ML14164A314

Project, Revision to Unit 1 and Unit 2 Emergency Action Levels, Evaluation of the Proposed Changes License Amendment Request (LAR) for Revision of EAL Scheme to NEI 99-01 Revision 6. Part 4 of 7

ML14164A314
Person / Time
Site:	South Texas
Issue date:	05/15/2014
From:	South Texas
To:	Office of Nuclear Reactor Regulation
Shared Package
ML14164A341	List: ML14164A300 ML14164A303 ML14164A305 ML14164A308 ML14164A311 ML14164A314 ML14164A318 NOC-AE-14003087, Revision to Unit 1 and Unit 2 Emergency Action Levels, Evaluation of the Proposed Changes License Amendment Request (LAR) for Revision of EAL Scheme to NEI 99-01 Revision 6. Part 7 of 7
References
NOC-AE-14003087
Download: ML14164A314 (511)

Text

CALC. NO. STPNOC013-CALC-002 tEN ER CO N Radiological Release Thresholds for Emergency Action Levels Ette5WiCA,&,yp~oJ.CL Everyday.

PAGE NO. 5 of 49 Table of Contents 1.0 OBJECTIVE/SCOPE

...........................................................................................................................

6 2.0 SUM M ARY OF RESULTS ..................................................................................................................

6 3.0 M ETH OD O F ANALYSIS ...................................................................................................................

7 4.0 INPUTS ...............................................................................................................................................

7

5.0 REFERENCES

.....................................................................................................................................

7 6.0 ASSUM PTIONS ...................................................................................................................................

8 7.0 STAM PEDE CALCULATIONS

...................................

6 .................................................................

9 7.1 Unusual Event -RU] ...................................................................................................................................

9 7.2 Alert, Site Area and General Emergencies

-RA 1, RSI, RG1 ...........................

...............................

10 Attachm ent 1 -H and Calculations

..................................................................................................................

12 Attachm ent 2 -Calculations'...........................................................................................................................

25 Attachm ent 3 -STAM PEDE O UTPUT .........................................................................................................

32 CALC. NO. STPNOC013-CALC-002 O .E N ER C.O-N Radilogical Rlese Thresholds v f4: ft .f for Emergency Action Levels PG O f4 1.0 OBJECTIVE/SCOPE The purpose of this calculation is to determine the Emergency Action Level (EAL) threshold values ofa radiological release from the Unit Vent or Main Steam Lines for an Unusual Event, Alert, Site ArekBmergency, or General Emergency.

The calculated threshold values ae to be included in the STP EAL Technical Basis document, which implements the new NEI 99-01, Revision 6, Emergency Action Level Sclehee and will be submittedto tIheNRC for appIrov"al. NRC approval, the values will be used in OERPO !-ZV-INO l, Revision 10,I Emergency Classification.

Both a hand calculation and the South Texas Assessment Model Projecting Emergency Dose Evaluation (STAMPEDE) software program Were used to generate the results. The hand calculation is included as Attachment 1.Revision 1 of this calculation incorporated decay for a release taking place one hour after reactor shutdown.

This was done to create continuity between the two methodologies prisent.2.0

SUMMARY

OF RESULTS The results of the calculationis for the radiation monitors specified in the STP EAL Basis Document and are listed in Table 2.1, below.Table 2.1: Summary of Calculation Results C Emergency Action Level RT-8010B, Unit Vent (ACi/sec)RT-8046 through 8049, Main Steam Lines (jiCi/cc)RU1 1, Unusual Event 1 I C U I "inerai iimergency 1"b 1AivirnUr,)

was not usea to cietermine ute tnresnuou ior l%.u 1 methodology should be used to determine the threshold value.This calculation will be used to establish the threshold values for abnormal radiation based emergencies in the STP EAL Technical Basis document.

CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds Mv. I pwied EFy ft for Emergency Action Levels PAGE NO. 7 of 49 3.0 METHOD OF ANALYSIS Previously, STAMPEDE was used to calculate the Emergency Action Level threshold values for effluent releases.

A hand calculation will verify the STAMPEDE calculations.

The hand calculation is described in Attachment 1 of this document STAMPEDE conforms to the requirements of STP Procedure OPGP07-ZA-0014, Software Quality Assurance Program. STAMPEDE was run at STP on an STP computer and under the supervision of an ENERCON employee with access to the STP site as a critical worker.4.0 INPUTS 4.1 Per NEI 99-0 1, Revision 6, Initiating condition AU1, EAL 1, the Notice of Unusual Event initiating condition is a release of gaseous or liquid radioactivity greater than two times the ODCM limit for sixty minutes or longer (Reference 5.10).4.2 The ODCM offsite dose limit is exceeded if the Xe-133 release concentration exceeds 7.41E-04 LCi/cc (Reference 5.6).4.3 The Unit Vent flow rate is 9.4E+07 cc/sec (Reference 5.1).4.4 The main steam line pressure and PORV choke flow rate are 1285 psig and 1.05E+06 lbmlhr, respectively (Reference 5.2).-4-.5-T he-speci-f-e-volume-ofsaturated-st-am'at-1285-gýir0;-3

-- .4.6 The concentration is varied to find the release concentration which correlates to each emergency action level. Emergency action levels are taken from NEI 99-01, Revision 6 (Reference 5.10) for initiating conditions AA1, ASI and AGI. EAL I is the EAL of interest in each initiating condition., The doses at the Site Boundary that correlate to the threshold concentrations are listed in Table 4.1.Table 4.1 EAL Offsite Dose Initiating Conditions Alert Site Area General Thyroid CDE S50 mrem L500 mrem L5000mrem

5.0 REFERENCES

5.2 Main Steam PORV Capacity Verification MC05591, Revision:

1 5.3 NIST Steam Tables, 2011 5.4 0ERP01-ZV-IN0l, Emergency Classification Draft Revision 10 5.5 OERPO 1-ZV-TP01, Offsite Dose Calculations, Revision 21 5.6 STP Calculation NC-9012, CRMS Rad Monitor Setpoints, Revision 7 5,7 STP Calculation NC-901 1, Revision 2 5.8 STAMPEDE Computer Program, Revision 7.0.3.3 5.9 STAMPEDE User's Manual 5.10 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors 5.11 OPGP07-ZA-0014 Quality Assurance Program 5.12 ITWMS Call Number 1000010987 Design Document, Revision 0 CALC. NO. STPNOCO13-CALC-002 E N E R.C O N Radiological Release Thresholds REV..I.,- .: .... for Emergency Action Levels ftPk- iEyty -1day PAGE NO. 8 of 49 6.0 ASSUMPTIONS 6.1 Unit Vent Noble Gas Monitor To .be c6nsistentfwith the ODCM methodolpgy, the unit vent release iS assumed t6 be entirely Xe-133. fThe unitvent noble gas monitor is calibraed to. Xe133 (Referenij 5.1)tfierefore; the mionitor reading accurately reflects the Xe-133arelea s e magn " " tude.To be consistent with ODCM methodology, the main steam line release is assunied to be entirely Xe- 133. The noble gas monitor is calibrated to Xe- 133 (Reference 5.6).6.2 Release Duration Per Reference 5.10, Sections IC AAI, ASi, and AGI developer notes, the release should be assumed to last one hour.6.3 Release followingReactorShliitdown The release initiates one hou r after react6r thutdown.

While ' release initilating atrreactor shutdown is likely, sighificant decay of short lived fiuclide occurs dur' g the migration time. A release at reactor shutdown Would have a sig'ificantly higher activity at the moftitor location than a.. ... I-a-reti anfoýi- th e- thr eshoid-to-not-be t .alculated-at-shutd6wn-as this----would crea.e avery high threshold Which would:notbe appropriate for releaseswhich occur shortly after shmtdown.

ne hour after reActOr slhutdown is sufficient time to decay short lived nuclides anid teate a conservative threshold.

6.4 Source

Term Per Reference.

5., ahy unit vent release-withlinreased RCS activity and no core melt should be calcuflated using tl6 gap inventory.

Theiffore.

the gap iinVentdi.

is used f6r all.unit vent releases.Per Reference 5.1, *for a main steam line release following a steam generator tube U rupture it is appropriate to use .an inventory of noble gases plus 0.2% iodine. A steam generator tube rupture is the only scenario which would create significant 0ffsite doses through a main steam line release.6.5 Default STAMPEDE Input Values Reference r.a .0 1-axid--AO-l-suggSt-isihg-the-ODCM or the site's emergency dose assessment mnethodology.

STAMPEDE is ttsed for emergency dose assessment.

Per Reference 5.1, when actual meterolog I is not available, the default STAMPEDE values should be used. HIad the ODCM methodolo6 been used,. the 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> peak ,/Q value would be used which is less conservative than the X/Q value produced by STAMPEDE using default meteorological cbndition0s'.

Therefore, the use of STAMPEDE default values provides a more conservative estimate than that of the alteriative method outlinqd in Reference 5.10.6.6 Average Effluent Concentration (X/Q).The same X /Q is used for the unit vent and main steam line release. Reference

5.1 applies

the same unit vent xIQ to Units I and 2 which would also be applicable to the main steam line. All releases are considered to be ground level releases.

CALC. NO. STPNOC0 3-CALC-002 E" N E R O Radiological Release Thresholds i vo4 for Emergency Action Levels c-Every pJt Evty doPAGENO.9o

7.0 STAMPEDE

CALCULATIONS

7.1 Unusual

Event -RU1 7.1.1 Unit Vent Monitor AUl recommends declaring an unusual event due to a release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer (Reference 5.10).STP sets the ODCM limit at 7.41E-04 jiCi/cc (Reference 5.6, pg. 16). Two times the limit would be 1.48E-03 pCi/cc. The threshold is listed in !.tCi/sec so that variations in flow rate do not change the threshold.

The normal flow rate from the unit vent is 9.4E+07 cc/sec (Reference 5.1).Concentration

• Flow Rate '; Release Rate )cc /ICIN* (9.4 + 07\\sec (1.48E -03) (9.4 + 7 1.4E + 051-...... ...... .... .-... ...Eq atzt6-n 7. 7]...... ...11.1 7.1.2 Main Steam Line Monitor The ODCM does not calculate a release corresponding to allowable limits for the main steam line monitors.

Since the unit vent release calculated in the ODCM was assumed to be primarily Xe-133, the assumption is made in the ODCM that other noble gases and iodine may be ignored in the calculation.

This assumption is equally justifiable for the main steam line and the same limiting release will be used.The magnitude of the release calculated for the unit vent Unusual Event applies to the main steam lines as well. The main steam line PORV's will create a dose exceeding two times the ODCM limit by releasing 1.4E+05 ptCi/sec of activity which is equivalent to the release from the unit vent.The steam lines hold saturated steam at 1285 psig, per Reference 5.2, which has a specific volume of 0.338 ft 3/lbm (Reference 5.3). The PORVs will release the steamn at 1.05E+06 lbm/hr per Reference 5.2. This creates a set flow rate of steam from the main steam lines of 2.79E+06 cc/sec as shown below.(lb m\/ft 3\ (se cc FDensity

• 28316.846 ( +) 3600 (-r sec 1.05E + 06 cb\ h 338 (ft--"n + 3sec LOSE+06h 0.338 E) 28316.846 (7t.. 36+ sec Equation 7.1.2.1 CALC. NO. STPNOC=13-CALC-002 0M E NERCO N Radiological Release Thresholds

.0 for Emergency Action Levels P C, Since the flow rate is set, the concentration will determine the limit. Equation 7.1.1.1 solves for the limiting concentration of 5.OOE-02 IXCi/cc as shown below.L Limiting Concentration

"-Release Rate(1.40o, 5mc'1.0 e/= 5.00E 02("cc)2.79106 1C( I cc\_sec)Equation 7.1.2.2 7.2 Alert, Site Area and 'General Emergencies

-RAl, RS t, RG1 7.2.1 Unit Vent Monitor Input The Alert EAL is set to 10 mrem TEDE and 50 mrem Thyroid CDE per Reference 5.10.__ _ _Theemergencyoffsite_dose calczulation-softwareSTAM DE.was-usedAtQ__Lcu-Iate the-__release which corresponds to this dose. A release concentration correlating to the EAL threshold value was calculated by varying the input The f0l6wing assumptions and inputs were used for'the calculation as described in Sections 4.9 and 6.0.* Release begins at reactor trip.* Release lastý for one hour C Gap inventory source termra*Defaul STAMPEDE input values o Windspeed

= 13.2 mph* o Stability.

class D Results Given a monitored unit vent release of 2.50E+06 gCi/sec, the Thyroid CDE is 51 em/ tthalo~se.

FA. Initiating-Condition is exceeded.Threshold, values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert. Since-the correlati6n between release concentration and dose is linear, threshold values for the steam line monitors are 2.50E+07 and 2.50E+08 gCi/sec for the SAE and GE respectively.

Both are also limited by Thyroid CDE. Additional STAMPEDE iterations were performed to confirm this and are attached.The input and output files can be found at the end of this document in Attachment

3.

CALC. NO. STPNOCOI3-CALC-002 ENERC0 N Radiological Release Thresholds for Emergency Action Levels PGN.Itlc ~ PAGE NO. I11 o f 49 7.2.2 Main Steam Line Monitor Input A release concentration correlating to the EAL threshold value was calculated by varying the input. The following assumptions and inputs were used for this calculation as described in Sections 4.0 and 6.0." Release begins at reactor trip* Release lasts for one hour* Noble gas + iodine with 0.2% iodine source term* Default STAMPEDE input values o Windspeed

= 13.2 mph o Stability class D Results Given a monitored main steam line release of 4.5 PCi/cc, the Thyroid CDE is 50 mrem/hr and the EAL Initiating Condition is exceeded.-7-:-The-input.and-output-files-can.

be foundaat-7the end of-this document in Attachment3-

.----7.3 Threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert. Since the correlation between release concentration and dose is linear, threshold values for the steam line monitors are 45 and 450 ýiCi/cc for the SAE and GE respectively.

Both are also limited by Thyroid CDE. Additional STAMPEDE iterations were performed to confirm this and are attached.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002

.E N E: A C. O for Emergency Action Levels REV. f-i' Attachment

'1 PAGE NO. 12 0f49 Attachment 1 -Hand Calculations 1.0 OBJECTIVE/SCOPE Each release calculated using STAMPEDE in the main document is calculated by hand in this attachment and the results compared to STAMPEDE.2.0

SUMMARY

OF RESULTS Table 2.1 is displayed again below showing ihe'results from all the calculations.

The minor difference is due to STAMPEDE using decay factors over a'one hourperiod after shutdgwn.

This also accounts for the change in the limiting dose being TEDE in the .ha:d calculationsý and Thyroid CDE in the STAMPEDE calculations.

The accuracy of the hand calculathon is considered sufficient and recommended for use in Emergency Action Levels.Tatibe 2. Results C Emergency Action-- -Le-vel RT-8010b, Unit Vent..... .( iIsec)-'--

RT-8046 through 8049,-Main Steam-Line--

-0(ici/cc)RAI I.J, kergency I RG1 I General. EmergencY I 3.0 METHOD OF ANALYSIS Using the limiting dose at the site boundary, the release is back calculated using atmospheric dispersion models. The XIQ value used is calculated from Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Rather than using the most conservative meteorology, average meteorological conditions are used as inputs Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 O E N E R C O N for Emergency Action Levels RBV. 1 i P"I- rmpdon Attachment 1 PAGE NO. 13 of 49 to most closely agree with STP emergency dose assessment methodology per the ODCM and STAMPEDE.

Assumed nuclide inventories are taken from Reference 5.4. The dose conversion factors are taken from Reference 5.2. A release concentration is used to find an initial projected dose at the Site Boundary.

Using the projected dose at the Site Boundary, the release concentration is scaled to find the limiting dose for each EAL.4.0 INPUTS The Unit Vent flow rate is taken from the Offsite Dose Calculation Manual; Revision 17, March 2011 and is 9.44E+07 cc/sec.The main steam line pressure and PORV choke flow rate were taken from Reference 5.5 and are 1285 psig and 1.05E+06 lbm/hr respectively.

The specific volume of saturated steam at this pressure is taken from the NIST steam tables and is 0.338 ft 3 ilbm.The release concentration is varied to find the release concentration which correlates to each emergency action level dose. Emergency action level doses are taken from NEI 99-01 Revision 6 for initiating conditions AA1, ASI and AGI. EAL 1 is the EAL of interest in each initiating condition.

The limiting doses are listed in Table 4.1. NEt 990-01 Revision 6 states that these values are:.basedn-ractions-of~the-BnvironmentalProte~tion-genei-e-Pra-t~iv tio .- -" -Guidelines (EPA PAGs) and the General Emergency represents the protective action values recommended by the EPA.Table 4.1 EAL Thresholds Alert Site Area General* A release lasting one hour is selected per NEI 99-01 Revision 6 developer notes.* Atmospheric dispersion factors are calculated per Regulatory Guide 1.145 (Reference 5.1). The reactor building dimensions used as inputs for this calculation are taken from Reference 513.* Nuclide inventories are taken from TGX/THX 3-1, (Reference 5.4) which is the source document for the nuclide inventories used in STAMPEDE.

The release inventories are a gap release and noble gases plus 0.2% iodine which are listed below. Each nuclide inventory was normalized to Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 0 E N E R CON for Emergency Action Levels *REV.Attachment I PAGE NO. 14 of 49 Table 4.2 Gap Inventory Nuclide Activity hitt rA Normalized Nuclide.Activity (U.cikOc Normalized 1-132 1ý.501B+/-05 1.53E.03 Xe-4137 1.900V+07

.4-O 1-134 2.40t+05 12.4513-03 Cs-134 3.70E+0~l1 3.78E-07, Kr-83ni 1.3OKR+06 1t~.33E'-w02 T12 4.BO+004,91-0 Kr-85~~~~~'

3,F+5 37E0 uQ .0-3 9QE1 Kr-88 78E0 .9E0 rS 1IO-2 l1El Xe-131m d.IOE+05 1.lE-0~C~4

.0~0l 75Bl Xe-133~~~~~~~~

_0.0 + 7 22E- 1 Sr 96 4 E0 5 1 C C Table 4.3 Noble Goses+0.2%

Iodine Inventory 1Nudilde Inventr Nd IVm'inlized 17-132 8.1E-'2 31E0 W-34 t .86E-02 69E Xer-1331n

.'25§0 .56-0 i Xe-138 5.80E-101 2.115E-3 Mr-85 7.60E+0-O 2,82E-02 Kr-87 9.80E-01 3.63E-03 Kr-89 8.40E 02 3. 12E04, The dose conversion factors taken from EPA 400R92001 (Reference 5.2) are listed in Tables 4.4 and 4.5 below.C Radiological Release Thresholds CALC, NO. STPNOC013-CALC-002 U E N E R C 0 N for Emergency Action Levels NO. I Exce"Ve-&,V

"*t EMY doAttachment 1 PAGENO. 15 of 49 Table 4.4 TEDE Dose Conversion Factors Dose Conversion Factor (rem ner Dose Conversion Factor (rem per Nuclide Nuclide 1-132 4.90E+03 Xe-137 11E0 1143.l1OE+03 Cs-134 6,30E+04 Kr-85 .30E+00 Ru1O3 13E0 Kr-881.30E+03 Zr9S .0+0 X~1~ni --4 .9-- -C ~144 -- .50E+05-Xe132.OOE+O 1 -.Sr89 5.O0E+04 Table 4.5 Thyroid CDE Dose Conversion Factors Thyroid CDE DCF Nuclide (rem per uCi*hr/cc)

I.70E+0 1-134 1.30E+03.ý M. HAMP"Ammm" The unit vent noble gas monitor energy efficiency by nuclide is taken from Offsite Dose Calculation Manual (Reference 5.3). The values are relative to Xe-133 efficiency since the monitor is calibrated to Xe-133. Table 4.6 displays the energy efficiency by nuclide relative to Xe-133.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC.002 O E N E R C 0 N for Emergency Action Levels REV._04,& PW.-rAttachment 1 PAGE NO. 16 of 49 Table 4.6 Energy Efficiency Relative to Xe-133 Efficiency Relative to Xe-133 e4i331fi:, .. 0'* .14 :X e -l in " Ic~,1*There is'no relative efficiency available for Kr-83m. Assumption

6.4 fiurther

justifies the omission.Table 4.7 Nuclide Half Lives Nuclide Half Life Nuclide Half Life (h 1 C 1-13 2. -O ' Xe-37' 6.3 8t-02 1-134 8.77E-O1' Cs14 1.80E+04 Kr-83mn 183+O00 Te132 7.79E+01 Kr-8 9.40E+04 RulO3 44E+Kr-88 2.84E+00 zr95 1,55E+03 Xe-131m 2.83E+02 Ce144 6.82E+03 Xe-133 1.27E+02 8r89 1.21E+03., A '. I 0 The half-lives are taken from Reference 5.15 which lists the input data used by STAMPEDE.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 EN ER C O N for Emergency Action Levels REV. I hekac-Fmyprqj,.cth~iy.

Attachment I PAGE NO, 17 of 49

5.0 REFERENCES

5.1 Regulatory

Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982.5.2 EPA 400R9200 1, Manual of Protective Action Guides and Protective actions for Nuclear Incidents, Revision 1, May 1992.5.3 Offsite Dose Calculation Manual, Revision 17, March 2011.5.4 TGX/THX 3-1, Revision 5, Westinghouse Radiation Analysis Manual.5.5 MC05591,.

Main Steam PORV Capacity Verification, Revision 1.5.6 NIST Steam Tables, 2011.5.7 0ERP0l-ZV-IN0I, Emergency Classification, Revision 10.5.8 OERPO 1 -ZV-TPO 1, Offsite Dose Calculations, Revision 21.5.9 STP Calculation NC-9012, Process and Effluent Radiation Monitor Set Points, Revision 7 5.10 STP Calculation NC-90 11, CRMS Rad Monitor Setpoints, Revision 2.5.11 STAMPEDE Computer Program, Revision 7.0.3.3.5.12 STAMPEDE User's Manual 5.13 STP Drawing 6C1 89N5007, General Arrangement Reactor Containment Building, Revision 6 5.14 NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors--5.15-. -ITWMS Call.N nbg r_1.0000-0987-Design-Doczumeat,-Revisipn 07: .. 7 7.. : 7 -.: 7 6.0 ASSUMPTIONS

6.1 Release

lasts for one hour Per NEI 99-01 (Reference 5.14), IC AA1, AS 1, AGl developer notes, the release should be assumed to last one hour.For this to be true for the main steam line, it is assumed that the PORV is open for one hour. To calculate the most limiting case, it is assumed that the maximum flow possible is being released from the PORV.6.2 Nuclide mix Per QERPO I -ZV-TP01, Offsite Dose Calculations (Reference 5.8) any unit vent release with increased RCS activity and no core melt should be calculated using a gap inventory.

It is conservative to assurme an increased RCS activity and not within the intended scope of the relevant initiating conditions to assume core melt. Therefore, a gap inventoly is used for all unit vent releases.Per 0ERP01-ZV-TP01, Offsite Dose Calculations (Reference 5.8) for a main steam line release following a steam generator tube rupture it is appropriate to use an inventory of 100 percent noble gases plus 0.2 percent iodine. Since a steam generator tube rupture releasing through the PORVs is the only steam generator tube rupture scenario which would create offsite doses large enough to meet or exceed the EALs, this assumption is made, Radiological Release Thresholds RALC. NO. I TNOC0I3-cALC-002 N N. E C 0:N for Emergency Action Levels " v..*Ki--&RtY oa. E ,Y .Attachment I PAGE NO. 18 of 49 (IF 6.3 Atmospheric Dispersion NEI 9901 .(Reference5.14) developer notes for initiating conditions AA1, ASl and AG1 suggest using the ODCM or the site's emergency doss-easessnient methodology.

Per 0ERPOi -ZV-TP01, Offsite Dose .Calculations (Reference 5.8), when actiual":meteorology is notavailable, the default STAMPEDE values should be used. The default STAMPEDE values assumea' stability class D for atmospheric dispersion and of 13.2 mph. These used as inputs for the atmospheric dispersion calculation.

It iS clear that STAMPEDE uses the sanm method for calculating atmospheric dispersion factor (X/Q) outlinedin section 7.1.1, ofthisAttachment.

However, does not follow the same logic in selecting the appropriate restilt from. the three zalcu!itions. .T S6A, MEDE value printed in the results foufid in attachment 3 is.consistent with the laig6st of the three hand calculated X/Q values. This suggest that STAMPEDE.

simply selects the largest of the three X/Q values resulting in a much more. c 5 nser.Vatj've estimateý This calculation, will deyiate from the recommendations of Regulatoiy Guide 1.145 and. cnform to the methodo.ogyASTAMPEDE uses.The close.'proximity of all relele points allows for a single atmiospheric dispersion coefficient to be u.sed. This ansiumptioh is .STAMPEDE,.::

" 6.4 Exposure Pathways The dose conversion factors used intable 4.4 and 4.5 represent a summation of dse conversion factors for external plume exposure, inhalation from the plume, and external exposure from C deposition.

Because the dose estimations are used for implementing early phase'protective actions, conversion factors usfign limited pathways.

e appropriate,'

.: The EPA does not provide a dose conversion factor for Kr-83m. Because the PAGs are based on EPA dose calculations, it s 'ppropriate .to only use the nuciides for which ose conversion factors are provided.

Additionialy kr-8i3m represents only "of the nurclide inyventory activity and its exclusion would not signifio hi 1, affect the final dose " 6.5 The release initiates one hour after reactor shutdown.

While a release initiating at reactor shutdown is ifely, significant decayý of short lived nuclides occurs during the migration time. A at the geceptioAi site. It is- important foi the threshold to not be calculated at shutdown as this would create a -veiy high t'shold Wh. ich i, uid hot be appiopriate for releass which occur shortly after One our Aer reactor shutdown is suff cient time to eiay short lived nuclides and create a conservative threshold.

Decay' is incorporated for one hour from reactor shutdown as well as migration time. Half-lives are taken from Reference 5.1 5.:Migration time is assumed to be the reciprocal of the wind speed.(I Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 EN E R C 0 N for Emergency Action Levels REV.I Fxkc-&ri EVVY ft Attachment I PAGE NO. 19 of 49 7.0 HAND CALCULATIONS 7.1 Unit Vent Monitor 7.1.1 X/Q The atmospheric dispersion factor, X/Q, determines the change in concentration between the unit vent discharge and the dose reception site. This value is based on meteorological conditions and will vary with wind speed and stability class. The ODCM uses the highest annual average X/Q value at the site boundary which is 5.3E-06 sec/r 3.However, for an accident related release STAMPEDE is used rather than the ODCM. STAMPEDE uses real time, user entered, or default meteorological conditions to calculate the X/Q for a specific accident.

Default values will be used as inputs into the Regulatory Guide 1. 145 method for calculating X/Q as described below. Default values are identified in section 6.0, Atmospheric Dispersion.

For a neutral atmospheric stability class, which is the default in STAMPEDE, X/Q values can be determined through the following set of equations.

QU 1 0 (Tryoz+)Equation 7.1.1.1 X 1 QU10(37rryor.)

Equation 7.1.1.2 X 1 Q U 1 Equation 7.1.1.3 Where X/Q = relative concentration (sec/mA3)X = 3.14159 U 1 0 = windspeed at 10 meters above plant grade (m/s)ay = larteral plume spread (in), a function of atmospheTric stability and distance, determined from Regulatory Guide 1.145 Figure 1 0z = vertical plume spread (m), a function of atmospheric stability and distance, determined from Regulatory Guide 1.145 Figure 2 ly = (M -1)uysoom + uy = lateral plume spread with meander and building wake effects (in), a function of atmospheric stability, windspeed U 1 0 , and distance; M is determined from Regulatory Guide 1.145 Figure 3 A = the smallest vertical-plane cross-sectional area of the reactor building (mA2), taken from Reference 5.13 and shown below Radiological Release Thresholds CALC. NO, STPNOC013-CALC-002

' E N .ER C 0 N for Emergency Action Levels .~ 1N.F --O-. ],y F doAttachment I PAGE NO. 20 of 49 C Figure 7.1.1.1: Reactor Building Dimensions.

EL 241*-0'EL 2 NS7" :') ' .. ..euOL 'CRIMM E L. .. ....... :. .. ... ". ....: .* ...ACCESS RE..'. ..,. ..El.R LEVEL S tt.tL'9".LL: " ©Is a, 8! E TUB IlVA RN DUCTKIN LUCRE SEIW I Ro .tý ... I..." ." .A ., OR:. .' YLV.N ,. -" EL t-18*'U2-Assuming the reactor building cross section to be a perfect rectangle and half sphere, the variables are defined as follows;U 1 0 13.2 iph. 5.9 m/s U), *zv ft'r crt :I'-r" CZ "4.2 m EY .9(M -.1)Clya m+ay;M". -Y1200 M A = (135'* 158') + -31128.37 The three equations become;..." " 1-= 5.398 10-6 Q 7- 9 2(i00

• 4.2*31128.37)

Radiological Release Thresholds CALC. NO. STPNOCO I 3-CALC-002-N ER CO N for Emergency Action Levels REV.I zelae-Evim-cthey

&y Attachment 1 PAGE NO. 21 of 49 X 1= 3.568

• 10-6 Q 5.9(37r

• 1200

• 4.2)X

• 1 1.07* 10-5 Q 5.9

• 7r * [(1 -- 1)cryaoom

+ 1200]

• 4,2 = .7*0-To select the appropriate X/Q value, the first two X/Q values should be compared and the higher value selected.

This value is then compared with the third X/Q value and the lower of those two is the appropriate X/Q value. The appropriate X/Q is 5.3 9E-06 sec/rn 3 for default meteorological conditions by the methodology recommended in Regulatory Guide 1.145.This calculated value is very similar to the ODCM highest average value of 5.3E-06 sec/m 3 which was not selected for use. Additionally, the value shown in the STAMPEDE output file at one mile is 1.032E-05 sec/m 3.This suggests that STAMPEDE uses the same methodology and simply selects the largest atmospheric dispersion value to remain conservative.

This methodology will be replicated and 1.07E-05 will be used as the X/Q.-7.T1.2

.. .As previously stated, a gap inventory is appropriate for this problem. The gap inventory is taken from TGX/THX 3-1 (Reference 5.4) which is used as the source term for STAMPEDE inventories.

The concentrations were then normalized so they could be scaled to the varying emergency classifications.

The values for the normalized inventory can be found in Table 4.2.7.1.3 Dose Conversion Factors As stated in NE199-01 (Reference 5.14) developer notes, the purpose of dose projections is to check if the Environmental Protection Agencies Protective Action Guidelines (EPA PAGs) have been exceeded.

The dose conversion factors provided by the EPA in EPA 400R92001 are used. These dose conversion factors account for external plume exposure, inhalation -from the plume, and external exposure from deposition and are listed Tables;S :W '9 vr_-;T W --a T Z 2; 1;; Fý i ,+.',+UU'tJ W1 UW~iIUI ~1 J.1 2~ alie i2-t% 'tUV1\74U elurenc.D.)

The EPA does not provide a dose conversion factor for Kr-83m. This nuclide contributes 1.33% of the inventory activity.

The lack of this nuclide's contribution to the final dose will not significantly affect the outcome.7.1.4 Decay Time One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown.

Decay is also incorporated for the duration of the migration time. The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.

Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 O E N E R CO N for Emergency Action Levels NO. f Exre q'-f '.. rAttachment 1 PAGENO.22of49 (7.1.5 Dose Calculations.

The dose rate at the site boundary is calculated using Equation 7.1.5.1.1.07575 D=~F Ci *O5 T 1/2 1 *DCFj: Equation 7.1.5.1 Where-.' do/ e rate per hour at thesite boundary X atmospheric dispersion coefficient as calculated in section 7.1.1 Q F unit..vent flow'.

• ae:. ......C' concentration of nuclide i.at the time. of shutdown 1. .

total decay timeo4'in' erest fr.rn section 7..4 Tj 1 j = the half-life of nuclide i'"C"." the dose conversion factor for nuclide i listed in tables 4.4 and..: ."." .:4.5 '.C!" "- " .... "'.. ' ": .The total concentratiofh of the ihuclides is varied t..find the dose rate of interest.Beginning with an arbitrary release concentration of 1 4Ci/ce, the dose rate is calculated.

Since the dose is linearly correlated to concentration, the release concentration may be Scaled to fmid the dose rate ofinterest.

The Alert EAL. Is ib- rem .TEDE or 50 mrarme Tyroid CDE. Using the above method to calculate TEDE, with the appropriate.

Odcoqýrsion fact6rs, a limiting release rate of'2.33E+06 gLCi/sec .riP the unitv4ent results in 5.7 nem TEDE. Using the calculated release rate. to find Thyroid CDEwith the cnverston factors, the same release resulj' inva 50 inremi. Thyroid.CDE at the site boundary.

Thus, 2.3 3E+06 j4Ci/sec is liniting release rate based on the 50 mrem Thyroid CDE EAL initiating conditin.The iimiting release rate threshold values for the Site Area Emergency and General Emergencies are multiples of 10 and 100 of the Alert release rate threshold value.These calculations can be found in Attachment 2.7.1.6 Monitor Response.The unit vent noble gas monitor is calibrated to Xe- 133. Monitor efficiencies relative to Xe4133 by nuclide are listed in ODCM Table B3-2. To find the monitor reading associated with each limiting release, the noble gas concefitriations must be multiplied by the monitor response and suimned. Table 4.6 shows the indicated response of the unit (.

Radiological Release Thresholds CALC. NO. STPNOCOI3-CALC-002 E N E R CO N for Emergency Action Levels REV.l f ttyo Attachment I PAGE NO. 23 of 49 vent noble gas monitor by nuclide and Equation 7.1.5.1 shows how the monitor response was calculated.

n Monitor Response = Ci

• Ret Equation 71.5.1 Where Ci = concentration of nuclide i (LCi/cc)Ret = monitor response to nuclide i (ptCi/cc)x,.

1 3 3 equivalent In the case of an Alert, the 2.33E+06 pCi/sec release rate will read as 1.57E+06 pCi/sec on the monitor. Kr-83m does not have an indicated monitor response coefficient.

Because Kr-83m is only 1.34% of the noble gases and does not contribute to the dose calculation, its exclusion is acceptable.

This again is a linear correlation and the SAE and GE scale by factors of 10 and 100 respectively.

These calculations can be found in Attachment 2.7.2 Main Steam Line Monitors 7.2.1 X/Q Since the atmospheric dispersion is independent of nuclide inventory or release rate and the close proximity of the releases, the X/Q value will be the same for a main steam line release as it is for a unit vent release. This assumption is also taken by STAMPEDE and outline in Assumption 6.3.7.2.2 Nuclide Inventory-er -KI'0-Z'--TP0 it Tie release path is the main steam iine w.tth a steam generator tube rupture, the nuclide inventory should be 100% noble gas and 0.2% of the iodine from the reactor coolant.The secondary steam concentration for noble gases and iodine after a steam generator tube rupture are taken from TGX/THX 3-1 (Reference 5.4). Values for the reactor coolant inventory are listed in table 4.3. All of the noble gases are used and the iodine concentration from the coolant inventory is scaled to total 0.2% of iodine in the-total coolant inventory.

These inventories are then normalized to one. These values are listed in Table 4.3.7.2.3 Dose Conversion Factors The dose conversion factors used are found in Tables 4.4 and 4.5, taken from tables 5-1, 5-2 in EPA 400R92001.

CALC, NO. STPNOC013-CALC-002 Radiological Release Thresholds

______________

ENERC0 N for Emergency Action Levels v. 1 V-i ft Attachment 1 PAGE NO, 24 of 49 A 7.2.4 Decay. Time, One hour of decay is incorporated to the monitor response due to the release initiating one hour after reactor shutdown.

Decay is also incorporated for the duration of the migration time. The total decay time is one hour plus the reciprocal of wind speed, or 1.07575 hours.7.2.5 Dose Calculations Equation 7.1.5.1 applies to the release from the main steam lines. The main stem line flow rate is Used instead of the unit vent flow rate for the valuie F. The main steam line flow rate was calculated in Equation 7.1.2.2 of the STAMPEDE CALCULATIONS section of this document as 2.79E+06 cc/sec.The Alert EAL thieshold is:l 0 or 50 rerem Thyroid CDE at the site.bo0undary (Table.4.2).'

thod 1.1.5.1 to caloulateTEDE with the appropriate conyetsidn factdors, a titi of shutdown of 4.10 iiCi/cc would" e:. in 6.ee.689 ,TBDE at, the ift ft "tem line'i0 Rv was open for an mrm TroiU 's steam line concentration Iciiate ThridCDR results in 50 at...

I. d--Thb-steam lifid-Chnc~ritratibons't the' time -of -shutdown for theý Site Area-Emergency and General Emergencies arenultiple' of 10 and 100 of the Alert. Since the correlation between release concentration and dose is linear, -values for the: siteam line concentration at time of shutdown Are 41.0 ad: 410 4iCi/cc for the SAE and GE respectively.

Both are also limited by Thyroid CDE.... ..These calculations can be f6und in Attachrient 2.7.2.6 Monitor Response main steam line Monitor is adjacentto the tnain.steam line, significant shielding takes place between and monitor. STP calculation NC-9011 Revision c2 alculates a conversion factor for the main steam lines for a noble gas inventory which is incorporated into'the monitor readout. No monitor response needs to be calculated.

The concentration of the main steam line one hour after shutdown given a concentration of 4.10 1 Cikc timieof sli td'wn is 3.90 I fCi/c". This "aleu~iaior" is also fourid in: Attaerhfrt2.

9 dditio eil iininit eings foj. the SAB and oEone hour after shutdown are 39,.0 and 390 gCi/cc respectively.

Thdse values are the thresholds for the main steam line monitor.

~Radiological Release Thresholds CALC. STPNOC013-CALC-002 E N E RCO N for Emergency Action Levels REV. I Attachment 2 PAGE NO. 25 of 49 Table A2- 1: Unusual Event Emergency Calculations

.I i S1.40E+05 I 2.79E+06 I"I 5.OOE-02 Table A2-2: Input Values for Calculations 1 -'.4Uh-UO 4Uk~-tJ ~ I ibOUl 11 1 V.42 1 / I.I i12-L .hI I ýk. Iuý-o4 1 1. /YZ-U.Z I ix/:ý/.) I i I Radiological Release Thresholds CALCNO. STPNOC013-002-EN 4:N.- ..C for Emergency Action L REV. I 0 Attachmnent 2 L ~~PAGE NO. 26of4 Table A2-3: Cal culafions for Boundary Concentrations and TEDE dose dueto Unit Vent Release 1i~ l.OE+105!

