE GENERATOR AND AUXILIARIES
- 1. Low Pressure Turbine Hood Sprays are available and in auto.
- 2. Generator Hydrogen Seal Oil System in operation.
N2-IOP-101A June, 1985
z t>> iia>>I ua vc
- 3. Generator filled with hydrogen; pressure 75 psig and purity greater than 94%,
/
/
- 5. Generator Hydrogen Coolers in operation.
- 6. Generator Exciter available.
- 7. Generator Core Monitor available.
- 8. Main Turbine oil tank level normal; vapor extractor, and continuous oil filter operating.
- 9. Turning Gear Oil Pump, Motor Suction Oil Pump and Bearing Lift 'fail Pumps operating.
- 10. Tur bine on Turning Gear.
ll. Emergency bearing oil pump (pump 5) available with control switch in AUTO (.2CEC*PNL851).
- 12. Main Steam aux. supply steam stop valves are open 2MSS-MOY19A and MOY198.
- 13. Moisture Separator Reheater System ready for operation.
- 14. EHC System operating.
- 15. Main turbine bypass valves closed (2CEC*PNL851).
NZ-IOP-101A June, 1985
Initials/Date
- 16. 8ypass opening Jack at 0% (closed) (2CEC*PNL851 ).
- 17. Chest warmfng selector set fn OFF posftfon (2CEC*PNL851).
- 18. Nafn stop valves, control valves, and combined intermediate valves closed (2CEC*PNL851).
- 19. ALL VALVES CLOSED speed set selected (2CEC+PNL851).
- 20. Pressure regulator setpofnt at 150 psfg (2CECAPNL851).
- 21. Maximum Combined Flow limit potentiometer set at 105% (2CEC*PNL851).
- 22. Load Limit set at ll0%. (2CEC*PNL851).
- 23. Load Set selector at 0% (2CEC*PNL851).
- 24. Turbine Supervisory Instruments fn servfce.
- 25. Isolated Phase Hus Duct Coolers available.
- 26. Cooling fans on main and auxiliary transformers fn service.
- 27. Steam Seal System available.
G. INSTRUMENTATION & CONTROL
- l. Yerffy the followfng unit surveillances are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of startup unless completed within the previous seven days (T.S.4.3.7.6, 4.3.1.1, 4.1.4.3 and 4.3.6):
N2- IOP-101A J une, 1985
i
- a. SRM channel functional test (N2-ISP-HMS-HX10).
- b. IRM flux high channel functional test (N2-IOP-F>S-WXOl).
- c. APRM channel functional test (N2-ISP-NMS-WK03),
- d. RBM channel functional (TS 4.1.4.3).
- e. Rx r ecirc flow channel functional test (N2-ISP-NMS-W8006).
- f. SRM detector not full in rod block functional test (N2- I SP-NMS-N009) .
- g. IRM detector not full in rod bl ock functi onal test (N2- I SP-HMS-WM09) .
- 2. Radiation Monitoring Systems in operation.
- 3. List any bypassed SRM channels, and enter these channels in the CSO log.
4.. List any bypassed IRM channels, and enter these channels in the CSO log.
- 5. List any bypassed APRM or flow unit channels, and enter these, channels in the CSO log.
- 6. Process Computer available.
- 7. Reactor Pressure Vessel metal temperature recorder on panel 2CEC-PNL614 in service and inking properly.
- 8. Recirculation Pump suction temperature recorder B33-R650 or control room panel 2CEC*PNL602 in service and inking properly.
N2-IOP-101A June, 1985
9, The Nuclear Bofler Instrumentation System fn operation.
- 10. Remote shutdown operable.
ll, The Reactor Pr otectfon System (RPS) fs operable.
- 12. Yfbratfon and loose parts monitoring system operable.
- 13. Seismic monitoring operable.
14, The Rod Block Monitoring System operable.
- 15. The Traversfng In-Core Probe System operable.
16, The Rod Sequence Control System (RSCS) operable.
- 17. Reactor Manual Control and Rod Position Indication System operable.
- 18. Feedwater Control System available.
H. FINAL CHECKS
- 1. CHECK the Equi pment Status Log to .make sure no equi pment is inoperative which would prevent plant startup.
- 3. Locked valve and open breaker checklist verified current.
N2- IOP-101A June, 1985
NOTE : Reactor Coolant System chemistry analysis results must be from within the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (T.S.4.4.4 and T.S.4.4.5).
Gross activity determination should be made from the most recent data.
- 4. Obtain-. current rad/chem daily chemistry analysis results of the Reactor Coolant System. Verify the following:
- a. Conductivity less than or equal to 2 umhos (T.S.4.4.4).
b, Chlorides less than or equal to 0.1 ppm (T.S.4.4.4).
- c. pH between 5.6 and 8.6 (T.S.4.4.4).
- d. Gross activity less than or equal to 0.2 uci/gram OOSE EQUIVALENT I-131 (T.S.4.4.5).
- e. Name of person/time contacted (for items a-d above):
- 5. Reactor Protection System reset.
Channel A Channel B
- 6. Check the SSS Jumper Log to verify that all unnecessary jumpers and blocks have been removed and required equipment fs not out of service.
- 7. REVIEM the Mark Ups for entries which may adversely impact on unit startup.
H
/
- 8. VERIFY that all Technical Specification surveillances are current.
tati on 1 Supervisor (SSS)
N2-IOP-101A June, 1985
- 9. List all inoperable or out of service control rods and accumulators,
- 10. Note any items on this checklist that could not be~ satisfied:
N2- IOP-101A June, 1985
ATTACHMENT III N2- I OP-101 A NINE MILE POINT - UNIT 2 MASTER STARTUP CHECKLIST Oate/Time Completed Reviewed By Approved By Items that cannot be satisfied shall be noted in Section H.13. Entry into Operational Condition Two (Startup) shall not be made unless the surveillance requirements associated with the Limiting Conditions for Operation have been performed within the required surveillance interval (T.S.4.0.4).
For a system/component to be considered as operable, the following must have been accomplished:
A. The required valve/electrical lineups have been performed on that applicable system/subsystem/component.
B. All periodic sur veill ances satisfactorily compl eted on that system/subsystem/component.
The MASTER STARTUP CHECKLIST shall be conducted/verified by a licensed Reactor Operator and Reviewed by the Chief Shift Operator. The Shift Supervisor shall approve the checklist after he has verified that items in H.13 are not required for startup.
The conditions are satisfied as listed in A through H.
N2- IOP-101A June, 1985
Ini ti al s/Date A. CONTAINMENT Prfmat y Containment established and integri ty maintained (T.S.3.6.1. I, N2-QSP-CNT-R8001, N2-OSP-CNT-8001) N2-0SP-CNT-SA8003, N2-0SP-CNT-8006),
- 2. Reactor Buil ding established, integrity maintained (T.S.3.6.5.1).
Reactor Building internal pressure less than or equal to -0.25 inches or water gauge (T.S.4.6.5.l.a,),
- 3. Suppression Pool operable wfth water between 24'l 5/8" and 23' 3/8" and the temperature is less than 90 oF (T,S.3.6,2.1).
- 4. All Drywell-Suppression Pool Vacuum Breakers are fully closed as 'per positfon indication at 2CEC*PNL628 (T.S.3.6,4).
- 5. Drywell Cooling System is.in operation. The drywell temperature will dictate the exact number and combination of drywell coolers (1 in the head, 1-4 general area, 1-4 recfrculatfng) operating at any gfven time. The reactor vessel flange and head f1ange metal temperatures must be greater than 70oF when the head bolts are tensioned (T.S.