1.12E-03 216-5 :iZE0 .793-08 1.93E-12 '2.78E-09 5.30B+04 1.47B.-03 1-132 1-133 1-135 1-135 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-13 Im Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 1.506+05 -1.5-3E-03I 2.20E+05:2.25-0 2.40E+05 2A5E-03 2.00E+05 2.05E-03 1.30E+06 1.33E-02 2.90E+06 2.97E-02 3.70E+05 3.78E-03 5.50E+06 5.62E-02 7.80E+06 7.98E-02 9.50E+06 9.72E-02 I.1OE+05 1.12E-03: 6.80E+05 6.95E-03 2.20E+07 2.25E-01 4.20E+06 4.30E-02 5.50E+06 5.62E-02-1.90E+07 1.94E-01: 1.80E+07 1.84E-01.3, 5 6 3 7 9 1 1 2, 1, 1.4 1:4 77E-05 28E-04 33E-04 33-05 39&-03 97E-03-40E-031 76E-05 71E,04,.55E -03 06F,03 39E-03*79E-03, 54tý3-03 1.015-03 1.01E-03 1.01E-03 1.01E-03 1.011E-03 1.011-03 1.0 1-03.1.013-03 1.O,1-03 1.O1E-03 1.013-03 1.OIE-03-1-01B,03: 1.01E-03,* .813-48 2.383+00 5.61E-08 2.03E0-01; 5.113-08 6.61E400 3,31E-07 1.83E+0.0 7,46F0-7 ý4.48+00 9.42E-09 9.40E+04 1.40E-06 1.273+00 1.99B-06 2.8431E-00 2.42E-06, 5.10_-02 2.79E&08:

2.83-F_02 5,61E-O0&

1.273+02 1.07E-06 -2.60E-01 1 .40E-06 9.08E+00'4.83.3-06 1 ,6.3W-02 4'593-06.

.23-6E-01 2.79AE-08' 4.56E-108 21,21E-07 6.27E-07 9A42&08.7.79E-07.1.53E-06'1.083 IM7E-07 5.57E-06 6.09E-08 1.29E-06 4,06E-171 1.0 (Y07, 1-.50E +04 3.10E+03 8.1OE+03 9.30E+01 1.30E+00 5.10E+02 1.30E+03 1.20E+03 2.50E+02 1.4OE+02:1.0E+02 7.20E+02 5.30E+04 4.90E+03 1.50E+04" 1.37E-04 8.11E-04-8.09E-05 3.70E-04 O.OOE+O0 5.833E-05 1.22E-07 3.97E-04 1.99E-03 1.30E-09 1.36F-07 2.90E-06 1.1-E-04 1.52E-05 1.81E-04 4.47E-09 1.40E-04 Radiological Release Thresholds taLC. NO. STPNOCO13-CALC-002 0 N E CO N for Emergency Action Levels REv. 1 F-1-CE't da Attachment 2 i PAGE NO. 27 of49 Cs-134 3.70E+01 3.78E-07 9.33E-09 L.O0E-03 9.42E-12 1.80E+04 9.42E-12 Cs-137 Te132 Mo99 Rul03 Rul06 Zr95 Lal40 Ce144 Ce-141 Sr89 Sr9o 2.90E+01 4.80E+00 1.22E+01 8.80E-03 2.90E-03 1.10E-02 1.90E-02 7.40E-03 1.00E-02 6.40E-02 3.20E-03 2.97E-07 4.91E-08 1.25E-07 9.00E-11 2.97-E- 11 1U12E-10 1.94E-10 7.57E3-11 1.02E-10 6.55E-10 3.27E-11 7.3: 1.2'3.0: 222 7.3: 2.7 4.7!1.81 2.51 1.61 8.0 E-09 E-09 E-09 6-12 E-13 E-12 B-12 E-12 B-12 E-11 E-13 1.01p-03 L.OIE-03 1.01E-03 1.01E-03 1.01E-03 I.OIE-03 I.01E-03 1.01E-03 1.013-03 1.01E-03 L.O1E-03 7.40E-12 1.22E-12 3.1 1E-12 2.24E-15 7-40E-16 2.79E-15 4.83E-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16 ,2.60E+05 7.79E+01 6.62E4-01 9.44E+02 8.84E+03 1.55E+03 4.03E+01 ,6.82E+03 7.77E+02;1.21E+03 2.50E+05 7.40E-12 1.21,-12 3.08E-12 2.24E-15 7.40E-16 2.79E-15 4.753-15 1.89E-15 2.54E-15 1.63E-14 8.15E-16 6.30E+04 4.10E+04 1.20E+04 5.20E+03 1.30E+04 5.70E+05 3.20E.+04 1.10E+04 4.50E+05 1. 10E+04 5.00E+04 1.60E+06 5.933E-07 3.03E-07 1.45E-08 1.60E-08 2.91E-11 4.22E-10 8.93E-11 5.22E-11 8.49E-10 2.79E-11 8.16E-10 1.30E-09 PrI~EDbeQ 5.7E0 Radiological Release Thresholds CALC. NO. STPNOa13-CALr-002 SE E *"for Emergency Action Levels REV. I Attachment 2 PAGE NO. 28 of 49 Table A2-4: Thyroid Dose Calculation for Unit Vent Release 1-131 1-132 1-133 1-134 1-135 2.78E-08 2.79E -08 5.40E-08 2.61E-08 4.56t-108 1.30E+06 7.70E+03 2.20E4+05 1 .3013+03 2.15E-04 1 .19E-02 3.39E-05 11 -'7"Tl' A-3 iTable A2-5': Unit Vent Monitor Response to Nuclide Inventory Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-I31m Xe-133m Xe- 133 Xe-135m Xe-135 Xe-137 Xe-138 L28E-04 1.83E+0 ' 2,25E-04 r.33E-04 4.48E+00 6.28E-04 1.9 033E-05 9A.OE+04 9.33E-05 2.4..39F_03 1.27E+00 8.03E-04 2.8 S975-03 2.84E+00 1.54E-03 23.40E-03.76E-05.71E-04-.55E-03.06E-03.39E-03.79E-03-.54E-03 5.10E-02 2.83E+02'5.42E.+01 1.27E3+02-2.601B-01 9.08E+00 6.38E-02 2.36E-01 3,00E-09 2.76F-05 1.69E-04 5.5213-03 1.28E-03 9. 15E-08 2.4 1E-04 2.8 0.015:0.14 1 0-.042, 2.5 2.8 2.8 0.OOE400 1. 19E-03 2.245-04 2.25E-03 3.55E-03 8AOE-09 4.13E-07 2-37E,051 5.52B-03",-3.10B-066 3.21E-03 2.56E-07 6.74E-04 Monitor Reading: (uCi/Ce) (uCi/sec)

N Radiological Release Thresholds CALC. NO. STNOC013-CALC-002 O ENERC0 N for Emergency Action Levels REV. 1 E-dkn--- ,tr E Attachment 2 PAGE NO. 29 of 49 Table A2-6: Input for Main Steam Line Release Calculation I 1-1-11 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe- 137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 O. 1 Un-Ul 8.61E-02 1.OOE-01 1.861E-02 2.73E-01 2.80E+00 2.40E+02 4.20E+00 7.60E+00 4.0013-01 1.60E-01 5.80E-01 3.70E-0F1 7.60E+00 1.50E+00 9.80E-01 2.80E+00 8.40E-02/.LOt_,-4Uf 3.19E-04 3.72E,-04 6.92E-05 1.01E-03 1.04E-02 8.90E-01 1.56E-02 2.82E-02 1.48E-03 5.93E-04 2.15E-03 1.37E-03 2.82E-02 5.56E-03 3.63E-03 1.04F-02 3.12E-04 Y./I / Jr-U'f 1.31E-03 1.53E-03 2.84E-04 4.14E-03 4.26E-02 3.65E+00 6.40E-02 1.16E-01 6.07E-03 2.43E-03 8.82E-03 5.62E,-03 1.16E-01 2.28E-02 1.49E-02 4.26E-02 1.28E-03 2.Y8OD53E2-U5 2.9853E-05 2.9853E-05 2.98531E-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853E3-05 2.9853E-05 2.9853E-05 2.9853E-05 2.9853F_,05 2.9853E-05 2.9853E-05 2.985313-05 2.9853E-05 2.9853E-05 2.985313-05

/1. 3.90E'08 4.5513-108 8.47E1'09 1 .24E-P07 1.27F 4 06 1.09E_404 1.91E-06 3.45E466 1.81E-:07 7.26-F,08 2.63E707 1.68E307 3.4 5 EiO 6 6.81B-07 4.441:07 1.27E1406 3.82E-08 1 .Y .iT.vz 2.38E+00 2.03E+01 8.77E,-01 6.61E+00 2.83E+02 5.42E+01 1.27E+02 2.60E-01 9.08E+00 6.38E-02 2.36-E-01 1.83E+00 4.48E+00 9.40E+04 1.27E+00 2.84E+00 5.10E-02 L. IOt.-UJ 2.85E-08 439E-08 3.62E-09 1.1013-07 1.27E-06 1.07E-04 1.90E-06 1.9613-07 1.67E-07 6.10E-13 1.12E-08 1,12E-07 2.92E-06 6.81E-07 2.47E-07 9.79E-07 1.71E-14 4.90E+03 1.5013+04 3.10E+03 8.1013+03 4.90E+00 2.00E+01 1.70E+01 1.40E+02 2.50E+02 1.40E+02 7.20E+02 I .1for--Ji 1.40E-04 6.58E-04 1.12E-05 8.95E-04 6.22E-06 2.1513-03 3.23E-05 2.75E-05 4.1713-05 8.53E-11 8.04E-06 0.0013+00 3.80E-06 6.337E-05 1.26E-04 1-2713-03 2.05E-11 1.3013+00 9.301E+01 5.10E+02 1.30E+03 1.2013+03 f-To6tal1Dosej

ý:6489E43,*Release Constant = X/Q

• duration

• release rat Radiological Release Thresholds CA. NO. STPNOCO13-CAULC0

IEN for Emergency Action Levels REV. I_f f r-.*aMd Attachment 2 PAGE NO. 30 of 49 Eable A2-8 Maia Steam Line ReleaseThyroid Dose Calculation 1-131 1-132 1-133 1-434-I-135 2.76-08-4;39E-08 1AIOE-07 1L30E+06 7.70E+'03 1 *30E4+03'-3.SOE0B 3.58E,02: 2.20E-04 91.66E,03 4:71E-06.4.2OBM3-I Radiological Release Thresholds CALC, NO. STPNOC013-CALC-002 0 E NE R CO N for Emergency Action Levels REV. .

Attachment 2 PAGE NO. 31 of 49 Table A2-9: Main Steam Line Reading at Release 1-131 1-132 1-133 1-134 1-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 9.27E-04 1.31E-03 1.53E-03 2.84E-04 4.14E-03 4.26E-02 3.65E+00 6.40E-02 1.1 6E-0 1 6.07E-03 2.43E-03 8.82E-03 5.62E-03 1.16E-01 2.28E-02 1.49E-02 4.26E-02 2.38E+00 2.03E+01 8.77E&01 6.6 1E+00 2.83E+02 5.42E+01 1.27E+02 2.60E-01 9.08E+00 6.38E-02 2.36E-01 1.83E+00_4.48E+00 9.40E+04 1.27E+00 2.84E+00 ,; i np..A'9.77E-04 1.47E-03 1.29E-04 3.73E-03 4.25E-02 3.60E+00 6.36E-02 8.04E-03 5.62E-03 4.65E-08 4.67E-04 3.85E-03 9.90E-02 2.28E-02 8.62E-03 3.34E-02 1 Anp-nO CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds E NEt--RC,0 N for Emergency Action LevelsNRV 2$.1rn-vjFc~vyd Attachment 3 PAGE NOý 32 of 49 C DRILL'STA flY~srT!AiuNmd ImifoufaDRL" ."3 .. .." --3 f I " t 1 " I ' * ....." ' " *.U...- .....---'-:-: -:- --I

• tyss tz~ m Ltaln!"..Id l-. Stan ux- ie b "wi Cm Ese a a .'u: '

5:24.-~u

• ... ..., , : :,G-. ..... .,. * .~LJ, (NOE XEQIS OI..I~ ..3*::.. n li... : .. 1. ,., .. .Q -..-Il: I3 0 U .....3.0 IT-I- AM=03

• 0...lt 5123e01*r I .. .i- .f.. .. ..* .v : : ..!.: PAMMIICAI* i t iissaoi Ln-419- 31349 7M4*- i.CECU* .IS: ' 25. .. 034t~" m-lUk *rOS 3s33.0 1*45321 k-SUM Ia-lit 1*-SUIt It-nt.ZK4,.113-9H " 2139430S 1923410 It-Nit 2104034 124M1U1A01-459M C

CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds SE NE RCO N for Emergency Action Levels REv. 1 Excetnce-Evyepet+d.

EMyday Attachment 3 PAGE NO. 33 of 49 DRILL s EResxut IorflorD IL D. r R ,aWMII PafJ cI DR 4L Da1m- ,nA SM lg~u n~s~x I"C.0D 0-5 10.0~2.0 Dhs, zc-2.0 7-5 1M.M.0 1.o 5.-.175 025&34 IL45 1:31 CNNI Vain 32513007 uLn?-M tUIB-IX XIBM=7373H-W0 3.54%M00 2.4IR-00 tMUNEO IYwraunwchflI Body%obi. gs3 ZMAu QBM aom GMOD 1sasrdes.Wha&BoAdy sAtCIn Mtonau.4mml amp Gas0 0101 aml amOI am30 OaD0 IPAr.D= atfesI IWrudt lyM....(.Th~m'rj

.-..... ,i-O.(l6X ami amo OL137 oMI'wig 9043 aom UKOI Q010 TM bdi" CUR estaru+Inftin Tyc ami CMr CMl 0101 amCo CA00 0"i 0.137 t0oM 12A7001331-0814 CALC. NO. STPNOC013-CALC-002 E8N.E.RCON 0M/jiie- wr yfl~ ft.yd Radiological Release Thresholds for Emergency Action Levels Attachment 3 REV. 1 j C PAGE NO. 34 of 49 DRILL 'Itt 1~1 DRILL us.. jn Ad Ofahlae~ a (r .. ..'" t Dl " !n.] ltl ... :" uitilee.ThUR 0A20!6 t00 0.001 MOOG0 CUE 0.451 HIS CLOW 02001 A General Emergency Rquires a Prutective Actiah Recommendation EVACUATE ZONE45 .. , SHELTER IN PLAkCE ZONB<SY 2 AFFECTED DOWNWIND SEOYOE& R, A, B AllRmann Zoned CIO ladoces AndMonitor EAS Ralio tafti&at Based on ifse Rate Vrojniono"f!.MrewftiIrmmiersiool Whoelaflody Noble Gas Gamma) a[ &6e Site BoundadyC Mile)rpTiI5Tintateq rlne i hegayltiteinitaig~niia~(AflERT)bams beein met.C PERYOMMEDBY RRVENU BYP Rad &kuager/Rad~olugkul Bfrctcr.1287(01S 24,4tM DatedTine 1:24:-2PM

(.

CALC. NO. STPNOCO3-CALCw002 Radiological Release Thresholds REV. I E E R C O N for Emergency Action Levels REv.

pqed: EvrYday Attachment 3 PAGE NO. 35 of 49 EUser Supplied 1forao DRILL D R. ........ e--a---mlA33 90, i 20 llufl-r U112013 0754. UwnrSmmnnliSbiin~I GImMaM MtardawsdD~bbph:

Qonhkayumdtt 112 udi Sbufiythsr

'D-Nerfrsr Mfmd)FaSfCT*Fwbsmz.

Biam Samm"MR6:

i-Mrwh IbatrhMtddmDatfihm*:

IZIWI*1MSOC BAlems.t"MieDaWm:

12fl19O3 fl NucN& 3tttur K&&Qme Ca late -jmetdno~e es NOMEGAS Nwi&t uQMK ionm Nndi& ncusef PARTICUAUM RuM&& mis/sc icr-am wk-M: me-lit X&-ibM 3ra-lit MB-liT 25SE4001 425B-MO 123-+00 i n-W 9-131: 1-133: 1-133: 322BM 415*OCI Is-Wi: lb!M-IU: En-IS: omseeoo O.021t0M 0.033B+=oAm4*Ex 0AS+01W 0 t2/t1I37i5519AM CALC. NO. STPNOC013-CALC-002 Radiological Release Thrffesholds E N E R CO: N for Emergency Action Levels .Attachment 3 PAGE NO. 36 of 49 C~r DRILL,;! S.TAIWLDE Eaw ,I°,U~~ tDR ,,IL lW IMt 2lftLW4 U1** tnfssw,'",: .. .' .. .' ..":...* .r .7 .... "' .'.0S SID 15 110 2U1 05'CD.20k 75 1.0 75 200 aam tam 0non 000 ram~x ann t ~ A VrIlA, tao Id ZQ (6w+M¶ , idm awl0 o4mE10 t*3d4Q and420 aftoemA'..

7.37fltM&IS. NA4ECtai w Issa 3SS38. am-5C0 CsmLOW tem&4 Ade C

CALC. NO. STPNOC0 I 3-CALC-002 Radiological Release Thresholds ENE RCO N for Emergency Action Levels REv.fi ,lf ,FrWI-,E.yd.

Attachment 3 PAGE NO. 37 of 49 D PILL ix7.3 9DW1 Fa~f DRINLL 11elhod IPripfrtiui STAXLPED RESULTS.Rdwa Rslti 1.-4-71t9tee 0ffdteDwe.rojertio (rem~r I =He Imflea 6mil"a 10miles TMD 0017 0.103~ (Loo1 0=00 C1DE 0.050 O.017 0.00 10DI0 Praju ai de~s~1.0 hwa A General Emergency Requires a Protective Action Recommendation EVA-CUATE ZONE(S)i 1 SHMELIE WN PLACE ZONE(S): 2 AMFECTf.D DOWNWIND SE&TORS& R, A. B -All Rmaining Zones Go, Inhooy And Moir EAS Radio Station 13ased an a Dons Rate Projection~

of > 3 iream/li (Immesion Whole Body Noble Ons Garmna) at &es Site Boizadary (I Mile) for 15 imintes or longer flie Eegency Canflcafian Initiating Coudition RAI (ALERI) kas been met.pEffG1MM By: 121UJ2013 7:55:14AM DoteTim.Dirb Ifi-ni i 1LIM2013 7:54-42AM CALC. NO. STPNOCO13-CALC-002 Radiological Release Thresholds

_______________

EN t Ct f ON r_______1 for Emergency Action Levels&tAWdeP .Ve a P.Y.Da Attachment 3 PAGE NO. 38 of 49 C DRILL: STA- E Us -man OR DRI"Da "."-ne " I 3 15. 5 *" ".. : ::'"..: t v. ...tkmm: d ;d : dtZ itfh m~und~d~Aoz 130 tgren k uw-ssn...1sbwl*y .z...ff.,ly .., .'*: htmi -tsrd...I..tet.

....

i Knf llflflnSU.'2S.

..K k , tat ae lt.llall .......ddns* uiilo"t" "a Pd cl " u * .ph * :.* " ." W S:WL. , f. lCA.S".: '. s"""""..' " " n'. 1 :7 ' "1& : e ",": C wonc'S WicKS u.Okec E&43M. IZSIIJS Dt4K 1035*05 n-KMt 7~O5iCG5 n-47~ 9D324015 fl4& 1.135+0W n-it flnw n-flat 3135*00*n-ut EZZ#OM fl-lint 1935*005 fl~l3t iAm4WE fl-hiM. tIS*C04 9.704101 a-at atmios mm IniSh tC i-ill: 312 s-rn: iii s-rn: &04&134: 1W s-Ut Ml*i-1U: 1031.A1 I". .Cr..:,.-t' 0 : 253 M." 5*0 5:.3.R1~f,.t03 l.O .t.B 313E40 J 12f17A013 32533. d (

CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds REV._I EN ERC0 N for Emergency Action Levels v'6wcellnce..-E t.pmJ&L vE.ydoy Attachment 3 PAGE NO. 39 of 49 DRILL TA PEDResults Iufrma~tion D IL nwrmt-n- IMflawO 15~25 U ae dVt5j m.5D 7.5 Ica0 0.0 0.23 1:31 C~s-0e)3MIPA07 L54IRM0 cmqw 7.3733WO0 7.44113-M 910mma ftI,'bmca 05 1.0 20.IsD 260 200 2IMaa , t Oam 0.93 0.012 aim mm Wine CDR exuramd~iouy7d UP OAS0 OLD41 6.602 0014 0.001 am00 u MDth~R Ifm d 018) 1364 0.08) 0.510 mim 0L176 0.023 0.021 Q.0m0O014 Ol00 0.00 I

CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds

.N E R C 0 N for Ernergency Action Levels. P NO o4-Attachment 3 PAGE*NO. 40 of 49 C RESUILTS 3.~~e~~ia .n~~dy 32 1 i~hr Rehi 1:n~eRM IM+ .E4OO1 f TID 0,060 'Moni ow0 00 E0-510 0.176 am4 a014 A Genirnl Em~ergency Requires a Pro-tective Action Recoinmendatibi EVACUATE ZqIIES)w 3 SH2ILTER IN PLACE ZONV,(S3p 2 AMFCTED DOWNWMhfl SE&tORS: R, A, 13 AU Remiiin Zwm G61n6~ ManitarlAS Rsdioqtiatn IBased diin Site ftuiw~Ai (I U):Diefoae Prnjedtiwi 0.1 remi TEDE And(% 0.5 mm .nThyroxid CDEbEtk 3"yCifucahionfifiafing ConA~tiongSJ (SITE AREA EMERGENCY) h"h been met.C PWOMDBY M'JIW, By1 Daw/Tih 12q7J2013S:2U.IFM C'

CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds REV. I E E R CO N for Emergency Action Levels REV. I Attachment 3 PAGE NO. 41 of 49 DRILL STAMPEDE User Suppflied Iormatian DRILL03 rA1 DRILL bzlfqe12fD013125 USffNwan SternUms SkAte Cmmmtred Qad Sutkm~a ftffilzwIfrDU Quhmdkdiumbli-a LOD 1Skaes wknrj&hsiffctstMFmliulffM!

S&T d vsytlasrnk~

?D-%5ers&u=I~W0WZGCAslqtrsmerate:Us&

uik NXMEACAS Nscli& tahsc Nmdhb PEEU-LATE WaS& sOCUer E, r-Uk: wgIJM xer-lan&-I5..7--Mt MISt 1.15f.005 5.S1tE',,tl 25113&0(15 991E+"1 392E43C2 3.10+0I05 1218 4 1 1-In: 1-133 I-LIT I-LIM: 103am0 3204MW 41PM*M0 4.22H+=B 124914M5 Cs-l": Ce-hil: Ia-UO: SrIY-lt Sr-%: 0.0msf0M 0.ll020=O.C(E-I!ODD a.umcOo GI.OZ+ODD aW340-10l 0D214-4E0l

-I rmrn7Ms329D3 Rd CALC. NO. STPNOC0I3-CALC-002 Radiological Release Thresholds________________, N E ,RC N for Emergency Action Levels ry.

ft Attachment 3 PAGE NO. 42 of 49 (I nL~et3WZE 12fltAM 3 15:21 ~ ine um raI~e 2D~.73 05 10.0D 7.5.20D 0~5 alp3 Ponl G.A1 um CWQA IU I. ~ Q aim 6.0ft Q001 DO~S30 3S511.WFf CUEtO~pa4i d..X 0.001 OW C L2f7MW03 332tS3W C CALC. NO. STPNOCOI3-CALC.002 Radiological Release Thresholds REV,_I E N E R C 0 N for Emergency Action Levels V F c-Fryqt. Fwdoy. Attachment 3 PAGE NO. 43 of 49 DRILL sTAET. DE Results Inormti-on DRILL DRHT J Rwifi=.10.3 90MM011 pap ~2of2D NL RESILTS 1Vad.velartity I32miAr liindDirtetims, ISO Methed~fPr~4edew STA1WThE Mei~i.Ratet I.2O11+008izCAM Offifte Dose Profrctiou (rem) I I alls 2mit" 6 5Infes 1m TEDE fX072 O 0.006 0.002 CME 0.506 0.175 (1041 0.013 Pzejeded dxumieofrelase:

L.0howj A General Emergency Reqjuires a Protective Action, Recommuendation EVACUATE ZONE(S: I SHELTER IN PLACE ZONE(S: 2*AFFECTED DOWNWVIND SECTOR&: R 5 A, B -*.All Remmining Zone Go Indoors And MomitorEAS Radio Staton Based on a Site Hlm:dary (I Mile) Dose reim TEDE emd/or U rem Thyroid CDE the Emergemy Clastfication RS1 (SITE AREA EMERGENC11) h been met II PEIFORAME BYt 12i11V213 3 2M0 IM Datefffhna Nama MM1EWU BY-1211712013 3:2&53PM SE hN. .R C.0. N Radiological Release Thresholds for Emergency Action Levels Attachment 3 CALC. NO. STPNOC013-CALC-002 REV. I PAGE NO. 44 of49 C DRIL ~UOO ~ R-ILL.DZWI~U1&42IU12W3 1S25 tap. ns~r. ijum w~tt~wt*b .R, ltwb aw d

• udai- " ...

alm l, ,llf , , -&24, Nds' elstzi nr, 1,,*I hhl,:.' -13:21 a~tnba1Idsse..13um.i.::...PO......

Nncl&lldnrc Opisskry C CI kO ..... .ELEGAS 9 t=. ..wows.&r-NM I~MXI6 gbm"6 KrBlk 73S4(X1 it-..ut X442 ',, ,MM.XrsZn M2.*I IifluWsec 10131K PAZI1CCLIJZ WudMe uCitse~ Ridit nd/sc 1-ni: 143k.1-135: i-nt 1-lit 3.050M5 SxmIa05.*ilkl4t..,,,. i , ,,] fl,.j :, Euffibil Brlfl,* it-S-i, 3120+01 1352+001-3MIu, 12A79,013 C CALC. NO. STPNOCOI 3-CALC-002 Radiological ReleaseThresholds E N ERCO N for Emergency Action Levels REV. I je Attachment 3 PAGE NO. 45 0 f49 DRILL STAMEDE.Resit LdI~mnafiouDRL D2gwtoue IMf7rA01 ISM~ Ir Mme- m2i~ Vmt Gkiad r~L.UUIIeii 7-5 10.0 0.5 10 75 2QD Pkwumaly.11ma 023 034 OAS 1:31 CzVQ Vsfta 17MYN-L009 37DMM CWYQ_ mm 7.373007 3X1454X1 7-44IER00 MOM=~e mmliam~vuE 61Y~03322 0.117 0.010 0.01 BubbI gm, frm. (rams)0.332 0.117 0.010 0.w1 PA, ,. &, if~UUTZ.L5SB omi 41017 CLOW0 0411 0I~ 0.411 ow13 0.214 OW01 OL135 OM6 am0 M73332t25P CALC. NO. STPNOC013-CALC-002 SRadiological Release Thresholds EN R Ofor Emergency Action Levels Attachment 3 PAGE NO. 46 of 49 (D LL STAMPDE RWl ORRILLlo~~10, 7777777ESUL77 TEE060.1 1)410 G.05M .017 CTZEL762, Pxe~d~PC~!:

Oi~f OL411 135 A General Emea'geiit-y .R~jire a PrateciVeefAion Reeommrned'toa EVACUATfE ZONE(St1 2 SIIELER IN PLACE ZONE(S)- 6,11 AFFECTE D DOWNjVVDSECTORS:

IR, A& 9 AJlRmIna Zones G Jdoor Axz&'b bonitorM5s Thdic SiaTdio Based aunaSie omduydar (Mile) Dose ftbleetiou!!

I rem.TWIE wndor5xuzmThyxoid CDE the Rbýargecy pCaiffmaati~nik~i~

>Condioi~Gi (GEN~ERAL EMER IEXCV) his beeu met C REV1MWfl Hp~U212f113 9:26:33PM~AIfl~4ut~fllGgVt~J .uir~ta~ UaWIbn?12J17(~O13 3~&25PM CALC. NO. STPNOCOI3-CALC-002 Radiological Release Thresholds CALC. NO. STPNOC013-CALC-002 E N E R C O N for Emergency Action Levels REV. I fxcellr- tFwrypqed.

Evn ft. Attachment 3 PAGE NO. 47 of 49 DRIL L STAID Usrwuied Taformation DRLDRILL, D 1dr 2/1fl3 15i30 Uwmae*:. eml Ceki Cashll: litwohagEtuiihtT lt::3a 3*Lmr-iiqftzsMilw One)Qm=ShaDmMbmt:

U2 ndftt RnleIMMStart Dstfluu: 121Th03 15:38 WOXZG&S MNust .cisec Nudik umer X&M&Xr-32M n-Np?: Xe-iiflSSg2B.r 4.1,,MEI I -'1238+M5 5434: I-IN: 329nI 0005+O*49MA 4243 CeiS4:f Os,-Ill: Oe#Pr-U&Ce-li]: SrM: mSr-" 0J10CM0X3 O~aEI+ornMR+=c wrstcoo DOMRMD 121701313:0:5O.M CALC. NO. STPNOC013-CALC-002 Radiological Release Thresholds N E.RCO .N " for Emergency Action Levels r ..ft-- 0rA ,... Mlyda. Attachment 3 PAGE NO. 48 of 49 J DRILL". R "X .:d" D_ ';i.nrfm I11ý?OL : .. J. : Th~ S~a~I~* .: ... ...., ... ..... ..... ..... ........ ......7.5 20 1.0 75 1.01 75.IOD w5 OL45 0Ab*.LU3 604,~ODII am0 QW4 0M6 7J7O 4,254o M2S 1347 a=B Am (1MM12013,0.I32M iM C:

CALC. NO. STPNOCO I 3-CALC-002 Radiological Release Thresholds REV._I EN ER CO N for Emergency Action Levels d-- Evyd Attachment 3 PAGE NO. 49 of 49 DRILL TAPDE RsTits Inor' D FILL-CnkeuIRIOU Com'1etq MEULTS WVhOVeIdity, nln2ift-rid~hatdi~mc 180 STAXMPFh Off1, I. O we 2171E Piojededauntonofrelease:

I.Dboux Relean Rntet 120E+009 e~ifsc I0 a woes M0.22 0.134 0.254 1.747 UMll~fI~L40 A 'General Emergency Reqaiires a Protective Acthiou Recommendation EVACUATE ZONE(S): 1, 2 SHELTER. IN PLACE ZONE(&): 56 11 AFF 7 ECTED D)OWNWIND SECTORIS: -R, A,-B.All Rewniding Zones Cki Tnuwss And Monitor EAS Rudii, Staficoi Based oii a Site Boundary (I Mýie) Dose Projectioi

ý, I remn TDEM =&or .5 rem Thyroid CDE the Emergenqy ClassifrntionlInifain*

Crindition RGI (dKMEAL FIdERGENCV) haa been met PUMoM4EDEY; 12fIlMO03 5,30.48 PM MMIVID Wo R~d~fi~wKfla1&

fi 12117J2013 1"3039PM STPEOS UFSAR ('" The particulate channel Is used as part of the Reactor Coolant Pressure Boundary (RCPB) leakage detection system. The sensitivity and response time of this part of the leakage detection system, which is used for monitoring unidentified leakage to the Containment, are sufficient to detect an increase in leakage rate of the equivalent of one gal/min within one hour, Elements of thi monitor, including, the indicator mounted in the RIMS CR cabinet, are designed -and qualified to remain ftmctional following a Safe Shutdown Earthquake (SS), in compliance with RI 1.45. Further information on the RCPB leakage detection system Is.presented in Section 5.2.5.11.5,2.33 Unit Vent Monitor: The unit vent monitor samples the plant vent stack prior to dis&charge to the environment and monitor for particulates, iodine, and noble gases.The unit vent particulate and iodine monitor draws representative ahi samples from the plant vent stack via isokinetic noizles in the stack, and directs them through a moving. filter paper monitored by a shielded beta-sensitive scintillation detector, The sample stream then passes through a charcoal collector,'

where collected iodine is mbnitored by a shielded gamma-sensitive scintillation detector, The sample is then returned to the vent stack.U*4 I I, I 4.I A separate wide-rango gas monitor is provided for the unit vent. The monitor has two isoldh etic-nozzles for sampling during both normal. and accident conditions, The. stack samples pass first through a sample conditioning unit which filters particulates and iodine and may be used to take grab samples, The samples then pass through the shielded detector assembly,.which uses three detectors to cover the complete range required, The low range detector usex:'a beta-sensitive plastic scintillator-photomnultiplier (PM) tube, The mid-range and high-range, detectors use cadmium telluride (CdTM), chlorine-doped, solid-state sensors,:

This wide-range gas monitor satisfies the reqplrements ofNUREO-0737, Item IfLF. I for provisions for sampling plant effluents for Modines and particulates and for noble gas offluents 1rom the plant vent.11.5.2.314 Control. Room Eleoical AuxiliaryBuildinz Ventilation Monitors:

The CR/EAB ventilation nhonitors are Class lB monitors which continuously assess the intake air to the.CR for indicatioi of abnormal airborne radioactivity concentration.

Each monitor assembly is.powered firom a separate electrical power source, In the event of high radiation CR emergency ventilation operation is initiated (Section 7,12), Failure of a monitor. is alarmed in the CR.Each monitor assembly is comprised of a pMxp,. beta-sensitive scintillation detector, four-pi lead ahielding,.

check source,. stainless steel sample gas receiving chamber, and associated electronics, 11,5.2.3.5 Condenser Vacuum Pump Mnitor: Gaseous samples are drawn through an off--iin'systemr-by-a-pnnp-from-the-discharge-ofthe-vaýunm-pump-exhaust-header-of-the-condeýnser, This channel monitors the gase6us sample for radioactivity which would be indicative of an S0 tube leak, allowing reactor coolant to enter the secondary, side fluid; this monitor complements the SOBD monitors in indication of a SO tube leak.' The gaseous radioaotivityý levels am monitored by awhigle detector In a manner similar to the 7nit vent wide range gas monitor, I I'5.2.3.6 pent Fuel Pool Exhaust Monitors:

The SFPB monitors are Class 1E and are identical to the CR/EAB ventilation monitors described in Section 11.5,2.3.4 except that they sample.the exhaust from the FHB.. In the event of high radiation the monitors initiate emergency operation-v 11.5-11-Rovtsion 14

$TPEGS UFSAR of the FHIB HVAC, causing the exhaust air to be filtered prior to release (Section 73.3). Failure of a monitoris alarmed in the CR, 11.5.2.3.7

&CQB Purgo Isolation Monitors:

The RCB/purge isolation monitors are Class 1B monitors that sample the Containment Normal Purge System or the Suplplemrntary Purgo System and are fdentical to the CR/EAB ventilation monitors described In Section 11.5.2,3.4.

In the event of high radiation the monitoi's send signals to the Solid-State Protoetion System (SSPS) for containment ventilation Isolation (Sectlon7.3.), 'Failure of a monitor Is alarmed in the CR.11,5.2.4 Liquid Monitors, Fixed, off-line moditois are provided for continuous detection and. measurement of radioictivity for liquid process streams. The d6stgn parameters for these monitors are summarized in Table 11.5-1. Bach monitor is provided with deminerallzed water for flushing.11,5.24..

1 imR & Devices,

• For each monitor, a sample is drawn from the process line, passed through a shielded sample chamber, through the sample pump and then returned to the system. Each sample pump is capable, of drawing at~lea t one gal/rain of liquid through the monitor.The sample flow rate is controlled by rieans of a manual valve.Each monitorjhas a

alarm.-The monitor IHiot and outlet lines have oompression fiftings.

The sample piping has isolation valves.so that the monitor -skid oarh be isolated and. the sampte. chamber dissembled for deoonTtamination,.

11.5.2.412 Deteoior.Uni't Eadh detector Is a NaI(TI) gamma-sensitive scintillation dotector.

The detectors are designed io remain. Mhly operational over a wide range of temperatures.

If they are exposed to high radiation transients exceeding th'.ohannel range, the channel maintains its operatlon and returns to normal 'inotlonihg when the transients have subsided, Since gm-ama.detectors ate used, oomparison of monitor readoit with the, resuits of grab samples Is possible, The grab samples are counted in the plant multichannel gamma pulse height speotrometer to .chec for proper moritor o~eratlon.

Solenold-op.ratod check sources are proVided to check detector response, S11.5.2.4,3 Rtim (Generator Blowdown ) tojuid 92onitog:

The SG blowdown liquid monitor samples _he liquid-fim_

e bith J._8 b-owdow f4lAafk sampJleis_

continuously mozitored by a shielded, gamma-ena.ltive detector.

Detection of high radiation by this*imonitor alerts the operator to tho possibility of prtimary-to-secondary leakage.In the event of activity aboye tho high alrmn setpoint or monitor failure, the.monitor initiates the automatic

ýfosuro 'f FV-5..19;.

the SG blowdown dlsohaa'ge to neutralization basin isolatlon valve,-- 1'5,2,4,4 Liquid Waste.Processina.System Monitors:

LWPS monitor no. 1 deteots" activity present in the liquid waste effluent being disoharged fitom the waste monitor tanks hi the LWPS. The monitor Is located upstream, of the LWPS diversion valve, FV-4077. Upon Initiation of a high radclWtion or monitor failure alarm, the monitor oauses the valve io automatically divert the back to the waste monitor tanks.11 .512 *Revision 14 STPEGS UFSAR 11.5-13 Revision 14 The sample is drawn from the CCW pump discharge line downstream of the CCW heat exchanger, monitored, and then returned to the CCW surge tank.

11.5.2.4.6 Boron Recycle System Monitor: This monitor is located in the BRS evaporator condensate line downstream of the recycle evaporator condensate filter.

Upon initiation of a high radiation or monitor failure alarm, the monitor initiates changeover of the BRS diversion valve, RCV-4202, causing the BRS condensate to be diverted from the Reactor Makeup Water Storage Tank (RMWST) back to the BRS recycle evaporator feed demineralizers.

11.5.2.4.7 Turbine Generator Building Drain Monitor: This monitor monitors the water in TGB drain sump no. 1. Upon detection of high radiation level or monitor failure, the monitor automatically stops the sump pumps and alarms the condition.

11.5.2.4.8 Failed Fuel Monitor: This monitor takes a sample downstream of the letdown heat exchanger (HX) of the CVCS and acts as a gross-failed fuel detector. The radiation alarms alert the operator to an abnormal increase in gross gamma activity in the CVCS letdown system possibly indicative of fuel cladding failure. Determination of the cause can be made by laboratory analysis.

The sample location provides a letdown sample point prior to filtration and demineralization.

11.5.2.4.9 Condensate Polishing System Monitor: This monitor is located on the discharge of the Condensate Polishing System to the neutralization basin. Upon detection of high radiation or monitor failure, the monitor sends a signal to automatically close FV-5804 to terminate discharge to the basin.

11.5.2.5 Adjacent-to-Line Monitors. Adjacent-to-line (ATL) monitors are used to monitor:

1. GWPS

2. MS System

3. Steam Generator Blowdown System (SGBS)

These monitors are mounted adjacent to the process line and do not require a sample stream to monitor for radioactivity.

11.5.2.5.1 Gaseous Waste Processing System Inlet Monitor: The GWPS inlet monitor employs a gamma (NaI crystal) scintillator/photomultiplier tube combination to measure the radioactivity level of the waste gases entering the GWPS. The monitor is used in conjunction with the GWPS discharge monitor to measure overall effectiveness of the GWPS.

11.5.2.5.2 GWPS Discharge Monitor: This monitor is similar to the GWPS inlet monitor and is installed upstream of the GWPS discharge valve. Upon detection of high radioactivity or monitor failure, the GWPS discharge valve, FV-4671, is automatically closed.

11.5.2.5.3 Main Steam Line Monitors: Each MS line is monitored by an ATL monitor consisting of a Geiger Mueller (GM) tube detector and an ion chamber detector with overlapping

ranges. The detectors are shielded by 3 in. of lead.

STPEGS UPSAR S1.5,2,5,1 Gaseous Waste Proce.ssin System Inlet oMnitor: The GWPS inlet monitor employs a gamma (Nal Crystal) sctntlator/photomultlpller tube combination to measute the radioactivity level of the waste gases entering the OWPN, The monitor is-used in conjunction with the GWPS discharge monitor to. measure overall effectiveness of the GWP S..11.5,2.5.2 GWPS. Disoharge' Mofittor:

This monitor is similar to the GWNS inlet monitor and is installed upstream of the GWPS discharge valve, Upon detection-of high radioaotivi'ty ot monitor falure, the OWPS discharge valve, FV-4671, is automatically closed.1 1.5.2.5.5 Main Steam Line Monitors:

Each MS line is monitored by an ATL monitor consistin~g of a Geiger Mueller (GM) tube detector and an ion chamber detector with overlapping ranges. The detectors are shielded by 3 in, of lead.The monitors are des1gnied to monitor gross gamma activify in the. steam line and provide a.basis for dotermining'posaible atmosph6rio releases flom the MS Power-operated relief valve (PORV), So safety valves, and/or auxiliary feedwater pump turbine.The monitors provide a dose :ate range equivalent to 10"' to i3 GLd/cm 3 xenon.133.

B1ased upon core nvyeetory; the ratio of xonon-133 to other nuclides in the fuel can be detenined, In order to obtain the above. concentrations of xenon- 133 in the main steam line, a large prlmary-.to-secondary leak must be present coincident with a large amount of fuel failure. The presence ofxeinon-133 indicates other radioactive isotopes are present.Using the Tela.tive ratios of isotopes present in the MS line, a computer model for determination of dose rates from these isotopes, detector response curves, the thickness of the MS line, and the geometry of the MS line relative to the detector, the. dose rate equivalent to' MS line concenOtration is obtained, T1he quantity of ridloaotlve effluents released is obtained by multiplying.the'xenon-133 equivalent M..line conoentrations by the isotope ratio tiies the steam release rate, These detecors are safety-related Class 1E and meet ýle requirements of RG 1.97 and NMEREG-0737.

11..2..4 Stam enrafi~ d~~Moh1V~4:Th-esumnitors-areid-entical-to.-the MS line monitors and are adjacent toe SO blowdown 1--s In solat on-Vailve-CiNblte (IVC) .... .-The monitois are used as an aid in-determining the source of SO blowdown radioactivity due to so tube rupture or a large iýrlmary-to-secondary leak.-T-heso-dotetors-Rre-safetyr-rlated-C-lass.-l-E-and-meet-th-e.requirements-of RG4-97-,.1 1.5.2.5.5 Main Steam Line Tigh Energy Gamma (N-16) Montor'n Each i=1n steam line, is monitored by an ATL NaI scintillation detector, These detectors were installed to monitor the status of steam generator, primary to secondary tube leaks and to provide a diagnostli tool for all individuals with steam generator condition, These detectors are designed to detect high i energy gamma activity in the 6 to 7.2 MEV energy range., High energy gamma actvity in the main..steam. lines indicates thepresece of N-16, The level of N.16 in the main steam lines is Used to 11.5-14 Revision 14 RA2 CALC. NO. STPNOCOI3-CALC-005 H EN ER CO N CALCULATION COVER SHEET REV. 1 PAGE NO. 1 of 23 Title: Fuel Handling Accident Monitor Response for Client: South Texas Project Emergency Action Levels Project: STPNOC013 Item Cover Sheet Items Yes No I Does this calculation contain any open assumptions that require confirmation? (If YES, Identify the assumptions)

X 2 Does this calculation serve as an "Alternate Calculation"? (If YES, Identify the design verified calculation.)

Design Verified Calculation No. X 3 Does this calculation Supersede an existing Calculation? (If YES., identify the superseded calculation.)

Superseded Calculation No. X Scope of Revision:

Alter assumptions to reflect the smallest reasonable filel handling accident rather than a design basis fuel-handling accideitT6tal-decayy tiif-asch-nge-d-fro-n-i42ti6ifs to-368 hlur--and

--fi ild-flu ihtan 1.2.Revision Impact on Results: The calculated dose rates decreased due to the increased decay time and decreased failed fulel bundles assumed. This is conservative for the purposes of developing the emergency action levels..Study Calculation Final Calculation Safety-Related Non-Safety Related (Print Name and Sign)Originator:

Caleb Trainor ,. / Date: 3/21/14 Design Verifier:

Chad Cruamer / / Date: 3/21/14 Approver:

Date: 3 21/14 Marvin Morris CALC. NO. STPNOC013-CALC-005 E EN E R C 0 N CALCULATION R REVISION STATUS SHEET REV. 1 PAGE NO. 2 of 23 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 03/04/20 14 Initial Issue 1 3/21/14 Assump tions changed to reflect a smaller fuel handling accident.PAGE REVISION STATUS PAGE NO. REVISION PAGE NO.. REVISION 1-13 1 ATTACHMENT REVISION STATUS ATTACHMENT NO. PAGE NO. REVISION NO. ATTACHMENT NO. PAGE NO. REVISION NO.1 14-22 1 2 23 1 (C C CALC. NO. STPNOC013-CALC-005 CALCULATIONE N E R C 0 N DESIGN VERIFICATION REV. I-'.',, PLAN AND

SUMMARY

SHEET PAGE NO. 3 of 23 Calculation Design Verification Plan: Calculation shall be verified by comparing the documented input with the references and checking the validity of the references for the intended use. As necessary, assumptions shall be evaluated and verified to determine if they are based on sound engineering principles and practices.

Verify the applicable methodology, inputs, results, and conclusions.(Print Name antd Sign for Approval -miark "N4" if not required)Approver:'-

.. .... .... ..Date- 3/21/14 M... ~arvin- MorrTis ..Calculation Design Verification Summary: Design inputs, assumptions, methodology, resuflts and conclusions were evaluated/verified and found to be acceptable.

All comments have been incorporated.

Based On The Above Summary, The Calculation Is Determined To Be Acceptable.(Print NVame and Sign)Design Verifier:

Chad Cramer Date: 3/21/14 Others: Date:

CALC. NO. STPNOC013-CALC-005

• E E C CALCULATIONE N E R CO N DESIGN VERIFICATION REV. I, ,o/c.;r Evey er o CHECKLIST PAGE NO. 4 of 23 Item CHECKLIST ITEMS Yes No N/A 1 Design Inputs -Were the design inputs correctly selected, referenced (latest revision), consistent with the design basis and incorporated in the calculation?

2 Assumptions

-Were the assumptions reasonable and adequately described, V/justified and/or verified, and documented?

3 Quality Assurance

-Were the appropriate QA classification and requirements V/assigned to the calculation?

4 Codes, Standard and Regulatory Requirements

-Were the applicable codes, standards and regulatory requirements, including issue and addenda, properly v identified and their requirements satisfied?

5 Construction and Operating Experience

-Have applicable construction and operating experience been considered?

6 Interfaces

-Have the design interface requirements been satisfied, including interactions with other calculations?

7 Methods -Was the calculation methodology appropriate and properly applied to V/satisfy the calculation objective?

8 Design Outputs -Was the conclusion of the calculation clearly stated, did it correspond directly with the objectives and are the results reasonable compared to V/the inputs?9 Radiation Exposure -Has the calculation properly considered radiation exposure s, to the public and plant personnel?

10 Acceptance Criteria -Are the acceptance criteria incorporated in the calculation sufficient to allow verification that the design requirements have been V satisfactorily accomplished?

11 Computer Software-Is a computer program or software used, and if so, are the V/requirements of CSP 3.02 met?COMMENTS: None..(P" Name and Sign)Desigii Verifier:

Chad Cramer Date: 3/21/14 Others: Date: C I CALC. NO. STPNOC013-CALC-005 , E N E R C a N Fuel Handling Accident Monitor Response for EAL Thresholds REV. I PAGE NO. 5 of 23 Table of Contents 1. P U R P O SE A N D SC O P E ....................................................................................................................

6 2.

SUMMARY

OF RESULTS AND CONCLUSIONS

..................................................................

6 3 .R E F E R E N C E S ....................................................................................................................................

6 4 .A S S U M P T IO N S ..................................................................................................................................

7 5. D E SIG N IN P U T S .............................................................................................................................

10 6. M ETH O D O LO G Y ..........................................................................................................

11 7. C A L C U L A T IO N S ............................................................................................................................

12 A T T A C H M E N T 1 ....................................................................................................................................

14 ATTACHMENT 2 ..............................................................

23 CALC. NO. STPNOCO13-CALC-005E N E c O N Fuel Handling Accident Monitor N E--, R- C, 0- N. , Response for EAL Thresholds REV. I PAGE NO. 6 of 23 1. PURPOSE AND SCOPE The purpose of this calculation is to determine the readings on the monitors RE-8090, RT-8035/8036, RE-8050, and RE-8099 during a fuel handling accident (FHA) at South Texas Project Electric Generating Station (STPEGS).

Monitor RE-8090 is located in the Fuel Handling Building adjacent to the Spent fuel pool. Monitors RT-8035 and RT-8036 are located in the exhaust system directly above the spent fuel pool. Monitors RE-8050 and RE-8099 are located on the 68' elevation of the reactor contaimnent building.

The accident occurs in either the Fuel Handling building (FHB) or the Reactor Containment Building (RCB) 368 hours0.00426 days <br />0.102 hours <br />6.084656e-4 weeks <br />1.40024e-4 months <br /> after shutdown.

The results are used as threshold values for Emergency Action Levels (EALs) in the STP EAL Technical Basis document, which implements the NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors".

Revision 1 of this calculation increased the decay time from 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> to 368 hours0.00426 days <br />0.102 hours <br />6.084656e-4 weeks <br />1.40024e-4 months <br /> and reduced the assumed number of fuel bundles fromn 1.2 to 1.0. This was done to produce conservative values when developing the emergency action levels.2.

SUMMARY

OF RESULTS AND CONCLUSIONS The results of this calculation are listed below.Table 2.1 Monitor Response Location Reading ,. ..........

.. ...... :. < .. ....... : ......... ... ..., ..... ......* ......-_ .RE-8090 1.56 R/hr RT-8035/8936.

3 A.78 Ci/cc RCB3 RE-8050 0.497 R/hr* ,:?.,. -: ., .;. ... : ...;. ...: .;, t .. , .'. ... ... ..". ..; , ' :~RE7-8099.

0.867R/hr.

Readings of these levels or above on the monitors listed will be indicative that a fuel handling accident has occurred.3. REFERENCES

3.1. STPEGS

Procedure 0ERP0O1-ZV-INO1, Emergency Classification, Rev. 10, Draft.3.2. NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors C C CALC. NO. STPNOC013-CALC-005 F-A E N E R C O N Fuel Handling Accident Monitor EV.r.; Response for EAL Thresholds

.~ ..C e,' PAGE NO. 7 of 23 3.3. STPEGS Calculation NC-6508, Fuel Handling Accidents, Revision 0 3.4. STPEGS Calculation NC-6006, Fuel Handling Accident (FHA) in Containment, Rev. 8 3.5. STPEGS Calculation NC-6007, Fuel Handling Accident (FHA) in Fuel Handling Building (FHB), Rev. 8.3.6. STPEGS Drawing 6C189N5007, RCB, Revision 6 3.7. STPEGS Drawing 9C129A81105, Radiation Zones Reactor Containment Building Plan at El. 68' 0", Revision 3 3.8. STPEGS Drawing 5-V-12-9-V-0433, Fuel Handling Building Partial Plan El 68' 0", Revision 7 3.9. STPEGS Drawing 9F-13-9-A-1054, Fuel Handling Building Sections, Revision 4 3.10. STPEGS Drawing 9F12-9-A81115, Radiation Zones Fuel Handling Building Plan at El 68' 0", Revision 2 3.11. MicroShield 6.20 3 12. ENERCON Services, Inc., MicroShield 6.20 Computer Code Verification, STPNOCO 13.3.13. Model RD-52, Offline Gas Detector Technical Infornation.

3.14. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.4. ASSUMPTIONS

4.1. Mixing

Volumes The fuel handling accident results in a release which is assumed to disperse instantaneously throughout the mixing volume. This method of dispersion will result in a conservative final result due to assuming the largest possible dilution volume. While the monitors may see higher concentrations initially, the lowest concentration is after full dispersal.

For the RCB, the buildings volume is assumed to be the entire structure above the 68' 0" operating floor consisting of akcylinder and half sphere and dimensions taken _from Reference

3.6. Using

only the free volume above the fuel handling deck is appropriate because the congestion in the lower containment compartments would delay mixing, this scenario is not conservative but is realistic.

The free volume above the operating floor is over estimated because it does not consider any mass occupying the volume in the building which would create a smaller dilution volume. Figure 4.1.1 shows the RCB dilution volume as it was modeled.

CALC. NO. STPNOC013-CALC-005 O E N E R C o N Fuel Handling Accident Monitor F4E ,ERC-N ,y ,: Response for EAL Thresholds REV. I PAGE NO. 8 of 23 134.9'Figure 4.1.1: RCB Mixing Volume For the FHB case, two volumes were assumed. The room dimensions were taken from References 3.8 and 3.9.The first volume is applicable to the area radiation monitor (ARM) and is calculated to be the entire volume of the FHB above the 68' 0" handling floor. This assumption for the free volume provides conservative results due to the large dilution volume. Figure 4.1.2 shows the FHB dilution volume as it was modeled.Y Figure 4.1.2: FHB Mixing volume (

CALC. NO. STPNOCO13-CALC-005 Fuel Handling Accident Monitor...

a,, Response for EAL Thresholds REV. I PAGE NO. 9 of 23 The volume used for the exhaust monitors is a cubic volume directly above the spent fuel pool and uip to the exhaust vent directly over the pool shown in Reference 3.8. This is the concentration which would be seen by the exhaust monitors and is consistent with the methodology used to calculate the offsite and control room doses for a fuel handling accident contained in Reference 3.3 4.2. Number of Damaged Fuel Rods This calculation assumes one damaged fuel bundle, or 264 rods, and that the same number of fuel rods would be damaged during a fuel handling accident in the RCB or the FHB. This is conservative as the fuel handling accident calculation (Reference 3.3)assumes that 1.2 fuel bundles are damaged (3 14 fuel rods). Reducing the number of fuel rods that are damaged will decrease the dose to the monitors and lower the EAL threshold setpoints.

4.3. Location

of Monitors The MicroShield models assume each monitor to be just outside the source volume and not immersed in the source volume. This is due to the limitations of the modeling capability of MicroShield but is expected to have a negligible effect.4.4. Geometry of RCB For the MicroShield model, the RCB is modeled as an equivalent volume cylinder rather than a cylinder with a dome. Due to the large volume, and the detectors being on the opposite side of the dome, these differences have a negligible effect.4.5. RT-8035 and RT-8036 These monitors measure the activity concentration in the fuel handling building exhaust vents. Since these monitors are calibrated toXe-133_,which is 90% of the release, it is assumed that this is the only activity seen by the monitors.

This will cause the threshold to be slightly lower than expected and is a conservative assumption.

4.6. Radial

Peaking Factor Reference 3.3 uses a worst case scenario to calculate offsite dose and assumes a radial peaking factor of 1.7 which causes the maximum amount of activity to be released.

For this calculation, a peaking factor of 1.0 will be used because this will release less activity and cause a lower monitor reading threshold.

CALC. NO.STPNOC013-CALC-005 NE o N Fuel Handling Accident Monitor E E C Response for EAL Thresholds REV.PAGE NO. 10 of 23 4.7. Decay A decay time of 368 hours0.00426 days <br />0.102 hours <br />6.084656e-4 weeks <br />1.40024e-4 months <br /> is assumed. This models a fuel handling accident at the end of the refueling period which would release the smallest amount of activity.

The time of 368 hours0.00426 days <br />0.102 hours <br />6.084656e-4 weeks <br />1.40024e-4 months <br /> is based on correspondence with STPEGS in Attachment

2.4.8. Buildup

Buildup was ignored, as the detector is immersed in the atmosphere.

This is conservative, as it produces a lower dose, which provides a lower detector setpoint level.5. DESIGN INPUTS 5.1. Source Term The source term is provided in Reference 3.3, Fuel Handling Accidents.

The decay constants are also taken from Reference 3.3.Table 5.1: Nuclide Gap Inventory C Core Inventory Decay Constant Activity after 368*~ ~ ~ t 1/Iiý) .(./M~) J hr (Ci/MWth).' 3 .41E 03 1 04 04 4.97&-57 Kr85m i 7.07E+03 ) 4.39E-05 3.90E-22....... .: .. ..

i.:. 1 .9 0 E + 4 '" .... .:.. : ....... .. " .: " .i,. ..: Kr8, 2'.9'E+02 2.04E-W;:9

' 2.92E+02.ý Kr87 1.3 4E+04 1.52E-04 4.71E-84 Kr88 L906.88E05" 45 ..95E-76 Kr89 2.)32E+04 3.).63E-03 O.OOE+00 XIlm 26E02 1.8E0 .1 IE+/-02--6.'68E .0 .j- .-e ' 1.66E+03 3.49E-06 -1.63E+01 Xel35mn 1.02E+/-04 7.40E-04 O.OOE+00;1.34Et04

.09E-05 l2E0 Xe1'37 4.63E+04 2.96E-03 ..OOE+00 Xe138..: 4..o9E04.

6.80E-04 ..OOE ".0 1131 2.59E+04 9.96E-07 6.92E+03 1123-.7--lEtO"'04' 9.27 E-05 :- 9'.72E-44' 1133 5.37E+04 9.22E-06 2.66E-01 1134 0.85E4 2.23E-04 .:2.91E"i24I 1i35 "3i88E04_

... 2.86E-05 .. i7E-i2 Note: Alkali metals are not included as they are fully scrubbed in the pool (Reference 3.14)C C CALC. NO. STPNOCOI3-CALC-005 E N E R C O N Fuel Handling Accident Monitor..... .,Response for EAL Thresholds REV.PAGE NO. 11 of 23 5.2 Volumes References 3.3 and 3.6-3.10 are used to calculate the mixing volumes for each accident.

The outer dimensions of each building were used to calculate a conservative mixing volume. The FHB cases used separate volumes for full dispersion as seen by the ARM and partial dispersion up to the point of contact with the exhaust vents. The volume for full dispersion for the ARM is 761,852 ft3 and the partial dispersion volume for the HVAC point is 71,136 ft3. The calculated mixing volume for the RCB is 2,384,437 ft3.5.3 Thermal Power Reactor thermal power is 4100 MWt (Reference 3.3)6. METHODOLOGY The reactor power, fraction of fuel failed, and radial peaking factor are combined to form a single multiplication factor since these parameters are independent of isotopes analyzed.This calculation uses a reactor power of 4100 MWt and an assumed number of failed fuel rods of 264 out of a total of 50,952 fuel rods (Reference 3.3). A radial peaking factor of 1.0 and a gap release of 0.05 are incorporated into the multiplication factor. A five percent gap release is appropriate to use for all nuclides because the core radionuclide inventory has been scaled appropriately for 1-131 and Kr-85 (Reference 3.3). The multiplication factor is calculated below: 264 failed Rods Multiplication Factor = 4100 MWt x 1.0 RPF x 0.05 x = 1. 06218 50,952 total rods Equation 6.1 was used to calculate the release concentrations of each nuclide.Release Concentrationi

= Ci

• 1. 06218

• 1 DF Equation 6.]28316.8466t-V cc Where: Cj is the gap inventory concentration of each nuclide from Reference 3.3, 1.06218 is the multiplication factor calculated above, 106 is a unit conversion to convert Ci to gCi, 283 16.8466 is a unit conversion to convert ft 3 to CC, V is the dilution volume for the case of interest; RCB, FHBI, and FHB exhaust (ft 3), and DF is the decontamination factor from Reference 3.14 and listed in Table 7.1 CALC. NO. STPNOCO 13-CALC-005E N E R c O N Fuel Handling Accident Monitor Response for EAL Thresholds PEN. 1 f PAGE NO. 12 of 23t After each concentration was calculated, the nuclide inventories were entered into MicroShield 6.20 to determine the dose rate associated with each detector.

MicroShield is appropriate for use in this calculation and has been approved and documented by ENERCON Services, Inc., MicroShield 6.20 Computer Code Verification, STPNOC013.

The FHB was modeled as a 80.5' X 182' X 52' volume evenly distributed source with a detector placed at (80.5001 ',67',5')

which correlates to the placement of detector RE-8090.The RCB was modeled as a cylinder with a volume equivalent to the free space calculated in the RCB above the refueling floor. The cylinder is-134.9' high with a 75' radius. A detector was placed at (75.0001',5',0')

which correlates to the placement of detector RE-8050. For detector RE-8099 the detector was placed on the end of the cylinder offset from center by 20'and 37.5'.For the exhaust vents in the FHB, the concentration is monitored, not the dose. This monitor is calibrated to Xe-133. Since this release is 90% Xe-133, it is assumed that this is all the monitor measures.7. CALCULATIONS C Table 7.1 contains the calculations used to find each release concentration.

Attachment 1 shows the input and results from each iteration of MicroShield 6.20.(

CALC. NO. STPNOCO I3-CALC-005E N E R c a N VFuel Handling Accident Monitor 1 E N E.,. :. C Response for EAL Thresholds REV. 1 of... ..PAGE NO. 13 of 23 Table 7.1: Calculations Nuclide... ....". ..... ........ ...Kr83m.... ...... ...........

.....Kr85m Kr87%'Kr88 Xc131m I ,Xe133m:5 i Xe133% .... .... .. ..............

.... ....L., -,,..

• 3 .r i iXe138 1131..... ... ....... .. :... t 1132..... ...........

..........

.....:1134 i135, Activity ]VIM after 368 hr decayz~: ....:.. ...:! : :. .........:.....4.97E-57 2.92E+-0'4 7.1E-84 4.95E-36* 000E-j-00 1.11E+02 1 1.63E.+01.. ........ -..5 ..k.... ......: 7.17E+03.. --... .... .7 E 0 .........0'.00QE+/-00:

...... ....o..... ... ...1.27E-08 0.OOE+00 6.92E+03'.!..Z ...Z ..:.... 7...9.72E-44...............

• 2.66E 2.91IE-124 1.7 E-.. 12...... .... ............ ... ...... .. I... ..... .. .... ...... ... .. ............

...........

... .......Release Itiplication Pool A Factor DF froci___from Pool Ci 1.06218 I1 5.27E-57 i~ 6 i ..........

..... ... .. .....o...o..1.062 18:ý 1 4.15E-22 1.06218 1 3. 1OE+02 1.06218.

• 1 '5.01E84, 1.06218 1 5.26E-36 1.06218. 1. ..:0.0.oE+007 1.06218 1 1.17E+02 1.0621 I .

1 1.06218 1 7.61E+03"v.:% 7 " .: " -! r.. .. ... .7 ...... .. ..;T C T .7 ... ......1.06218 1 0.00E+00... .... 8 ..........

............

.-+ 0 1.06218 1' 1.3'" 4E0 A.06218 I 0.OOE+00.1.06218 1PO O .68E+01>1.06218.'

1/.200. 3;68E+01= .:.L ...:... ... ... = ...= -@ ..=2 ..... ... .. -.,....L....:..

.. .1.06218 1/200 *5.16E-46 1.06ý2i 8 1 :,1/200 ",:.1.4 1. Em l'~-3* : .:- ..... .. f ...... I, A... ... ... ............

1.06218 1/200 1.54E-126 1.06218:'.

1/200 .9.09E- 15 FB IIVAC I Concentration IUCicc.2.62E-60 i...-:.06E-25 1.54E-01 2.61E-39 0 OE00t-6 5.83E-02.E ..........

' .' ..... ...... .= -;8.59E-03 ' , 3.78E+00 q , ...... .... ..... :..........

...... .... " '0.00oE+0.0;.'

6.67E-12 S0.OOE+00 0.0013+00 1.83E-02 ., 2.56E-49 7.2-07 7 66E-130....... ... ........ 7 ~ ..-..........

8 CALC. NO. STPNOC013-CALC-005 r EN ER CON Fuel Handling Accident MonitorRE 1 Response for EAL Thresholds RAEV NO1 f2 ATTACHMENT 1 Case SummaxY of FHB FHA Page I of 3 tin Microshield 6.20 (05-MSD-6.20-l 158)Enercon DOS rm. ra.C'.R~n Dt M',Ch17. 2014 R~rn~~u 151:4 F',~,w~t~oa 405:I (1828 1042 a-. 1 t 204 1, II A~wftv ft~t"bwvo He qm~ SU p Mt qj~ 1.9o.p fh W 3/17'201 CALC. NO. STPNOCOI3-CALC-005i E N E R c N Fuel Handling Accident Monitor REV. 1 Response for EAL Thresholds PAGE NO. 15 of 23 Cane Sumamn of FHB FHA Page. 2 of 3 1 .4 +-A look+W1 Jlos 04. 110.. ... '~ 4~12 1 ~+L2+02 1 '6'e+~~ 5~..+02~ ~+L12 4<"' 2 S 2~S~+~S~ ~ 12'II ".1 113 4~ 4. 1~2 ~~-2i 161 ~12.41 1.01 2.21' ~-11~141~.4 '~-i1 6' ~..-C0 04 1~-O'4<'~400 1~4~'-CA'4 2 22~2e+o2~1e:A 1 C:,Po2 inf F~e22Otx6)~icr SLe14x~m~e..C2 e~ie~ T~TIFH%2G...3!1i7!"201.4 rENERCON Fuel Handling Accident Monitor Response for EAL Thresholds CALC. NO. STPNOCO13-CALC-005 REV. 1 PAGE NO. 16 of 23 Case Sumrmiqu of FHB FHA Page 3 of 3 Y (fi~e///9Prgra%2Oile%~O~S6)crobieXam~e~~a~eik~~l~fJF{BQF..

31 7,2014 (

CALC. NO. STPNOCO03-CALC-005 E N E R C a N Fuel Handling Accident Monitor Response for EAL Thresholds REV. I PAGE NO. 17 of 23 Case SumnaarN of RCB ELA.tlicroShield 6.20 (05-MSD-6.20-1156)

Enercon Page 1 of 3 P~ag DOS File Run Date RuJn Tkrno AC HA 8M5 3611 hr- ns 1;ý;'Urf 0 _V4 10: 57 r1P M~e ker Daty Chocked Cose TRtk A4C9 h Geometiry.

-Cy n dr Vohmce -Slee Shed1 Sourcetu tms&Height 4.1L+3 cmn RadhUa 2-3e+3c toQse ftints A 1 21qc-03m an 152_4 6 ft 0.0 in S It 0 CoLrce Iraiutio Ar L0zp ShielAds Dimesioncm Nuclde I-11 Kr -'A 4cr XLc- 13A 2.01 1.11 1 4(1S-~0.2.I 4.1 5.0: 0.20 Source Input: Gtouping Method -Startda 8 Indlcca Number af roups 2 Lowver Encergy Cutoff. 0.413 Phtns- .0.1S Included XL_+001 1.215116+01 5 'Vil~-34-12 5.100-116 2.213-131 2-1 .cr4 43-1%,^C-057 1.1,4rc-0734ý L4.3-1 12404 (1,34 ft 1M8 in cm 0 am 0In 0A0 in del Oe"It 4 1 1 U 2+.01.2. 43 le-OlO 1.1+002 7.411.34+002

1. 2300c+001 L 441e-t02 2.24435222 2.234+0'14 L.2rC BeIdup : The mxterial refrence Is -Yasto integratiorn ftwanwmlr~s Y birrctimz M~+Energy Activity 1`4 V Photonn/sec flunce Ratee Fluence Rato~MeVIcmnlroc Moic~nil/soc Nft Suildup With SuIldup tx~oosure Ram4 Eposure Rate mf/hr mRt/hr No Bu~dup With 8uiddup, ffie1`/,UC:,Pro~mTaiam% "OFilesý'20(xS6)/\licroSbield..Examples!CasýeFileisHThýLJJRCB%20F...

!~17i2014 CALC. NO. STPNOCO13-CALC-005 , E N E R C O N Fuel Handling Accident Monitor R..-

Ev,.r e Response for EAL Thresholds PAGE NO. 18 of 23 C Case Summnry of RCB FHA-Page 2 of 3 00L,~ 00~~ 00~01 04 10 Tot~1 ..........' 24ff Cý?tý-,......... ..5S ý2V 27 6, -241-2.. ... .. ...... .14 C2 -0 1 S-+C.; .. ...... .. .... ... ... ... ..j, 4 1~0.........~1eih~CHProanrn%

~ S!1712014 C CALC. NO. STPNOC013-CALC-005 F; rE N E R C 0 N Fuel Handling Accident Monitor Response for EAL Thresholds REV. I PAGE NO. 19 of 23 Case Suimmarv of RCB FI{A Page 3 of 3 Z fil e]`C: TFroia 20Hies3/42O ,XS6)M~icroShieldEx,-napl elCaseFiltýýie T-Iff!RC'B%)2GF...

117t~2014 CALC. NO. STPNOC013-CALC-005 EN E R CO N Fuel Handling Accident Monitor Response for EAL Thresholds PAGE NO. 20 of 23 Case Summary of RCB FHA Page I of 3 PTrint COS f14.Run Date Run Tiu.Micro~hield 6.20 (05-MSD--6.20-1158)

Envercen ftcS PM14A5 3614ý ýt ma-th 1,. 2014 Dle~LV,110 U 01.ke8 Case Tift, iý F+tA Geomeotry:

9 -Optxreer V~lwtn -14rd ShieM ""ig.Iht =A x Y~ 1 ~ 9.LI i 4118 Iy (134 ft 10.8 tn~(Tha~1143 ci Sh~ield N Aý 0.10122 Source Input, Grouping $01edhod -ftwudar 1ndlce Numbe F Gromps z21 LowerftnwV)

Cutoff t0I I'motom4 <0.01 : Included (Mudtde 1-111 1-12..1- 1fl 1-154 1-115 Kr-8Si~i Kr~87 Kr- 814 Kr-80 Xo-131n~c-133 l~-133rn Xe-131 Xe- liSrn Xe- 13 Xe~- 1314 3.6141X1e+0 1 5.1600e-046 1~41D0e-Q03 tS4tXle-125 9fA00e-eJ15

3. lOOSe+002 4,ISOOe-0 2 501-0144 S 2600+-016 2.1 VOe+0132.41?Xiu+003

1. 73EXk+OOi 1 ,1400c-0014

~5448e-~00

~ ~44,8e-009

.34414e-019 becquarrk 5,1.16+1 4, 9 c4ý+ C07 2814 L- CY L7322-C033

51. D I05 2-1714701.

2,14f53-A46 4 1 4 142it-Olt auld w m at: ea rmceis-Suc R46eear~tr Enery Ar Fluenc.. Raft 14v photon"s/SeC N Bu~ildupa Thuenc Rat&9lcpoum uieRte wRhr etp-raltate mR/hr With 86dup... -+02 fileJNýýC::Proa-ram%ý'2OFilees%20~x86)Ai~croShiel&ExampleoiCa.seFilesIfHr]hfL.RCB%20F...

3117,,'214

(

Fuel Handling Accident Monitor CALC. NO. STPNOCOI3-CALC-005 Respons f" "I ALcThesholds REV. 1 E.N..R..0.

Response for EAL Thresholds

....... NO. 21 of 23 Case Suimnaar ofRCB FHA Paae o-0.4 1 2~C 0 ..... ...Ci 04 ) 2L~+C C~f2L~MaI5iM+

o ~i .)C~+1C 4~.+L~2 I ~ 21.C.' .t + .... ......... ..... .. ...Ci .~0 ... ....... .. M* C4o4 1-9840+C 1,4C0 9.6,65+01C I AV3+03 CALC. NO. STPNOCO13-CALC-005 0 EN ER CO N Fuel Handling Accident Monitor REV. I Response for EAL Thresholds En PAGE NO. 22 of 23 (Case Summary of RCB FHA Page 3 of 3 (z 31,1712014 CALC. NO. STPNOC013-CALC-005 , N E RCO N Fuel Handlin Accident Monitor EV. I Response for EAL Thresholds PE N.

NO. 23 of 23 ATTACHMENT 2 From: "Gore, Duane" <degore@STPEGS.

COM<aiIlto:degore@STFEGS.C)M>>

To: "Mikus, Alan" <amikus@STPECS.COM<maiIto:amikus@STPECS.COM>>

Cc: "Domal, Michael" <mjdomal@STPE S.COM<maiitc:rinjdcrna l@STPEGS.COM>>, "Huang, Shih-Fang" <s+/-huang 2@STPEGS.COM<naiito:sfthuanoi CSTPEGS.COM>>

Subject:

Comments on the Enercon EAL calcs See attached.

I need to get Shih to review parts of -0003 and -0004.Duane Gore, P.E.Reactor Analysis Supervisor STPNOC (361) 972-8909 (979) 318-6314 (c)Below is an excerpt containing all relevant information from the attached document: STPNOC013-CALC-005.

Rev 0 1. Assumption 4.2: use only 264 rods, not 315. The 315 is for worst case. Here we want a lower release; for a lower monitor threshold.

2. Assumption 2: Use 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for decay. Again, this will give a lower release; for a lower monitor threshold.

Actually the fuel movement campaign last about 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, a defueled window from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 7 days, then a 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> reload window. So, if we wanted the smallest release we would use: I 00h + 50h +(7days*24h/d)

+ 50h = 368h This may also fix the "off-scale High" problem 3. Section 5.2: provide the isotopic half lives for the decay calc 4. Need to fully develop the building dimensions from the drawings.

Provide dimensioned sketches.Table 211A Need to state where, in the calc, the values come from. I could not find them. Also, should justify the statements about the monitor ranges. It would be best not to include them. That way we do not have to revise the calc if the monitors changes.

STPEGS UFSAR The new fuel assemblies are transported to the new fuel storage pit or to the new fuel elevator by the 15/2-ton, dual-service FHB crane. The 2-ton hoist of this crane is designed to handle new fuel assemblies.

New fuel handling is discussed in detail in Section 9.1.4. Use of the 2-ton hoist of the 15/2-ton crane or of the fuel-handling machine to handle new fuel ensures that the design uplift of the racks will not be exceeded.The new fuel storage pit is situated in the approximate center of each FHB. The floor of the new fuel storage pit is at El. 50 ft-3 inches. The new fuel storage pit access hatch is provided with a three-section protective cover at El. 68 ft. The fuel assemblies are loaded into the new fuel storage racks through the top and stored vertically.

9.1.1.3 Safety Evaluation.

Units I and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1. Flooding of the new fuel storage pit from fluid sources inside either FHB is not considered credible since all fluid systems components are located well below the elevation of the new fuel storage pit access hatch. A floor drain is provided in the new fuel storage pit to minimize collection of water.The applicable design codes and the ability of the FHB to withstand various external loads and forces are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7.Missile protection of the FHBs is discussed in Section 3.5. Failure of nonseismic systems or structures will not decrease the degree of subcriticality provided in the new fuel storage pit.In accordance with American National Standards Institute (ANSI) N 18.2, the design of the normally dry new fuel storage racks is such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum moderation (under dry or fogged conditions).

For the unborated flooded condition, assuming new fuel of the highest anticipated enrichment in place, the effective multiplication factor does not exceed 0.95. Credit may be taken for the inherent neutron-absorbing effect of the materials of construction.

The new fuel assemblies are stored dry, the 21-in. spacing ensuring a safe geometric array. Under these conditions, a criticality accident during refueling and storage is not considered credible.Consideration of criticality safety analysis is discussed in Section 4.3.Design of the facility in accordance with RG 1.13 ensures adequate safety under both normal and postulated accident conditions.

The new fuel storage racks also meet the requirements of General Design Criterion (GDC) 62.9.1.2 Spent Fuel Storage 9.1.2.1 Design Bases. The spent fuel pool (SFP) is a stainless steel-lined reinforced concrete pool and is an integral part of each FHB. All spent fuel racks are classified as seismic Category 1, as defined by RG 1.29, and as ANS SC 3.The spent fuel storage facility provides storage capacity for 1,969 high density absorber spent fuel racks in a honeycomb array in each unit. Two storage regions are provided in the SFP. Two of the 9.1-2 Revision 16 STPEGS UFSAR Region 2 rack modules on the south end of the pool (modules #12 and #16) have not been installed.

A Fuel Ultrasonic Cleaning system may be used in the open space designated for modules #12 and#16. The Fuel Ultrasonic Cleaning system is freestanding and is seismically qualified.

It has no adverse effect on the fuel assemblies that are selected for cleaning; nor does it have an adverse effect on the design function of the spent fuel pool or its associated support systems. Figure 9.1.2-2 shows the pool layout for both Units I and 2. The six Region 1 rack modules are located in the northwest corner of the spent fuel pool.The Region 1 racks have 10.95-in.

nominal center-to-center spacing between the cells. This region is conservatively designed to accommodate unirradiated fuel at enrichments to 4.95 weight percent.Region I storage cells are each bounded on four sides by a water box except on the periphery of the pool. The Region 1 spent fuel racks include a lead-in-guide to assist in depositing fuel assemblies into the fuel cell. Figure 9.1.2-3 shows a typical Region 1 spent fuel rack.The reactivity characteristics of fuel assemblies which are to be placed in the spent fuel storage racks are determined and the assemblies are categorized by reactivity.

Alternately, if necessary, all assemblies may be treated as if each assembly is of the highest reactivity class until the actual assembly reactivity classification is determined.

Section 5.6 of the Technical Specifications provides the definitions of the reactivity classifications and the allowed storage patterns.

Fuel assemblies are loaded into the racks in a geometrically safe configuration to ensure rack subcriticality.

Fuel assembly reactivity requirements for close packed storage and checkerboard storage are specified in the Technical Specifications.

The boron concentration of the water in the spent fuel pool is maintained at or above the minimum value needed to ensure that the rack Keff is less than or equal to 0.95 in the event of misplaced assemblies in the close packed storage areas or in checkerboard storage areas. Consideration of criticality safety is discussed in Section 4.3.The Region 2 racks have a 9.15-in. nominal center-to-center spacing with fixed absorber material surrounding each cell. A sheet of neutron absorber material is captured between the side walls of all adjacent boxes. To provide space for the absorber sheet between boxes, a double row of matching flat round raised areas are coined into the side walls of all boxes. The raised dimension of these locally formed areas on each box wall is half the thickness of the absorber sheet. The boxes are fusion welded together at all these local areas. The absorber sheets are scalloped along their edges to clear these areas. Figure 9.1.2-4 shows a typical Region 2 spent fuel rack.The axial location of the absorber with respect to the active fuel region is provided and maintained by the structure of each box. At the outside periphery of each rack, a sheet of absorber material is captured under thin stainless sheets which are intermittently welded all around to the box.All rack surfaces that come into contact with fuel assemblies are made of annealed austenitic stainless steel. These materials are resistant to corrosion during normal and emergency water quality conditions.