3/4.4.6) and the .drywell average temperature limit (T.S. 3/4.6.1.7) is 150oF
- 6. Both Standby Gas Treatment Trains ready for operation per N2-IOP-61B.
7, Yerffy all high radiation area doors are locked.
- 8. Dyrwell average afr temperature is not greater than 150oF (T.S.3.6.1.7).
- 9. Orywell 8 Suppression chamber internal pressure is between -0.5 and
+.75 psig (T.S.3.6.1.6, N2-0SP-LOG-SOOl).
V
- 10. Containment Atmosphere Monftoring System is in operation per N2- I OP-82.
/
N2- IOP-101A June, 1985
- 11. The Post LOCA Oxygen/Hydrogen Monf torf ng Trains in operatf on to monitor Containment oxygen .per N2-IOP-87,
/
- 12. Switch over the appropriate gas operated instrument hook ups fn the Drywell. Pneumatic Systems from instrument air to instrument nitrogen per N2-IOP-61A, Section E,1.3.
/
- 13. Place the Automatic Depressurfzatfon System fn the standby mode with nftrogen supplying the safety relfef valves per N2-IOP-34, Section E.l through E.8.
/
- 14. Pr imary Contaf nment Vent Pur ge and Nf trogen System lined up per N2- I OP-61A.
/
/
- 16. Drywell Equipment and Floor Drains System fn operation per N2-IOP-67.
- 17. Containment Leakage Monitoring System lined up per N2-IOP-81.
- 18. Primary Containment Isolation System operable per N2-IOP-83.
- 19. Primary Containment System Instrument Penetration Valve Lineup verfffed per Table 1, N2-IOP-99.
20, Verify that the Under Vessel Service Platform is: 1) platform is properly oriented for SRM/IRM withdrawal; 2) SRM/IRM detector cables are started through their respective platform openings; and 3) platform fs pinned and locked in place.
B. REACTOR AND EMERGENCY SYSTEMS
- 1. Shutdown Margin has been determined to be adequate per T.S.3.1.1.
/
N2- IOP-101 A June, 1985
ln~ t~ a i square Reactor Vessel temperature greater than the minimum shown on T.S.
Figure 3.4.6-1C (T.S.4.4.6.1.1).
3, Reactor Pressure Vessel head in place and bolts tensioned.
- 4. Estimated critical rod position has been received from the Reactor Analyist Department.
- 5. Low Pressure Core Spray system (CSL) operable and lined up in accordance with N2-IOP-32 (T.S. 3/4.5.1).
- 6. Three independent Low Pressure Coolant In)ection subsystems operable and lined up in accordance with N2-IOP-31 (T.S. 3/4.5.1).
- 7. High Pressure Core Spray system (CSH) operable and lined up in accordance with N2-IOP-33 (T.S. 3/4..5.1).
- 8. All Control Rod Drives are oper able (T.S. 3.1.3.1).
- 9. Reactor Core Isolation Cooling system (ICS) is available in accordance with N2-IOP-35 (T.S. 3/4.7.4).
3/4.4.2).
/
- 11. Control Rod coupling integrity has been demonstrated (T.S.3.1.3.6, N2-OSP-RMCN02) .
- 12. Control Rod Drive Housing Support is in place (T.S.3.1.3.8).
E
/
- 13. Control Rod Drive Hydraulic System in operation in accordance with N2-IOP-30. CHECK for CRD Accumulator Inop lights.
N2- IOP-101A June, 1985
- 15. Reactor- Recirculation Pumps in operation in slow speed in accordance with N2-IOP-29 (T.S. 3/4.4.1).
- 16. Reactor Mater Cleanup System in operation and capable of rejecting water to Radwaste facilities or to the main condenser in accordance wi th N2- IOP-37.
- 17. Drain the Main Steam Lines per N2-IOP-l.
- 18. Main Steam System lined up per N2-IOP-1.
- 19. Standby Liquid Control System is operable per N2-IOP-36A (T.S.
3/4.1.5).
- 20. Redundant Reactivity Control System operable per N2-IOP-368.
- 21. Veri fy that the foll owing val ves are shut and verify that their associated Motor Breakers are open and that both valves and breakers are marked up and tagged out. (This is a NMP-2 Appendix R Conmittment.)
VALVE 8 DESCRIPTION 2RHS*MOV32A RHSA Stm Cond Mode to ICS 2RHS*MOV328 RHSB Stm Cond Mode to ICS 2RHS MOV37A Ht Exch A Drain to Sup pool 2RHS~V378 Ht Exch 8 Drain to Sup pool 2DER+MOY129 DER Drain Isol Ylv. from btm head 2RHS MOV22A RHS "A" Steam Line Isolation 2RHS"MOY228 RHS "8" Steam Line Is lation 2RHS*MOVBOA Stm. Line A Isol. Vlv Bypass 2RHS*MOVSOB Stm. Line 8 Isol. Ylv Bypass 2RHS"MOV67A RHS A Sht Dn Clg CV Bypass 2RHS*MOV678 RHS 8 Sht Dn Clg CV Bypass N2-IOP-101A June, 1985
Ini ti a I s/Date C ELECTRICAL
- l. Two physically fndependent circuits Network and the Onsfte Class between the Offsfte Transmission lE Distribution System operable per N2-OSP-LOG-'%01 (T.S.3.8.1.l.a). (Lined up according to N2-IOP-70:)
- 2. VERIFY . Main Transformer (2MTX-XMlA, 1B and 1C), Reserve Station Service Transformers (2RTX-XSRlA, 1B) and Auxiliary Boiler Transformers (2ABS-Xl) are lined up for normal operation per applicable pot tions of N2-IOP-68 and N2-IOP-70.
/
- 3. The followfng Po~er Distribution System division shall be energized with the breakers open between redundant busses within the unit in accordance wf th Technf cal Specf ffcatf ons 3.8.3.1 as per N2-0SP-LOG-M001,
- a. Standby and Emergency A.C. Power Distribution (lfned up
.according to N2-IOP-72).
- 1) Division I, consisting of:
a) 4160 Volt A.C. Bus.
b) 600 Volt A.C. Load Center/MCC's/Dfst. Panels.
c) 240/120 volt and 208/120 volt A.C. distribution panels.
- 2) Dfvfsfon II consisting of:
a) 4160 Volt Bus A.C. Bus.
b) 600 Volt A.C. Load Center/MCC's/Dist. Panels, c) 240/120 volt and 208/120 volt A.C. distribution panels.
- 3) Division III consisting of:
a) 4160 Volt Bus A.C. Bus.
b) 600 Volt A.C. MCC's.
c) 240/120 volt and 208/120 volt A.C. distribution panels.
- b. Emergency D.C. Power Distribution (TS 3/4.8,2, lined up according to N2-IOP-74A and N2-IOP-74B).
- 1) Division I consisting of 125 Volt D.C. swftchgear, MCC and associated distribution panels. Panels 2BYS*PNL-201A, 2BYS*PNL-202A, 2BYS*PNL-204A.
- 2) Division II consisting of 125 Volt D.C. swi tchgear, MCC and associated distribution panels. Panels 2BYS*PNL-201B, 2BYS~PNL-202B, 2BYS*PNL>>204B.
3.) Division III consisting of 125 Volt D.C. switchgear, MCC and associated distribution panels. Panels 2BYS*PNL-201C, 2BYS"PNL-202C, 2BYS*PNL-204C.
N2- IOP-101A June, 1985
Jll I IrlOI 4/ VC'4C
- 4. The onsf te normal el ectr ical dfstrfbutf on system is energf zed and lined up fn accordance with N2-IOp-71 followfng a Refuelfng Outage.
/
5, The Normal D.C. Distribution fs energized,and lined up in accordance with N2-IOP-73A followfng a Refuelfng Outage.
6, The 24 volt D.C. Dfstrfbutfon System fs energized and lined up per N2- I OP-73C.
- 7. The EDG 1 (Div. I ) Df esel Generator fs 1 f ned up for standby operation. This includes the following:
- a. EDG-1 (Div. I) Cooling Mater System fs lined up per N2-IOP-11 following a Refuel fng Outage.
- b. EDG-1 (Dfv. I) Ventilation System fs lined up per N2-IOP-57.
'/
- c. EDG-1 (Df v, I) Auxf1 f aries are 1 fned up per N2- I following Refueling Outage.
OP-100.0'.
a The Diesel Oil Transfer system fs lined up per NZ-IOP-100.0 following a Refueling Outage.
- e. EDG-1 (Div. I) f s 1 fned up per N2-IOP-100.0 for standby operation..
/
- f. 50,148 gal. of fuel fn the EDG-1 (DIV. I) storage tank.
8, The EDG-3 (Dfv. I I) Df esel Generator is lined up for standby operation. This includes the following:
- a. EDG-3 (Div. II) Cooling Water System is lined up per N2-IOP-11 following a Refueling Outage.
/
b.'DG-3 (Div. II) Ventilation System fs lined up per N2-IOP-57.
N2- IOP-101A June, 1985
Ini u a s/Oate I
- c. EDG-3 'Div. II) Auxiliaries are lined up per N2-I OP-100.0 following a Refueling Outage.
- d. The Diesel Oil Transfer system is lined up per N2- I OP-100. 0 following a Refueling Outage.
- e. EDG-3 (Div. II) is lined up per N2-IOP-100,0 for standby operation.
- f. 50,148 gal. of fuel in the EDG-3 (DIV. !I) storage tank.
- 9. The EDG-2 (Div. III) Di esel Generator is 1 ined up for standby operation. This includes the following:
- a. EDG-2 (Div. III) Cooling Mater System is lined up per N2-IOP-11 following a Refueling Outage.
- b. EDG-2 (Div. III) Ventilation System is lined up per N2-IOP-57.
- c. EDG 2 (Div. III) Auxiliaries are lined up per N2 IOP 100 1 following a Refueling Outage.
- d. The Diesel Oil Transfer system is lined up per N2-IOP-100.1 following a Refueling Outage.
- e. EDG-2 (Div, III) is lined up per NZ-IOP-100.1 for standby operation.
- f. 33,879 gal. of fuel in the EDG-2 (DIY. III) storage tank.
- 10. Station Lighting System in operation per N2-IOP-75.
ll. Plant Conmunications System in operation per N2-IOP-76.
N2- IOP-101A June, 1985 1
Lnl Cldl 5/UdCH
- 12. VERIFY that the Main Generator and Netering Potent)al Transformer fuses are INSTALLED and the cabinet doors are lOCKED.
D. CONDENSER AND FEEDMATER SYSTEM
- 1. Mafn Condenser manways closed.
- 2. Circulating Mater System and cooling tower are operating fn accordance with N2-IOP-10A, 10B, 10C with a minimum of one circulating water pump per condenser section running, bypassing the cooling tower.
/
- 3. Nafn Condenser Vacuum Breakers closed (2CEC*PNL851).
- 4. Mechanical vacuum pump avaflable per N2-IOP-9.
- 5. Off Gas System lined up per N2-IOP-42.
- 6. Steam Packing Exhauster available per N2-IOP-25.
- 7. Condensate Demineralizer System avaflab]e per N2-IOP-5.
- 8. .Condensate Demineralfzer Resin Regeneration System available'er N2- I OP-5.
At 1 east orie condensate pump running wi th fl ow (500-3000 gpm/demfneralfzer) established through one or two condensate demfneralfzers.
The purpose of the Condensate Fl ow Path is to establ fsh feedwater qual i ty in accordance wf th N2-CSP-2. Chemi cal Analysis and Corrective Action are to be taken prior to startup.
- 9. Condenser hotwell level is approximately 256 feet, with the hotwell level control in auto, and makeup and reject stations in service in accordance wi th N2-IOP-4.
N2-IOP-101A June, 1985
Initiais/Oate lO. Start up the Condensate/Feedwater System(s) in short Cycle or Long Cycle cleanup per N2-IOP-3.
ll. place the Extraction Steam/Heater N2-IOP-8, E.l.l. through E.l.4.
Drain System in operation per 12, Condensate Booster pumps and aux oil pumps available.
- a. 2CNM-P2A and aux oil pump 2A
- b. 2CNM-P2B and aux oil pump 2B
- c. 2CNM-P2C and aux oil pump 2C
- 13. Reactor feedwater pumps and aux oil pumps available.
- a. 2FWS-PlA and aux oil pump 2A.
- b. 2FWS-PlB and aux oil pump 2B.
- c. 2FWS-PlC and aux oil pump 2C.
- 14. VERIFY that the heater drain tank level controllers on 2CEC*PNL851 are in MANUAL and the level control valves (LY24A, B, C) are CLOSED.
E.~ TURBINE GENERATOR AND AUXILIARIES l.~ Low Pressure Turbine Hood Sprays are available and in auto.
- 2. Generator Hydrogen Seal Oil System in operation in accordance with N2-IOP-22D.
3, Generator filled with hydrogen; pressure 75 psig and purity greater than 94'X (N2-IOP-27).
- 5. Generator Hydrogen Coolers in operation in accordance with N2-IOP-14.
N2- IOP-101A June, 1985 .
- 6. Generator Exciter available.
- 7. Generator Core'onitor has power and is available (N2-IOP-27).
8, Main Turbine oil tank level normal; vapor extractor operating, and continuous oil filtet operating in accordance with N2-IOP-228.
- 9. Turning Gear Oil Pump, Motor Suction Oil Pump and Bearing Lift Oil Pumps operating in accordance with N2-IOP-22A.
- 10. Turbine on Turning Gear in accordance with NZ-IOP-21.
11, Emergency Bearing Oil Pump (pump 5) available with control switch in AUTO (2CEC*PNL851).
- 12. The following startup drains are open:
a; Steam lead drain valves for high pressure turbine.
- 1) Low point drains - 2MSS AOV87A, B, C, 0.
- 2) Low point drain ori fice bypass val ves 2MSS-AOY85A, B, C, 0 which are to be closed as soon as the low point drain pot level alarm clears.
- 3) 48" steam header drains, 2MSS-AOVBBA, B.
- 4) Lines down stream of 48" header, 2MSS-AOV209.
- b. Main Stop Yalve before seat drains (2MSS-MOV21A, B, C. 0).
- c. Combined control valve before seat drains (2MSS-MCV147).
- d. High pressure Turbine Steam lead drain valves down stream of the control valves: 2MS'MOY10A, 10B, 10C, 100).
- e. High oressure exhaust. lines to MSR's ElA and ElB drains (2MS: 'V9A, AOV180, V5 for ElA MSR; 2MSS-MOV199, MOV9B for ElB MSR)
N2- I OP-101A June, 1985
In] t] al s/Date NOTE: VERIFY the fol 1 owing heater extract] on steam stop val ves c1osed: 2ESS-MOY34A, 8, C for 2FMS-E6A, B, C heater s, 2ESS-MOV28A, B, C for 2FMS-E5A, B, C heaters, 2ESS-MOV23A, B, C for 2FMS-E4A, B, C heaters, 2ESS-MOY15A, B, C for 2FMS-E3A, B, C heaters.