The racks are designed to withstand normal operating loads as well as to remain functional with the occurrence of an SSE. The racks are designed with adequate energy absorption capabilities to withstand the impact of a dropped spent fuel assembly from the maximum lift height of the spent fuel pit bridge hoist. The racks are designed to withstand a maximum uplift force equal to the uplift force of the bridge hoist. The 14-in. and 16-in. racks also meet the requirements of ASME Code,Section III, Appendix XVII. The high-density spent fuel racks meet the criteria of Appendix D to Standard Review Plan (SRP) 3.8.4.9.1-3 Revision 16 STPEGS UFSAR Shielding for the SFP is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1. A minimum depth of approximately 13 ft of water over the top of an array of 193 (full core) assemblies with 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of decay is required to limit radiation from the assemblies to 2.5 mR/hr. or less.The FHB Ventilation Exhaust System is designed to limit the offsite dose in the event of a significant release of radioactivity from the fuel, as discussed in Sections 12.3.3,15.7.4, and 9.4.2. However, no credit for the FHB Ventilation Exhaust System is taken in the LOCA and Fuel Handling accident in Chapter 15.The FHB is designed to prevent missiles from contacting the fuel. A more detailed discussion on missile protection is given in Section 3.5.In addition, space is provided for storage of fuel during refueling inside the RCB for 64 fuel assemblies in four 4 x 4 modules having 16-in. center-to-center spacing (Figure 9.1.2-1A).

These modules are firmly bolted in the floor.9.1.2.2 Facilities Description.

The FHB abuts the south side of the RCB and is adjacent to the west side of the MEAB of each unit. The locations of the two FEIBs are shown in the station plot plan on Figure 1.2-3. For general arrangement drawings of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2-48 as listed in Table 1.2-1.The spent fuel storage facilities are designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor vessel. The spent fuel is transferred to the FHB and handled and stored in the spent fuel pool underwater.

The fuel is stored to permit some decay, then transferred offsite. For a detailed discussion of spent fuel handling, see Section 9.1.4.The SFP is located in the northwest quadrantof each FHB. The floor of the pool is at El. 21 ft-1I in., with normal water level at El. 66 ft-6 inches. The top of a fuel assembly in a storage rack does not extend above the top of the storage rack which is El. 39 ft-10 in. maximum. The fuel assemblies are loaded into the spent fuel racks through the top and are stored vertically.

9.1.2.3 Safety Evaluation.

Units I and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1.A detailed discussion of missile protection is provided in Section 3.5.The applicable design codes and the various external loads and forces considered in the design of the FHB are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7.Design of this storage facility in accordance with GDC 62 and RG 1.13 ensures a safe condition under normal and postulated'accident conditions.

The Keff of the spent fuel storage racks is maintained less than or equal to 1.00, even if unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administrative procedures to control the placement of burned and fresh fuel and control rods.Under accident conditions, the Keff is maintained well below 0.95 assuming 2200 ppm borated water.The boron concentration of the water in the spent fuel pool is maintained at or above the minimum 9.1-4 Revision 16 REQUIREMENTS FOR RELIABLE SPENT FUEL POOL LEVEL INSTRUMENTATION AT OPERATING REACTOR SITES AND CONSTRUCTION PERMIT HOLDERS All licensees identified in Attachment I to this Order shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:

(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.1. The spent fuel pool level instrumentation shall include the following design features: 1.1 Instruments:

The instrumentation shall consist of a permanent, fixed primary instrument channel and a backup instrument channel. The backup instrument channel may be fixed or portable.

Portable instruments shall have capabilities that enhance the ability of trained personnel to monitor spent fuel pool water level under conditions that restrict direct personnel access to the pool, such as partial structural damage, high radiation levels, or heat and humidity from a boiling pool.1.2 Arrangement:

The spent fuel pool level instrument channels shall be arranged in a manner that provides reasonable protection of the level indication function against missiles that may result from damage to the structure over the spent fuel pool. This protection may be provided by locating the primary instrument channel and fixed portions of the backup instrument channel, if applicable, to maintain instrument channel separation within the spent fuel pool area, and to utilize inherent shielding from missiles provided by existing recesses and corners in the spent fuel pool structure.

1.3 Mounting

Installed instrument channel equipment within the spent fuel pool shall be mounted to retain its design configuration during and following the maximum seismic ground motion considered in the design of the spent fuel pool structure.

1.4 Qualification

The primary and backup instrument channels shall be reliable at temperature, humidity, and radiation levels consistent with the spent fuel pool water at saturation conditions for an extended period. This reliability shall be established through use of an augmented quality assurance process (e.g., a process similar to that applied to the site fire 15i-otection program).1.5 Independence:

The primary instrument channel shall be independent of the backup instrument channel.1.6 Power supplies:

Permanently installed instrumentation channels shall each be powered by a separate power supply. Permanently installed and portable instrumentation channels shall provide for power connections from sources independent of the plant ac and dc power distribution systems, such as portable generators or replaceable batteries.

Onsite generators used as an alternate power source and replaceable batteries used for instrument channel power shall have sufficient capacity to maintain the level indication function until offsite--'Aftachment 2 resource availability is reasonably assured.1.7 Accuracy:

The instrument channels shall maintain their designed accuracy following a power interruption or change in power source without recalibration.

1.8 Testing

The instrument channel design shall provide for routine testing and calibration.

1.9 Display

Trained personnel shall be able to monitor the spent fuel pool water level from the control room, alternate shutdown panel, or other appropriate and accessible location.

The display shall provide on-demand or continuous indication of spent fuel pool water level.2. The spent fuel pool instrumentation shall be maintained available and reliable through appropriate development and implementation of the following programs: 2.1 Training:

Personnel shall be trained in the use and the provision of alternate power to the primary and backup instrument channels.2.2 Procedures:

Procedures shall be established and maintained for the testing, calibration, and use of the primary and backup spent fuel pool instrument channels.2.3 Testing and Calibration:

Processes shall be established and maintained for scheduling and implementing necessary testing and calibration of the primary and backup spent fuel pool level instrument channels to maintain the instrument channels at the design accuracy.

(7 NEI 12-02 (Revision 1).A. gust 2012 The three critical levels that must be monitored in a spent fuel pool are discussed below.It should be noted that continuous indication from a single instrument over the entire span from level 1 to level 3 is not required but could be used. If more than one instrument is used to monitor the entire span, that set of instruments constitutes a single channel satisfying either the primary or backup instrument channel requirement (refer to Figure 1 below).A visual representation of monitoring levels 1, 2 and 3 and the associated requirements for monitoring between the points are presented in Figure 1. The minimum requirements apply to the separation distance between level indications and support development of appropriate response procedures.

These requirements are separate from the instrument channel design accuracy discussed in section 3, which apply to either discrete or to continuous instruments.

Figure 1 2.3.1. Level 1 -level that is adequate to support operation of the normal fuel pool cooling system A typical fuel pool cooling system design includes a combination of weirs and/or vacuum breakers that prevent siphoning of the pool water level, below a minimum level, in the event of a piping rupture that can affect the SFP level.Level 1 represents the HIGHER of the following two points: 3 NEI 12-02 (Revision 1)August 2012 (

• The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis.This level will vary from plant to plant and the instrument designer will need to consult plant-specific design information to determine the actual point that supports adequate cooling system performance.

2.3.2. Level

2 -level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck Level 2 represents the range of water level where any necessary operations in the vicinity of the spent fuel pool can be completed without significant dose consequences from direct gamma radiation from the stored spent fuel. Level 2 is based on either of the following:

• 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or 0 a designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while perfonning local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components.

Additional guidance can be found in EPA-400 (Reference 4), USNRC Regulatory Guide 1.13 (Reference

5) and ANSI/ANS-57.2-1983 (Reference 6).Designation of this level should not be interpreted to imply that actions to initiate water make-up must be delayed'until SFP water levels have reached or are lower than this point.2.3.3. Level 3 -level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.Level 3 corresponds nominally (i.e., +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner to provide the maximum range of information to operators, decision makers and emergency response personnel.

Designation of this level should not be interpreted to imply that actions to initiate water make-up must or should be delayed until this level is reached.4 Nuclear Operagring Company South T~ ftlhoed Eltctric Cenamt/ng Statlon PO. Box 28,9 ,adswaorth.

s 77483 -February 28, 2013 NOC-AE-1 3002959 10 CFR 50.4 10 CFR 2.202 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1&2 Docket Nos, STN 50-498, STN 50-499.Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

References:

1. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051)

2. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Revision 0, August 29, 2012 3, Letter D. W. Rencurrel to NRC, "Initial Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)", dated October 24, 2012 On March 12, 2012, the Nuclear Regulatory Commission (NRC) issied an 'order (Reference 1)to STP Nuclear Operating Company (STPNOC).

Reference 1 directs STP Nuclear Operating Company to provide a reliable indication of the water level in associated spent fuel storage pools. Specific requirements are outlined in Attachment 2 of Reference 1, Reference I required submission of an overall integrated plan, including how compliance will be achieved.

The final Interim staff guidance (Reference

2) was issued August 29, 2012 providing licensees an acceptable approach for complying with the order, The purpose of this letter is to provide the overall integrated plan, including a description of how compliance will be achieved pursuant to Section IV, Condition 0.1 .a, of Reference 1 in accordance with the guidance in Attachment 2 to Reference I and the guidance in Reference

2. See the Enclosure for STPNOC's response to the requested information.

There are no new commitments In this letter.33650640 NOC-AE-13002959 Page 2 of 3 If there are any questions regarding this letter, please contact Robyn Savage at (361) 972-7438.I declare under penalty of perjury that the foregoing is true and correct.Executed on: CX//c /8 Dennis L. Koehl President and CEO/CNO

Enclosure:

South Texas Project (STP) Overall Integrated Plan for Implementation of Unit 1 &Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 33650640 NOC-AE-13002959 Page 3 of 3 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K..Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 B1)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN116 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 U. S. Nuclear Regulatory Commission Director of Office of Nuclear Regulation One White Flint North (MS 13 H 16M)11555 Rockville Pike Rockvllle, MD 20852-2738 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services 33650640 I.ENCLOSURE NOC-AE-13002959 South Texas Project (STP)Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 ( _Page 1 of 12 STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 1.0 OVERALL INTEGRATED PLAN INTRODUCTION This document provides the overall Integrated Plan (the "Plan") which the STP Nuclear Operating Company ("STPNOC")

will implement for Units I and 2 to comply with the requirements of NRC Order EA-12-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" (Ref.2), (the "ORDER"), NRC Interim Staff Guidance JLD-ISG-2012-003

[Rev.0] (Ref.3), (the "ISG"), and NEI Report 12-02[Rev.1] ("NEI 12-02").This Plan follows the format and provides all of the information on the STP 1 & 2 Integrated Plan that is required in NEI 12-02 [Rev.1] (Ref.i), Section A-2-2. Throughout this Plan, any reference to NEI 12-02 and the ISG will be based on the revisions above.Any reference to NEI 12-02 will include compliance to the clarifications and exceptions to NEI 12-02 required by the Interim Staff Guidance, Rev. 0.In response to the NRC requirements, STPNOC will provide two channels of independent, permanently-installed, wide-range spent fuel pool level instrumentation

("SFPLI"), for the spent fuel pool ("SFP") of each unit. The spent fuel pool for each unit is independent and not interconnected in any way. For each SFP, the instrumentation provided for each channel will utilize the same technology, as permitted by the NEI 12-02 [Rev. 1]. The spent fuel pool level instrumentation will provide continuous level indication for each SFP on both the Primary and Backup Channels.Both the Primary and Backup Channel/Instrument location and display of the SFP level will be independently mounted in each unit's Radwaste Control Room in the Mechanical Electrical Auxiliary Building (MEAB), which is an accessible post-event location.

Other locations are still being considered.

Both the Primary and- Backup Channel remote, non-safety related indication of the SFP level will also be provided in each unit's Control Room via input to the Plant Computer.The instrumentation systems will not be safety-related, but will meet the requirements for augmented quality in accordance with NEI 12-02 [Rev. 1] and the ISG as described below.Since all of the potential suppliers have not completed development, the information in this Plan is based on the overall strategy and on information which, based on current information from potential suppliers, is thought to envelope the systems being developed for this application.

If there are any changes to the requirements in NRC JLD-ISG-2012-003

[Rev.0] and NEI.12-02 [Rev.1], relief from the requirements and schedule documented in this Plan may be required, in accordance with Section 12.0. Any required changes to this Plan will be defined in the periodic status reports submitted to the NRC.2.0 APPLICABILITY:

This Plan applies to the spent fuel pools for South Texas Project Unit 1 and Unit 2.Page 2 of 12 STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 3.0 SCHEDULE: The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit .1 is scheduled for completion prior to 10/28/2015, which is the end of the second refueling outage (1REi9) following submittal of this Plan.The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit 2 is scheduled for completion prior to 4/29/2015, which is the end of the second refueling outage (2RE17) following submittal of this Plan.Unit 1 Milestones are as follows: 0 Design/Engineering-September of 2014 0 Purchase of instruments

& equipment

-February of 2015 a Receipt of equipment

-June of 2015 0 Unit 1 Installation

& Functional Testing -October of 2015 Unit 2 Milestones are as follows:.* Design/Engineering

-December of 2013* Purchase of instruments

& equipment

-August of 2014* Receipt of equipment-November of 2014* Installation

& Functional Testing*-April of 2015 Consistent with the requirements of the ORDER and the guidance from NEI 12-02 [Rev.1], status reports will be generated in six (6) month intervals from the submittal of this Plan.4.0 IDENTIFICATION OF SPENT FUEL POOL WATER LEVELS: The STP Unit 1 and 2 spent fuel pools are essentially identical.

The following SFP elevations are identified:

• The bottom of the pool is at Plant El. 21 ft. 11 in.a The top of the SFP racks is approximately at Plant El. 39 ft. 10 in.a The minimum Limiting Condition for Operation SFP level is Plant El. 62 ft.* Normal. SFP water level is at Plant El. 66 ft. 6 in.a Non-safety related level switch alarms are activated at Plant El. 67 ft. on high level and Plant El. 66 ft. on low level.* The spent fuel pool deck is at Plant El. 68 ft.The required key SFP water levels, per guidance of NEI 12-02 [Rev.1] and ISG (with clarifications and exceptions), are as follows: Page 3 of 12 STP Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 4.1 LEVEL 1: Level adequate to support operation of the normal fuel pool cooling system.LEVEL 1 represents the HIGHER of either the level at which reliable suction loss to the cooling pump occurs, or, the required NPSH (Nominal Pump Suction Head) of the cooling pump.Loss of reliable suction to SFP cooling pumps. For the purposes of this Plan, this level will conservatively be placed at Plant El. 64 ft. 2 in. This allows for just over 1 ft. of SFP water level above the top of the suction inlet flange (SFP Cooling Pump 14 in. suction line with centerline of suction inlet flange at Plant El. 62 ft. 6 in.)which will be sufficient for NPSH. (Ref. 9)Therefore, considering the top of SFP fuel storage rack is at Plant El. 39 ft. 10 in., the indicated level on either the Primary or Backup Instrument Channel of greater than 24 ft. 4 in. above the top of the SFP fuel storage racks based on the design accuracy for the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, is adequate for normal SFP cooling system operation.

LEVEL I = Plant El. 64 ft. 2 in or 24 ft. 4 in. water level above the top of the SFP fuel storage rack 4.2 LEVEL 2: Level adequate to provide substantial radiation'shielding for a person standing on the SFP operating deck.Indicated level on either the Primary or Backup Instrument Channel of greater than 10 ft. above the top of SFP stored fuel assemblies based on current guidance in NRC RG 1.13 [Rev.2] (Ref. 4) will achieve substantial radiation shielding.

Requirements on substantial SFP radiation shielding is also given in ANSI/ANS-57.2-1983 (Ref. 5), and states that radiation shall not exceed 2.5 mRem/hr, but the minimum water depth to achieve this is undefined.

NRC RG 1.13 [Rev.2] took exception to using dose rates as design input for minimum SFP water level, and instead defined the minimum level as 10 ft. above the stored fuel assemblies.

STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

Therefore, indicated level on either the Primary or Backup Instrument Channel of greater than 10 ft. above the top of the SFP fuel storage rack, based on the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, ensures there is adequate water level to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck.Page 4 of 12 S-TP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 LEVEL 2 = Plant El 49 ft. 10 in. or 10 ft. water level above the top of the SFP fuel storage rack.4.3 LEVEL 3: Level where the fuel remains covered.As stated above, STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

The installation of the SFPLI sensor will be such that it will measure as close as possible to the top of the SFP fuel rack. Indicated level on either the Primary or Backup Instrument Channel of greater than % ft. above the top of SFP fuel storage racks based upon the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, satisfies the NEI 12-02 [Rev.2] requirement of +/- 1 ft.from the top of the fuel rack. This monitoring level ensures there is adequate water level above the stored fuel seated in the SFP fuel storage rack.LEVEL 3 = Plant El 40 ft. 4 in. or 6 in. water level above the top of the SFP fuel storage rack.5.0 INSTRUMENTS:

Both the Primary and Backup Instrument Channels will utilize permanently-installed instruments.

The design of the primary and backup instruments will be consistent with the requirements by NEI 12-02 (Rev.1], the ISG, and this Plan.The current plan is for both channels to utilize Guided Wave Radar, which functions according to the principle of Time Domain Reflectometry (TDR). A generated pulse of electromagnetic energy travels down the probe. Upon reaching the liquid surface the pulse is reflected and based upon reflection times level is inferred.

The measured range will be continuous from the high pool level elevation (67') to the top of the spent fuel racks (Ref. 8). However, STP is still evaluating other designs for this application.

Any changes to the current plan will be reported in the 6 month update letter.The supplier for the SFP instrumentation will be chosen by a competitive bidding process completed after submittal of this Plan, so the information in this Plan is based on the overall strategy and on available information from potential supplier's information on systems being developed for this application.

5.1 Primary

(fixed) Instrument Channel The Primary Instrument Channel level sensing components will be located in the northeast corner of the Spent Fuel Pool, as shown in Attachment

1. The primary instrument channel will provide continuous level indication over a range from Plant El. 40 ft. 4 in. (LEVEL 3) to Plant El. 67 ft. (SFP high level alarm) or a range of 26 ft.8 in. In addition, the capability for discrete level indications at LEVEL1, LEVEL 2 and LEVEL 3, as described in Section 4.0, will be available.

Page 5 of 12 Nuclear Operating Company Sotith Tea7s ProJcl EecdrIc Genert/b Station P0. Box 289 Wadswortlh, TeXs 77483 June 25, 2013 NOC-AE-13003008 File No.: G25 10 CFR 2.202 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Unit I & 2 Docket Nos. STN 50-498, STN 50-499 Response to Request for Additional Information Regarding the Overall Integrated Plan in Response to Order EA-12-051,"Reliable-Spent Fuel Pool Instrumentation" (TAC Nos. MF0827 and MF0828)

References:

1. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051) (ST-AE-NOC-12002271) (ML12054A679)

2. Letter, D. L. Koehl to NRC Document Control Desk, "Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)," dated February 28, 2013..... .. .......-

3002959)-(ML4I30-70A006)

....-..........

....3. NRC letter dated June 7, 2013, "South Texas Project, Units 1 and 2 -Request for Additional Information RE: Overall Integrated Plan in Response to Order EA-12-051, "Reliable Spent Fuel Pool Instrumentation" (TAC Nos. MF0827 and MF0828) (ST-AE-NOC-13002439) (ML131149A09)

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued an Order (Reference 1)modifying licenses with regard to requirements for reliable spent fuel pool instrumentation.

On February 28, 2013, STP Nuclear Operating Company (STPNOC) submitted an Overall Integrated Plan (OIP) (Reference

2) in response to the NRC Order. By a letter (Reference 3)dated June 7,. 2013, the NRC staff determined that additional information is needed to complete their review of the OIP. The STPNOC response to Reference 3 is provided in the attachment to this letter.There are no regulatory commitments in this letter.STI:33704694 NOC-AE-13003008 Page 2 If there are any questions, please contact Ken Taplett at 361-972-8416.

I declare under penalty of perjury that the foregoing is true and correct.Executed on: .Z013 G. T. Powell Site Vice President

Attachment:

Response to Request for Additional Information Regarding Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) kit NOC-AE-13003008 Page 3 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal Senior Project Manager U,S. Nuclear Regulatory Commission One White Flint North (MS 8 BI)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MNI16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 U.SI Niclear Rgltfy-C6-fission Director, Office of Nuclear Reactor Regulation One White Flint North (MS 13 H 16M)11555 Rockville Pike Rockville, MD 20852-2738 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pe~a City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin Texas Department of State Health Services Robert Free Texas Department of State Health Services Attachment NOC-AE-13003008 Page I of 23 Response to Request for Additional Information Regarding Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

References:

1. Letter, D. L. Koehl to NRC Document Control Desk, "Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)," dated February 28, 2013 (NOC-AE-1 3002959) (ML13070A006)

2. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051) (ST-AE-NOC-12002271) (ML12054A679)

3. NRC Japan Lessons-Learned Project Directorate Interim Staff Guidance JLD-ISG-2012-03, "Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Revision 0, August 29, 2012 (ML12221A339)

4. NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses.

with Regard to Reliable Spent Fuel Pool Instrumentation," Revision 1, dated August 2012 (ML122400399)

Reference I provided the Overall Integrated Plan (OIP) which-the STP Nuclear.Operating Company ("STPNOC")

will implement for Units 1 and 2 to comply with the requirements of NRC Order EA-12-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" (Reference 2), NRC Interim Staff Guidance JLD-ISG-2012-003, Revision 0, (Reference

3) and NEI Report 12-02, Revision 1 (Reference 4).As discussed in Reference 1, any changes to the requirements in NRC JLD-ISG-2012-003 or NEI 12-02 may require relief fromthe requirements and schedule documented in the OIP.As provided in the OIP, the Milestones for completing the design and engineering work for Unit 1 are September 2014 and for Unit 2 is December 2013.The following responses to the request for additional information .are based on information developed to date. Any changes to the following information that occur after completing and approving the final design for reliable spent fuel pool instrumentation will be provided in the periodic 6-month status reports submitted to the NRC required by Order EA-12-051.

Attachment NOC-AE-1 3003008 Page 2 of 23 REQUEST FOR ADDITIONAL INFORMATION OVERALL INTEGRATED PLAN IN RESPONSE TO ORDER EA-12-051, "RELIABLE SPENT FUEL POOL INSTRUMENTATION" STP NUCLEAR OPERATING COMPANY SOUTH TEXAS PROJECT, UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499 1.0 Introduction By letter dated February 28, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13070A006), STP Nuclear Operating Company (STPNOC, the licensee), submitted an Overall Integrated Plan (OIP) in response to the March 12, 2012, U,S. Nuclear Regulatory Commission (NRC), Commission Order modifying licenses with regard to requirements for Reliable Spent Fuel Pool (SFP) Instrumentation (Order Number EA-12-051; ADAMS Accession No. ML12054A679) for South Texas Project (STP), Units 1 and 2. The NRC staff endorsed Nuclear Energy Institute (NEI) 12-02, "Industry Guidance for Compliance with. NRC Order EA-12-051, to Modify Licenses with Regard to Reliable SFP Instrumentation,".Revision 1-,.dated August -2012-(ADAMS-Accession -No. ML12240A307-).,with-exdeptions-as

-documented in Interim Staff Guidance (ISG) 2012-03, "Compliance with Order EA-12-051, Reliable .SFP Instrumentation," Revision 0, dated August 29, 2012 (ADAMS Accession No. ML12221A339).

The NRC staff has reviewed the February 28, 2013, response by the licensee and determined that the following request for additional information (RAI) is needed to complete its technical review. Please provide the response to the following RAIs.

Attachment NOC-AE-13003008 Page 3 of 23 2.0 Levels of Required Monitoring The OIP states, in part, that LEVEL 1: Level adequate to support operation of the normal fuel pool cooling system..Plant El. 64 ft. 2 in or 24 ft. 4 in. water level above the top of the SFP fuel storage rack.LEVEL 2: Level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck.Plant El 49 ft. 10 in. or 10 ft. water level above the top of the SFP fuel storage rack.LEVEL 3: Level where the fuel lemains covered.Plant El 40 ft. 4 in. or 6 in. water level above the top of the SFP fuel storage rack....The installation of the SFPLI [spent fuel pool level instrumentation]

sensor will be such that it will measure as close as possible to. the top of the SFP fuel rack.Indicated level on either the Primary or Backup Instrument Channel of greater than 1/2 ft. above the top of SFP fuel storage racks based upon the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, satisfies the NEI 12-02 [Rev.2] requirement of+/-1 ft. from the top of the fuel rack. This monitoring level ensures there is adequate water level above the stored fuel seated in the SFP.fuel storage rack.NRC RAI-la Please provide the following:

.a) For Level 1, please specify how the identified location represents the HIGHER of the two points described in the NEI 12-02 guidance for this level.STPNOC Response LEVEL 1 represents the HIGHER of either the level at which reliable suction loss to the spent fuel pool (SFP) cooling pump occurs, or the required net positive suction head (NPSH) of the SFP cooling pump Required NPSH.The SFP cooling pumps were analyzed for the conservative worst case operation of the SFP cooling pumps. Maximum values for line resistance, fluid temperature, suction flow Attachment NOC-AE-13003008 Page 4 of 23 and static head were used to calculate NPSH parameters for both required and available NPSH (NPSHR and NPSHA). It was determined that for the worst case scenario, the NPSHA was significantly higher than NPSHR. The NPSHA was calculated to be 4267 feet (ft) and NPSHR was calculated to be 18.75 ft.Therefore, NPSHR is not the determining value to be used for LEVEL 1.Loss of reliable suction to SFP cooling pumps.For the purposes of the OIP, this level is conservatively placed at Plant elevation (El.) 64 ft, 2 inches (in). This level provides for more than one foot of water above the top of the SFP cooling pump suction inlet flange (the centerline of the 14 Inch suction line flange to the pump is at Plant El. 62 ft. 6 in.) which will be sufficient for NPSH.A vortex calculation shows 0.134% air entrainment at an elevation one foot above the suction pipe centerline.

Level I at 64 ft. 2 in. is adequate for normal SFP cooling system operation.

Therefore, Level 1 represents the HIGHER of the two points described in the NEI 12-02 guidance.NRC RAI-lb b) A clearly labeled sketch depicting the elevation view of the proposed typical mounting arrangement for the portions of instrument channel consisting of permanent measurement channel equipment (e.g., fixed level sensors and/or stilling wells, and mounting brackets).

Please indicate on this sketch the datum values representing Level 1, Level 2, and Level 3 as well as the top of the fuel. Indicate on this sketch the portion of the level sensor measurement

... range. that is sensitivetomeasu.rement oftthe.fue.ulpo-l-level,wi.threspect-to

....the Level 1, Level 2, and Level 3 datum points.STPNOC Response See Figures 1 and 2 of this Attachment.

3.0 Instrumentation

and Design Features 3.1 Instruments and Arrangement The OIP states, in part, that Both the Primary and Backup Instrument Channels will utilize permanently-installed instruments....

The Primary Instrument Channel level sensing components will be located in the northeast corner of the Spent Fuel Pool, as shown in Attachment 1....

Attachment NOC-AE-13003008 Page 5 of 23 The Backup Instrument Channel level sensing components will be located in the northwest corner of the Spent Fuel Pool, as shown in Attachment 1 ....The current Plan isto mount the supporting electronic instruments outside of the spent fuel pool area, to provide a more benign radiation and environmental conditions, and also provide for reasonable and accessible locations for operators.

SFP Primary and Backup Channel Level Instruments are currently planned to be located in Radwaste Control Room of the Mechanical Auxiliary Building (MAB);however, STPNOC is still evaluat.ing other possible.

locations (i.e. relay.room).

NRC RAI-2 Please provide a clearly labeled sketch or marked-up plant drawing of the plan view of the SFP area, depicting the SFP inside dimensions, the planned locations/

placement of the primary and back-up SFP level sensor, and the proposed routing of the cables that will extend from the sensors toward the location of the read-out/display device.STPNOC ResQonse See Figure 3 of this Attachment.

3.2 Mounting

The OIP states, in part, that Consideration will be given to the maximum seismic ground motion that occurs at the installation location for the permanently installed equipment which is documented in the UFSAR [Updated Final Safety Analysis Report] Section 3.7.The mountings shall be designed consistent with the highest safety or seismic classification of the SFP. The level sensors will be mounted on seismically qualified brackets.NRC RAI-3a Please provide the following:

a) The design criteria that will be used to estimate the total loading on the mounting device(s), including static weight loads and dynamic loads. Please describe the methodology that will be used to estimate the total loading, inclusive of design basis maximum seismic loads and the hydrodynamic loads that could result from pool sloshing or other effects that could accompany such seismic forces.

ASFED. DCN RADAR HORNS (SEE FIG. 2)(5 TOP OF MEASUREMENT-&.

-RANGE EL. 64 -2"-EL. 49'-10'-EL. 40'-4-" GRADE EL. 28'-0" 1 ,-OPERAT I NG FL. EL. 68--ýG"" :-1 W.L..EL. 66'-1/2 " LEVEL I m/-LEVEL 2_-.LEVEL 3 SPENT FUEL POOL TOP OF THE TOP NOZZLE OF THE FUEL ASSEMBLIESS SPENT FUEL STORAGE EL. 2r-1V-ze IEL.1 (3 2f m 0 CD 0 CD 0 0 0 CD CL CD CD EL.(n xn U'5,b1C SPENT FUEL POOL COOLING PUMP 3 R 2 11N P A 10 1A 7 I.E .3 9fE I I* ii I I f I NOTE: THE SENSOR MEASUREMENT RANGE IS 0-10YZ. CORRESPONDING TO LEVEL 3 THROUfGH TOP OF MEASUREMENT RANGE.z 0 CD -=C)C CD C w) co Attachment NOC-AE-1 3003008 Page 21 of 23 Figure 2 Proposed Mounting Arrangement

/, RADAR HORN ANTENNA VIEWA-A

-RA3 NRC: 10 CFR Appendix A to Part 50--General Design Criteria for Nuclear Power Plants Page 6 of 13 The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions.

A switchyard common to both circuits is acceptable.

Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating curient power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded.

One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.Criterion 18-Inspection and testing of electric power systems. Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components.

The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.Criterion 19-Control room. A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design approvals or certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses or manufacturing licenses under part 52 of this chapter who do not reference a standard design approval or certification, or holders of operating licenses using an alternative source term under § 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 K.) Sv (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 for the duration of the accident.III. Protection and Reactivity Control Systems httD ://www.ilre.aov/readinz-rir-doc-collections/cfr/DpatO5O/part050-aDDa.html 1/17/2014

• Contamination of the control room envelope atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility, Contamination of the control room envelope atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope.4.2.2 The radioactive material releases and radiation levels used in the control room envelope dose analysis should be determined using the same source term, in-plant transport, and release assumptions used for determining the EAB and the LPZ dose values, unless these assumptions would result in nonconservative results for the control room envelope.4.2.3 The models used to transport radioactive material into and through the control room envelope," 0 and the shielding models used to determine radiation dose rates from external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.

4.2.4 Credit

for engineered safety features that mitigate airborne radioactive material within the control room envelope may be assumed. Such features may include control room isolation or pressurization, or intake or recirculation filtration.

Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP (Ref. 14); Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants" (Ref. 22);and Generic Letter 99-02 (Ref. 23) for guidance.

The control room envelope design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageous.

In most designs, control room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents.

Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.4.2.5 Credit should generally not be taken for the use of personal protective equipment or use of potassium iodide (KI) as a thyroid prophylactic drug.4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room envelope for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days 10 The iodine protection factor (IPF) methodology of Reference 19 may not be adequately conservative for all DBAs and control room arrangements since it models a steady-state control room condition.

Since many analysis parameters change over the duration of the event, the IPF methodology should only be used with caution. The NRC computer codes HABIT (Ref. 20) and RADTRAD (Ref. 21) incorporate suitable methodologies.

1.195-17

". to 30 days. 1 For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10" cubic meters per second (Ref. 25).4.2.7 Control room envelope doses should be calculated using dose conversion factors identified in Regulatory Position 4.1 above for use in offsite dose analyses.

The calculation should consider all radionuclides that are significant with regard to dose consequences and the release of radioactivity.

The whole body dose from photons may be corrected for the difference between finite cloud geometry in the control room envelope and the semi-infinite cloud assumption used in calculating the dose conversion factors using a compartment geometry correction factor. This factor is incorporated in Equation 10 of Regulatory Position 2.8. This correction is not applied to the beta skin dose estimates, as the range of beta particles in air is less than the typical control room dimensions.

The skin dose DCFs presented in Federal Guidance Report 12 (Ref. 17) are based on both photon and beta emissions.

Doses should be calculated using the factors in the column headed "Skin" in Table 111. 1 of Federal Guidance Report 12.4.3 Other Dose Consequences The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737 (Ref. 2).4.4 Offsite Acceptance Criteria The radiological criteria for the EAB and the outer boundary of the LPZ are given in 10 CFR 100.11. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA.For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 4. The criteria provided in Table 4 are the same criteria provided in the Standard Review Plan (Ref. 14).4.5 Control Room Acceptance Criteria The following guidelines may be used in lieu of those provided in SRP 6.4 (R.ef. 14) when showing compliance with the dose guidelines in GDC-19 of Appendix A to 10 CFR Part 50. The following guidelines relax the thyroid and skin acceptance criteria from that given in SRP 6.4.Currently, 10 CFR 20.1201 limits organ dose to 50 rem annually.

The release duration is specified in Table 4. The exposure period is 30 days for all accidents.

The criterion in GDC-19 applies to all accidents.

These occupancy factors are already included in the determination of the X/Q values using the Murphy and Campe methodology (Ref. 19) and should not be credited twice. The ARCON96 Code (Ref. 24) does not incorporate these occupancy factors in the determination of the x/Q values. Therefore, when using ARCON96 X/Q values, occupancy factors should be included in the dose calculations.

1.195-18 RS1 STP'OS UFSAR The partioulate channel is used as part ofthe Reactor Coolant Pressure Boundary (RCPB) leakage detection system. The sensitivity and response time of this part of the leakage detection system, which'is used. for monitoring unidentified leakage to the Containment, are sufficient to detedt an increase -in leakage rate of the equivalent of one gal/min within one hour, Elements of this monitor, including the indicator mounted In the RMS CR cabinet, are designed and qualified to remain functional following a Safe Shutdown Earthquake (SSE), in compliance with GO 1.45, Further information on the RCPB leakage detection system is presented in Section 5,2,5.11.5.2.3.3 U n Ver otor: The unit vent monitor samples the plant vent stack prior to.discharge to the environment and monitor for particulates,'iodjne,, nd noble'gases

.. .The 'unit vent particulate and iodine monitor draws rpresentative air samples from the plant vent stack via Isokinetic nozzles in the stack, and directs them through a moving filter paper-monitored by a shielded beta-sensitive scintillation detector.

Tfe sample-stream then passes through a charcoal collect6r,.where collected iodine ismonitored by a shielded gamma-sensitive scintillation detector.The sample is then. returned to the vent stack.A separate wide-range gas monitor is provided ftr the unit vent. The monitor has two isoldnetio nozzles foi' sampling during both normal .and 'accident conditions.

Thu stack samples pass first through a sample conditioniug unit which filters particulates and iodine and may be used to take grab *samples. The samples then pass through the shielded detector assembly, which uses three detectors to ooverlthec~~n

_lj r~angoe required.

The low range detector uses a bet-sensiive plastic sbintillator-photomultlplier (PM) tube. The mid-range and, high-range detectors use cadmium telluride (OdTe), dhlorine-doped, solid-state sensors, This wide-range gas monitor satisfies the requirements of NtUREO-0737, Item .Fl, 1 for provisions for sampling plant effluents .for lodines and Partlculates and f6r noble gas effluents'from the plant vent, 11.5.2.3.4' Control Room Eleical A.dliarv BIulding Ventila.tion Monitors; The CR/EAB ventilation frronitors are Class IE monitors which continuously assess the intake air to the .CR for indioation of abnormal airborne radioactiVity concentration, Each monitor assembly Is powered fron a separate electrical power source, In. the evewt of high radiation CR emergency ventilation operation is Initiated (Section 7.312), Failure of a monitor is alarmed inthe CR.Each monitor assembly Is comprised, of a-redioulatlon pump, beta-sensitive stintilation-detector, ..... .four-pi lead shielding, check source, stainless steel sample gag receiving chamber, and associated electronics, 11.5.2.3.5 Condenser Vacdmp Pump Montor: Gaseous samples are drawn through an -off-line system by a pump from the dis-hr.of the v&6uum pump exhaust header of the condenser.

I This channel monito's th. gaseous sampi radioactivity ww ouidtiiiatiof-an tStube leak, allowing reactor ooolant'to enter the secondary side. fluid, this monitorcomplements the $GBD monitors in -indication of a SQ tube leak, The gaseous. radioactivity levels are monitored by a single detector in a. manner similar to the unit vent wide gas monitor.S11.5.2.3.6 .Snent. Fuel Pool chaust Monitrs: The SFP3 monitors are Class IB and are identical to the CRJEAB ventilation monitors described in Section 1l125.2.3.4 except that they sample the exhai'st from the FHB. In the event of high radiation

'he monitors initiate emergency operation 11.5-11 Revision 14

.STPEGS UFSAR i 11.5.2.5.1 G-aseous Wasto Processing System Inlet Monitor: The GWPS inlet monitor erploys a gafima (Nal crystal) sclntillator/photomultiplier tube combination to measure the radioactvity level of the waste gases entering the OWP.S. The monitor is used In corjunction with the GWP S discharge monitor to measure overall effeotiveness of the GWP S.11.5.2.5.2 GWPS Dlsoharme Monitor: This monitor is similar to the GWPS inlet monitor and is hiMalled upstream of the discharge valve, Upon detection of high radioactivity or monitor failure, the OWPS discharge valve, FV-4671, is automatically closed.11.5.2,5.3 Main Steam Line.Monitors:

Each MS line is monitored by an ATL monitor consisng of a Geiger Mueller (GM) tube deteotor and an Ion chamber detector with overlapping ranges. The detectdrs, are shielded by 3 in. of lead, The monitors are designed to monitor gross gamma activity in the steam line and provide a basis for determining possible atmospheric releases from the MS power-operated relief valve (PORV), SO safety valves, and/or auxiliary feedwater pump turbine, ,1 Thenaionitors.

provide a dose rate range equivalent to 1.0.1 to 10 3 ýCiicm3 xenon-133.

Based upon coo.. inventory, the ratio of xenon-133 to other nuclides in the fael can be detemined.