- f. Moisture separator/reheaters A and B extraction steam first stage drains . 2DSR-MOV86A, B and 2DSR-AOV82A, B, and main steam second stage drains OPEN 2CRS-MOV18A, B and 2CRS-MOY9A, B and 2CRS-MOV8A, B and 2CRS-MOV7A, B.
- g. Moisture separator/reheaters A and B drain rece]ver drains OPEN, 2DSM-LY76A for 20SM-TK-4A, 2DSM-LY768 for 20SM-TK-48.
- 13. Main Steam aux. supply steam stop valves are open 2MSS-MOY19A and MOV19B.
- 14. Moisture Separator Reheater System lined up per N2-IOP-2.
- 15. EHC system operating in accordance with N2-IOP-23.
- 16. Main turbine bypass valves closed (2CEC*PNL851).
- 17. Bypass opening jack at O'X (closed) (2CEC*PNL851).
- 18. Chest warming selector set in OFF position (2CEC*PNL851).
- 19. Ma]n stop valves, control valves, and comb]ned intermediate valves closed (2CEC*PNL851).
- 20. ALL VALVES CLOSED speed set selected (2CEC*PNL851).
- 21. Pressure regulator setpoint at 150 psig (2CECAPNL851).
N2- IOP-101A June, 1985
- 22. Maximum" Combined Flow limit potentiometer set at 10M (2CEC*PNL851 ),
- 23. Load Limit set at 110%. (2CEC*PNL851).
- 24. Load Set selector at 0% (2CEC*PNL851 ),
- 25. Turbine Supervisory Instruments in service.
- 26. Isolated Phase Bus Duct Coolers available (N2-IOP-24).
- 27. Cooling fans on the main and auxiliary transformers available.
- 28. Steam Seal System available (N2-IOP-25).
'E F.~ STATION AUXILIARY SYSTEMS ating in accordance wi th N2-IOP-ll
~
1.~ Service Water System oper (T.S.
3/4.7.1).
~ ~ ~
- 2. Travel ing Mater Screens and Mash Di sposal System in oper ati on, N2- I OP-12.
- 3. CCS System operating in accordance with N2-IOP-14.
- 4. CCP System operating in accordance with N2-IOP-13.
- 5. Process Sampling System in operation in accordance with H2-IOP-17.
N2- IOP-101A June, 1985
<n> nay.siuate
- 6. Instrument/Service Afr System operating fn accordance wf th N2-IOP-19.
- 7. Breathfng Afr fn operatfon per k2-IOP-20.
- 8. Ffre Protection System operating fn accordance with N2-IOP-43, -44,
-45, -46, -47 (T.S.: 3/4.7.7.1, 3/4,7.7.2, 3/4.7.7.3, 3/4.7.7,4, 3/4.7.7.5, 3/4.7.7.6).
- 9. The Spent Fuel Pool Cool f ng and Cleanup System avail abl e per N2-IOP-38.
- 10. Turbine Building Ventilation System operating fn accordance wf th N2-IOP-55.
- 11. Reactor Building Yentflatfon System operating fn accordance with N2- IOP-52.
/
- 12. Radwaste Buf1 df ng Ventil at) on System operating f n accordance wf th N2-I OP-56.
e
/
- 13. Control Room, Computer Room, Auxf1 f ary El ectrf c Equi pment Room Yentil atf on Systems, operating f n accordance wf th,N2-IOP-53A and N2- I OP-53B.
- 14. Control Building Special Filter Train System operable per N2-IOP-53A (T.S. 3/4.7.3).
- 15. Controlled Building Chilled Water System in operation per N2-IOP-53C and ventilation-chilled water (lithium bromide) in operation per k2-IOP-54B.
- 16. Liquid Radwaste System is available and capable of receiving water from reactor startup (N2-IOP-40).
N2-IOP-101A June, 1985
- 17. Solid Radwaste System available as required to support the Liquid Radwaste System per N2-IOP-41.
/
18.~ Standby- Swi tchgear/Battery Room Ventil ati on and Normal Swi tchgear Building Ventilation operating in accordance with N2-IOP-53K and N2-IOP-84A.~
- 19. Makeup Water Treatment and Makeup Water Storage and Transfer available per N2-IOP-15 and N2-IOP-16,
- 20. Auxiliary Boiler Steam System in operation to support reactor startup per N2>>I OP-48.
- 21. Hot Water and Glycol Heating System available per N2-IOP-49.
- 22. Screenwell Building & Fire Pump Room Ventilation System in operation per N2-I OP-58.
- 23. Control Building - Reactor Building electrical tunnels ventilation in operation per N2-IOP-59A.
- 24. Auxiliary Building South Air Conditioning/Carbon Oioxide Tank Room Ventilation System in operation per N2-IOP-59B.
- 25. Auxiliary Boiler Room Ventilation System in operation per NZ-IOP-59.C.1
- 26. Condensate Storage Tank Building Ventilation System in operation per N2-.0P-59.C.2.
- 27. Oemineralized Storage Tank Building Ventilation System in operation
.per N2-IOP-59.C.3.
N2-IOP-101A June, 1985
Ini ti al s/Date
- 28. Electrical Bay Ventilation and Screen House Ventilation Systems in operation per N2-IOP-59.C.4.
- 29. Chiller Building Ventilation System in oper ation per N2-IOP-59.C.5.
30, Service Building Ventilation System in operation per N2-IOP-59.C.6.
- 31. Reactor Building Drains in operation per N2-IOP-63.
- 32. Turbine Building Drains in operation per N2-IOP-64.
- 33. Radwaste Building Drains in operation per N2-IOP-65.
- 34. Miscellaneous Building Drains in operation per N2-IOP-66.
- 35. Reactor Building Crane stored per N2-IOP-84.
G. INSTRUMENTATION 5 CONTROL
- l. Verify the following. unit sur veillances are completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of startup unless completed within the previous 'even days (T.S.4.3.7.6, 4.3.1.1, 4.1.4.3 and 4.3.6):
- a. SRM channel functional test (N2-ISP-NHS-HM10).
- b. IRM flux high channel functional test (N2-IOP-NHS-W8001).
- c. APRM channel functional test (N2-ISP-NHS-WX03).
- d. RBM channel functional (TS 4.1.4.3).
- e. Rx recirc flow channel functional test (N2-ISP-NHS-WM06).
- f. SRH detector not full in rod block functional test (N2-ISP-NHS-W8009) .
- g. IRM detector not full in rod block functional test (N2- ISP-NHS-WK09) .
N2- IOP-101A J une, 1985
3.1.4.2).
NOTE: The following may be performed with any half scram surveillance.
- 3. Check SCRAM solenoid circuit continuity as'ollows:
- a. Scram reset with all rods in.
- b. Manually scram A channel and observe the full core display for blue scram lights. If a blue light is found, proceed to f.
below.
- c. Reset A channel.
- d. Manually Scram B channel and observe full core display for Blue Scram Lights. If a blue light is found, proceed to f. below.
- e. Reset B channel.
- f. Check 'nd replace fuses of any rods which scrammed. If rod scrammed and fuses are good, write a WR to check the scram solenoids.
- 4. Radiation Monitoring Systems operable per N2-IOP-79 (T.S.: 3/4.3.7,1, 3/4.3.7.12, 3/4.3.7.13).
- 5. SRM detector s fully inserted.
- 7. SRM counts:
A C Meter B D Reading /
- 8. List any bypassed SRM channels, and enter these channels in the CSO log.
N2-IOP-101A 57- June, 1985
Initials/Date 9~ IRM detectors fully inserted and on range 1 and selected for recor df ng.
- 10. List iny bypassed IRM channels, and enter these channels fn the CSO log.
- 11. Li st any bypassed APRM or flow unf t channel s, and enter these channels in the CSO log.
- 12. Process Computer available.