In order to.obtain the above conoentratoný of xenon-133, in the main steam line, a large primary-to-secondary leak inust presence of xenon-I 33 indicates other radioactive Isotopes are present.Using the relative-ratios of isotopes present in the MS line, a computer model for determination of dose rates from these isotopes, detector response eceves, the thiolmess of the MS line, and the geometry of the MS line relative to the detector, the dose rate equivalent to MS line concentration is 3 obtained.

The quantity of radioactive effluents released is obtained by multiplying the xenon-133 equivalent MS line concentrations by the isotope ratio times the steam release rate.These detectors are safety-related Class lB: and meet the requirements of RO 1.97 and NUREG-0737, 11,5.2.5,4 Stea- Geneoraor B1owdown Monitors:

These monitors are identical to the MS..... -line trinilooi-aiit aie adj -6nt~fihb-SG-blowdbwn l iuie In the sbdlahi~h Valve Ct}ibili&(VC)7, The monitors are used as an aid in determining the source of SO blowdown radioactivity due.to So tube rupture or a large primary-to-secondary leak.-jV-Tees o-deteetors-are safoe ,related-Clasa-1-E-ind-m~etthe-requirements-of-ROl

.9-7.S 11,5,2,55 Main Steam Line Nigh EnerAg Gamma 4N- 6) Mentors: Each main steam line is monitored by an ATL Nat sointlllation-detector, These detectors were installed to monitor th1 status of ste~am generator primary to seoondary tube'leals and'to provide a diagnostic tool for all individuals concerned wilt steam generator condition.

These detectors are designed to detect high energy gamma activity in the 6 to 7.2 MBV energy range. High energy gamma activity in the main steam lines indicates the presence of N-16. The level of N-16 in the main steam -lines is used to 11.5-14 Revision 14 RS2 STPEGS UFSAR Shielding for the SFP is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1. A minimum depth of approximately 13 ft of water over the top of an array of 193 (full core) assemblies with 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of decay is required to limit radiation from the assemblies to 2.5 mRlhr. or less.The FHB Ventilation Exhaust System is designed to limit the offsite dose in the event of a significant release of radioactivity from.the fuel, as discussed in Sections 12.3.3, 15.7.4, and 9.4.2. However, no credit for the FHB Ventilation Exhaust System is taken in the LOCA and Fuel Handling accident in Chapter 15.The FHB is designed to prevent missiles from contacting the fuel. A more detailed discussion on missile protection is given in Section 3.5.In addition, space is provided for storage of fiel during refueling inside the RCB for 64 fuel assemblies in four 4 x 4 modules having 16-in. center-to-center spacing (Figure 9.1.2-1A).

These modules are firmly bolted in the floor.9.1.2.2 Facilities Description.

The FHB abuts the south side of the RCB and is adjacent to the west side of the MEAB of each unit. The locations of the two FIBs are shown in the station plot plan on Figure 1.2-3. For geneial arrangement drawings of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2-48 as listed in Table 1.2-1.The spent fuel storage facilities are designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor vessel. The spent fuel is transferred to the FHB and handled and stored in the spent fuel pool underwater.

The fuel is stored to permit some decay, then transferred offsite. For a detailed discussion of spent fuel handling, see Section 9.1.4.The SFP is located in the northwest quadrant of each FHB. The floor of the pool is at El. 21 ft-1I in., with normal water level at El. 66 ft-6 inches. The top of a fuel assembly in a storage rack does not extend above the top of the storage rack which is El. 39 ft-10 in. maximum. The fuel assemblies are loaded into the spent fuel racks through the top and are stored vertically.

9.1.2.3 Safety Evaluation.

Units 1 and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1.A detailed discussion of missile protection is provided in Section 3.5.The applicable design codes and the various external loads and forces considered in the design of the FHB are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7.Design of this storage facility in accordance with GDC 62 and RG 1.13 ensures a safe condition under normal and postulated accident conditions.

The Keff of the spent-fuel -storage racks is maintained less than or equal to 1.00, even if unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administrative procedures to control the placement of burned and fresh fuel and control rods.Under accident conditions, the Kety is maintained well below 0.95 assuming 2200 ppm borated water.The boron concentration of the water in the spent fuel pool is maintained at or above the minimum 9.1-4 Revision 16

'0 r'-33 -REQUIREMENTS FOR RELIABLE SPENT FUEL POOL LEVEL INSTRUMENTATION AT OPERATING REACTOR SITES AND CONSTRUCTION PERMIT HOLDERS All licensees identified in Attachment I to this Order shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:

(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.1. The spent fuel pool level instrumentation shall include the following design features: 1.1 Instruments:

The instrumentation shall consist of a permanent, fixed primary instrument channel and a backup instrument channel. The backup instrument channel may be fixed or portable.

Portable instruments shall have capabilities that enhance the ability of trained personnel to monitor spent fuel pool water level under conditions that restrict direct personnel access to the pool, such as partial structural damage, high radiation levels, or heat and humidity from a boiling pool.1.2 Arrangement:

The spent fuel pool level instrument channels shall be arranged in a manner that provides reasonable protection of the level indication function against missiles that may result from damage to the structure over the spent fuel pool. This protection may be provided by locating the primary instrument channel and fixed portions of the backup instrument channel, if applicable, to maintain instrument channel separation within the spent fuel pool area, and to utilize inherent shielding from missiles provided by existing recesses and corners in the spent fuel pool structure.

1.3 Mounting

Installed instrument channel equipment within the spent fuel pool shall be mounted to retain its design configuration during and following the maximum seismic ground motion considered in the design of the spent fuel pool structure.

1.4 Qualification

The primary and backup instrument channels shall be reliable at temperature, humidity, and radiation levels consistent with the spent fuel pool water at saturation conditions for an extended period. This reliability shall be established through use of an augmented quality assurance process (e.g., a process similar to that applied to the site fire pirotection program).1.5 Independence:

The primary instrument channel shall be independent of the backup instrument channel.1.6 Power supplies:

Permanently installed instrumentation channels shall each be powered by a separate power supply. Permanently installed and portable instrurnentation channels shall provide for power connections from sources independent of the plant ac and dc power distribution systems, such as portable generators or replaceable batteries.

Onsite generators used as an alternate power source and replaceable batteries used for instrument channel power shall have sufficient capacity to maintain the level indication function until offsite--Attachment 2 resource availability is reasonably assured.1.7 Accuracy:

The instrument channels shall maintain their designed accuracy following a power interruption or change in power source without recalibration.

1.8 Testing

The instrument channel design shall provide for routine testing and calibration.

1.9 Display

Trained personnel shall be able to monitor the spent fuel pool water level from the control room, alternate shutdown panel, or other appropriate and accessible location.

The display shall provide on-demand or continuous indication of spent fuel pool water level.2. The spent fuel pool instrumentation shall be maintained available and reliable through appropriate development and implementation of the following programs: 2.1 Training:

Personnel shall be trained in the use and the provision of alternate power to the primary and backup instrument channels.2.2 Procedures:

Procedures shall be established and maintained for the testing, calibration, and use of the primary and backup spent fuel pool instrument channels.2.3 Testing and Calibration:

Processes shall be established and maintained for scheduling and implementing necessary testing and calibration of the primary and backup spent fuel pool level instrument channels to maintain the instrument channels at the design accuracy.

(i NEI 12-02 (Revision 1),. Awust 2012 The three critical levels that must be monitored in a spent fuel pool are discussed below.It should be noted that continuous indication from a single instrument over the entire span from level 1 to level 3 is not required but could be used. If more than one instrument is used to monitor the entire span, that set of instruments constitutes a single channel satisfying either the primary or backup instrument channel requirement (refer to Figure 1 below).A visual representation of monitoring levels 1, 2 and 3 and the associated requirements for monitoring between the points are presented in Figure 1. The minimum requirements apply to the separation distance between level indications and support development of appropriate response procedures.

These requirements are separate from the instrument channel design accuracy discussed in section 3, which apply to either discrete or to continuous instruments.

Figure 1 2.3.1. Level 1 -level that is adequate to support operation of the normal fuel pool cooling system A typical fuel pool cooling system design includes a combination of weirs and/or vacuum breakers that prevent siphoning of the pool water level, below a minimum level, in the event of a piping rupture that can affect the SFP level.Level 1 represents the HIGHER of the following two points: 3 NEI 12-02 (Revision 1)August 2012* The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or* The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis.This level will vary from plant to plant and the instrument designer will need to consult plant-specific design information to determine the actual point that supports adequate cooling system performance.

2.3.2. Level

2 -level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck Level 2 represents the range of water level where any necessary operations in the vicinity of the spent fuel pool can be completed without significant dose consequehces from direct gamma radiation from the stored spent fuel. Level 2 is based on either of the following:

4 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or 0 a designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components.

Additional guidance can be found in EPA-400 (Reference 4), USNRC Regulatory Guide 1.13 (Reference

5) and ANSI/ANS-57.2-1983 (Reference 6).Designation of this level should not be interpreted to imply that actions to initiate water make-up must be delayed'until SFP water levels have reached or are lower than this point.2.3.3. Level 3 -level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.Level 3 corresponds nominally (i.e., +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner to provide the maximum range of information to operators, decision makers and emergency response personnel.

Designation of this level should not be interpreted to imply that actions to initiate water make-up must or should be delayed until this level is reached.4 N~uclear OperangCopn Souti a P 'ro/Tct E/ed/lc Generakng S/a/ton PO. 6ox 28,9 Wdsporlh, Texas 77483 A/AMv February 28, 2013 NOC-AE-1 3002959 10 CFR 50.4 10 CFR 2.202 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1&2 Docket Nos. STN 50-498, STN 50-499.Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool instrumentation (Order Number EA-12-051)

References:

1. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051)

2. NRC Interim Staff Guidance JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Revision 0, August 29, 2012 3. Letter D. W. Rencurrel to NRC, "Initial Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)", dated October 24, 2012.On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued an order (Reference 1)to STP Nuclear Operating Company (STPNOC).

Reference 1 directs STP Nuclear Operating Company to provide a reliable indication of the water level in associated spent fuel storage pools, Specific requirements are outlined in Attachment 2 of Reference 1, Reference I required submission of an overall integrated plan, including how compliance will be achieved.

The final interim staff guidance (Reference

2) was issued August 29, 2012 providing licensees an acceptable approach for complying with the order. The purpose of this letter is to.provide the overall integrated plan, Including a description of how compliance will be achieved pursuant to Section IV, Condition C.1.a, of Reference 1 in accordance with the guidance in Attachment 2 to Reference 1 and the guidance in Reference

2. See the Enclosure for STPNOC's response to the requested information.

There are no new commitments, in this letter.L.,', 33650640 NOC-AE-1 3002959 Page 2 of 3 If there are any questions regarding this letter, please contact Robyn Savage at (361) 972-7438.I declare under penalty of perjury that the foregoing is true and correct.Executed on: c' --Dennis L. Koehl.President and CEO/CNO

Enclosure:

South Texas Project (STP) Overall Integrated Plan for Implementation of Unit 1 &Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 Q(,.33650640 NOC-AE-1 3002959 Page 3 of 3 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 B1)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MNI16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 U. S. Nuclear Regulatory Commission Director of Office of Nuclear Regulation One White Flint North (MS 13 H 16M)11555 Rockville Pike Rockville, MD 20852-2738 A. H, Gutterman, Esquire Morgan, Lewis & Bockius LLP.Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services 33650640 ENCLOSURE NOC-AE-13002959 South Texas Project (STP)Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Level Instrumentation to Meet NRC Order EA-12-051 (~Page 1 of 12 r STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 C' Revision:

00 1.0 OVERALL INTEGRATED PLAN INTRODUCTION This document provides the overall Integrated Plan (the "Plan") which the STP Nuclear Operating Company ("STPNOC")

will implement for Units 1 and 2 to comply with the requirements of NRC Order EA7 12-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" (Ref.2), (the "ORDER"), NRC Interim Staff Guidance JLD-ISG-2012-003

[Rev.0] (Ref.3), (the "ISG"), and NEI Report 12-02[Rev.1] ("NEI 12-02").This Plan follows the format and provides all of the information on the STP I & 2 Integrated.

Plan that is required in NEI 12-02 [Rev.1] (Ref.1), Section A-2-2. Throughout this Plan, any reference to NEI 12-02 and the ISG will be based on the revisions above.Any reference to NEI 12-02 will include compliance to the clarifications and exceptions to NEI 12-02 required by the Interim Staff Guidance, Rev. 0.In response to the NRC requirements, STPNOC will provide two channels of independent, permanently-installed, wide-range spent fuel pool level instrumentation

("SFPLI"), for the spent fuel pool ("SFP") of each unit. The spent fuel pool for each unit is independent and not interconnected in any way. For each SFP, the instrumentation provided for each channel will utilize the same technology, as permitted by the NEI 12-02 [Rev.1]. The spent fuel pool level instrumentation will provide continuous level indication for each SFP on both the Primary and Backup Channels.Both the Primary .and Backup Channel/Instrument location and display of the SFP level will be independently mounted in each unit's Radwaste Control Room in the Mechanical Electrical Auxiliary Building (MEAB), which is an accessible post-event location.

Other locations are still being considered.

Both the Primary and Backup Channel remote, non-safety related indication of the SFP level will also be provided in each unit's Control Room via input to the Plant Computer.The instrumentation systems will not be safety-related, but will meet the requirements for augmented quality in accordance with NEI 12-02 [Rev. 1] and the ISG as described below.Since all of the potential suppliers have not completed development, the information in this Plan is based on the overall strategy and on information which, based on current information from potential suppliers, is thought to envelope the systems being developed for this application.

If there are any changes to the requirements in NRC JLD-ISG-2012-003

[Rev.0] and NEI 12-02 [Rev.1], relief from the requirements and schedule documented in this Plan may be required, in accordance with Section. 12.0. Any required changes to this Plan will be defined in the periodic status reports submitted to the NRC.2.0 APPLICABILITY:

This Plan applies to the spent fuel pools for South Texas Project Unit 1 and Unit 2.Page 2 of 12 STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 3.0 SCHEDULE: The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit 1 is.scheduled for completion prior to 10/28/2015, which is the end of the second refueling outage (1 REI 9) following submittal of this Plan.The installation of reliable spent fuel pool level instrumentation for the spent fuel pool associated with Unit 2 is scheduled for completion prior to 4/29/2015, which is the end of the second refueling outage (2RE17) following submittal of this Plan.Unit 1 Milestones are as follows: 0 Design/Engineering

-September of 2014.Purchase of instruments

& equipment-February of 2015* Receipt of equipment

-June of 2015* Unit 1 Installation

& Functional Testing -October of 2015 Unit 2 Milestones are as follows:* Design/Engineering

-December of 2013 ( Purchase of instruments

& equipment

-August of 2014* Receipt of equipment

-November of 2014* Installation

& Functional Testing -April of 2015 Consistent with the requirements of the ORDER and the guidance from NEI 12-02 [Rev.], status reports will be generated in six (6) month intervals from the submittal of this Plan.4.0 IDENTIFICATION OF SPENT FUEL POOL WATER LEVELS: The STP Unit I and 2 spent.fuel pools are essentially identical.

The following SFP elevations are identified:

• The bottom of the pool is at Plant El. 21 ft. 11 in.* The top of the SFP racks is approximately at Plant El. 39 ft. 10 in.* The minimum Limiting Condition for Operation SFP level is Plant El. 62 ft.* Normal SFP water level is at Plant El. 66 ft. 6 in.* Non-safety related level switch alarms are activated at Plant El. 67 ft. on high level and Plant El. 66 ft. on low level.* The spent fuel pool deck is at Plant El, 68 ft.The required key SFP water levels, per guidance of NEI 12-02 [Rev. 1] and ISG (with clarifications and exceptions), are as follows: (k, Page 3 of 12 I STP Overall Integrated Plan for Implementation of Unit 1 & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051 Revision:

00 K 4.1 LEVEL 1: Level adequate to support operation of the normal fuel pool cooling system.LEVEL 1 represents the HIGHER of either the level at which reliable suction loss to the cooling pump occurs, or, the required NPSH (Nominal Pump Suction Head) of the cooling pump.Loss of reliable suction to SFP coolinQ pumps. For the purposes of this Plan, this level will conservatively be placed at Plant El. 64 ft. 2 in. This allows for just over I ft. of SFP water level above the top of the suction inlet flange (SFP Cooling Pump 14 in. suction line with centerline of suction inlet flange at Plant El. 62 ft. 6 in.)which will be sufficient for NPSH. (Ref. 9)Therefore, considering the top of SFP fuel storage rack is at Plant El. 39 ft. 10 in., the indicated level on either the Primary or Backup Instrument Channel of greater than 24 ft. 4 in. above the top of the SFP fuel storage racks based on the design accuracy for the instrument channel per NEI 12-02 [Rev. 1], for both the Primary and Backup Instrument Channels, is adequate for normal SFP cooling system operation.

LEVEL I = Plant El. 64 ft. 2 in or 24 ft. 4 in. water level above the top of the, SFP fuel storage rack 4.2 LEVEL 2: Level adequate'to provide substantial radiation shielding for a* person standing on the SFP operating deck.Indicated level on either.the Primary or Backup Instrument Channel of greater than 10 ft. above the top of SFP stored fuel assemblies based on current guidance in NRC RG 1.13 [Rev.2] (Ref. 4) will achieve substantial radiation shielding.

Requirements on substantial SFP radiation shielding is also given in ANSI/ANS-57.2-1983 (Ref. 5), and states that radiation shall not exceed 2.5 mRem/hr, but the minimum water depth to achieve this is undefined.

NRC RG 1.13 [Rev.2] took exception to using dose rates as design input for minimum SFP water level, and instead defined the minimum level as 10 ft. above the stored fuel assemblies.

STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

Therefore, indicated level on either the Primary or Backup Instrument Channel of greater than 10 ft. above the top of the SFP fuel storage rack, based on the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, ensures there is adequate water level to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck.Page 4 of 12 I STP Overall Integrated Plan for Implementation of Unit I & Unit 2 Spent Fuel Pool Instrumentation to Meet NRC Order EA 12-051/,.Revision:

00 LEVEL 2 = Plant El 49 ft. 10 in. or 10 ft. water level above the top of the SFP fuel storage rack.4.3 LEVEL 3: Level where the fuel remains covered.As stated above, STPNOC elects to use the conservative approach of defining the top of the fuel rack as a basis for measurement.

The installation of the SFPLI sensor will be such th-at it will measure as close as possible to the top of the SFP fuel rack. Indicated level on either the Primary or Backup Instrument Channel of greater than 1, ft. above the top of SFP fuel storage racks based upon the design accuracy of the instrument channel per NEI 12-02 [Rev. 1], for both the Primary and Backup Instrument Channels, satisfies the NEI 12-02 [Rev.2] requirement of +/- 1 ft.from the top of the fuel rack. This monitoring level ensures there is adequate water level above the stored fuel seated in the SFP fuel storage rack.LEVEL 3 = Plant El 40 ft. 4 in. or 6 in. water level above the top of the SFP fuel storage rack.5.0 INSTRUMENTS:

Both the Primary and Backup Instrument Channels will utilize permanently-installed instruments.

The design of the primary and backup instruments will be consistent with the requirements by NEI 12-02 [Rev.1], the ISG, and this Plan.The current plan is for both channels to utilize Guided Wave Radar, which functions according to the principle of Time Domain Reflectometry (TDR). A generated pulse of electromagnetic energy travels down the probe. Upon reaching the liquid surface the pulse is reflected and based upon reflection times level is inferred.

The measured range will be continuous from the high pool level elevation (67') to the top of the spent fuel racks (Ref. 8). However, STP is still evaluating, other designs for this application.

Any changes to the current plan will be reported in the 6 month update letter.The supplier for the SFP instrumentation will be chosen by a competitive bidding process completed after submittal of this Plan, so the information in this Plan is based on the overall strategy and on available information from potential supplier's information on systems being developed for this application.

5.1 Primary

(fixed) Instrument Channel The Primary Instrument Channel level sensing components will be located in the northeast corner of the Spent Fuel Pool, as shown in Attachment

1. The primary instrument channel will provide continuous level indication over a range from Plant El. 40 ft. 4 in. (LEVEL 3) to Plant El. 67 ft. (SFP high level alarm) or a range of 26 ft.8 in. In addition, the capability for discrete level indications at LEVELI, LEVEL 2 and LEVEL 3, as described in Section 4.0, will be available.

Page 5 of 12 Nuclear Operating Company Soath Tevas P'ioject Electric Generathi Station P[O. BoX289 Wadswaotth Texas 77483 1 June 25, 2013 NOC-AE-13003008 File No.: G25 10 CFR 2.202 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 1 & 2 Docket Nos. STN 50-498, STN 50-499 Response to Request for Additional Information Regarding the Overall Integrated Plan in Response to Order EA-12-051,"ReliableSpent Fuel Pool Instrumentation" (TAC Nos. MF0827 and MF0828)

References:

1. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-1 2-051) (ST-AE-NOC-1 2002271) (MLI 2054A679)2. Letter, D. L. Koehl to NRC Document Control Desk, "Overall Integrated Plan Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)," dated February 28, 2013 (NOC-AE-1 3002959) (ML13070A006)

3. NRC letter dated June 7, 2013, "South Texas Project, Units 1 and 2 -Request for Additional Information RE: Overall Integrated Plan in Response to Order EA-12-051, "Reliab[e Spent Fuel Pool Instrumentation" (TAC Nos. MF0827 and MF0828) (ST-AE-NOC-13002439) (ML131149A09)

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued an Order (Reference 1)modifying licenses with regard to requirements for reliable spent fuel pool instrumentation.

On February 28, 2013, STP Nuclear Operating Company (STPNOC) submitted an Overall Integrated Plan (OIP) (Reference

2) in response to the NRC Order. By a letter (Reference 3)dated June 7,. 2013, the NRC staff determined that additional information is needed to complete their review of the OIP. The STPNOC response to Reference 3 is provided in the attachment to this letter.There are no regulatory commitments in this letter.STI: 33704694 NOC-AE-13003008 Page 2 If there are any questions, please contact Ken Taplett at 361-972-8416.

1 declare under penalty of perjury that the foregoing is true and correct.Executed on: U ve5 25o, 3 G. T, Powell Site Vice President

Attachment:

Response to Request for Additional Information Regarding Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) kit NOC-AE-13003008 Page 3 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 BI)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MNI 16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 U.S. Nuclear Re-gulatdry Cnriirission.

Director, Office of Nuclear Reactor Regulation One White Flint North (MS 13 H 16M)11555 Rockville Pike Rockville, MD 20852-2738 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Peha City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele City of Austin Rihf-ard-A.

Ratliff Texas Department of State Health Services Robert Free Texas Department of State Health Services Attachment NOC-AE-1 3003008 Page 1 of 23 Response to Request for Additional Information Regarding Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

References:

1. Letter, D. L. Koehl to NRC Document Control Desk, "Overall Integrated Plan. Regarding Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)," dated February 28, 2013 (NOC-AE-1 3002959) (MLI307OA006)

2. Letter, Eric Leeds to E. D. Halpin, "Issuance of Order to Modify Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation," March 12, 2012 (EA-12-051) (ST-AE-NOC-1 2002271) (MLI 2054A679)3. NRC Japan Lessons-Learned Project Directorate Interim Staff Guidance JLD-ISG-2012-03, "Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Revision 0, August 29, 2012 (ML12221A339)

4. NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation," Revision 1, dated August 2012 (ML1 22400399)Reference 1 provided the Overall Integrated Plan (OIP) which the STP Nuclear Operating Company ("STPNOC")

will implement for Units 1 and 2 to comply with the requirements of NRC Order EA-12-051, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" (Reference 2), NRC Interim Staff Guidance JLD-ISG-2012-003, Revision 0, (Reference

3) and NEI Report 12-02, Revision 1 (Reference 4).As discussed in Reference 1, any changes to the requirements in NRC JLD-ISG-2012-003 or NEI 12-02 may require relief from the requirements and schedule documented in the OIP.As provided in the OIP, the Milestones for completing the design and engineering work for Unit I are September 2014 and for Unit 2 is December 2013.The following responses to the request for additional information are based on information developed to date. Any changes to the following information that occur after completing and approving the final design for reliable spent fuel pool instrumentation will be provided in the periodic 6-month status reports submitted to the NRC required by Order EA-12-051.

Attachment NOC-AE-1 3003008 Page 2 of 23 REQUEST FOR ADDITIONAL INFORMATION OVERALL INTEGRATED PLAN IN RESPONSE TO ORDER EA-12-051, "RELIABLE SPENT FUEL POOL INSTRUMENTATION" STP NUCLEAR OPERATING COMPANY SOUTH TEXAS PROJECT, UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499 1.0 Introduction By letter dated February 28, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML1 3070A006), STP Nuclear Operating Company (STPNOC, the licensee), submitted an Overall Integrated Plan (OIP) in response to the March 12, 2012, U.S. Nuclear Regulatory Commission (NRC), Commission Order modifying licenses with regard to requirements for Reliable Spent Fuel Pool (SFP) Instrumentation (Order Number EA-12-051; ADAMS Accession No. ML12054A679) for South Texas Project (STP), Units 1 and 2. The NRC staff endorsed Nuclear Energy Institute (NEI) 12-02, "Industry Guidance for Compliance with NRC Order EA-12-051, to Modify Licenses with Regard to Reliable SFP Instrumentation,"*Revision I-,-dated August 2012 -(ADAMS-Accession-No.

ML-1-2240A307-) -with exceptions as documented in Interim Staff Guidance (ISG) 2012-03, "Compliance with Order EA-12-051, Reliable .SFP Instrumentation," Revision 0, dated August 29, 2012 (ADAMS Accession No. ML12221A339).

The NRC staff has reviewed the February 28, 2013, response by the licensee and determined that the following request for additional information (RAI) is needed to complete its technical review. Please provide the response to the following RAls.

Attachment NOC-AE-13003008 Page 3 of 23 2.0 Levels of Required Monitoring The OIP states, in part, that LEVEL 1: Level adequate to support operation of the normal fuel pool cooling system.Plant El. 64 ft. 2 in or 24 ft. 4 in. water level above the top of the SFP fuel storage rack.LEVEL 2: Level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck.Plant El 49 ft. 10 in. or 10 ft. water level above the top of the SFP fuel storage rack.LEVEL 3: Level where the fuel remains covered.Plant El 40 ft. 4 in. or 6 in. water level above the top of the SFP fuel storage rack....The installation of the SFPLI [spent fuel pool level instrumentation]

sensor will be such that it will measure as close as possible to the top of the SFP fuel rack.Indicated level on either the Primary or Backup Instrument Channel of greater than 1/2 ft. above the top of SFP fuel storage racks based upon the design accuracy of the instrument channel per NEI 12-02 [Rev.1], for both the Primary and Backup Instrument Channels, satisfies the NEI 12-02 [Rev.2] requirement of+1 ft. from the top of the fuel rack. This monitoring level ensures there is adequate water level above the stored fuel seated in the SFP fuel storage rack.NRC RAI-la Please provide the following:

a) For Level 1, please specify how the identified location represents the HIGHER of the two points described in the NEI 12-02 guidance for this level.STPNOC Response LEVEL 1 represents the HIGHER of either the level at which reliable suction loss to the spent fuel pool (SFP) cooling pump occurs, or the required net positive suction head (NPSH) of the SFP cooling pump Required NPSH.The SFP cooling pumps were analyzed for the conservative worst case operation of the SFP cooling pumps. Maximum values for line resistance, fluid temperature, suction flow Attachment NOC-AE-13003008 Page 4 of 23 and static head were used to calculate NPSH parameters for both required and available NPSH (NPSHR and NPSHA). It was determined that for the worst case scenario, the NPSHA was significantly higher than NPSHR. The NPSHA was calculated to be 42.67 feet (ft) and NPSHR was calculated to be 18.75 ft.Therefore, NPSHR is not the determining value to be used for LEVEL 1.Loss of reliable suction to SFP cooling pumps.For the purposes of the OIP, this level is conservatively placed at Plant elevation (El.). 64 ft, 2 inches (in). This level provides for more than one foot of water above the top of the SFP cooling pump suction inlet flange (the centerline of the 14 inch suction line flange to the pump is at Plant El. 62 ft. 6 in.) which will be sufficient for NPSH.A vortex calculation shows 0.134% air entrainment at an elevation one foot above the suction pipe centerline.

Level 1 at 64 ft. 2 in. is adequate for normal SFP cooling system operation.

Therefore, Level 1 represents the HIGHER of the two points described in the NEI 12-02 guidance.NRC RAI-1b b) A clearly labeled sketch depicting the elevation view of the proposed typical mounting arrangement for the portions of instrument channel consisting of permanent measurement channel equipment (e.g., fixed level sensors and/or stilling wells, and mounting brackets).

Please indicate on this sketch the datum values representing Level 1, Level 2, and Level 3 as well as the top of the fuel. Indicate on this sketch the portion of the level sensor measurement range that is sensitive tomeasurement of the fuel pool level, with respect to the Level 1, Level 2, and Level 3 datum points.STPNOC Response See Figures 1 and 2 of this Attachment.

3.0 Instrumentation

and Design Features 3.1 Instruments and Arrangement The OIP states, in part, that Both the Primary and Backup Instrument Channels will utilize permanently-installed instruments....

The Primary Instrument Channel level sensing components will be located in the northeast corner of the Spent Fuel Pool, as shown in Attachment 1...

Attachment NOC-AE-13003008 Page 5 of 23 The Backup Instrument Channel level sensing components will be located in the northwest corner of the Spent Fuel Pool, as shown in Attachment 1....The current Plan is to mount the supporting electronic instruments outside of the spent fuel pool area, to provide a more benign radiation and environmental conditions, and also provide for reasonable and accessible locations for operators.

SFP Primary and Backup Channel Level Instruments are currently planned to be located in Radwaste Control Room of the Mechanical Auxiliary Building (MAB);however, STPNOC is still evaluating other possible locations (i.e. relay room).NRC RAI-2 Please provide a clearly labeled sketch or marked-up plant drawing of the plan view of the SFP area, depicting the SFP inside dimensions, the planned locations/

placement of the primary and back-up SFP level sensor, and the proposed routing of the cables that will extend from the sensors toward the location of the read-out/display device.STPNOC Response See Figure 3 of this Attachment.

3.2 Mounting

The OIP states, in part, that Consideration will be given to the maximum seismic ground motion that occurs at the installation location for the permanently installed equipment which is documented in the UFSAR [Updated Final Safety Analysis Report] Section 3.7.The mountings shall be designed consistent with the highest safety or seismic classification of the SFP. The level sensors will be mounted on seismically qualified brackets.NRC RAI-3a Please provide the following:

a) The design criteria that will be used to estimate the total loading on the mounting device(s), including static weight loads and dynamic loads. Please describe the methodology that will be used to estimate the total loading, inclusive of design basis maximum seismic loads and the hydrodynamic loads that could result from pool sloshing or other effects that could accompany such seismic forces.

m 0 0 CD 0 0 0 0.(D CD ID z 0 C~)CD -(Z) CD C)N) CD(

Attachment NOC-AE-13003008 Page 21 of 23 Figure 2 Proposed Mounting Arrangement o 0 O 0 A RADAR HORN ANTENNA VIEW A-A SOUTH TEXAS PROJECT ELECTRIC GENERATING STATIOD0 D0527 STI 33754102 OPOP03-ZG-0008 I Rev. 56 1. Page I of 84 Power Operations Quality Safety-Related Usage: IN HAND Effective Date: 09/12/2013 D. Rohan N/A Crew I C Operations PREPARER TECHNICAL USER COGNIZANT DEPT.Usage Table 4 4 3 3 3 1 4 3 3 3 3 3 3 4 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 of Contents Page Purpose .........................................

...........

.3 R eferences

......................................................................................

.... .......................

....3,>Preerencsies

./.6 Prerequisites

............................................................................................................

6 N otes and P recautions 7.................................................................................................

7 Steady State Operation

.. ....... .. ..17 P ow er R eduction 33.......................................................................................................

33 Pow er E scalation

............................................

..................

38 A lternate Power Increase to 100% Power ..........

.............................................

48 Coastdow n...........................

..... , .... .............................................

55 R ecords R eview ........................................................

................................................

60 Support D ocum ents ...............................

.................

.............................................

60 Addendum 1, Maximum Generator Loading Requirements With Heater Strings Isolated/B ypassed .........; ..............................................................................

61 Addendum 2, Main TuiliIine OperatihGuidelines

...............................................

63 Addendum 3, Fuel Conditioning Requirements/Recommendations

......................

64 Addendum 4, Generator Exciter Ope'rating Guidelines

.................................

67 Addendum 5, Nbminal. ,d Temperature Limit vs Power Level .........................

68 Addendum 6, Example of Indications and Actions For a Feedwater Ultrasonic Flow m eter System Failure ...........................................................................

69 ,Addendum

/7, Control-Loop Alignment

................................................................

71 Addendum 8, Percent Power vs Program RCS Tavg .............................................

75 OPOP03-ZG-0008 Rev. 56 Page 2 of 84 Power Operations 1 Form 1, Turbine Load Changes using the Limiter .................................................

76 1 Form 2, Turbine Load Changes using the Setpoint Controls ...............................

77 3 Form 3, (Automatic/Normal Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment

...................

79 3 Form 4, (Manual/Test Mode ONLY) Online Main Generator Voltage/Reactive Load A djustm ents .......................................................................................

81 1 Form 5, Routine at Power Minor Rod Movements

...............................................

83 Usage 1 -IN HAND 2 -IN HAND CONTROLLING STATION 3 -REFERENCED 4 -AVAILABLE This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 3 of 84 Power Operations

1.0 Purpose

1.1 This procedure provides guidance for operation at a normal steady state power level.1.2 This procedure provides instructions for reducing reactor power to less than 100% and for returning to 100% reactor power following a power reduction.

1.3 This procedure provides guidelines for Coastdown Operations when RCS boron concentration lowers to less than 5 ppm or at the discretion of the Plant Operations Manager.1.4 Forms 1 -5 may be used in Modes 1 and 2 (as applicable) in conjunction with 0POP03-ZG-0005, Plant Startup to 100% and 0POP03-ZG-0006, Plant Shutdown from 100% to Hot Standby.2.0 References

2.1 Technical

Specifications

2.1.1 Technical

Specification 3.1.3.1 2.1.2 Technical Specification 3.1.3.2 2.1.3 Technical Specification 3.1.3.6 2.1.4 Technical Specification 3.2.1 2.1.5 Technical Specification 3.2.2 2.1.6 Technical Specification 3.2.4 2.1.7 Technical Specification 3.3.1 2.1.8 Technical Specification 3.4.8 2.1.9 Technical Specification Table 4.4-4 Item 4.b 2.1.10 Technical Specification 3.3.5.1, 3.7.1.6 2.2 Plant Curve Book 2.3 Westinghouse Precautions, Limitations and Setpoints, 5ZO0OZSI 101 2.4 STPEGS Updated Final Safety Analysis Report 2.5 Core Operating Limits Report (COLR)This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 4 of 84!I Power Operations I 2.6 Plant Procedures:

2.6.1 0POP02-MS-0001, Main Steam System 2.6.2 0POP03-ZG-0005, Plant Startup to 100%2.6.3 0POP03-ZG-0006, Plant Shutdown from 100% to Hot Standby 2.6.4 OPSP03-NI-0001, Power Range NI Channel Calibration 2.6.5 OPSP10-NI-0001, Target Axial Flux Difference Determination 2.6.6 OPEP02-CU-0001, Calorimetric Verification 2.6.7 OPOPO1-ZA-0021, AC Electrical Notes and Precautions 2.6.8 OPEP02-ZX-0007.

100% Power Instrument Alignments 2.6.9 OPOPOI-TM-0001, Main Turbine/Generator Operations Guidelines 2.6.10 Conduct of Operations 2.6.11 OPOP02-CV-0001, Makeup to the Reactor Coolant System 2.7 Commitments:

2.7.1 MATS Item 8500029-860 (ST-HS-HL-4465 and OMR 84-214)2.7.2 MATS Item 8500034-866 (ST-UB-HL-047/GNL 84-02 1), Long Term Low Power Operation in PWRs 2.7.3 MATS Item 8500035-866 (OMR 84-214), Positive Moderator Temperature Coefficient During Power Ascension 2.7.4 MATS Item 8500051-866 (NRC Question 430.10), Assurance the Facilities Real and Reactive Power, Voltage and Frequency are Operated Within Limits 2.7.5 MATS Item 8801896-936 (ST-HS-HL-4465/SER 88-02 1), PWR Startups With High Control Rod Worth 2.7.6 MATS Item 8901156-936 (SPR 89-0487), Licensed Power Level May Have Been Exceeded 2.7.7 MATS Item 9000966-936 (SPR 90-0380), Overpower Transient Occurred After Placing a Mixed Bed Demineralizer in Service This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 5 of 84!I Power Operations I 2.7.8 PIE 96-060-8, Feedwater Heater Operating Recommendations

2.8 Power

Coastdown Operating Procedures (ST-UB-HL-01 130)2.9 CVCS Pump Operation (ST-HS-HS-2 1262)2.10 Response to SOER 90-003 2.11 ST-HS-HS-31173, Reactor Power Ramp Rates 2.12 CR 95-1543, Excessive Reactor Power Increase 2.13 ST-P2-P2-249, Exceeding the Full "Steady-State" Licensed 100% Power Level 2.14 CREE 96-2054-13, Feed Temperature Reduction to Achieve 100% Power 2.15 CREE 96-16047, Evaluation on main generator output for Unit I when the Aux Transformer is out of service.2.16 USQE 97-0030 Feedwater inlet temperature to the Steam Generators (SG) to be operated in a range between 440'F and 420'F 2.17 CREE 97-13664-4, Evaluation of OPOP03-ZG-0008 Addendum 1 and 5 setpoints 2.18 ST-UB-HL-434, Limitations and Conditions for Westinghouse Fuel Operation 2.19 ST-UB-NOC-02002222, NT-TG-02-3, Jan 29, 2002, Limitations and Conditions for Westinghouse Fuel Operation, Rev 6 2.20 ST-WN-NOC-02000226, South Texas Feedwater Temperature Loop Variations 2.21 Westinghouse Instruction Book, 1,311,838 KW Steam Turbine Operation and Control, (VTD-W 120-0003)2.22 Siemens Turbine Generator Instruction Book 1387-MW Turbine Generator 13A5281 /1 3A5291 (ETB-TMOO-9001 through ETB-TMOO-9031) 2.23 CR 07-3018, North American Electric Reliability Corporation (NERC) standards for Bulk Electrical Power Systems.2.24 CREE #06-14665-1, Evaluate current method of operating the Main Turbine Load Control 2.25 NC-07088, Allowable Feedwater Operating Temperatures At Various Power Levels This procedure section, when completed, SHALL be retained.

2.26 NRC RIS 2007-21 Rev. 1, Adherence To Licensed Power Limits 2.26.1 NEI Position Statement for Guidance to Licensees on Complying with the Licensed Power Limits (ADAMS ML081750537) 2.26.2 Safety Evaluation Regarding Endorsement of NEI Guidance for Adhering to the Licensed Thermal Power Limits (ADAMS ML082690105) 2.27 CREE 11-5655-1, FWH 15/16(25/26')

strings have the potential to isolate during plant shutdown 2.28 CR 13-6385-1, Potential violation of North American Electric Reliability Corporation (NERC) Reliability Standard VAR-002-2b Requirement R3.1 due to an administrative Unit 1 Reactive Power limitation.