- 13. Reactor Pressure Vessel metal temperature recorder on panel 2CEC-PNL614 fn service and inking properly.
/
- 14. Recirculation Pump suction temperature recorder 833-R650 or control room panel 2CEC*PNL602 fn service and inking properly.
- 15. The Nuclear 8ofler Instrumentatfon System fn operation per N2-IOP-28.
- 16. Remote shutdown operable per N2-IOP-78 (T.S. 3/4.3.7.4).
- 18. Vibration and loose parts monitoring system operable per N2-IOP-86 (T.S. 3/4.3.7.11).
- 19. Seismic monitoring operable per N2-IOP-90 (T.S. 3/4.3.7.2).
- 20. The Neutron Moni tot f ng System operabl e per N2-IOP-92 (T.S. Tabl e 3.3.1-1).
N2- I OP-101A June, 1985
<<>> isa>>iu~ce
- 21. Yerify that the shorting links are in place.
/
- 22. The Rod Block Monitoring System operable per N2-IOP-93 (T.S. 3/4.3.6).
- 23. The Tr aversing In-Core Prube 'System operable per N2-IOP-94 (T.S.
3/4.3.7.7).
- 24. The Rod Morth Minimizer (ROl) System operable per N2-IOP-95A (T.S.
3/4.1.4).
- 25. Reactor Manual Control available and Rod Position Indication System operable per N2-IOP-96 (T.S. 3.1.3.7).
- 26. Feedwater Control System available per N2-IOP-7.
H. FINAL CHECKS
- l. CONTACT the planning coordinator to determine that all ma3or modifications that could affect startup are complete. Obtain a list of uncompleted and/or unsatisfactory items.
- 2. CHECK the Equipment Status Log to make sure no equipment is inoperative which would prevent plant startup.
- 3. Feedwater control lineup as follows:
- a. Place the master level controller for LY10 A, B, C in manual and set the output signal to zero with the close button. Place the set point tape at the desired operating level between 178.3 in.
and 187.3 in.
- b. Place the individual M/A controllers for LY10 A, B, C in manual and set the output signal to zero with the close button.
N2- I OP-101 A June, 1985
Initials/Date Place the set point tape for the low flow master controller at the desired operating level between 178 in. and 187 in.
- d. Place the M/A controllers for the low flow control valves (LV55 A, B and 2CNM-LV137) in manual and set the output signal to zero with the close button.
e, Place the level column selector switch for the desired level in put A or B.
- f. Place the'ingle Element/Three Element selector switch in single element control.
- g. Reset the set point set down signal on panel 603 and verify that the amber light on panel 603 is out.
- h. Reset the control signal failure lockout for each of the main feedwater flow control valves (LV10 A, B, C) on panel 603 and verify that the amber lights on panel 603 are out.
Verify that there are not high reactor vessel water level trip signals present on panel 603. If necessary, reset the high reactor vessel water level trips on panel 603 and verify that the associated amber lights are out.
- 4. Locked valve and open breaker checklist verified current.
NOTE: Reactor Coolant System chemistry analysis results must be from within the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (T.S.4.4.4 and T.S.4.4.5).
Gross activity determination should be made from the most recent data.
- 5. Obtain current rad/chem daily chemi stry analysis resul ts of the Reactor Coolant System. Verify the following:
- a. Conductivity less than or equal to 2 umhos (T.S.4.4.4).
N2- I OP-101A June, 1985
b, Chlorides less than or equal to O,l ppm (T.S.4.4.4).
- c. pH between 5.6 and 8.6 (T.S.4,4,4).
- d. Gross activity less than or equal to 0.2 uci/gram DOSE EQUIVALENT I-131 (T.S.4.4.5).
- e. Name of person/time contacted (for items a-d above):
- 6. Reactor protection System reset.
Channel A Channel B
- 7. Check the SSS Jumper Log to verify that all unnecessary jumpers and blocks have been removed and required equipment is not. out of service.
- 8. REVIEM the Mark Ups for entries which may adversely impact unit star tup.
- 9. After a Refueling Outage VERIFY that the reactor vessel hydro has been per formed.
- 10. The ISI examinations are complete and satisfactory.
Super>nten ent ll. VERIFY that all Technical Specification surveillances are current.
a ion Supervisor (SSS)
N2-IOP-101A June, 1985
- 12. List all inoperable or out of service control rods and accumulators.
- 13. Note any items on this checklist that could not be satisfied:
N2-? OP-101A June, 1986
NINE MILE POINT NI':i AR STATION UNIT f2 INTERIM OPERATING PROCEDURES CHANGE LOG procedure Mo. N2-10p- <o>"
CHANGE NUMBER SECTION NUMBER PAGES BRIEF DESCRIPTION OF CHANGE
Enclosure II'18 NINE MILE POINT NUCLEAR STATION UNIT 2
~ e~
/ ~ ~ ~
'~>Fr REACTOR ANALYST PROCEDURE I PROCEDURE NO. N2-RAP-6 POST REACTOR SCRAM ANALYSIS AND EVALUATION DATE AND INITIALS APPROVALS SIGNATURES REVISION 0 REVISION 1 REVISION 2 Reactor Analys t Supervisor R. G. Smith Station Superintendent NMPNS Unit 2 R. B. Abbott General Superintendent Nuclear Generation T. J. Perkins Summer of Pa es Revision 0 (Effective P~ae a Date 1-26 March 1986 NIAGARA MOHAWK POWER CORPORATION THIS PROCEDURE NOT TO BE USED AFTER SUBJECT TO PERIODIC REVIEW ~
1 N2-RAP-6 POST REACTOR SCRAM ANALYSIS AND EVALUATION 1,0 PDBPOSE The purpose of this procedure is to provide the review and evaluation of specific parameters associated with a Reactor Scram from all operating conditions. This procedure is designed to evaluate system performance from an initiation or isolation standpoint. The determination of safety system initiation, proper flow paths and system operation will be done using post trip logs, control room instrumentation, recorders, alarms and indicating lights. A secondary purpose of this procedure is to evaluate proper functioning of a system using the General Electric Transient Analysis REcording System (GETARS).
2.0 DESCRIPTION
Following a Reactor Scram various systems associated with maintaining Core Coolant Inventory and Reactor Containment Integrity must be directly monitored for proper sequential actuation and operation.
3.0 ACCEPTANCE CRITERIA All parameters monitored must satisfy either Tech Spec or expected system performance.
- 2) Scram Discharge volume surveillance requirements must be satisfied per Technical Specification 4.1 .3.1 .4.
- 3) Max cooldown of 100'F/hr not exceeded pex Technical Specification 3.4.6.1.
- 4) Support/Snubber inspections satisfactory per Tech Specs 4.7.5. d when required.
4.0 RESPONSIBILITIES AND CONDUCT 4.1 The Reactor Analyst Department will be directly responsible for data gathering and process evaluation. The analysis will be completed by the Unit Reactor Analyst or Site Reactor Analyst. In the event that those individuals are unavailable, the analysis will be conducted by a senior member of Technical Services or Operations.
4.2 At the conclusion o f the scram analysis, a repor t summary wit h recommendation will be included in this procedure. The scram report and this report will be sent to SORC for review.
5.0 PREREQUISITES
- 5. 1 Reactor scrammed.
N2-RAP-6 -1 March 198 6
6.0 ATTACHMENTS A. Scram Summary Sheet B. Pre-Scram Information Sheet C. System Response Sheets D. Plant Personnel Statements E Logic Check Sheet F. Evaluation Check Shee t G. Final Assessment H. Procedure Closeout Sheet 7.0 REFERENC E A. Generic Letter 83-28 8.0 PROCEDURE 8.1 After a Reactor Scram, with the permission of the SSS and knowledge of the CSO, collect the following when available:
a) Sequence of Events Log b) NSSS Post Trip Log c) BOP Post Trip Log d) Alarm typer printout e) Turbine Trip Recall Log f) Trend recording of various parameters needed to support analysis'ut the original out, tape to a blank sheet and attach to this procedure. Mark on the remaining trend paper that the missing section is with RAP-6.
g) GETAR S 8.2 Reactor Analyst Technician should complete the "Scram Summary", "Pre Scram Information" and "System Response Sheets" as specified and include comments in the appropriate locations as required.
8.3 Include all supporting graphs, trends, alarm printouts and reference material with the complete data sheets and forward to the Reactor Analyst or alternate per 4.1 for his review and analysis ~
8.3. 1 Originals of trends should be removed, at a convenient time for the CSO, and attached to this procedure.
In general, the following trends (arranged by panel) are desired (only select trends affected by the scram):
N2-RAP-6 -2 March 1986
8.3.1 (Cont'd)
F601 Service Water/RHR Temperature Post Accident Monitor (only if tripped to fact speed )
F602 Total Recirc Flow Recirc Pump Suction Temperature (speed 1"/hour)
F603 APRM/IRM (only one necessary, unless all are different)
SRM Reactor Pressure/Turbine Steam Flow Core Pressure Drop/Total Core Flow Reactor Steam Flow/Feedwater Flow Reactor Water Level P652 Turbine Bearing Metal Temperatures Turbine Bearing Drain and Thrust Brg. Temp .
Turbine Temperatures Turbine Vibration F875 DW and Suppression Chamber Temp SBGTS Discharge Flow/Filter 1B Diff Pressure F870 SBGTS Discharge Flow/Filter 1A Diff Pressure DW Equip Drain Leak Rate DW Floor Drain Leak Rate DW Equip Drain Pump Flow 3A, 3B DW Equip Drain Tank Level DW Equip Drain Pump Flow 1A, 1B DW Equip Drain Tank Level P612 Vessel Temperature PT 1 Vessel Head Flange 2 Vessel Bottom Head 3 Bottom Head Drain 4 Shell Flang e N2-RAP-6 -3 March 198 6
8.4 The Reactor Analyst or alternate will review the "Pre-Scram Information" and "System Response Sheets" and supporting information and will then complete the "Logic Check Sheets".
will 8 ~ 5 The Reactor Analyst Check and Final Assessment.
then complete the Logic Check, Evaluation 8.6 Reactor Analyst and Department Technician will closeout the procedure per Procedure Closeout Sheet.
N2-RAP-6 -4 March 1986
Scram Summer Scram 0 (obtain 8 from Rx. Analyst's file)
Date Time Sensor(s) causing scram:
Short narrative:
N2-RAP-6 -5 March 1986
Pre-Scram Information Date 6 Time of Scram Mode Switch Position Power Level Rx. Water Level Rx. Pressure Recirc. Temperature Loop A Flow (60Hz) or (15Hz)
Loop B Flow (60Hz) or (15Hz)
Total Core Flow Control Mode Drywell Pressure Drywell Temperature Suppression Pool Level Suppression Pool Temp Service Water Pumps on Line A B C D E F Circ. Water Pumps on Line A B C D E F Feedwater:
A Feed Pump Flow B Feed Pump Flow C Feed Pump Flow Total Feedwater Flow Feedwater Control Mode
( check) 3 Element Single Element Startup Controller Man Auto Master Controller Man Auto Were condensate and feedwater lineups normal'Yes) (No)
If no, explain:
Were electrical lineups normal? (Yes) (No)
If no, explain:
Were any surveillance procedures, tests or other evolutions in progress that may have effected station operationP (Yes) (No)
If yes, explain:
List any ECCS Systems that were running-and lineups at time of scram:
N2-RAP-6 -6 March 1986
S stem Res onse Reactor Vessel (System 28):
~ Highest water level attained (source)
Lowest water level attained (source)
~
Level control comments:
Highest Reactor Pressure attained (source)
~
Lowest Reactor Pressure attained (source)
~ Pressure control comments:
~
Comments Main Steam (System 01):
~
MSIV 's (Indicate open or closed)
Inside: HYV6C (F022C) 6D (F022D) 6A (F022A) 6B (F022B )
Outside: HYV7C (F028C) 7D (F028D) 7A (F028A) 7B (F028B )
Comments:
~
Relief/Safety Valves (open or closed)
Before After Time Time Elapsed Scram Scram Opened Closed Time PSV 133 (F013A) f SVVBC01 ]
PSV 128 (F013B)
[SVVBC02]
PSV 137 (F013C)
[SVVBCO3]
PSV 123 (F013D)
[SVVBC04]
N2-RAP-6 -7 March 1986
~ Relief/Safety Valves (Cont'd)
(open or closed)
Before After Time Time Elapsed Scram Scram Opened Closed Tim e PSV 136 (F013E)
[svvBc05 ]
PSV 122 (F013F)
[SVVBC06]
PSV 132 (F013G) fSVVBC07]
PSV 127 (F013H)
[svvBc08]
PSV 131 (F013J)
[svvBc09]
PSV 126 (F013K)
[SVVBC10]
PSV 135 (F013L)
[SVVBCll]
PSV 121 (F013M)
[SVVBC12]
Psv 134 (F013N) fSVVBC13]
PSV 120 (F013P) f SVVBC14 ]
PSV 130 (F013R)
[SVVBC15]
PSV 125 (F013S)
[SVVBC16]
Psv 129 (F013U)
[SVVBC17]
Psv 124 (F013V)
[SVVBC18]
N2-RAP-6 -8 March 1986
System Response (Cont'd)
(Cont'd)
~
Any relief valve leaking7 (Yes) (No)
If yes, comment:
~ Did any por tion o f ADS actuate 7 (Yes) (No)
If yes, comment:
~
Highest value attained on Main Steam Flow Area Temp Area d T
~ Comments:
Recirculation System (System 29):
~
Were recirc pumps downshifted7 (Yes) (No )
If yes, Auto Manual.
~
Were recirc flow control valves runback7 (Yes) (No)
If yes, Auto Manual.
~
Were recirc pumps tripped7 (Yes) (No )
If yes, Auto Manual.
~ Were any problems experienced with the pumps? (Yes) (No)
If yes, explain:
~ Were any problems experienced with recirc flow control valves7 (Yes) (No)
If yes, explain:
~ Comments:
Feedwater (System 06,07,08):
~
Max flow attained on feed pumps C
N2-RAP" 6 -9 March 198 6
(Cont'd)
~
Did any feed pump tripP (Yes) (No )
If yes, check which pumps tripped and provide cause. A cause:
B cause:
C cause:
Were high level tripe actuatedP (Yes) (No)
If yes, did all 3 high level trip
,lights come onP (Yes (No)
Was there any control problems associated with either the controller or flow control valvesP (Yes) (No)
If yes, explain:
Did fourth point heater drain pumps tripP (Yes) (No)
~ Comments:
Condensate (System 03):
~
Did any condensate pumps tripP (Yes) (No)
If yes, explain:
Did any condensate booster pump tripP (Yes) (No)
If yes, explain:
Comments:
Turbine (System 21):
~
Trip? (Yes) (No)
If yes: Auto Manual If auto trip, explain the cause:
~
Did bypass valves openP (Yes) (No)
If yes, did they function smoothly control reactor pressureP to (Yes) (No)
Comments:
N2-RAP-6 -10 March 1986
~ Coast down time.