3.0 Prerequisites

3.1 Forms

1, 2 and 5 SHALL be used in Modes 1 and 2, as required.3.2 Forms 3 and 4 may be used in Modes I and 2, as required.3.3 This procedure is to be entered above 65% reactor power, unless plant operation is being governed by a different operating procedure.

3.4 This procedure may be entered at any reactor power level during Coastdown Operations.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 7 of 84 Power Operations

4.0 Notes

and Precautions 4.1 The principles of OPGP03-ZO-0042, Reactivity Management Program, are in effect at all times during Operations in this procedure.

4.2 Operating

PROCEDURES are written based on a defined set of plant conditions and equipment availability.

PROCEDURE changes are NOT required to document alternate performance based on conditions different from those assumed if the PROCEDURE can be performed safely. The decision to proceed lies with the Shift Manager/Unit Supervisor and is based on knowledge of system design and operation and the impact of omitting or re-sequencing steps.4.2.1 The Shift Manager/Unit Supervisor may authorize alternate performance for operating PROCEDURE sequence, including omitting steps, based on plant operating conditions.

The Shift Manager/Unit Supervisor ensures such an alternate performance does NOT adversely impact the safety of personnel or equipment, and documents the alternate method in the appropriate PROCEDURE or logbook. See OPGPO3-ZA-00 10, Performing and Verifying Station Activities for specific details.4.2.2 The Shift Manager/Unit Supervisor may authorize early start of procedure steps to enhance plant performance, when the early start is of no safety impact for current plant conditions.

Documentation is NOT required for an early start as long as the step is completed before moving past this step in the overall sequence.4.2.3 Steps within this procedure SHALL be performed in order listed or in order provided in an authorized early start (Step 4.2.2) or alternate performance (Step 4.2.1). Steps that are authorized to be omitted SHALL be designated by placing "N/A" in the signoff or initial blanks. See OPGPO3-ZA-0010, Performing and Verifying Station Activities, for specific details.4.3 IF Tavg lowers to < 561 'F, THEN Comply with Technical Specification 3.1.1.4. IF Tavg cannot be restored to GREATER THAN 561 'F within 15 minutes, THEN manually trip the reactor and GO TO OPOP05-EO-EOOO, Reactor Trip Or Safety Inljection.

4.4 The Shift Manager/Unit Supervisor SHALL signoff or initial all steps unless otherwise designated within this procedure.

This does not include Forms 1 through 5 which are performed by Reactor Operators.

4.5 Control

Rod Insertion limits SHALL be observed at all times. (Technical Specification 3.1.3.6)This procedure section, when completed, SHALL be retained.

4.6 Reactor

Power monitoring during a power reduction or escalation:

• WHEN changing Reactor Power, THEN RCS loop Delta-T power indication should be used until Reactor Power and secondary conditions have stabilized.

Diverse power indications should also be monitored as a backup to confirm RCS loop Delta-T power (e.g. Nis and U 1118).* WHEN steady state conditions are reached at any power level and U 1118 has been confirmed to be accurate, THEN UI 118 should be primary power indication used while also monitoring diverse indications.

4.7 Changes

in Tavg, axial and radial power distribution, and control bank position affects the accuracy of the Power Range NI detectors. (Reference 2.10)* Alternate indications of reactor power should be compared to NI power. Alternate indications include average RCS loop Delta-T, plant computer calorimetric estimates, turbine impulse pressure, and turbine generator load.* IF the Power Range NI detectors and calorimetric power indicate a difference greater than 2%, THEN reactor power should be stabilized and the respective NI detectors calibrated. (Technical Specification 4.3.1.1.2.a.2) 4.8 The Main Generator SHALL be operated within the limits of the Main Generator Capability Curve in the Plant Curve book. (Reference 2.7.4)4.9 Reactor power, control bank position and RCS boron concentration SHALL be maintained within the limits established by the Rod Withdrawal Limits Curve contained in the Plant Curve Book, as applicable.

4.10 RCS Average Temperature (Tavg) SHALL be maintained in accordance with Step 5.2"RCS Temperature (Tavg) Control" utilizing rod movement, dilution, boration or turbine load changes, as necessary.

During "Coastdown Operations" RCS Average Temperature (Tavg) SHALL be maintained in accordance with Section 9.0.4.11 Addendum 3, Fuel Conditioning Requirements/Recommendations, SHALL be referenced for guidance on Control Rod movement and Main Turbine loading rates.4.12 Control Rod alignment SHALL be monitored between control rods within banks AND between the bank group step counters. (Technical Specification 3.1.3.1 and 3.1.3.2)4.13 Caution SHALL be exercised when moving control rods in regions of high differential rod worth. Small changes in control rod position can produce large reactivity changes in these regions. (Reference 2.7.5)This procedure section, when completed, SHALL be retained.

4.14 Critical parameters SHALL be adjusted, as necessary, to ensure greater than or equal to 5%margin to trip (OTDT) including lowering RCS temperature, rising RCS pressure and/or lowering reactor power.4.15 Pressurizer boron concentration SHALL be maintained within 50 ppm of Reactor Coolant System (RCS) boron concentration.

4.16 Chemistry SHALL be coordinated with, as necessary, to maintain RCS and secondary chemistry specifications.

4.17 RCS degassification may be performed as determined necessary by the Shift Manager and Chemistry Supervisor, by placing a Nitrogencover gas on the VCT.4.18 IF Reactor Thermal Power changes of greater than 15% Rated Thermal Power are performed within any one (1) hour period, THEN Chemistry SHALL be notified to perform an isotopic analysis of the RCS to satisfy the requirements of Technical Specification 3.4.8, Table 4.4-4, Item 4.b.4.19 WHEN steady state operation at less than 85% of Rated Thermal Power for greater than two weeks is anticipated, THEN Reactor Engineering SHALL be notified to recommend an AFD control strategy.

Prior to raising Reactor Thermal Power to greater than 90%following operation at less than 85% of Rated Thermal Power for greater than two weeks, an Core Power Distribution Measurement SHALL be performed to ensure that core thermal limits are NOT exceeded per Technical Specification 3.2.2. (Reference 2.7.2)4.20 Main Generator reactive load SHALL be adjusted as requested by the Systemii Load Dispatcher, NOT to exceed the limits of the Main Generator Capability Curve in the Plant Curve Book.4.21 The Main Turbine Governor Valves should be placed at the valve position limiter or slightly below the valve position limiter to minimize the potential for unexpected power excursions.

This condition may be relaxed for any of the following conditions:

• During testing* At the discretion of the Shift Manager/Unit Supervisor

• When EHC system problems occur 4.22 WHEN a Feedwater Heater string is isolated and/or bypassed, THEN the requirements of Addendum 1, Maximum Generator Loading Requirements With Heater Strings Isolated/Bypassed, SHALL be observed.4.23 Maintain feedwater temperature within limits of Addendum 5 "Nominal Feed Temperature Limit vs Power Level".This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 10 of 84 Power Operations 4.24 (UNIT 1 ONLY) IF the Auxiliary Transformer is NOT in service, AND the Main Generator is operating in a lead direction, THEN Main Generator output SHALL be limited to less than or equal to 1304 MWe with MVARs less than 515 MVARs. This limit is due to a lower transformer rating on MST IA. (CREE 96-16047)4.25 (UNIT 1 ONLY) IF the Auxiliary Transformer is NOT in service, AND the Main Generator is operating in a lag direction, THEN Main Generator output SHALL be limited by the Generator Capabilities Curve. (CREE 96-16047)4.26 IF both Isolated Phase Bus Duct Cooling fans are inoperable, THEN Main Generator load SHALL be reduced at 5% per minute until generator current is less than or equal to 18.5 K amps, as indicated on CP007.4.27 The Main Turbine SHALL NOT be operated with Main Turbine Exhaust Pressure (Condenser Pressure "inHG") in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines.

IF the Main Turbine is operating in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines, THEN the following SHALL be performed within 5 minutes: (Reference 2.22)Main Turbine Exhaust Pressure (Condenser Pressure "inHG") SHALL be returned to a value outside the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines OR Main Turbine load (MWe) SHALL be lowered to a value outside the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines 4.28 IF operation outside the "Restrictive Zone" of Addendum 2, Main Turbine Operating Guidelines can NOT be established within 5 minutes, THEN the Main Turbine SHALL be removed from service (placed on the jacking gear) in a safe and expeditious manner.(1-5% per minute) and remain out of service until an Engineering Evaluation can be performed.

4.29 Main Turbine Exhaust Pressure should be at the lowest value attainable (i.e., Best Vacuum) during Turbine Startup and Low Load Operation to prevent the overheating of the LP turbine blading and excessive thermal expansion. (Reference 2.22)4.30 Steady state power control is normally using the Setpoint Controls for Turbine Load Changes. Limiter control is only to be used if needed when Setpoint Control in not available or during transient conditions. (Ref. 2.24)This procedure section, when completed, SHALL be retained.

4.31 Main Turbine load may be raised and lowered in either OPERATOR AUTO or MANUAL control at the Shift Manager/Unit Supervisors discretion.

Refer to OPOPOI-TM-0001, Main Turbine/Generator Operations Guidelines.

4.32 Using the "SETPOINT CONTROL" and Load Rate provides a far more controlled load adjustment when responding to present operational considerations instead of using the Limiter that moves the valves at a rate of 133% per minute to make the desired load change. (Ref. 2.24)4.33 Steady state power control is normally in IMP-OUT, with the limiter at a position at or slightly above the current operating demand and relying on the use of the "SETPOINT CONTROL" and Reference counters (Turbine Actual and Setpoint), with the ramp rate set to 0.25% MW/MIN to make the Main Turbine adjustments required to maintain steady state Reactor Power. (Ref. 2.24)4.34 Evolutions and activities that have the potential to raise reactor power above 100%SHALL be controlled by the expectations of Conduct of Operations Chapter 10 "On-Line Maintenance Expectations".

4.35 Each SG PORV SHALL remain OPERABLE, in AUTO with a nominal setpoint of 1225 +/- 7 psig. (MODES 1, 2) (REF 2.1.10)4.36 IF SG PORVs are required to be controlled in manual operation, OR in automatic operation with reduced setpoints, THEN an OAS entry is required to ENSURE compliance with Technical Specifications 3.3.5.1, and 3.7.1.6. (REF 2.1.10)4.37 Raising power using 1(2)-MOV-0 108, "FEEDWATER HEATER I IA/I IB(21A/21B)

BYPASS ISOLATION MOV" SHALL NOT be perfon-ned other than described in Section 8.0 "Alternate Power Increase to 100% Power". (USQE 97-0030)4.38 Opening the 1(2)-MOV-0108, "FEEDWATER HEATER I A/ 11B(21A/21B)

BYPASS ISOLATION MOV" may have adverse consequences on the Feedwater Ultrasonic Flowmeter (FW UTF) System. A rise in streaming in the FW Header related to bypassing cool water around the HP FWH may cause the standard deviation of flow measured by FW UTF to exceed Acceptance Criteria.

IF standard deviation criteria are NOT met, THEN FW UTF will roll out on the affected Loop. IF Opening the 1(2)-MOV-0108 is required, THEN request an EVALUATION by engineering (vendor AMAG) prior to bypassing the high pressure FWHS and leaving UTF in service (CR 04-7089).This procedure section, when completed, SHALL be retained.

4.39 IF it becomes necessary to operate Turbine Controls in Manual Mode: THEN all Turbine Control procedural steps referencing "Auto" Control Modes SHOULD be interpreted as "Manual" Control Mode.* Since the burden on the operator is greatly raised during turbine manual operation, the operator SHOULD NOT start (rollup) the turbine in the Manual Mode unless it is unavoidable and approved by the Shift Manager/Unit Supervisor.

• Manual Operation of Turbine Controls including transitioning between "Manual" and"Auto" Control Modes is considered "Skill of the Craft" Refer to OPOPO I -TM-0001, Main Turbine/Generator Operations Guidelines.

4.40 IF the Turbine Controls automatically transfers from "Auto" to "Manual" Control Modes, THEN Controls SHALL NOT be returned to "Auto" until the cause of the transfer has been evaluated and the Shift Manager/Unit Supervisor authorized returning to "Auto" control. Refer to OPOP01-TM-0001, Main Turbine/Generator Operations Guidelines.

4.41 Use of the "PREFERRED" control channels is an Operations good practice.

Addendum 7, Control-Loop Alignment may be performed at any time as directed by the Shift Manager/Unit Supervisor.

4.42 IF the steam dump system does NOT operate properly to maintain Tavg, THEN the Shift Manager/Unit Supervisor MAY direct actions to control RCS Tavg until the malfunctioning steam dump(s) can be repaired or isolated.4.43 WHEN Steam Dump or the Steam Chest drains open, THEN CLOSE the Steam Dump or the Steam Chest drains, as desired to reduce secondary heat loads.4.44 RCS/PZR Pressure Control during steady state and transients, SHOULD be performed by the following: (WHEN controls are NOT responding as expected, THEN Manual intervention for pressure control is AUTHORIZED)

• Automatic control of Pressurizer Pressure Controller RC-PK-0655A (PZR Press Master Controller) to Control the output of the PZR CONTROL HTRS* Automatic control of Pressurizer Spray Valve Controller "PRZR SPR PCV-0655B" and "PRZR SPR PCV-0655C" to Control RCS/PZR Pressure* Backup Heaters may be cycled as necessary to aid in PZR turnover flow or pressure control.This procedure section, when completed, SHALL be retained.

4.45 The following is the Master Controller outputs in Manual Operation:

Function Controller Output VDC Controller Output % Signal Direction PCV-0655A OPENS 8.75 87.5 [NC (4)PCV-0655A CLOSES 7.50 75.0 DEC(+)PZR PRESS DEV HI 7.19 72.0 INC (4)ALARM SPRAY FULL OPEN 7.19 72.0 INC (4)PRES PRESS DEV HI 6.99 70.0 DEC (4)ALARM RESETS SPRAY FULL 4.06 40.5 DEC (4)CLOSED CONT HTRS 0% PWR 3.44 34.5 INC (4)CONT HTRS 50% 2.50 25.0 N/A PWR CONTROL HTRS 1.56 15.5 DEC (40)100% PWR PZR PRESS DEV LO 0.94 9.5 DEC (4)ALARM & BU HTRS ON 4.46 For NORMAL plant operations, THEN do NOT transfer the Main Turbine IMP PRESS FEEDBACK mode of operation WHEN Reactor Power is greater than 98%. Transferring the Main Turbine to IMP PRESS FEEDBACK to "IN or "OUT" mode, may result in a momentary load swing. Refer to OPOPOI-TM-0001, Main Turbine/Generator Operations Guidelines.

4.47 Placing Excess Letdown in service will cause Plant Computer point U 1118 to become inaccurate in the NON-Conservative direction due to bypass flow causing approximately 2 MWth to bypass instrumentation.

This bypass flow and heat is NOT accounted for in the calorimetric.

This procedure section, when completed, SHALL be retained.

4.48 The following forms provide guidance for aid in the operation of the Main Turbine/Generator and Reactor:* OPOP03-ZG-0008, Form 1, Turbine Load Changes using the Limiter* OPOPO3-ZG-0008, Form 2, Turbine Load Changes using the Setpoint Controls* OPOP03-ZG-0008, Forn 3, (Automatic/Normal Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment* OPOP03-ZG-0008, Form 4, (Manual/Test Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

• OPOPO3-ZG-0008, Form 5., Routine at Power Minor Rod Movements* OPOP02-CV-0001, Form 1, Modes 1-2 RCS Boration Checklist* OPOP02-CV-0001, Form 2, Modes 1-2 RCS Dilution Checklist OPOP02-CV-0001, Form 3, Modes 1-2 RCS Alt Dilution Checklist* OPOP02-CV-0001, Form 4, Modes 1-2 Automatic Operation Checklist 4.49 IF using the Plant Computer Point UI 118, THEN VERIFY Each loop's 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average Loop Mismatch Powers ( Computer points U1 121P, UI 122P, Ul 123P, and Ul 124P)equals zero ( 0 ) MW plus or minus two (+2) MW.4.50 Minimize operation in the VARs IN, Leading (Underexcited) region to limit eddy-current heating of the turbine end stator core. Operation in this region, or below the drawn-in curve (Main Generator Capability Curve in the Plant Curve Book) may cause a rise in heating and subsequent damage proportional to the depth and duration of operation below the curve. (CREE 07-4972-2) 4.51 For detailed information concerning Calorimetric Computer Instruments and Constants, Feedwater UTF Correction Information, U 1118 Root Points, Use of U 1118 Root Points and Time Averaged Root Points and Ul 118 Reliability Verification REFER to OPEP02-CU-0001, Calorimetric Verification.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 15of 84 Power Operations 4.52 Due to differences in the steam leakage from the HP turbine glands as load is raised/lowered.

Manual adjustments may be performed, as necessary, on the following components to balance HP turbine gland steam pressure to acceptable values LAW OPOP02-GS-0001, Turbine Gland Seal Steam System:* 1(2)-GS-0201, GLAND STEAM ALTERNATE SPILLOVER BYPASS VALVE* 1(2)-GS-PC-6153, TURBINE STEAM SEAL HIGH PRESSURE TURBINE GLANDS PRESSURE CONTROLLER 1(2)-GS-MOV-0079, GLAND STEAM SPILLOVER PV-6156 BYPASS MOV OPERATOR 1(2)-GS-PC-6156, TURBINE STEAM SEAL SPILLOVER PRESSURE CONTROLLER 4.53 During Coastdown Operations or plant conditions where calculated Tref is in error (e.g., instrument failure, turbine offline, etc), Program Tavg should be utilized for Tavg vs. Tref comparisons. (REFER to Addendum 8, Percent Power vs Program RCS Tavg.)4.54 In this procedure, WHEN the Main Turbine is online and the Tref instrumentation is functional, THEN the terms "Tref' and "Program Tavg" are to be considered synonymous.

The indicated "Tref' may be substituted for "Program Tavg". WHEN using the indicated"Tref', THEN referral to Addendum 8, Percent Power vs Program RCS Tavg is NOT mandatory.

4.54.1 "Tref' and "Program Tavg" are only truly equal at 100% power as "Tref' is not linear with "Program Tavg" once you go below 100% power ("Program Tavg" is linear from 0 to 100% power and "Tref' is NOT linear from 0 to 100% power).When stabilizing at a power level less than 100%, there may be some degree of temperature mismatch which needs to be evaluated prior to returning control rods to 'auto'. Past history has shown that automatic control rod motion may occur with as little as 0.3°F mismatch between "Tref' and "Program Tavg".This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 16 of 84 Power Operations 4.55 WHEN the Reactor Power or Turbine Load values/ranges provided within this procedure are NOT obtained, THEN the Shift Manager/Unit Supervisor may authorize expanded ranges for the associated procedure step(s) based upon the following rules:* (Rx Pwr = all) Expanded values/ranges for Rx Pwr are NOT allowed when meeting Technical Specification required Power Levels (Example:

Power level to meet a QPTR requirements)

• (Rx Pwr > 98%) Expanded values/ranges are NOT allowed* (Rx Pwr 10% -98%) Rx Pwr values/ranges may be expanded by + 2% Rx Pwr* (Rlx Pwr 10% -98%) Turbine Load values/ranges are approximations and may vary as seasons and systerms change. If specific Turbine Load values are required they will be specified without the approximation symbol.* (Rx Pwr < 10%) Expanded values/ranges are NOT allowed.4.56 FWH 15/16(25/26) strings have the potential to isolate during plant shutdown. (Refer to CREE 11-5655-1) 4.57 The STP Coordinator and the TDSP SHALL be notified within 30 minutes when tile Main Generator Voltage Regulator is in Manual, and the expected duration (Use 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if unknown). (REF 2.28)4.58 IF there is a change to the reactive capability of the Main Generator, THEN the STP Coordinator and the TDSP SHALL be notified within 30 minutes what the chan2e is to the reactive capability of the Main Generator, and the expected duration (Use 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if unknown). (REF 2.28)4.59 For planned engineering changes to the Main Generator reactive capability, the required 30 minute notification will be coordinated by Engineering with the Control Room. (REF 2.28)This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 17 of 84 Power Operations

5.0 Steady

State Operation 5.1 Pressurizer Heater Operation 5.1.1 During steady state operation, Pressurizer Heaters should be maintained in AUTO with pressure being controlled by the variable heater output.5.1.2 Pressurizer Back-Up Heaters may be energized, as required, for the following reasons:* RCS pressure control* Boron equalization

• Mitigate transients

• At the discretion of the Operating Crew 5.1.3 The controlling pressurizer pressure transmitter with the lowest reading should be selected.

This will provide a larger margin to the OTDT setpoint and DNB.This procedure section, when completed, SHALL be retained.

5.2 RCS Temperature (Tavg) Control 5.2.1 If Tavg lowers to < 561"F, Comply with Technical Specification 3.1.1.4. IF Tavg cannot be restored to GREATER THAN 561'F within 15 minutes, THEN manually trip the reactor and GO TO OPOP05-EO-EOOO, Reactor Trip Or Safety Injection.

NOTE* Maintaining RCS Tavg outside the band specified in Step 5.2.2 during steady state conditions SHALL require a cycle specific evaluation (documented by an approved CREE) and implementation by a CROE or procedure revision.* During periods of maintenance on the highest reading Tavg channel, monitor the next highest reading channel and maintain it at a constant temperature.

5.2.2 During

steady state conditions with NO Xenon transients in progress, maintain RCS Tavg within the following band relative to Programmed Tavg. The Auctioneered High Tavg SHALL be the controlling channel: UNIT 1(2): + 0.5°F of Programmed Tavg (at all times)5.2.2.1 IF RCS Tavg can NOT be maintained within the bands specified in Step 5.2.2 during steady state conditions, THEN ENSURE Engineering has been notified to evaluate the condition.

5.2.3 During

plant transient conditions lasting no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Tavg may be maintained at a lower value (more negative than -0.5°F of Programmed Tavg), BUT within 1.5°F of Programmed Tavg for any of the following reasons:* To support testing or maintenance for raising the OTDT setpoints* To minimize spurious alarms When deemed necessary by the Shift Manager This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 19 of 84 Power Operations 5.3 Rod Control System 5.3.1 The Rod Control System will normally be maintained in Automatic, however, Rod Control may be placed in Manual for any evolution, as deemed necessary by the Reactor Operator.5.3.2 Control rod position SHALL be monitored to ensure the following:

• Rod Insertion Limits are met at all times, AND* Rod Withdrawal Limits are maintained, when applicable.

5.3.3 USE caution during Control Rod movement in regions of the core with relatively high differential rod worth. (Reference 2.7.5)5.4 Axial Flux Difference (AFD) Control 5.4.1 AFD should be maintained at or near the target.5.4.2 Reactor Engineering may periodically request AFD to be controlled away from the target in order to meet requirements for an upcoming procedure or in order to establish a new target. AFD may be controlled away from the target under these conditions with concurrence of the Shift Manager.5.4.3 To minimize Rod Control System usage, RCS temperature may be adjusted to control AFD. This is done by turbine adjustments to raise/lower power or dilution/boration as desired. Less than a 0.5% power change should be sufficient to control AFD.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 20 of 84 Power Operations 5.5 Main Turbine Generator Controls 5.5.1 During steady state power operations the Main Turbine will normally be in the IMP PRESS FEEDBACK "OUT" mode of operation.

5.5.2 IF required by a particular procedure or testing, THEN the IMP PRESS FEEDBACK "IN" mode of operation may be used with concurrence of the Shift Manager/Unit Supervisor lAW OPOPO0-TM-0001, Main Turbine/Generator Operations Guidelines.

5.5.3 PLACE

the valve position limiter at a position at the current operating demand or slightly above the current operating demand as directed by the Shift Manager/Unit Supervisor to minimize the potential for unexpected power excursions. (Ref. 2.24)5.5.4 CONTROL bus voltage as directed by the System Load Dispatcher.

MVAR loading should be coordinated between the units.5.5.5 Minimize operation in the VARs fN, Leading (Underexcited) region to limit eddy-current heating of the turbine end stator core. Operation in this region, or below the drawn-in curve (Main Generator Capability Curve in the Plant Curve Book) may cause a rise in heating and subsequent damage proportional to the depth and duration of operation below the curve. (CREE 07-4972-2)

5.5.6 Generator

MVAR output should be maintained as close to a balanced MVAR load between Units I and Unit 2 as possible, while maintaining MVAR load IAW the Main Generator Capability Curve (in the Plant Curve Book), grid stability and minimizing vibration on the Turbine Generator bearings.This procedure section, when completed, SHALL be retained.

I 0POP03-ZG-0008 Rev. 56 Page 21 of 84 E Power Operations

5.6 Reactor

Power Controls NOTE For detailed information concerning Calorimetric Computer Instruments and Constants, Feedwater UTF Correction Information, Ul 118 Root Points., Use of Ul 118 Root Points and Time Averaged Root Points and UI 118 Reliability Verification REFER to 0PEP02-CU-0001, Calorimetric Verification.

5.6.1 Reactor

power should normally be controlled based on Reactor Total Thermal Power indication from Plant Computer point U 1118 provided the computer point is operable and consistent with other primary plant indications (e.g., calorimetric power, delta-T's, etc). (Reference 2.7.6)5.6.2 IF Plant Computer point U 1118 is used to obtain calorimetric power, THEN VERIFY the following:

Each loop's 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average Loop Mismatch Powers ( Computer points U 1121P, Ul 122P, Ul 123P, and UI 124P ) equals zero (0) MW plus or minus 2 MW.* Reactor power as indicated by all operable Delta-Ts is greater than 30%,* Plant Computer point U 1118 has been verified operable per OPEP02-CU-0001, Calorimetric Verification within the last eight days, IF reactor power has lowered to less than 30% since the last performance of OPEP02-CU-0001, Calorimetric Verification, THEN VERIFY U 1118 operable per OPEP02-CU-0001, Calorimetric Verification, prior to use for calibration, and Prior to use for NI calibration following a 30 day or longer outage, THEN VERIFY the Calorimetric Constant values are as defined in OPEP02-CU-0001, Calorimetric Verification, Addendum 3.5.6.3 IF Plant Computer point U 1118 is inoperable, THEN Reactor Power should be controlled based on Average Power Range Channel NI, U 1169, or other available power range channel NI indications and other primary plant indications (e.g., Delta-T's). (Reference 2.7.6)This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 22 of 84 Power Operations 5.6.4 WHEN Plant Computer point U 1118 is unreliable, THEN Plant Computer point UI 118 SHOULD be placed in Test Mode with POOR quality. {"P" clarifier should appear after the U 118 value indicating "POOR" quality and "T" clarifier should appear after the U 1118 value indicating "TEST MODE"}5.6.4.1 WHEN the condition(s) which caused Plant Computer point Ul 118 to be unreliable is restored, THEN Plant Computer point U 1118 SHOULD be placed in Test Mode "OFF".5.6.4.2 WHEN Plant Computer point UI 118 is "POOR" quality (for a sufficient duration), THEN the time average points of Plant Computer point U 1118 will go to BAD quality.5.6.4.3 WHEN Plant Computer point U 1118 is POOR quality, THEN Plant Computer point UI 118 will NOT be used by BEACON.5.6.5 [F Plant Computer Mode goes to BAD quality, THEN U 1118 will go POOR.5.6.5.1 IF the Plant Computer Mode is BAD for longer than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, THEN the UTF will roll out. After Plant Computer Mode goes to GOOD quality, it may take up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for Ul 118 to return to GOOD quality. This time required is for the UTF to roll back in and the smoothing buffer to refill with good data.5.6.5.2 IF Plant Mode goes to BAD quality in Modes 1 or 2, THEN manually enter the applicable Plant Mode using the Plant Mode Nuclear Application.

5.6.6 Operating

at Rated Thermal Power (RTP).5.6.6.1 WHEN any portion of the Feedwater Ultrasonic Flowmeter (FW UTF) System is out of service or unavailable, THEN reactor power SHALL be reduced to less than or equal to 99.6% RTP prior to the next Technical Specification required performance of OPSPO3-NI-0001, Power Range NI Channel Calibration.

5.6.6.2 WHEN reactor power is less than or equal to 99.6% RTP AND any portion of the FW UTF System is out of service or unavailable, THEN reactor power SHALL be limited to less than or equal to 99.6% RTP.This procedure section, when completed, SHALL be retained.

5.7 Feedwater

Ultrasonic Flowmeter (FW UTF) System NOTE* Each secondary feedwater loop has an ultrasonic flow measurement device for monitoring feedwater.

flow in addition to the venturi flow monitoring system. The venturi flow monitors are typically conservative in their estimate of feedwater flow (erroneously high feedwater flow) due to fouling. The FW UTF information is used to generate a correction factor which is applied to the loop feedwater flow measurement value. The correction factor is NOT applied below reactor power levels of 80% RTP.* The FW UTF Correction Factor can be automatically removed from the UI 118 calculation by the FW UTF program in cases of bad data, excessive data deviation statistics or loss of communication with the FW UTF System. In all cases, the correction factor is removed by linearly ramping it out of the calculation over a period of time.* For detailed information conceming Calorimetric Computer Instruments and Constants, Feedwater UTF Correction Information, U 1118 Root Points, Use ofUl 118 Root Points and Time Averaged Root Points and Ul 118 Reliability Verification REFER to OPEP02-CU-0001, Calorimetric Verification.

5.7.1 IF reactor power is greater than or equal to 80% RTP AND the correction factor generated from the FW UTF is NOT applied at its maximum allowable programmed value for any I or more feedwater loops, THEN the following actions will occur:* a "C" clarifier will appear after the UI 118 value (conditionally calculated)

• UI 118 will be reading conservatively higher since the maximum allowable programmed correction factor value is NOT being applied" the U 1118 value and its trendline will change to a color of white" Annunciator Window 4M08-C5 "Plant Computer System Alarm" will alarm as a result of the appropriate alarm on the Primary Alarm Page as shown below:* HIGH1 U1118 REACTOR THERMAL OUTPUT MW AND one or more of the following alarms:* ICCCROO06 LP A UI 118 ULTASON FLOW REM* ICCCROO07 LP B U 1118 ULTASON FLOW REM* ICCCROO08 LP C UI 118 ULTASON FLOW REM" ICCCROO09 LP D Ul 118 ULTASON FLOW REM This procedure section, when completed., SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 24 of 84 Power Operations

5.7.2 Feedwater

Ultrasonic Flowmeter (FW UTF) System out of service or unavailable.

5.7.2.1 The FW UTF System SHALL be considered out of service or unavailable for determining the applicable power limit when the following conditions occur:* Any of the current FW UTF Correction Factors have a "S" or"P" quality tag:* ICCUR0411* ICCUR0431* ICCUR0451* ICCUR0471" Any of the current FW UTF Correction Factors have a "C" clarifier with reactor power greater than 99.6%.* ICCUR0411 0 ICCUR0431* ICCUR0451* ICCUR0471" Any portion of the FW UTF System is known to be inoperable (i.e., one of the required inputs out of tolerance).

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 7 Rev. 56 Page 25 of 84 Power Operations NOTE Prior to performing OPSP03-NI-0001, reduce reactor power enough to ensure the calculated reactor power will NOT be greater than the applicable power limit.IF a reactor power reduction is required because the FW UTF System went out of service and correction factors rolled out, THEN the power reduction as indicated by NI power will need to be greater to ensure the calorimetric power with the Correction Factors rolled out is less than the reactor power limit.5.7.2.2 WHEN any portion of the FW UTF System is out of service or unavailable, THEN reactor power SHALL be reduced to less than or equal to 99.6% RTP prior to the next Technical Specification required performance of OPSP03-NI-0001, Power Range NI Channel Calibration.

5.7.2.3 REFER to Addendum 6 for an example of expected indications and acceptable actions assuming a failure of one of the FW UTF System transducers.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 26 of 84 I : Power OperationsI NOTE A "PC" clarifier can be caused by symptoms other than the FW UTF program detecting a sudden defouling event. A "P" signifies a poor quality value and is a result of several different sources. Only a sudden defouling event will give the Primary Alarm Zone alarms as shown in the following step.5.7.3 IF sudden defouling of the venturis is detected THEN the following actions will occur:* the U1118 value may be reading non-conservative since the perceived mass 0 0 0 0 flow rate value is now reduced a "P" clarifier will appear after the UI 118 value (poor quality)a "C" clarifier will appear after the Ul 118 value (conditionally calculated) the Ul 118 value and its trend line will be colored white Annunciator Window 4M08-C5 "Plant Computer System Alarm" will alann as a result of the appropriate alarm on the Primary Alarm Page as shown below: 0 HIGHI U1118 REACTOR THERMAL OUTPUT MW AND one or more of the following alarms: 0 0 0 0 ICCCR0010 ICCCROOI ICCCROO12 ICCCROO13 LP A U 1118 FEED FLOW DEFOUL LP B U 1118 FEED FLOW DEFOUL LP C U 1118 FEED FLOW DEFOUL LP D UI 118 FEED FLOW DEFOUL This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 27 of 84 Power Operations 5.7.4 IF sudden defouling of the venturis is detected THEN Ul 118 should NOT be used as an indication of primary reactor power and alternate indications of reactor power other than U 1118 should be used until the correction factor is removed fr'om Ul 118 or until Engineering has determined that Ul 118 is acceptable for use.5.7.4.1 5.7.4.2 Reactor power should NOT be raised in response to the non-conservative lower U 11 18 indication.

Notify Engineering and Maintenance of the occurrence to perform an investigation and institute any corrective actions.NOTE"Ultrasonic Feedwater Flow Status" may be viewed on Plant Computer display FW-16.5.7.5 A predefined report may be printed from the Plant Computer showing the"Ultrasonic Feedwater Flow Status" per the following steps: 5.7.5.1 Select the "menu" icon.5.7.5.2 Select "Predefined Reports".5.7.5.3 Select "UTF Status".This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 28 of 84 Power Operations 5.7.6 WHEN the ICS Plant MODE indicator is NOT functioning the effect on UI 118 and the UTF correction factor are as follows: 5.7.6.1 Plant Mode uses an average NI and average IR power from QDPS to determine mode.5.7.6.2 Flow corrections for FW Flow, Steam Flow, Blowdown Flow only get temperature and pressure corrected in Mode I and 2.Therefore, all the corrected flows go to FAIR quality when Mode goes to BAD.5.7.6.3 After approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of feeding UTF FW Flow with FAIR quality, the UTF 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> buffer does NOT have enough GOOD quality data and the UTF corrections is rolled out.5.7.6.4 WHEN Mode is manually set to Mode 1, THEN the corrected flows should go to GOOD quality.5.7.6.5 IF Mode goes BAD quality, THEN U 1118 will go POOR.5.7.6.6 IF Mode is BAD for longer than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, THEN the UTF will roll out.5.7.6.7 Depending on how long Mode was bad, it may take up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for U 1118 to return to normal. Approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the Mode is manually reset, the UTF correction should roll in. Over the next hour the smoothing buffer should fill with good data and the UI 118 should return to normal.5.7.6.8 The Plant Mode is manually set and reset using the Plant Mode Nuclear Application.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 29 of 84 Power Operations

5.8 Guidelines

for Maintaining Reactor Power at 100%NOTE Step 5.8.2 is intended to prevent temporary power excursions above the applicable power limit.Examples of activities which may warrant reducing Reactor Thermal Power are: Placing a CVCS Demineralizer in service with newly installed resins that are NOT boron saturated.

0 Placing a CVCS Demineralizer in service that was last utilized with a Boron Concentration LESS than RCS boron concentration by more than 25 ppm (Dilution of RCS).O Initially placing BTRS in service at the end of a fuel cycle.0 Changing Feedwater Heater alignment.

0 Starting OR Stopping a LP HTR DRIP PUMPs, MSR DRIP PUMPs, S/U SGFP 14(24), S.G.F.P. Turbine or Feedwater Booster Pump.0 Maintenance activities with high potential to result in a positive reactivity addition.(Isolation of extraction steam to a high pressure feedwater heater)0 Routine activities such as placing a CVCS Cation bed in service would NOT typically warrant a reduction in Reactor Themaal Power.0 Placing Excess Letdown in service will cause Plant Computer point U1 118 to become inaccurate in the NON-Conservative direction due to bypass flow causing approximately 2 MWth to bypass instrumentation.

This bypass flow and heat is NOT accounted for in the calorimetric.

The behavior of the reactor causing a brief power excursion above full power limits (i.e. NI splits, RCS flow anomalies, delta I swings) is not considered intentional because it is understood to be a transient condition that will recover to a power level less than the full power limit. These power swings should be closely monitored for the expected behavior.5.8.1 No actions are allowed that would intentionally raise power above the licensed power limit of 3853 MWth for any period of time.This procedure section, when completed, SHALL be retained.

5.8.2 IF an evolution is infrequently performed that could affect Reactor Power, THEN PERFORM the following:

• A Pre-Job brief.0 Review OPGP03-ZO-0042, Reactivity Management Program.REDUCE Reactor Thermal Power by an appropriate amount prior to the performance of the evolution.

5.8.3 IF an evolution has the potential to raise reactor power above 3853 MWth, THEN actions should be taken to reduce power prior to performing the evolution.

Refer to Conduct of Operations Chapter 10 "On-Line Maintenance Expectations".

NOTE The following list of Reactor Thermal Power Limits represents the power level and time where tile overpower condition MAY be reportable.

IF any of the limits set forth in Step 5.8.4 are exceeded, THEN the occurrence SHALL be documented per 0PGP03-ZX-0002, Condition Reporting Process, and evaluated for reportability.

5.8.4 The Rated Thermal Power (RTP) (3853 MWth) Limit is exceeded under any of the following conditions:

Intentionally raising power above Rated Thermal Power (RTP) (3853 MWth) Limit Failure to ensure that Thermal Power is less than or equal to the Rated Thermal Power (RTP) (3853 MWth) Limit when the two hour average exceeds Rated Thermal Power (RTP) (3853 MWth) Limit. {refer to ICS computer point Ul 1 18R for an indication of the two hour average of reactor power} (Ref. 2.26)Permitting the core thermal power average for a shift (12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period) to exceed Rated Thermal Power (RTP) (3853 MWth) Limit {refer to ICS computer point U I 118S for an indication of the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> average of reactor power, CR 05-7363, CR 05-6963} (Ref. 2.26)Failure to take prudent action prior to a pre-planned evolution that could cause power to exceed Rated Thermal Power (RTP) (3853 MWth)Limit.This procedure section, when completed, SHALL be retained.

5.8.5 Normal

steady state reactor power will be targeted for 3852 MWth with the two hour power average being controlled less than or equal to 3853 MWth. {refer to ICS computer point U1 118R for an indication of the two hour average of reactor power} (Ref. 2.26)5.8.5.1 Controls used to maintain normal steady state reactor power could include any and all of the following:

• RCS temperature adjustments" Rod repositioning

• Turbine output adjustment 5.8.6 IF reactor power unexpectedly exceeds the applicable power limit due to a plant transient, THEN operator actions SHALL be taken (e.g., turbine load reduction and/or control rod insertion) to reduce reactor power to less than the applicable power limit. This action may be supplemented by boron addition, as necessary.(Reference 2.7.7)5.8.7 IF any of the limits set forth in Step 5.8.4 are exceeded., THEN the occurrence SHALL be documented per OPGP03-ZX-0002, Condition Reporting Process, and evaluated for reportability.

5.9 Miscellaneous

Information and Responsibilities 5.9.1 WHEN performing plant evolutions, THEN MONITOR available instrumentation to ensure the expected response is obtained.5.9.2 Various secondary system automatic temperature control valves may NOT be functioning properly at all times. This requires close monitoring and manual adjustments, as necessary.

5.9.3 WHEN the Start-Up Steam Generator Feedwater Pump is operating, THEN the correct combination of SGFP Turbines and Feedwater Booster Pumps SHALL be ensured to be operating, in order to maintain adequate feed pump suction pressure.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page32of 84 Power Operations

5.9.4 COORDINATE

with the Condensate Polisher Watch and Radwaste Operator to ensure the following:

Proper number of condensate demineralizers in service.Adequate volume exists in the Recycle Hold-Up Tanks for boration/dilution evolutions required for plant operations.

5.9.5 During

normal steady state operation, only one (1) Centrifugal Charging Pump (CCP) should be operating.

Two (2) CCPs may be operated when required to maintain pressurizer water level, during the switching of one (1) CCP to the other, or for surveillance testing. (Reference 2.9)5.9.6 WHEN problems occur with the Main Feedwater Regulating Valve(s) requiring manual operation, THEN the low power feedwater regulating valve(s) may be placed in service to allow for automatic steam generator level control until the Main Feedwater Regulating Valve(s) are repaired.5.9.7 WHEN manual control of an automatic system and/or component (e.g., manual/auto controller, bypass valve or controls, etc.) is required, THEN the Reactor Operator may do so provided all applicable Technical Specifications are met or the required Action Statements are entered.5.9.8 Various controls and control switches may be operated at the discretion of the operating crew. Examples of these controls and switches are those used for monitoring different phases of electrical buses, changing controlling instrument channels for troubleshooting/testing, and chilled water valve control switches to maintain a more positive control over area temperatures.

5.9.9 IF any procedure conflicts OR situations with NO procedure guidance occur, THEN the Shift Manager should be notified to determine a plan of action.5.9.10 The MSR Reheat Controller should be operated in the "AUTO" mode of operation, whenever possible, to minimize the effects of hydraulic transients on the Moisture Separator Reheater and associated piping This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 33 of 84 Power Operations Initials 6.0 Power Reduction___CAUTION Do NOT place the Main Turbine in IMP PRESS FEEDBACK "IN" WHEN Reactor Power is greater than 98%.6.1 RECORD the unit, date and time this procedure was entered.Unit: Date: Time: Hrs 6.2 REVIEW OPGP03-ZO-0042, Reactivity Management Program.NOTE The following Step is a CONTINUOUS ACTION Step, the applicable Forms may be implemented as many times as required.6.3 REVIEW and IMPLEMENT the following Forms as applicable in the operation of the Main Turbine/Generator and Reactor:* OPOP03-ZG-0008, Form 1, Turbine Load Changes using the Limiter OPOP03-ZG-0008, Form 2, Turbine Load Changes using the Setpoint Controls OPOP03-ZG-0008, Form 3, (Automatic/Normal Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment OPOP03-ZG-0008, Form 4, (Manual/Test Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments OPOPO3-ZG-0008, Form 5, Routine at Power Minor Rod Movements* OPOP02-CV-0001, Form 1. Modes 1-2 RCS Boration Checklist* OPOP02-CV-0001, Form 2, Modes 1-2 RCS Dilution Checklist* OPOP02-CV-0001, Form 3, Modes 1-2 RCS Alt Dilution Checklist* OPOP02-CV-0001, Form 4, Modes 1-2 Automatic Operation Checklist This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 34 of 84 Power Operations Initials NOTE* Pressurizer boron concentration should be maintained within 50 ppm of RCS concentration.

• Equalize the boron concentration, as necessary, by energizing at least two (2) sets of Pressurizer Heaters to force additional spray.* Efforts to maintain Pressurizer boron concentration within 50 ppm of RCS concentration are secondary to safe Pressurizer Pressure control.6.4 IF directed by the Shift Manager/Unit Supervisor and Plant conditions can support it, THEN the Backup Heaters may be energized as necessary to aid in PZR turnover flow.6.5 MAINTAIN Main Generator cold gas temperature greater than or equal to 90'F during the Main Generator load reduction.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 35 of 84 Power Operations Initials CAUTION" The Main Turbine SHALL NOT be operated with Main Turbine Exhaust Pressure (Condenser Pressure "inHG") in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines.

IF the Main Turbine is operating in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines, THEN the following SHALL be performed within 5 minutes: (Reference 2.22)Main Turbine Exhaust Pressure (Condenser Pressure "in HG") SHALL be returned to a value outside the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines OR Main Turbine load (MWe) SHALL be lowered to a value outside the"Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines" IF operation outside the "Restrictive Zone" of Addendum 2, Main Turbine Operating Guidelines can NOT be established within 5 minutes, THEN the Main Turbine SHALL be removed from service (placed on the jacking gear) in a safe and expeditious manner.(1-5% per minute) and remain out of service until an Engineering Evaluation can be performed.

NOTE Main Turbine Exhaust Pressure should be at the lowest value attainable (i.e., Best Vacuum)during Turbine Startup and Low Load Operation to prevent the overheating of the LP turbine blading and excessive thermal expansion. (Reference 2.22)6.6 ENSURE the Main Turbine Exhaust Pressure is within the limits of Addendum 2, Main Turbine Operating Guidelines.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 36 of 84 Power Operations Initials NOTE* Control Rods may also be used for assisting in control of Tavg provided AFD is maintained within the limits of Technical Specifications.

FWH 15/16(25/26) strings have the potential to isolate during plant shutdown. (Refer to CREE 11-5655-1)

6.7 COMMENCE

RCS boration to establish the desired Tavg ramp rate to the desired Reactor Power level.6.8 COMMENCE Main Turbine load reduction at the desired ramp rate to the desired Reactor Power level.6.9 ADJUST Main Turbine load reduction or RCS boron concentration, as necessary, to maintain Tavg within 1.5°F of Program RCS Tavg (Refer to Addendum 8, Percent Power vs Program RCS Tavg).6.10 WHEN Reactor Power less than or equal to 98%, THEN PERFORM the following:

6.10.1 ARM the Modulate Signal for the Main Steam to DA valves by performing the following: (A single handswitch controls both valves)PLACE 1(2)-MS-PV-7174 and 1(2)-MS-PV-7174A handswitch to the "MOD" position and return to "AUTO".CAUTION PLACING the Main Turbine in the IMP PRESS FEEDBACK "IN" mode, may result in a momentary load swing. Refer to OPOPO 1 -TM-000 1, Main Turbine/Generator Operations Guidelines.

6.10.2 IF directed by the Shift Manager/Unit Supervisor, THEN PLACE Main Turbine in the IMP-IN mode by depressing the IMP PRESS FEEDBACK "IN" push-button, OTHERWISE N/A. {CP007}This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 37 of 84 Power Operations I Initials NOTE The valve position lights on 1(2)-MOV-0108, "FEEDWATER HEATER 1 A/I IB(21A/21B)

BYPASS ISOLATION MOV" are operated off an internal limit switches designed to change indication at 95% from full open or 5% from full closed.* To fully position 1 (2)-MOV-0108, "FEEDWATER HEATER 1 A/I IB(21A/21B)

BYPASS ISOLATION MOV" continue to hold/jog the valve handswitch until the torque switch electrically shuts the valve motor off. (If valve is operating properly, this jog should be for less than 30 sec.)* Full closure of 1(2)-MOV-0108, "FEEDWATER HEATER 1 A/ 11B(21A/21B)

BYPASS ISOLATION MOV" may require local operator check of valve position.6.11 {Mark N/A if feedwater heater 11 (21 )A/B removed for maintenance}

WHEN Reactor Power goes below 95%, THEN ENSURE CLOSED 1(2)-MOV-0108,"FEEDWATER HEATER 1 A/I IB(21A/21B)

BYPASS ISOLATION MOV".6.12 IF it is desired to reduce Reactor Power load to less than 65% Rx Pwr (; 901 MWe), THEN PERFORM the appropriate sections of OPOP03-ZG-0006, Plant Shutdown from 100% to Hot Standby.CAUTION PLACING the Main Turbine in the IMP PRESS FEEDBACK "OUT" mode, may result in a momentary load swing. Refer to OPOPO I-TM-0001, Main Turbine/Generator Operations Guidelines.

6.13 IF directed by the Shift Manager/Unit Supervisor, THEN PLACE Main Turbine in the IMP-OUT mode by depressing the IMP PRESS FEEDBACK "OUT" push-button, OTHERWISE N/A. {CP007 }6.14 STABILIZE Tavg within 1.5°F of Program RCS Tavg (Refer to Addendum 8, Percent Power vs Program RCS Tavg).This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 38 of 84 Power Operations I Initials 7.0 Power Escalation NOTE The principles of OPGP03-ZO-0042, Reactivity Management Program are in effect at all times during Operations in this procedure.

7.1 IF Reactor Power is less than z 65% Rx Pwr (z 901 MWe), THEN GO TO OPOP03-ZG-0005, Plant Startup to 100%, to restore reactor power to 100%.7.2 RECORD the Unit, date and time this procedure was entered.Unit: Date: Time: Hrs CAUTION PLACING Main Turbine in the IMP PRESS FEEDBACK "IN" mode, may result in a momentary load swing. Refer to 0POPO1-TM-0001, Main Turbine/Generator Operations Guidelines.

7.3 IF directed by the Shift Manager/Unit Supervisor, THEN PLACE Main Turbine in the IMP-IN mode by depressing the IMP PRESS FEEDBACK "IN" push-button, OTHERWISE N/A. {CP007}7.4 MAINTAIN Main Generator cold gas temperature greater than or equal to 90 0 F, WHEN raising Main Generator load.This procedure section, when completed., SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 39 of 84 Power Operations

_Initials CAUTION" The Main Turbine SHALL NOT be operated with Main Turbine Exhaust Pressure (Condenser Pressure "inHG") in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines.

IF the Main Turbine is operating in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines, THEN the following SHALL be performed within 5 minutes: (Reference 2.22)Main Turbine Exhaust Pressure (Condenser Pressure "inHG") SHALL be returned to a value outside the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines OR Main Turbine load (MWe) SHALL be lowered to a value outside the"Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines

• IF operation outside the "Restrictive Zone" of Addendum 2, Main Turbine Operating Guidelines can NOT be established within 5 minutes, THEN the Main Turbine SHALL be removed from service (placed on the jacking gear) in a safe and expeditious manner.(1-5% per minute) and remain out of service until an Engineering Evaluation can be performed.

NOTE Main Turbine Exhaust Pressure should be at the lowest value attainable (i.e., Best Vacuum)during Turbine Startup and Low Load Operation to prevent the overheating of the LP turbine blading and excessive thermal expansion. (Reference 2.22)7.5 ENSURE the Main Turbine Exhaust Pressure is within the limits of Addendum 2, Main Turbine Operating Guidelines.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 40 of 84 Power Operations Initials NOTE Control rods may also be used to assist in control of Tavg provided AFD is maintained within the Limits of Technical Specifications.

7.6 DILUTE

the RCS to establish the desired Tavg ramp rate NOTE Addendum 3, Fuel Conditioning Requirements/Recommendations, SHALL be referenced for Requirements on Loading Rates.7.7 COMMENCE raising Reactor Power and Main Turbine load to z 90% Rx Pwr (z 1248 MWe).7.8 ADJUST the Main Turbine loading rate or RCS dilution rate to maintain Tavg within 1.5°F of Program RCS Tavg (Refer to Addendum 8, Percent Power vs Program RCS Tavg).7.9 WHEN reactor power is 90%, THEN PERFORM the following:

7.9.1 STABILIZE

Reactor Power.7.9.2 VERIFY loop Delta-T, power range indications and U 1118 (if available) are consistent.

7.9.3 IF loop Delta-Ts, power range indications and U 11 8 (if available) are NOT consistent, THEN PERFORM OPSP03-NI-0001, Power Range NI Channel Calibration.

7.9.4 Maintain

feedwater temperature within limits of Addendum 5"Nominal Feed Temperature Limit vs Power Level".This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 41 of 84 Power Operations Initials CAUTION" (UNIT 1 ONLY) IF the Auxiliaty Transformer is NOT in service, THEN the Turbine Generator output SHALL be limited in accordance with Steps 4.24 and 4.25. This limit is due to a lower transformer rating on MST IA." Generator MVAR output should be maintained as close to a balanced MVAR load between Units I and Unit 2 as possible, while maintaining MVAR load IAW the Main Generator Capability Curve (in the Plant Curve Book), grid stability and minimizing vibration on the Turbine Generator bearings.Minimize operation in the VARs IN, Leading (Underexcited) region to limit eddy-current heating of the turbine end stator core. Operation in this region, or below the drawn-in curve (Main Generator Capability Curve in the Plant Curve Book) may cause a rise in heating and subsequent damage proportional to the depth and duration of operation below the curve.(CREE 07-4972-2) 7.10 RAISE reactor power to 98%, while closely monitoring the following critical parameters associated with OTDT and OPDT:* Pressurizer Pressure* Loop Average Temperature

• NI Power* Loop Delta-Ts* OTDT Setpoints* OPDT Setpoints This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 I Rev. 56 Page 42 of 84 Power Operations Initials 7.11 WHEN Reactor Power is 98% AND stable, THEN COMPLETE the following:

7.11.1 PERFORM Addendum 7, Control-Loop Alignment.

7.11.2 ENSURE the operating Steam Generator Feedpump Recirc Valves are in AUTO.I CAUTION PLACING the Main Turbine in the IMP PRESS FEEDBACK "OUT" mode, may result in a momentary load swing. Refer to OPOPO I -TM-0001, Main Turbine/Generator Operations Guidelines.

7.11.3 PLACE the Main Turbine impulse pressure feedback to IMP-OUT by, DEPRESSING the IMP PRESS FEEDBACK "OUT" pushbutton.

{CP007}7.11.4 PERFORM a calorimetric per OPSP03-NI-0001, Power Range NI Channel Calibration.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 43 of 84 Power Operations Initials 7.11.5 RECORD the following data: LOOP LOOP LOOP LOOP 1A(2A) 1B(2B) 1C(2C) 1D(2D)LOOP AVERAGE TEMPERATURE LOOP DELTA-T OTDT SETPOINT OPDT SETPOINT PI-0455 PI-0456 PI-0457 PI-0458 PRESSURIZER PRESSURE NI-0041 NI-0042 NI-0043 NI-0044 NI POWER 7.11.6 VERIFY the margin to trip between all four loop Delta-Ts and their respective OTDT setpoints greater than or equal to 7%.* LOOP A* LOOP B* LOOP C* LOOP D This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 44 of 84 Power Operations Initials 7.11.7 VERIFY the margin to trip between all four loop Delta-Ts and their respective OPDT setpoints greater than or equal to 7%.* LOOP A* LOOP B* LOOP C* LOOP D NOTE The approach to 100% Rated Thermal Power may continue provided the margin to trip between all four loop Delta-Ts and their respective OTDT setpoints remains at greater than or equal to 5%.See Plant Curve Book Table 5.2 for Rated Thermal Power (RTP)7.12 Normal steady state reactor power will be targeted for 3852 MWth with the two hour power average being controlled less than or equal to 3853 MWth. {refer to ICS computer point UI I 18R for an indication of the two hour average of reactor power} (Ref. 2.26)7.12.1 Controls used to maintain normal steady state reactor power could include any and all of the following:

• RCS temperature adjustments

• Rod repositioning" Turbine output adjustment 7.12.2 IF Reactor power unexpectedly exceeds the applicable power limit due to a plant transient, THEN operator actions SHALL be taken (e.g., turbine load reduction and/or control rod insertion) to reduce Reactor power to less than the applicable power limit. This action may be supplemented by boron addition, as necessary.(Reference 2.13)This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 45 of 84 Power Operations Initials 7.13 Prior to raising reactor power above 99.6% RTP VERIFY all loops of FW UTF are in service and available per Section 5.7.7.14 RAISE reactor power until one of the following limits are reached while monitoring critical parameters associated with OTDT and OPDT: 99.6% RTP if any portion of FW UTF is out of service or unavailable

• 100% Rated Thermal Power* Controlling Governor Valve [normally GV #4] 60% Open* Less than 5% margin to trip on associated OTDT setpoints* Less than 5% margin to trip on associated OPDT setpoints 7.15 IF 100% Rated Thermal Power is NOT reached in Step 7.14, THEN PERFORM one (1) of the following, OTHERWISE NA: HOLD Reactor power at current level GO TO Section 8.0 Alternate Power Increase to 100% Power IF ONLY limited by Controlling Governor Valve [normally GV #4], THEN RAISE reactor power until Controlling Governor Valve 100%Open.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 46 of 84 Power Operations Initials NOTE Steady state power control is normally in IMP-OUT, with the limiter at a position at or slightly above the current operating demand and relying on the use of the "SETPOINT CONTROL" and Reference counters (Turbine Actual and Setpoint), with the ramp rate set to 0.25% MW/MIN to make the Main Turbine adjustments required to maintain steady state Reactor Power. (Ref. 2.24)7.16 PLACE the governor valves valve position limiter at the current operating demand or slightly above current operating demand of the governor valves as directed by the Shift Manager/Unit Supervisor to minimize the potential for unexpected power excursions.

{CP007}7.16.1 MOMENTARY DEPRESS the governor valve limiter (lower)pushbutton.

7.16.2 REPEAT Step 7.16.1 as required until valve position limiter position is at the current operating demand or slightly above current operating demand of the governor valves.7.16.3 USE the Main Turbine "SETPOINT CONTROL" and Reference counters (Turbine Actual and Setpoint) to make fine adjustments as needed to Maintain 100% Reactor Power as describe in Step 5.8.5.7.17 ENSURE the MSR Reheat Control Valves in the "AUTO" control mode.7.18 ENSURE the LP Turbine Inlet Steam Inlet Temperatures (MSR outlet temperature) can be maintained with the MSR Reheat Control Valves in"AUTO".7.19 Maintain feedwater temperature within limits of Addendum 5 "Nominal Feed Temperature Limit vs Power Level".This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 47 of 84 Power Operations Initials NOTE* Changes to bypass valve position will change core Delta-T.* Monitor the difference between Delta-T indications and actual reactor power after bypass valve adjustments have been made. IF steady state Delta-T indications (1 minute average) are indicating greater than 2% different than actual reactor power, THEN request engineering perform 0PEP02-ZX-0007, 100% Power Instrument Alignments. (USQE 97-0030)7.20 IF required, after FW HTR bypass valve adjustments have been made, THEN REQUEST Engineering perform OPEP02-ZX-0007, 100% Power Instrument Alignments, OTHERWISE NA. (USQE 97-0030)This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 48 of 84 Power Operations Initials 8.0 Alternate Power Increase to 100% Power NOTE This Section provides a method to raise Reactor Thermal Power to 100% while lowering Secondary Thermal efficiency to allow the Main Turbine Controlling Governor Valve a small amount of throttling bite.The principles of 0PGP03-ZO-0042, Reactivity Management Program are in effect at all times during Operations in this procedure.

IF Opening the 1(2)-MOV-0108 is required, THEN request an EVALUATION by engineering (vendor AMAG) PRIOR to bypassing the high pressure FWHs and leaving UTF in service (CR 04-7089).IF this Section "Alternate Power Increase to 100% Power" is performed for greater than 7 days per cycle, THEN write a Condition Report requesting an Engineering Evaluation of the affected components and lines for excessive wear and steam erosion.For each SGFP on the HP Governor (main steam) versus the LP Governor (extraction steam) Secondary Thenrnal efficiency will be reduced to allow the plant to raise reactor power by 0.5% with governor valves wide open.8.1 SELECT the desired method(s) to reduced Secondary Thermal efficiency: (N/A method(s)

NOT used)* (Preferred)

Reduce the Secondary Thennal Efficiency using the HP Governor for the Main Feed Pumps.* (Alternate

1) Reduce the Secondary Thennal Efficiency using the HP Main Turbine and MST Crossunder Drains.* (Alternate

2) Reduce the Secondary Thermal Efficiency using the SG'BLWDN FLASH TANK VENT" to the CNDSR. {CPO18}* (Alternate

3) Reduce the Secondary Thermal Efficiency using"1(2)-FW-MOV-0108, FEEDWATER HEATER 11 A/I lB(2 IA/2 I B)BYPASS ISOLATION MOV".This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 49 of 84 Power Operations

:___I Initials CAUTION Reducing the Secondary Thermal Efficiency using Main Feed Pumps Governors may cause a rise in Reactor Power.* Plant Computer point U 1118 SHALL be OPERABLE (Ref 2.16)* ENSURE Reactor Power is maintained within limits.* IF Reactor power unexpectedly exceeds the applicable power limit due to a plant transient, THEN operator actions SHALL be taken (e.g., turbine load reduction and/or control rod insertion) to reduce Reactor power to less than the applicable power limit.This action may be supplemented by boron addition, as necessary. (Reference 2.13)8.2 IF reducing the Secondary Thermal Efficiency using Main Feed Pumps Governors, THEN PERFORM the following, OTHERWISE N/A: NOTE WHEN reducing the Secondary Thermal Efficiency using the HP Governor for the Main Feed Pumps, the SGFPT's should be placed on HP steam sequentially, in the order of the least recently to the most recently rebuilt HPGV. (Contact the System Engineer for recommendation).

One (1) Main Feed Pump(s) on the High Pressure Governor Valve is expected to be adequate for reducing the Secondary Thermal Efficiency to allow for Power to rise to 100% Power.IF Reactor power unexpectedly exceeds the applicable power limit due to a plant transient, THEN operator actions SHALL be taken (e.g., turbine load reduction and/or control rod insertion) to reduce Reactor power to less than the applicable power limit.This action may be supplemented by boron addition, as necessary. (Reference 2.13)8.2.1 PLACE the desired number of Main Feed Pump(s) on the High Pressure Governor Valve JAW 0POP02-FW-0002., S.G.F.P. Turbine.8.2.2 ENSURE a Condition Report(s) is written requesting an inspection/overhaul of the effected HP Governor Valves during the next outage.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 50 of 84 Power Operations Initials CAUTION Reducing the Secondary Thermal Efficiency using the HP Main Turbine and MST Crossunder Drains may cause a rise in Reactor Power.* Plant Computer point U 1118 SHALL be OPERABLE (Ref 2.16)* ENSURE Reactor Power is maintained within limits.* IF Reactor power unexpectedly exceeds the applicable power limit due to a plant transient, THEN operator actions SHALL be taken (e.g., turbine load reduction and/or control rod insertion) to reduce Reactor power to less than the applicable power limit.This action may be supplemented by boron addition, as necessary. (Reference 2.13)8.3 IF reducing the Secondary Thermal Efficiency using the HP Main Turbine and MSR Crossunder Drains, THEN PERFORM the following, OTHERWISE N/A: 8.3.1 PLACE the HP Main Turbine and MSR Crossunder Drain Handswitch HS6173, "TURB STM LN DRN VLV" in the OPEN position {CP007}.CAUTION Reducing the Secondary Thermal Efficiency using the SG "BLWDN FLASH TANK VENT" to the CNDSR may cause a rise in Reactor Power.* Plant Computer point UI 118 SHALL be OPERABLE (Ref 2.16)* ENSURE Reactor Power is maintained within limits.* IF Reactor power unexpectedly exceeds the applicable power limit due to a plant transient, THEN operator actions SHALL be taken (e.g., turbine load reduction and/or control rod insertion) to reduce Reactor power to less than the applicable power limit.This action may be supplemented by boron addition, as necessary. (Reference 2.13)8.4 IF reducing the Secondary Thermal Efficiency using the SG "DBLWDN FLASH TANK VENT" to the CNDSR, THEN PERFORM the following, OTHERWISE N/A: 8.4.1 PLACE SG "BLWDN FLASH TANK VENT" switch to the CNDSR position.

{CP018}This procedure section, when completed, SHALL be retained.

I OPOP03-ZG-0008 Rev. 56 Page 51 of 84 Power Operations Initials CAUTION Opening of"I (2)-MOV-0 108, FEEDWATER HEATER 1 A/ 11B(21A/21B)

BYPASS ISOLATION MOV" will cause a rise in Reactor Power.* Plant Computer point Ul 169 SHALL be OPERABLE and monitored to ensure reactor power limits are maintained within limits.IF 1(2)-FW-MOV-0108 is opened locally, THEN continuous communications SHALL be established between the Reactor Operator in the Control Room and the Plant Operator manipulating 1(2)-FW-MOV-0108.

Allow the plant to stabilize for approximately 5 minutes between each manipulation of 1(2)-FW-MOV-0108.

• A prejob brief SHALL be performed prior to the start of this evolution.

• Maintain feedwater temperature within limits of Addendum 5 "Nominal Feed Temperature Limit vs Power Level".IF Opening the 1(2)-MOV-0108 is required, THEN request an EVALUATION by engineering (vendor AMAG) PRIOR to bypassing the high pressure FWHs AND leaving UTF in service (CR 04-7089).

Plant Computer point U 1118 should be declared inoperable unless OPERABILITY is supported by an Engineering Evaluation.

NOTE* The valve stroke time for 1(2)-FW-MOV-0 108 is approximately I minute and is a "JOG" open valve.* EITHER use Plant Computer point U0490 "SG FW AVG TEMP", OR calculate the average of Plant Computer points T7200, T7203, T7206, and T7209 to determine average Feedwater temperature.

• 1(2)-FW-MOV-0108 Opening increments should be in movements of LESS THAN 10%Stroke. (Typically 5 sec jog times have functioned well, with 4-5 jogs required)* Opening the "l(2)-MOV-0108, FEEDWATER HEATER I A/I lB (21A/21B)

BYPASS ISOLATION MOV" may have adverse consequences on the Feedwater Ultrasonic Flowmeter (FW UTF) System. A rise in streaming in the FW Header related to bypassing cool water around the HP FWH may cause the standard deviation of flow measured by FW UTF to exceed acceptance criteria.

IF standard deviation criteria are NOT met, THEN FW UTF will roll out on the affected Loop.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 52 of 84 Power Operations Initials 8.5 IF reducing the Secondary Thermal Efficiency using "1(2)-FW-MOV-0108, FEEDWATER HEATER 1 A/ 11B(21A/21B)

BYPASS ISOLATION MOV", THEN PERFORM the following, OTHERWISE N/A: 8.5.1 REQUEST an EVALUATION by engineering (vendor AMAG) prior to bypassing the high pressure FWHS and leaving UTF in service (CR 04-7089).8.5.2 SLOWLY THROTTLE OPEN "1 (2)-FW-MOV-0 108, FEEDWATER HEATER 1 A/ 11B(21A/211B)

BYPASS ISOLATION MOV" using jog or local manual.8.5.3 MONITOR Delta-T between feedwater lines to each SG. Plant Computer point U0490C SG FW TEMP MAX-MIN.8.5.4 IF Delta-T between feedwater lines to each SG is greater than 0.9°F, THEN PERFORM the following: (Reference 2.20)8.5.4.1 STOP the power rise.8.5.4.2 CONTACT Reactor Engineering.

8.6 ALLOW

approximately 5 minutes for plant stabilization.

8.7 Prior

to raising reactor power above 99.6% RTP VERIFY all loops of FW UTF are in service and available per Section 5.7.8.8 RAISE reactor power until one of the following limits are reached while monitoring critical parameters associated with OTDT and OPDT:* 99.6% RTP if any portion of FW UTF is out of service or unavailable

• 100% Rated Thermal Power* Controlling Governor Valve [normally GV #4 ] 100% Open* Less than 5% margin to trip on associated OTDT setpoints* Less than 5% margin to trip on associated OPDT setpoints 8.9 REPEAT Steps 8.2 through 8.8 as necessary to achieve 100% Rated Thermal Power.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 53 of 84 Power Operations Initials 8.10 IF steam pressure rises, THEN PERFORM the following based upon the method selected in Step 8.1, MARK Steps for methods NOT used N/A: CAUTION IF Reactor power unexpectedly exceeds the applicable power limit due to a plant transient, THEN operator actions SHALL be taken (e.g., turbine load reduction and/or control rod insertion) to reduce Reactor power to less than the applicable power limit. This action may be supplemented by boron addition, as necessary. (Reference 2.13)8.10.1 PERFORM a prejob brief prior to the start of this evolution.

8.10.2 ENSURE the desired Main Feed Pump(s) on the Low Pressure Governor Valve JAW 0POP02-FW-0002, S.G.F.P. Turbine.8.10.3 ENSURE the HP Main Turbine and MSR Crossunder Drain Handswitch HS6173, "TURB STM LN DRN VLV" is in the CLOSE position {CP007}: 8.10.4 ENSURE SG "BLWDN FLASH TANK VENT" switch to the FW HTR position.

{CP018}8.10.5 ENSURE CLOSED "1 (2)-FW-MOV-0108, FEEDWATER HEATER I IA/i IB(21A/21 B) BYPASS ISOLATION MOV".NOTE iThe valve position lights on 1(2)-MOV-0108, "FEEDWATER HEATER 1 A/I 1B(21A/21B)

BYPASS ISOLATION MOV" are operated off an internal limit switches designed to change indication at 95% from full open or 5% from full closed." To fully position l(2)-MOV-0108, "FEEDWATER HEATER 1 A/ 11B(21A/21B)

BYPASS ISOLATION MOV" continue to hold/jog the valve handswitch until the torque switch electrically shuts the valve motor off. (If valve is operating properly, this jog should be for less than 30 sec.)" Full closure of 1(2)-MOV-0 108, "FEEDWATER HEATER I IA/i IB(21A/21 B) BYPASS ISOLATION MOV" may require local operator check of valve position.8.10.6 IF reducing the Secondary Thermal Efficiency using"1(2)-FW-MOV-0108, FEEDWATER HEATER I lA/I IB(21A/21B)

BYPASS ISOLATION MOV", WHEN "1 (2)-FW-MOV-0108, FEEDWATER HEATER IlA/i IB(21A/21B)

BYPASS ISOLATION MOV" is fully closed, THEN EXIT this section of the procedure.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 54of 84 Power Operations

=Initials 8.11 IF Section "Alternate Power Increase to 100% Power" is performed for greater than 7 days per cycle, THEN write a Condition Report requesting an Engineering Evaluation of the affected components and lines for excessive wear and steam erosion, OTHERWISE N/A.8.12 WHEN 100% Rated Thermal Power is reached, THEN GO TO Step 7.16.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 55 of 84 Power Operations

9.0 Coastdown

Operations NOTE The principles of OPGP03-ZO-0042, Reactivity Management Program are in effect at all times during Operations in this procedure.

9.1 Coastdown

Operations General Information and Guidelines

9.1.1 Coastdown

Operations are required when burnup reaches a point where full power can no longer be maintained, i.e., power operation beyond the end of full power capability.

To compensate for fuel depletion, power (load) reductions are required.9.1.2 Coastdown Operations are NOT allowed unless a Coastdown Safety Evaluation has been performed for the current fuel cycle.9.1.3 Coastdown Operations should commence when RCS boron concentration reaches approximately 5 ppm or, as deemed necessary by the Plant Operations Manager.9.1.4 During Coastdown Operations, all evolutions that change steam demand and/or feedwater flow should be performed at a slow controlled rate.9.1.5 RCS Tavg and Steam Generator Feedwater temperature changes should be slow and deliberate.

Avoid performance of, or exercise caution when performing any evolution that could cause RCS Tavg and Steam Generator Feedwater temperatures to change such as: 9.1.5.1 Removing or returning Feedwater Heater strings from/to service.9.1.5.2 Changing Steam Generator Feedwater Pump or Booster Pump combinations.

9.1.5.3 Any evolutions that would cause feedwater flow to oscillate or change.9.1.5.4 Placing or removing CVCS demineralizers and/or filters in service.9.1.5.5 Adding chemicals to the RCS.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 56 of 84 Power Operations 9.1.5.6 Any evolution or testing that could cause a RCS boron concentration change.9.1.5.7 Any other testing on systems or components that could cause a transient on either the primary or secondary plant such as; the Turbine-driven Auxiliary Feedwater Pump, the Main Turbine Throttle and/or Governor valves, or Control Rods.9.1.6 Raising power should only occur when recovering the unit after a reactor trip or other significant forced power reductions.

9.1.7 IF a large power reduction OR reactor trip should occur, THEN reactor power SHALL NOT be raised above the previous steady state (equilibrium xenon)power level. IF reactor power is raised above the previous steady state value (because of lower xenon concentrations), THEN the following will occur:* Xenon instability.

• The Coastdown Safety Evaluation will be violated.9.1.8 Prior to raising reactor power to greater than 50% following a reactor trip or forced power reduction, reasonable AFD stability should exist. Reactor power should NOT be raised to greater than 50% until AFD can be maintained stable inside the Target Band.9.1.9 Reactor Engineering should be contacted for any of the following:

Prior to any reactor power rise.AFD oscillations exceeding 2.0% peak to peak.Any condition which require deviations from Coastdown Operations guidelines.

9.1.10 WHEN Coastdown Operations are in progress, THEN the following parameters should be maintained:

Tavg within 1.5°F of Program RCS Tavg (Refer to Addendum 8, Percent Power vs Program RCS Tavg).AFD oscillations less than or equal to 1.0% peak to peak.Control Bank D position at 240 to 250 steps withdrawn.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 57 of 84 Power Operations 9.1.11 Significant power reductions should be avoided during Coastdown Operations.

Inability to dilute the RCS makes reactor power stabilization difficult.

9.1.12 IF significant power reductions are required, THEN Tavg may be maintained within 3'F of Program RCS Tavg (Refer to Addendum 8, Percent Power vs Program RCS Tavg) provided the Margin to trip setpoints for both OPDT and OTDT are maintained greater than or equal to 5%.9.1.13 Control rod motion should be minimized.

Movement of control rods can cause large shifts in AFD and xenon oscillations to occur. Control Bank D usage should be limited to 3 steps in one direction per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period to maintain AFD less than or equal to 1.0% peak to peak value, or a value within the Target Band as determined by Reactor Engineering with Shift Manager concurrence.

9.1.14 Allow AFD to trend slowly in the positive direction while controlling oscillations to less than or equal to 1.0% peak to peak. WHEN AFD is approximately 1.0% from the positive Technical Specification limit, THEN NOTIFY Reactor Engineering to update the target AFD per OPSP I 0-NI-000 1, Target Axial Flux Difference Determination.

Target AFD updates will normally occur weekly.9.1.15 AFD should NOT be allowed to rise to within 0.5% of the positive Technical Specification limit. Control rod insertion should be used to keep AFD at least 0.5% below the positive Technical Specification limit.9.1.16 A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> minimum period between power reductions is recommended to allow xenon to peak and start lowering before the next power reduction.

Without this minimum time period, the coast down rate may exceed the desired rate due to xenon build-in.9.1.17 WHEN Coastdown Operations begin. THEN Turbine Load will be reduced 1% Turbine Load 14 MWe) every 2 day(s) or reactor power by approximately z 38 MWth (z1% reactor power) z every 2 day(s) to maintain Tavg.9.1.18 AFD will become more positive by approximately 0.6% for each 1% reduction in power. This change in AFD assumes no rod motion.9.1.19 During Coastdown operations, Auctioneered High Tavg should be used for RCS Tavg control.9.1.20 Plant Computer points U I 18 and Q0340 should be used as indications for the periodic power reductions.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 58 of 84 Power Operations 9.1.21 Once Coastdown Operations have commenced, THEN RCS boron dilution should be terminated, except during transient recovery.

Reactor Makeup Water additions for VCT level control are acceptable.

9.1.22 During Coastdown Operations calculated Tref (program Tavg) should be utilized for Tavg vs. Tref comparisons.

9.2 Commencing

Coastdown Operations

9.2.1 Reactor

power, Tavg, Control Bank D position and AFD should be plotted hourly. This information will be used for the following:

To identify a developing xenon oscillation so that the least amount of Control Rod movement can be used to minimize the oscillation.

• Aid Operations personnel in timing power reductions.

NOTE Due to the small load changes performed and the potential for causing transients during transferring of control modes, the Main Turbine should be operated in the IMP PRESS FEEDBACK OUT mode.Maintain feedwater temperature within limits of Addendum 5 "Nominal Feed Temperature Limit vs Power Level".9.2.2 ALLOW Tavg to lower below Program RCS Tavg (Refer to Addendum 8, Percent Power vs Program RCS Tavg) by approximately 0.5°F. WHEN Tavg is approximately 0.5°F below Program RCS Tavg, THEN REDUCE turbine load by approximately 1% Turbine Load (z 14 MWe) or reactor power by approximately

38 MWth (z 1% reactor power). This reduction should be performed slowly over an entire one hour period.9.2.3 The z 1% {z 14 MWe ori 38 MWth} power reduction should raise Tavg initially by approximately 0.5°F (before xenon builds in).9.2.4 ALLOW Tavg to respond naturally to the power reduction.

9.2.5 PERFORM

subsequent power reductions when AFD is stable or lowering; this will aid in the control of AFD oscillations.

The optimum time to start the ramp is shortly after AFD has reached its positive peak and is moving in the negative direction.

A power reduction should NOT be initiated if AFD is approaching its negative peak or has already crossed the midpoint of the downward oscillation trend. The Reactor Operator should review the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> history trend for AFD prior to lowering power.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 59 of 84 Power Operations 9.2.6 As Coastdown Operations continue, the magnitude of the daily power reductions will lower. This reduction is due to the lowering in fuel burnup rate as Reactor Power is reduced. For example, the daily power reduction after 40 days in coast down operations will be 0.75% Rx Pwr (= 10.6 MWe).9.2.7 CONTINUE the load (power) reductions periodically when Tavg lowers below Program RCS Tavg (Refer to Addendum 8, Percent Power vs Program RCS Tavg) by approximately 0.5°F. The normal power reductions should occur every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.9.2.8 IF a significant oscillation occurs, THEN NOTIFY Reactor Engineering as soon as possible.

MAINTAIN AFD within the Technical Specification Target Band by moving control rods, as necessary.

Subsequent strategies will be developed by Reactor Engineering and the Plant Operations Department.

9.2.9 MONITOR

Pressurizer level to ensure it follows program level.9.2.10 MONITOR Pressurizer heaters and spray valves for proper operation.