~
Was turning gear oil pump started? (Yes) (No)
Auto Manual
~ Was emergency bearing oil pump started? (Yes) (No)
Au to Manual Did turning gear engage? (Yes). (No)
~
Comments:
Service Water (System 11):
~
Did any pumps trip? (Yes) (No)
If yes, what pumps?
Cause:
~
Comments:
Neutron Monitorin (System 92):
~
Highest power level attained APRM IRM SRM
~ Comments:
Reactor Core Isolation Cooling (RCIC) (System 35):
Was RCIC started? (Yes) (No)
If started: Auto Manual
~
Answer the following only if the system was started:
System flow GPM Controller setting GpM Turbine Speed RPM Turbine Erhaust Pressure Pump Suction Pressure Pump Discharge Pressure Steam Line Flow Did RCIC trip at any time? (Yes) (No)
If tripped, what was the trip signal?
Explain trip circumstances:
~ Comments:
N2-RAP-6 -ll March 1986
Residual Heat Removal (System 31):
Indicate if any pumps vere started:
A, if initiated Manual or Auto B, if initiated Manual or Auto C, if initiated Manual or Auto
~
If any system was started, indicate the following parameters:
RHR SW h T h T Radiation S stem Flov Flov RHR SW Level on SW NA NA NA
~ Other comments Lov Pressure Core S ra (System 32):
~
Was LPCS i.nitiatedP (Yes) (No)
If yes, how vas it donee Auto Manual
~
System flov rate
~
Comments:
N2-RAP-6 -12 March 1986
Hi h Pressure Core S ra (HPCS) (System 3 3):
~
Was HPCS initiatedt (Yes) (No)
If yes, how was it donee Auto Manual
~
System flow rate Pump suction: CST or Supp. Pool CST level Other cotttents:
Containment (System 81,82,83):
Highest Suppression Pool Water Level attained Lowest Suppression Pool Water Level attained Highest Suppression Chamber Air Temp Highest Drywell Pressure Highest Drywell Air Temp Drywell Oxygen Concentration Drywell Hydrogen Concentration Highest Drywell Radiation Level attained Drywell Floor Drain leakage rate
~ Drywell Equipment Drain leakage rate Has there been a change as a result of the scram'Yes) (No)
If yes, explain:
Other comments:
Control Rod Drive (CRD) (S stem 30):
~ Valve Closure Time Vent-AOV124 Vent AOV132 Drain AOV123 Drain AOV 130 Valve: (F010) ( F180) (F011) ( F181)
Comp Pt ~ RDSZC01 RDSZC02 RDSZC03. RDSZC04 Time of Comp. Pt ~
Scram Time Closure Time Did CRD air pressure vent off' (Yes) (No)
Final air pressure after the scram chert recorder N2-RAP-6 -13 March 1986
Time that scram dump volume vent and drain valve open:
tTime of Comp pt.:
RDSZC01 RDSZC02 RDSZC03 RDSZC04 Scram Dump Vol Drain Down Time Time of Comp Pt Time of Scram Reset Drain Down Time
~
Check on Scram Dump Vol Hi Lvl Rod Block Level Switches Time in RDSLC103 (LS125) RDSLC104 (LS127)
Time Cleared RDSLC103 RDSLC104
~
Comments:
C~leemu (System 37):
~
Was cleanup in service at time of scram'f (Yes) (No) yes, which pump was in serviceP A B Did cleanup pump(s) tript If yes explain:
Did cleanup system isolatet (Yes) (No)
If yes, explain:
Comments:
N2-RAP-6 -14 March 1986
Electrical (Systems 69-74, 100A & B):
~
If Normal Station Transformer was supplying 2NPS-SWG001 and 2NPS-SWG003, was a normal fast transfer observedP (Yes) (No)
Were any problems encountered, on transfer l from normal to reserve? (Yes) (No )
If yes, explain:
~
Was Div. I diesel startedP (Yes) (No)
Initiation Auto Manual
~
Was Div II diesel startedP (Yes) (No)
Initiation Auto Manual
~
Was Div III diesel initiated (Yes) (No)
Initiation Auto Manual
~
If any diesel generator was supplying its respective bus, record the, following:
RPM VOITAGE MAX LOAD Div I
~
Did any diesel auto tripP (Yes) (No)
If yes, explain:
~
Was any problem encountered with normal DC power suppliesP (Yes) (No )
If yes, explain:
N2-RAP-6 -15 March 1986
(Cont'd)
~
Was any problem encountered with Emergency DC power suppliesP (Yes) (No)
If yes, explain:
~
Comments:
~
Was Standby Gas System startedP (Yes) (No)
Initiation Auto Manual If it was running, did the system tripP (Yes) (No)
If yes, explain:
~
Comments:
Radiation Levels (System 79,80):
NORMAL ABNORMAL ARM' Drywell Stack Gas Off Gas Control Room Service Water RBCLC TBCLC Circ. Water Turb. Bldg. Vent
~ Comments:
Su orts/Snubbers Was a standby emergency system actuatedP (Yes) (No)
If yes, what systems were actuatedP
~
If a system were actuated, notify the SSS that inspections may be required to meet the T. S. Surveillance requirements of 4.7.5.d.
SSS Notified Time/Date
~
Comments:
N2-RAP-6 -16 March 1986
PLANT PERSONNEL STATEMENTS
- 1) On-shift STA should disperse these sheets and solicit comments from personnel involved in the scram.
- 2) Prepare a handwritten statement describing the trip event sequenc e and plant response as you remember it. Include your indications that a problem existed, your actions as a result of those indications, noted equipment malfunctions or inadequacies, and any identified procedure deficiences. Also include any information you consider important to review this unscheduled reactor trip.
Comments:
(Use additional sheets if necessary)
Signature Date Time Position N2-RAP-6 -17 March 1986
LOGIC CHECR SHEET LOGIC APPROACH Using information available from process computer, trend recorder and GETARS, check that the below listed actions occurred.
Reactor Water Level Trip Verification Explanation Set oint Allowable Action (Y/N or NA) If No L8 202.3 209.3 Turbine Trip Feed Pump Trip HPCS In). Valve Closed RCIC Turbine S/D MOV 120 Steam Admiss
+ MOV 128 L7 187.3 Hi Water Level Alarm L4 178.3 Low Level Alarm FCV Runback with loss of 2 of 3 Feed Pumps L3 159.3 157.8 Scram Set Point Set Down ADS Confirm L3 Shutdown Cooling Isolation MOV 40 A/B, 67 A/B, 104>
112, 113 Recirc Pump Downshift to LFMG L2 108.8 101.8 Control Bldg. Special Vent SBGTS Auto Start Normal Rx. Bldg. Supply 6 Exhaust Trip Rx. Bldg. Recirc. Fan Start s HPCS Starts; HPCS MOV 107 Reopens Div III D,G.
Recirc Pump Trip ARI Initiation RCIC Initiation Prim. Cont. Isolation (except MSIVs 8 Ll 6 SDC 8 L3) (refer to T.S .
for groups and valves)
Post Accid. Monitoring to Fast 1" /Hr 1" /Min Ll 17.8 Div I, II DG Start LPCI, LPCS MSIV Closur e ADS Initiation N2-RAP-6 -18 March 198 6
LOGIC CHECK SHEET
~
Reactor Pressure Trip Verification Explanation Set oint Allowable Action (Y/N or NA) If No 1205 8 + 1Z 4 Safety Relief Valve Lift in Safety Mode (PSV 137,127,134,129) 1195 0 + 1Z 4 Safety Relief Valve Lift in Safety Mode (PSV'126,135,121,130) 1185 0 + 1Z 4 Safety Relief Valve Lift in Safety Mode (PSV 122,132,120, 125) 1175 8 + 1Z 4 Safety Relief Valve Lift in Safety Mode (PSV 123,136,131,124) 1148 8 2 Safety Relief Valve Lift in Safety Mode (PSV 128,133) 1116 0 4 Safety Relief Valve Lift in Relief Mode 1106 4 4 Safety Relief Valve Lift in Relief Mode 1096 8 4 Safety Relief Valve Lift in Relief Mode 1086 4 Safety Relief Valve Lift in Relief Mode 4'076 8 2 Safety Relief Valve Lift in Relief Mode 1050 8 Recirc Pumps downshift to LFMG 1050 8 and 25 sec TD and power greater than 4Z, recirc trips to zero and FW control valves close 1037 8 1057 8 Rx Scram 766 8 746 8 MSIV closure when MSS in run 128 8 148 8 SDC mode of RHR isolates MOV 40 A/B, 67 A/B and 104, 112, 113 70 8 RCIC isolation ICS MOV 121, 128, 170 RCIC Vac. Bkr. isolation (MOV 148) coincident with Hi DW Press .
N2-RAP-6 -19 March 1986
~
Core Power (Neutron Monitoring)
Trip Verification Explanation Set oint Allowable Action (Y/N or NA) If No 118Z 120Z Fixed Neutron Flux Upscale Scram
.66W + .66W + Flow Biased Upscale Scram 51X with 54Z with max of max of 113.5Z 115.5X 15Z 20Z APRM Setdown Scram in Startup 120/125X 122/125 IRM Sera m 2 xa05 Scram on SRM with shorting links removed Drywell Pressure 1.68 8 1.88 8 Rx Scram Group 3,4,8,9 isolation Group 11 with low RCI C Steam Pressure of 754 Initiates SBGTS Trips Rx Bldg. Ventilation Star t Div. I, II, to r III Diese 1 Genera Actuate Div. I RHLA and LPCS systems Actuate Div. II RHR 6B C Systems Actuate Div. III (HPCS) System Hi DW Press. Alarm Ct Purge FCV 125 closes if open
~
Condenser Vacuum 25" Hg Alarm 22.1" Hg Turbine Trip 8.5"Hg 7.6"Hg MSIV Closure MSL Drains Closed N2-RAP-6 -20 March 198 6
LOGIC CHECR SHEET Turbine Trip Verification Explanation Set oint Allowable Action (Y/N or NA) If No 8 psig Turbine Trip (TT) on Thrust Brg Wear 1100 psig TT on Low EHC Fluid 2 25oF TT on High Exhaust Hood Temp.
TT on Moisture Sep.
Level with time delay 105.psig TT on low lube oil pressure from shaft pump when > 1300 RPM.
No Speed Feedback TT 10 mila'2 TT with 15 min. Time Relay mila TT immediately 22.1" Hg Vac TT 110% Speed TT on mech. overspeed 112% Speed TT on elec. overspeed 202.3" TT on Rx. Water Hi Level 800 psig TT on low ETS oil pressure 8 psig TT on low bearing oil pressure Elec. Fault TT on vari'ous elect. faults Runback failure TT if armature current is > 24551 on loss of stator amps after 2 minutes time delay or water cooling < 7006 amps after 3.5 minutes.
RCIC Initiation TT 13 psig Turbine runback on low pressur e stator water 180oF Turbine runback on high stator water temp.
15% mismatch Turbine runback is if cooling flow 15% less than req'd flow based on generator load .
N2-RAP-6 -21 March 198 6
LOGIC CHECK SHEET Turbiae (Cont'd)
Trip Verification Explanation Set oint Allowable Action (Y/N or NA) If No 190 psig Turning gear oil pump starts on low oil pressure 180 psig Emerg. Brg. oil pump starts 15 psig Turning gear oil pump starts on low bearing oil pressure 10 psig Em. Brg oil press starts on low bearing oil press.
Motor suction oil pump aut o start on low brg oil pressure .
Reactor Protection System (RPS) - not previously covered 6Z closed 7Z MSIV Closure Scram 3X NFPB 3.6X Main Steam Line Hi Rad Scram.
46.5" 79.5" .Scram Dump Volume High Level Scram Transmitter 46.5" 79.5" Scram Dump Volume High Level Float Switch 5X TSV Closure Scram 530 psig 465 psig TCV Fast Closure Scram on TCV Low Oil Pressure N2-RAP-6 -22 March 198 6
EVALUATION CHECK SHEET SAFETY LIMITS Did Reactor Pressure exceed 1325 psig while in operating cond. 1, 2, 3, or 4t Yes No Did Reactor Water Level drop below top of irradiated fuel (-14.4" on fuel zone instrument )
while in operating cond. 3, 4, or 5P Yes No Did thermal power exceed 25Z of rated thermal power with reactor vessel steam dome pressure
< 785 psig or core flow < 10Z of rated flow (10.85 MLB/hr) while in operating cond. 1 or 2P Yes No
~
Did the minimum critical power ratio drop to
< 1.06 while the reactor vessel steam dome pressure was > 785 psig and core flow greater than 10Z of rated flow (10.85 MLB/hr) while in operating cond. 1 or 27 Yes No If any of the above questions were answered yes, immediately notify the SSS for action.
SSS Notified Time /Dat e
~
If any of the above were answered yes, then explain:
~
Did the transient response of systems perform as expected according to Technical Specification, FSAR, Reload Licensing Analysis and Post ScramP Yes No If no, explain:
~
If any relief/safety valves did lift, did they lift and reseat at correct setpointsP Yes No N2-RAP-6 -23 March 1986
EVALUATION CHECK SHEET COOLDOWN
~
Was a maximum water cooldown of 100oP in any hour exceeded7 (ref. T.S. 3.4.6.1) Yes No If yes, advise the SSS of the action statement in T.S. 3.4.6.1.
SSS Notified Time /Dat e If yes, also notify Syracuse Engineering to perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system, Initial Dat e SCRAM DISCHARGE VOLUME Did the scram discharge volume drain and vent valves close within 30 seconds7 Yes No per T.S. 4.1.3.1.4.
Did the valves open upon scram reset7 Yes No Was SSS notified to have I&C verify proper float response. This is a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action .
(NA for scram initiated from < 900 psig) per T.S. 4.1.3.1.4 Yes No SUPPORTS/SNUBBERS
~
Vere there any supports or snubbers that failed inspection as a result of inspections per T.S.
4.7.5. d Yes No If yes, explain:
N2-RAP-6 -24 March 1986
FINAL ASSESSMENT SHEET FINAL ASSESSMENT Attach scram summary write-up with recommendations.
Initials
~ Did Plant Systems Function as designed? Yes No If Plant Systems did not function as designed, what was the abnormality and why?
Is there a condition not understood? Yes No NOTE: If there is a condition not understood, the Station Superintendent should be so notified and appropriate staff members called in to assist in evaluation.
If after further evaluation the scram is still not understood, SORC must review this report before authorization to restart.
If yes, ezplain conditions of concern.
'f after further evaluation the scram is still not understood, revie~ this report before authorization to restart.
SORC must Will SORC need to review this before restart? Yes No N2-RAP-6 -25 March 1986
PROCEDURE CLOSEOUT SHEET PROCEDURE CLOSEOUT
~
Evaluation areas of this procedure completed satisfactorily.
Rx. Analyst or Alternate Dat e /Tim e
~
SSS notified of completion.
Initials
~
Forward Copies of this procedure to:
Reactor Analyst Scram File, and Operations Superintendent. Initials
~
Record Scram information in the Scram History file and update Scram Timing History. Initials
~
Forward original of this procedure to the SORC Secretary.
Initials N2-RAP" 6 -26 March 1986