This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 60 of 84 Power Operations 10.0 Records Review 10.1 REVIEW procedure package to ensure all applicable sections are completed as required.Shift Manager/Unit Supervisor Date 11.0 Support Documents 11.1 Addendum 1, Maximum Generator Loading Requirements With Heater Strings Isolated/Bypassed 11.2 Addendum 2, Main Turbine Operating Guidelines 11.3 Addendum 3, Fuel Conditioning Requirements/Recommendations 11.4 Addendum 4, Generator Exciter Operating Guidelines 11.5 Addendum 5, Nominal Feed Temperature Limit vs Power Level 11.6 Addendum 6, Example of Indications and Actions For a Feedwater Ultrasonic Flowmeter System Failure 11.7 Addendum 7, Control-Loop Alignment 11.8 Addendum 8, Percent Power vs Program RCS Tavg 11.9 Form 1L Turbine Load Changes using the Limiter 11.10 Form 2, Turbine Load Changes using the Setpoint Controls 11.11 Form 3, (Automatic/Normal Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment 11.12 Form 4, (Manual/Test Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments 11.13 Form 5. Routine at Power Minor Rod Movements This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 61 of 84 Power Operations Addendum 1 Maximum Generator Loading Requirements With Heater Page 1 of 2 Strings Isolated/Bypassed CAUTION" IF Tavg lowers to < 561 OF, THEN Comply with Technical Specification 3.1.1.4. IF Tavg cannot be restored to GREATER THAN 561 OF within 15 minutes, THEN manually trip the reactor and GO TO OPOP05-EO-EOOO, Reactor Trip Or Safety Injection." Maximum available Condensate Flow may be more plant limiting than Generator Loading when multiple feed heaters are removed from the flow path.* These "Maximum Generator Loading Requirements" are for equipment protection.

Power reduction to these values SHALL be performed promptly "As Soon As Possible", BUT in an orderly and safe manner. Trouble Shooting SHALL NOT take priority over equipment protection. (Ref 2.21)* IF possible., always reduce the load before removing the heater from service. (Ref 2.21)* Removal of feedwater heaters from service can distort the steam flow distribution in the turbine and change the loading, flow, temperature, and pressure drop across the stages of turbine blading. These changes could cause abnormally high blade stress. (Ref 2.21)* To remove feedwater heater(s) from service, REDUCE the electrical power to that shown on the MWe power column shown below. Follow the MWe power column shown below for any additional power reductions if additional feedwater heater(s) are to be removed from service.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 62 of 84 Power Operations Addendum I Maximum Generator Loading Requirements With Heater Page 2 of 2 Strings Isolated/Bypassed EXTRACTION STEAM ISOLATED/HEATERS MAXIMUM TURBINE/GENERATOR LOADING ISOLATED/BYPASSED Ref 2.22 MWe % of 1387 MWE -Design Rated MWe Load (NOT Based on Rx PWR)I IA(21A) or llB(21B) 1387 100 IIA(21A), I1B(21 B) 1387 100 IIA(21A).

1IB(21B), 12*(22*) 1387 100 11A(21A), 1 IB(21B), 12*(22*), 13A(23A) 1352 97.5 IlA(21A), IlB(21B), 12*(22*), 13A(23A), 13B(23B) 1318 95.0 I lA(21A), 1IB(21B), 12*(22*), 13A(23A), 1283 92.5 13B(23B), 14A(24A)1 IA(21A), 1 IB(21B), 12*(22*), 13A(23A), 1248 90.0 13B(23B), 14A(24A), 14B(24B)11A(21A), 1 IB(21B), 12"(22"), 13A(23A), 1179 85.0 13B(23B), 14A(24A), 14B(24B), {15A(25A) or 15B(25B) or 15C(25C)}I IA(21A), 1 B(21B), 12*(22*), 13A(23A), 1110 80.0 133B(23B), 14A(24A), 14B(24B), { 15A(25A) or 15B(25B) or 15C(25C)}, {16A(26A) or 16B(26B) or 16C(26C)}16A(26A) 1387 100 16A(26A), 16B(26B) 1387 100 16A(26A), 16B(26B), 15A(25A) 1248 90.0 16A(26A), 16B(26B), 15A(25A), 15B(25B) 1248 90.0 15A(25A) 1387 100 15A(25A), 15B(25B) 1387 100 16A(26A), 15A(25A) 1248 90.0 16B(26B), 15A(25A) 1387 100 14A(24A) 1387 100 14A(24A), 14B(24B) 1387 100 14A(24A), 13A(23A) 1318 95.0 14A(24A).

13B(23B) 1318 95.0 14A(24A), 13A(23A), 12*(22*) 1248 90.0 14A(24A), 13A(23A), 12*(22*), 11 A(21A) 1318 95.0 14A(24A), 13B(23B), 12*(22*) 1318 95.0 14A(24A), 13A(23A), 13B(23B), 12*(22*) 1179 85.0* Deaerator This procedure section, when completed, SHALL be retained.

I OPOP03-ZG-0008 Rev. 56 Page 63 of 84 SEiPower Operations Main Turbine Operating Guidelines Exhaust Pressure Limitations Maximum Permissible Condensing Pressure I C a)2~U)U)0)a...0..1~~1)U)C 0)C 0 0 20 21 22 23 24 25 25.26 27 28 29 30 not permissible 777a~erl Ntn~ue 2'1.00 in Hg Ninstance in this in region; with a total of ___ ______ __M300 minutes allowed I during the total working life of last orxsblade row. 5-26.00 in Hg mend: Trip Alarm permissible without limit 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 Load -MW CD14035 U1(2/2/12)

This procedure section, when completed, SHALL be retained.

Power Operations Addendum 3 Fuel Conditioning Requirements/Recommendations Page 1 of 3 NOTE* Reactor Engineering SHALL be contacted if interpretations of this Addendum are necessary.

• Loading Rates and Axial Flux Difference requirements for initial power ascension and for power ascension after extended periods at low power are cycle specific and contained in Plant Curve Book Figure 5.18, Power Ascension Guidelines.

1.0 Requirements

on Loading Rates and Axial Flux Difference A. Unconditioned Fuel 1.1 During the initial return to power following a Refueling Shutdown or following a Cold Shutdown where fuel assemblies have been handled (e.g., inspection), the loading and Axial Flux Difference requirements are in Plant Curve Book Figure 5.18, Power Ascension Guidelines.

1.1.1 This ramp rate requirement applies during the initial startup of a reload cycle for that period of time until full power is achieved for 72 cumulative hours out of any 7 day operating period at power. It may also apply for any other rise in power during that time period, depending on the maximum power level achieved and length of operation at that power level.1.1.2 Specifically, the requirements from Plant Curve Book Figure 5.18, Power Ascension Guidelines can be removed for Reactor Power levels at or below a given power level PROVIDED the plant has operated at or above that power level for at least 72 cumulative hours out of any 7 day (168 hr) operating period at power.1.1.2.1 Time at shutdown (Mode 2 below POAH and Modes 3, 4 and 5) is NOT considered as operating time.1.1.3 IF Axial Flux Difference is NOT controlled within the requirements of Plant Curve Book Figure 5.18, Power Ascension Guidelines, THEN HOLD power ascension until limits are met.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 65 of 84 Power Operations Addendum 3 Fuel Conditioning Requirements/Recommendations Page 2 of 3 1.2 After extended periods (longer than 27 days) of low power (in all cycles after the initial startup requirements in Step 1.1 are met), the loading and Axial Flux Difference requirements in Plant Curve Book Figure 5.18, Power Ascension Guidelines, SHALL apply: 1.2.1 The rate of Reactor Power rises above the highest power level sustained for at least 72 cumulative hours during the preceding 30 days of operation at power SHOULD follow the guidance in Plant Curve Book Figure 5.18, Power Ascension Guidelines.

1.2.2 This requirement only applies when power has been reduced for a period of time longer than 27 days.1.2.3 Time at shutdown (Mode 2 below POAH and Modes 3, 4 and 5) is NOT considered as operating time and is NOT counted toward the thirty days as stated in Step 1.2.1.1.2.4 IF Axial Flux Difference is NOT controlled within the requirements of Plant Curve Book Figure 5.18, Power Ascension Guidelines, THEN HOLD power ascension until limits are met.B. Conditioned Fuel 1.3 At all other times, unless otherwise approved by Reactor Engineering and the Plant Operations Manager, the rate of Reactor Power rise between 0% and 100% of full power SHOULD be less than or equal to 10% of full power in one hour, but SHALL NOT exceed an rise of 12% over any I hour.1.4 IF the rate of Reactor Power rise exceeds 15% of full power in an hour, THEN CONTACT Reactor Engineering for evaluation.

1.5 IF Reactor Power changes of greater than 15% power are performed in any one hour period, THEN Chemistry SHALL be notified to perform an isotopic analysis of the RCS to satisfy the requirements of Technical Specification 3.4.8, Table 4.4-4, Item 4.b.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 66 of 84 Power Operations Addendum 3 ) Fuel Conditioning Requirements/Recommendations Page 3 of 3 2.0 Recommendations for Control Rod Movement 2.1 During the initial return to power following a Refueling Shutdown or following a Cold Shutdown where fuel assemblies have been handled (e.g., inspection), the following applies: 2.1.1 Control rod motion during the initial return to power should be kept to a minimum, and the startup implemented with the control rods withdrawn.

2.1.2 Control

rod withdrawal rate should be limited to 3 steps per hour above 50%power WHEN raising power greater than 1% per hour. Rod withdrawal may occur concurrently with power rises.2.1.3 Rod withdrawal may exceed 3 steps per hour to maintain Axial Flux Difference control, but should NOT exceed 6 steps per hour.2.1.4 WHEN the control rods have been withdrawn to some position at a given power level, THEN during subsequent maneuvers there is no restriction on rod withdrawal to the previous position tip to that power level.212 During any incore/excore calibration maneuver during initial startup, control rods should NOT be rapidly withdrawn beyond their initial position.

However, during a calibration up to 75% of full power, the rods may be inserted as many as 40 steps and subsequently withdrawn to their initial position.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 67 of 84 Power Operations Addendum 4 Generator Exciter Operating Guidelines Page 1 of I NOTE The Generator Exciter SHALL be Operated within the limits of the Generator Operability Curve at all times.NOTE A sudden rise in field voltage NOT accompanied by a rise in reactive VAR may be an indication of fuse-diode assembly failure.CONDITION RECOMMENDED ACTION 1 fuse out of service per phase per polarity PERFORM daily fuse monitoring (4)2 fuses out of service per phase per polarity PREPARE plan to replace bad rectifier modules (1) and fuses.MONITOR fuse condition hourly.3 fuses out of service per phase per polarity REDUCE reactor power to less than "P9" as (2) fquickly as possible, then CONTINUE with a (2) controlled shutdown of the Unit.IF the Exciter condition worsens. THEN TRIP the Reactor/Generator as necessary to mitigate equipment damage.REDUCE power to less than "P9" and TRIP the Greater than 3 fuses out of service per phase per polarity turbine/generator.

(3)IF the Exciter condition worsens, THEN TRIP the Reactor/Generator as necessary to mitigate equipment damage.(1)(2)(3)(4)Two (2) fuses per phase (either: red, white, or blue) per polarity (North or South Pole), NOT a total of two (2) fuses.Three (3) fuses per phase (either: red, white, or blue) per polarity (North or South Pole), NOT a total of three (3) fuses Greater than three (3) fuses per phase (either red, white, or blue) per polarity (North or South pole)Monitoring may be allowed to return to original schedule with Engineering's approval.This procedure section, when completed, SHALL be retained.

0POP03-ZG-0008 Rev. 56 Page 68 of 84 !Power Operations Addendum 5 Nominal Feed Temperature Limit vs Power Level Page 1 of I FIGURE 1 Main Feedwater Operating Temperature at Various Power Levels[Source: NC-07088]450 440 430 420 410 400 390-Z6 380-370 F--360 350 340 330 320 310" Oe rati o .O ....... J ...... ....I ...................

........" 0-001 Operation NOTwe Aloe.° .... ...... ....... ...................

.................

.30 40 50 60 70 80 90 100 Power (percent)Feedwater Temperature is the average of at least three loops of ICS pohits T7200, T7203, T7206 and T7209, OR U0490 IF feedwater temperature cannot be maintained as required in FIGURE I (above), EXCEPT as noted below, THEN contact Engineering for evaluation (Ref 2.25).IF Rated Thermal Power cannot be achieved due to governor valve limitations, THEN this procedure Section 8.0 "Alternate Power Increase to 100% Power" may be used.EITHER use plant computer point U0490 "SG FW AVG TEMP", OR calculate the average of plant computer points T7200, T7203, T7206, and T7209 to detennine average Feedwater Outlet temperature.

A nominal variance of_+/- 2 'F from the Design Feedwater Temperature is within the normal allowed range. Variances exceeding the -2 'F from the Design Feedwater Temperature are permitted, but unless performing Section 8.0 "Alternate Power Increase to 100% Power" variances exceeding

-2 'F may indicate secondary plant performance issues.The "Operations NOT Allowed" area of FIGURE 1 SHALL remain in effect at all times and is NOT dependant upon the feed heaters configuration.

If Tavg decreases to 561 'F, Comply with Technical Specification 3.1.1.4. IF Tavg cannot be restored to GREATER THAN 561 'F within 15 minutes, THEN manually trip the reactor and GO TO OPOP05-EO-EOOO,.

Reactor Trip Or Safety Injection.

WHEN a Feedwater Heater string is isolated and/or bypassed, THEN the requirements of Addendum 1, Maximum Generator Loading Requirements With Heater Strings Isolated/Bypassed, SHALL in addition to the limits of this Addendum, be observed. (Ref. 2.17)This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 69 of 84 Power Operations Addendum 6 Example of Indications and Actions For a Feedwater Page 1 of 2 Ultrasonic Flowmeter System Failure NOTE The values in the following examples represent a failure of a single loop of FW UTF. Multiple failures will cause values to change in magnitude proportional to the number of failures.

For example, if four loops of FW UTF fail, then the indicated power in Step 1.3 would rise to approximately 3893 MWt (101%).1.0 The following is an example of expected indications and acceptable actions assuming a failure of one the Feedwater Ultrasonic Flowmeter (FW UTF) System transducers:

1.1 Assume

the last performance of OPSP03-NI-0001 was completed on Wednesday at 0200.The next Technical Specification required performance of 0PSP03-NI-0001 is Thursday at 0800 (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> plus 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of grace).1.2 Assume reactor power is stable at 3853 MWt, U 1118 is equal to 3853 MWt, NI power (Ul 169) is equal to 100%, and the Loop B Correction Factor (ICCUR0451) is equal to 0.9900.1.3 On Wednesday at 1000 UI 118 receives a "C" clarifier and rises to approximately 3863 MWt (100.25%).

All other independent power indications remain stable.1.4 Investigation of the FW UTF System Plant Computer displays and reports indicate that a Defouling event did NOT occur and Loop B Correction Factor (ICCUR0451) rolled out to a value of 1.00 and is displaying a "C" clarifier.

1.5 Based

on this information, U 1118 is declared unreliable and power is maintain at 100%using NI power indications.

1.6 Investigation

by I&C and the system engineer determine that one of the Loop B FW UTF System transducers failed and will NOT be available for 3 days.1.7 Since it is NOT necessary to use available grace and the next scheduled performance of OPSP03-NI-0001 is Thursday at 0200, the required power reduction to less than or equal to 99.6% (3838 MWt) is commenced on Thursday at 0000.1.8 With Loops A, C, and D Ultrasonic Feedwater Flow System operating normally, U 1118 with a "C" clarifier will be the best indication of power for the power reduction.

A Manual Calorimetric power would be expected to equal U 1118 power.1.9 U1 118 power is reduced from 3863 MWt (100.25%)

to 3838 MWt (99.6%) as indicated by Ul 118. During the power reduction NI power drops from 100% to 99.35%.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 70 of 84 Power Operations Addendum 6 Example of Indications and Actions For a Feedwater Page 2 of 2 Ultrasonic Flowmeter System Failure 1.10 The next scheduled performance of 0PSP03-NI-0001 is completed by Thursday at 0200. The reactor power was determined to be 3838 MWt (99.6%) and NI power was adjusted Lip from 99.35% to 99.6%. The OPSP03-NI-0001 was completed using the U 1118 Method even though U 1118 had a "C" clarifier.

1.11 Power is maintained less than or equal to 3838 MWt with a portion of the FW UTF System out of service or unavailable.

1.12 On Sunday at 1200 the Loop B FW UTF System transducer is replaced and put in service.1.13 On Sunday at 1600 the Loop B Ultrasonic Feedwater Flow System buffer has refilled with enough good quality data and the "C" clarifier is removed from both the Loop B Correction Factor (ICCUR0451) and UI 118. At the same time, the Loop B Correction Factor rolls in to a value of 0.9900 and UI 118 lowers from 3838 MWt (99.6%) to 3828 MWt (99.35%).1.14 As expected an investigation shows all other independent power indications remained stable and the Ultrasonic Feedwater Flow System Plant Computer displays and reports indicate that a Defouling event did NOT occur.1.15 On Sunday at 1700, OPSP03-NI-0001 is completed.

Reactor power is determined to be 3828 MWt (99.35%) and NI power is adjusted down from 99.6% to 99.35%.1.16 The FW UTF System is verified to be fully in service and available.

1.17 On Sunday at 1800 reactor power is raised from 3828 MWt to 3853 MWt using U 1118.During the power rise, NI power rises from 99.35)5% to 100%.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 71 of 84 Power Operations Addendum 7 Control-Loop Alignment Page 1 of 4 NOTE* Use of the "PREFERRED" control channels is an Operations good practice.* Alternate channels may be selected without Shift Manager/Unit Supervisor approval WHEN required by an approved procedure or work package.* The Shift Manager/Unit Supervisor MAY authorize use of Alternate control-loop alignments.

• Use of Alternate control-loop alignments should be recorded in Control Room Logs.* WHEN the use of Alternate control-loop alignments is NO LONGER required, THEN the"PREFERRED" control channels should be selected.* This Addendum may be performed at any time as directed by the Shift Manager/Unit Supervisor.

• Steps in this Addendum may be performed in any order.* IF performing this Addendum due to POST Surveillance Testing OR POST Maintenance Testing, THEN channels that are NOT part of these test/work activities may be marked N/A.Initials 1.0 ENSURE the following "Steam Generator" related control-loops are aligned as follows: 1.1 ENSURE the "SG LVL" control-loops are aligned to the "PREFERRED" control channels OR to Alternate control channels as directed by the Shift Manager/Unit Supervisor."SG LVL" Instrument Channel/Power Supply Preferred Alternate , Performer Initials S Initials to use Alternate 1 (DP1201) II (DP]202) .___:".:."_.'

SG A LT-571 LT-519 SG B LT-572 ",LT-529 S.. .. ., ..:i..:,": S Ii (DP1202) 1I(DPI20)SG C LT-539 ...LT-573 SG D LT-549 LT-574 This procedure section, when completed, SHALL be retained.

I OPOP03-ZG-0008 Rev. 56 Page 72 of 84 Power Operations I Addendum 7 Control-Loop Alignment Page 2 of 4 Initials 1.2 ENSURE the "FW FLOW" control-loops are aligned to the "PREFERRED" control channels OR to Alternate control channels as directed by the Shift Manager/Unit Supervisor."FW FLOW" Instrument Channel/Power Supply'..* >" P C.. .." , .i ,i ,"i,:.. .:::i'."..

.. .. Perform er .. SS/US#,s iuS ials ~to.Preferred iiiaAlternate rftias Suse Al te a Initias useAlternate

.I(DPI201)

I (DPL2..).SG A FT-510 FT--51i SG B FT -520 FT.-521_____11 (.DP1202)

I (DPI20 : ' ' " ," : : ?i::: SG C FT-531 FT.53 .SG D FT-541 FT -540 1.3 ENSURE the "STM FLOW" control-loops are aligned to the "PREFERRED" control channels OR to Alternate control channels as directed by the Shift Manager/Unit Supervisor."STM FLOW" Instrument Channel/Power Supply Preferred efoterate .rm r Y SI nitials to eeeAlternate

..... s ... ,: e Alternate.

I (DP 1201 ) 11. i (DP1202*.), .::'SG A FT -512 FT 513 SG B FT-522 FT -523I (D P 1202 ) I (D)P l20 1) .....,.:........

...: .SG C FT -533 FT SG D FT -543 542_This procedure section, when completed, SHALL be retained.

I OPOP03-ZG-0008 IRev. 56 Page 73 of 84 Power Operations Addendum 7 Control-Loop Alignment Page 3 of 4 Initials 2.0 ENSURE the following "Pressurizer" related control-loops are aligned as follows: 2.1 RECORD PRESSURIZER PRESSURE in the following table: PI-0455 PI-0456 PI-0457 PI-0458 PRESSURIZER PRESSURE 2.2 DETERMINE the "PREFERREW PRZR PRESS channel by perfonming the following:

2.2.1 SELECT

the lowest pressure channel in order to obtain the highest OTDT setpoint margin for the remaining 3 OTDT loops.2.2.2 PLACE a Check Mark (V) in the "Preferred" column for the selec-ted PRZR PRESS channel in the table of Step 2.3.2.3 ENSURE the "PRZR PRESS CONT SEL" control-loops are aligned to the"PREFERRED" control channels OR to Alternate control channels as directed by the Shift Manager/Unit Supervisor."PRZR PRESS CONT SEL" Instrument Channel/Power Supply Preferred Position II II1 IV Perfonner SS/US Initials to (,/) (DP1201) (DP1202) (DP1203) (DP1204) Initials use Alternate P457/456 N/A P456 P457 N/A P455/456 P455 P456 N/A N/A P455/458 P455 N/A N/A P458 This procedure section, when completed, SHALL be retained.

1OPOP03-ZG-0008 Rev. 56 Page 74 of 84 Ad 7Power Operations P; Addendum 7 Control-Loop Alignment Page 4 of g 4 Initials 2.4 ENSURE the "PRZR LEVEL CONT SEL" control-loops are aligned to the"PREFERRED" control channels OR to Alternate control channels as directed by the Shift Manager/Unit Supervisor."PRZR LEVEL CONT SEL" Instrument Channel/Power Supply"I II 111 IV Performer SS/US Initials to Position .:,(DP1201) (DP1202) (DP1203) (DPI204.)

Initials use Alternate Preferred L467/466 N/A L466 L467 N/Aa *T , L4 5 ,5/4 6 6 _L 4 6 5 ý .L 4 6 6 N /A N /A A.t.nat./4

.. N/A Alternate (n 465/467 L465 ' " L467. " N/A " ... -.. .. .Note (1) -WHEN Channel III is the "PREFERRED" PRZR PRESS channel, THEN the selection of an Alternate channel for "PRZR LEVEL CONT SEL" should be considered the "PREFERRED" channel.This procedure section, when completed, SHALL be retained.

OPOP03-ZG-0008 Rev. 56 Page 75 of 84 Power Operations Addendum 8 Percent Power vs Program RCS Tavg Page I of I Auctioneered High RCS Tavg = 592°F Percent Program Percent Program Percent Program Percent Program Power Tavg Power Tavg Power Tavg Power Tavg 1 567.25 26 573.50 51 579.75 76 586.00 2 567.50 27 573.75 52 580.00 77 586.25 3 567.75 28 574.00 53 580.25 78 586.50 4 568.00 29 574.25 54 580.50 79 586.75 5 568.25 30 574.50 .55 580.75 80 587.00 6 568.50 31 574.75 56 581.00 81 587.25 7 568.75 32 575.00 57 581.25 82 587.50 8 569.00 33 575.25 58 581.50 83 587.75 9 569.25 34 575.50 59 581.75 84 588.00 10 569.50 35 575.75 60 582.00 85 588.25 11 569.75 36 576.00 61 582.25 86 588.50 12 570.00 37 576.25 62 582.50 87 588.75 13 570.25 38 576.50 63 582.75 88 589.00 14 570.50 39 576.75 64 583.00 89 589.25 15 570.75 40 577.00 65 583.25 90 589.50 16 571.00 41 577.25 66 583.50 91 589.75 17 571.25 42 577.50 67 583.75 92 590.00 18 571.50 43 577.75 68 584.00 93 590.25 19 571.75 44 578.00 69 584.25 94 590.50 20 572.00 45 578.25 70 584.50 95 590.75 21 572.25 46 578.50 71 584.75 96 591.00 22 572.50 47 578.75 72 585.00 97 591.25 23 572.75 48 579.00 73 585.25 98 591.50 24 573.00 49 579.25 74 585.50 99 591.75 25 573.25 50 579.50 75 585.75 100 592.00 This Form when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 76 of 84 USAGE:- IN HAND Power Operations Form 1 Turbine Load Changes using the Limiter Page I of 1 E~h -s IFCAUTION L is Form assumes that Turbine Load is being controlled onl the GOV CONT "VLV LIMIT" (Valve Limiter).NOTE* Normal Turbine Load control for steady state power is using "SETPOINT CONTROL" and Reference counters (Turbine Actual and Setpoint).

• Steady state power control is normally in IMP-OUT, with the limiter at a position at or slightly above the current operating demand and relying on the use of the "SETPOINT CONTROL" and Reference counters (Turbine Actual and Setpoint), with the ramp rate set to 0.25% MW/MIN to make the Main Turbine adjustments required to maintain steady state Reactor Power. (Ref. 2.24)1.0 ENSURE permission to perform a Turbine Load Change is obtained from the Shift Manager/Unit Supervisor.

2.0 IF desired to Raise Turbine/Generator load, THEN PERFORM the following:

2.1 RAISE

the GOV CONT "VLV LIMIT" (Valve Limiter) {CP007- Operators Display panel} by momentary depressing (TAP) the GOV VLV LIMIT CONT"A" pushbutton

{CP007- AEH panel}.2.2 MONITOR for the proper load changes.2.3 As required for the current authorized adjustment, REPEAT the above steps, as required, until desire Load is obtained.3.0 IF desired to Lower Turbine/Generator load, THEN PERFORM the following:

3.1 LOWER

the GOV CONT "VLV LIMIT" (Valve Limiter) {CP007- Operators Display panel} by momentary depressing (TAP) the GOV VLV LIMIT CONT"Y" pushbutton

{CP007- AEH panel}3.2 MONITOR for the proper load changes.3.3 As required for the current authorized adjustment, REPEAT the above steps, as required, until desire Load is obtained.This Form when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 77 of 84 USAGE -IN HAND Power Operations Form 2 Turbine Load Changes using the Setpoint Controls Page 1 of 2 mIO CAUTIONI Load changes MAY be stopped, as needed, by DEPRESSING the SETPOINT CONTROL"HOLD" pushbutton.

I 1.0 ENSURE permission to perform a Turbine Load Change is obtained from the Shift Manager/Unit Supervisor.

2.0 ENSURE

the LOAD RATE -PRCT MW/MIN" Thumbwheel is set to the ".25" position OR as authorized by the Shift Manager/Unit Supervisor.

{CP007}3.0 IF the "LIMITS" light "VLV POS LIMIT" {CP007- Operators Display panel} is lit, THEN PERFORM the following, OTHERWISE N/A: 3.1 LOWER the value of the Setter (as read on the "SETPO1NT" display) 0.2% by momentary depressing (TAP) the SETPO1NT CONTROL "V" pushbutton, as required.NOTE WHEN the "LIMITS" light "VLV POS LIMIT" extinguishes, THEN DEPRESS the SETPOINT CONTROL "HOLD" pushbutton.

3.2 DEPRESS

the SETPOINT CONTROL "GO" pushbutton and MONITOR for the "VLV POS LIMIT" light to extinguish. (IF required, THEN repeat above Steps until the "VLV POS LIMIT" light extinguishes.)

3.3 RAISE

the GOV CONT "VLV LIMIT" (Valve Limiter) to value selected by the Shift Manager/Unit Supervisor

{CP007- Operators Display panel} by momentary depressing (TAP) the GOV VLV LIMIT CONT "A" pushbutton, as required.

{CP007- AEH panel}4.0 IF PWR is < 98% and directed by the Shift Manager/Unit Supervisor, THEN PLACE Main Turbine in the IMP-IN mode JAW OPOP01-TM-0001, Main Turbine/Generator Operations Guidelines, OTHERWISE N/A.This Form when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 78 of 84 USAGE -IN HAND Power Operations Form 2 Turbine Load Changes using the Setpoint Controls Page 2 of 2 5.0 RAISE/LOWER (ADJUST) the value of the Setter (as read on the "SETPOINT" display) to the desired value by momentary depressing (TAP) the SETPOINT CONTROL "A" or "Y" pushbutton, as required.NOTE* The SETPOINT CONTROL "GO" Lamp should extinguish when TURBINE -ACTUAL equals SETPOINT, or if "HOLD" pushbutton is pressed." Steady state power control is normally in IMP-OUT, with the limiter at a position at or slightly above the current operating demand and relying on the use of the "SETPOLNT CONTROL" and Reference counters (Turbine Actual and Setpoint), with tile ramp rate set to 0.25% MW/MIN to make the Main Turbine adjustments required to maintain steady state Reactor Power. (Ref. 2.24)6.0 DEPRESS the SETPO1NT CONTROL "GO" pushbutton and MONITOR for the proper load changes.7.0 IF directed by the Shift Manager/Unit Supervisor, THEN PLACE Main Turbine on the GOV CONT "VLV LIMIT" (Valve Limiter) by performing the following:

7.1 IF the limiter is required to be lowered, THEN ADJUST the GOV CONT "VLV LIMIT" (Valve Limiter) to a position that the "VALVE POS LIGHT" is just Lit or Valve Limiter slightly above the current operating demand by momentary depressing (TAP) the GOV VLV LIMIT CONT "A" or "Y" pushbutton, as required.

{CP007- AEH panel}7.2 IF Setpoint Control requires adjustment, THEN PERFORM the following:

7.2.1 RAISE/LOWER (ADJUST) the value of the Setpoint counter (as read on the "SETPOINT" display) to the desired value by momentary depressing (TAP) the SETPOINT CONTROL "A" or "4*" pushbutton, as required.7.2.2 DEPRESS the SETPOINT CONTROL "GO" pushbutton and MONITOR for the proper load changes.7.2.3 MONITOR the limiter at a position that "VALVE POS LIGHT" is just Lit or Valve Limiter slightly above the current operating demand.This Form when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 79 of 84 USAGE -REFERENCED Power Operations Form 3 (Automatic/Normal Mode ONLY) Online Main Page 1 of 2 Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment CAUTION* Main Generator voltage SHALL be adjusted slowly in order to prevent a volts/hertz trip.* Main Generator voltage response to manual voltage adjustments considerably lags switch operation." Operate the "VOLTAGE ADJUSTER" and the "BASE ADJUSTER" using the Bump and Wait method, (i.e. 1 second bump then a 10 seconds wait before next bump to see results of previous change).* Voltage adjustments SHALL be coordinated between Units to prevent excessive circulating currents.* This Form is for use when the Main Generator voltage regulator is in the Automatic (Normal)Mode only.* Minimize operation in the VARs IN, Leading (Underexcited) region to limit eddy-current heating of the turbine end stator core. Operation in this region, or below the drawn-in curve (Main Generator Capability Curve in the Plant Curve Book) may cause a rise in heating and subsequent damage proportional to the depth and duration of operation below the curve. (CREE 07-4972-2)

• IF performing an ESF DG surveillance (either Unit online or outage), THEN MONITOR DG load while performing the Voltage or Reactive Load adjustment. (Prevents invalidating a PSP or DG overload.)

1.0 ADJUST

the Main Generator Voltage or Reactive Load as directed by performing the following:

{CP007}1.1 ENSURE permission to perform a Main Generator Voltage or Reactive Load Change is obtained from the Shift Manager/Unit Supervisor.

1.2 CONTACT

the other Unit Control Room to coordinate Voltage or Reactive Load adjustment.

1.3 MONITOR

Switchyard voltage and Reactive Power.1.4 UTILIZING the "VOLTAGE ADJUSTER" and USING the Bump and Wait method, THEN slowly ADJUST Main Generator voltage to the desired Voltage or Reactive Load.This Form when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 80 of 84 USAGE -REFERENCED Power Operations Form 3 (Automatic/Normal Mode ONLY) Online Main Page 2 of 2 Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment I a 1.5 UTILIZING the "BASE ADJUSTER" and USING the Bump and Wait method, THEN slowly ADJUST the "VOL REG NULL" meter to ýC"0".1.6 ENSURE the limits of the Main Generator Capability Curve in the Plant Curve Book are NOT exceeded.1.7 IF requested by the STP Coordinator, THEN NOTIFY the STP Coordinator the Main Generator Voltage or Reactive Load adjustment is complete, OTHERWISE N/A.This Form when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 81 of 84 USAGE -REFERENCED Power Operations Form 4 (Manual/Test Mode ONLY) Online Main Generator Page 1 of 2__Voltage/Reactive Load Adjustments CAUTION* Main Generator voltage SHALL be adjusted slowly in order to prevent a volts/hertz trip.* Main Generator voltage response to manual voltage adjustments considerably lags switch operation.

• Operate the "BASE ADJUSTER" using the Bump and Wait method, (i.e. 1 second bump then a 10 seconds wait before next bump to see results of previous change).* Voltage adjustments SHALL be coordinated between Units to prevent excessive circulating currents.* This Form is for use when the Main Generator voltage regulator is in the Manual/Test Mode ONLY.* Minimize operation in the VARs IN, Leading (Underexcited) region to limit eddy-current heating of the turbine end stator core. Operation in this region, or below the drawn-in curve (Main Generator Capability Curve in the Plant Curve Book) may cause a rise in heating and subsequent damage proportional to the depth and duration of operation below the curve. (CREE 07-4972-2)

• IF performing an ESF DG surveillance (either Unit online or outage), THEN MONITOR DG load while performing the Voltage or Reactive Load adjustment. (Prevents invalidating a PSP or DG overload.)

• With the voltage regulator in Manual/Test mode:* The STP Coordinator and the TDSP SHALL be notified within 30 minutes when the Main Generator Voltage Regulator is in Manual/Test mode, and the expected duration (Use 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if unknown). (Ref. 2.23 & 2.28)* A rise in operational awareness is required to maintain the desired MVAR loading.* There is NO automatic compensation for changes in system grid voltage.* Primary generator relay protection and the voltage regulator protective drawer are NOT affected, however certain automatic controls associated with generator transient stability (maximum and minimum excitation limiter, maximum volts/hertz limiter, reactive current compensator, excitation system damping) are NOT functioning.

• IF possible., THEN the load control unit should be shifted to the other Unit.This Form when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 82 of 84 USAGE -REFERENCED Power Operations Form 4 (Manual/Test Mode ONLY) Online Main Generator Page 2 of 2 Voltage/Reactive Load Adjustments

1.0 NOTIFY

STP Coordinator and the TDSP within 30 minutes of the Main Generator Voltage Regulator being placed in Manual/Test and the expected duration (Use 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if unknown). (Ref. 2.28)2.0 ADJUST the Main Generator Voltage or Reactive Load as directed by performing the following:

{CP007}2.1 ENSURE penrmission to perform a Main Generator Voltage or Reactive Load Change is obtained from the Shift Manager/Unit Supervisor.

2.2 CONTACT

the other Unit Control Room to coordinate Voltage or Reactive Load adjustment.

2.3 MONITOR

Switchyard voltage and Reactive Power.2.4 UTILIZING the "BASE ADJUSTER" and USING the Bump and Wait method, THEN slowly ADJUST Main Generator voltage to the desired Voltage or Reactive Load.2.5 ENSURE the limits of the Main Generator Capability Curve in the Plant Curve Book are NOT exceeded.2.6 IF requested by the STP Coordinator, THEN NOTIFY the STP Coordinator the Main Generator Voltage or Reactive Load adjustment is complete AND Main Generator Voltage Regulator is in Manual/Test Mode. (Ref. 2.23)This Fonn when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 83 of 84 USAGE -IN HAND Power Operations Form 5 Routine at Power Minor Rod Movements Page 1 of 2 CAUTION" NEVER pull Control Rods except in a deliberate, carefully controlled manner, while closely monitoring the Reactor's response." Control Rod Insertion limits SHALL be observed at all times. (Technical Specification 3.1.3.6)* Addendum 3, Fuel Conditioning Requirements/Recommendations, SHALL be referenced for guidance on Control Rod movement and Main Turbine loading rates.* Control Rod alignment SHALL be monitored between control rods within banks AND between the bank group step counters. (Technical Specification 3.1.3.1 and 3.1.3.2)Caution SHALL be exercised when moving control rods in regions of high differential rod worth.Small changes in control rod position can produce large reactivity changes in these regions.(Reference 2.7.5)" During Coastdown Operations, Control rod motion should be minimized.

Movement of control rods can cause large shifts in AFD and xenon oscillations to occur. Control Bank D usage should be limited to 3 steps in one direction per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period to maintain AFD less than or equal to 1.0%peak to peak value, or a value within the Target Band as determined by Reactor Engineering with Shift Manager concurrence.

1.0 ENSURE

permission to perform a Minor Rod Movement is obtained from the Shift Manager/Unit Supervisor.

2.0 3.0 ENSURE "ROD BANK SEL" switch is in the MANUAL position.WITHDRAW/INSERT Control Banks by holding Rod Control Switch in OUT/IN position.NOTE The Rod Control System will normally be maintained in Automatic; however, Rod Control may be placed in Manual for any evolution, as deemed necessary by the SM/US.4.0 PLACE the "ROD BANK SEL" switch in the "AUTO" position or as deemed necessary by Shift Manager/Unit Supervisor.

5.0 MONITOR

the condition of the reactor and report to the Shift Manager/Unit Supervisor (Control Room) if any unexpected results from the movement of control rods are realized.This Foma when completed, has NO retention.

OPOP03-ZG-0008 Rev. 56 Page 84 of 84 USAGE -IN HAND Power Operations Form 5 Routine at Power Minor Rod Movements Page 2 of 2 6.0 VERIFY the DRPI and "STEP DEMAND" indications on the Plant Computer display agree with the actual DRPI and Group Demand indications on CP005 for each rod within the selected bank.6.1 IF Plant Computer has the wrong height for step counters, THEN PERFORM the following:

6.1.1 From the Nuclear Applications Program, select the Rod Supervision Menu (i.e.click on the "M" next to "rod supervision")

6.1.2 From the RS menu -rod supervision, select "RS0410 -Rod Bank Step Counter Data Entry" 6.1.3 Enter the actual step counter data on the appropriate line.6.1.4 Click on the apply button.6.2 IF Plant Computer has the wrong height for DRPI, THEN DECLARE Rod Position Deviation Monitor alarm inoperable and INITIATE a Condition Report.7.0 INFORM the Shift Manager/Unit Supervisor (Control Room) at the completion of the evolution and that control systems are returned to their normal expected configuration.

This Form when completed, has NO retention.

SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION D0527 STI 33780132 0POP03-ZG-0006 Rev. 54 Page 1 of 102 Plant Shutdown From 100% to Hot Standby Quality Safety-Related Usage: IN HAND Effective Date: 11/07/2013 CONTROLLING STATION D. Rohan N/A Crew IC Operations PREPARER TECHNICAL USER COGNIZANT DEPT.Usage_ Table of Contents Page 4 1 P u rp o se ................................................................................................

..............

..............

2 4 2 R eferences

......................................................................................

................

.... ...... 2 3 3 Prerequisites

....................................................................................

4 3 4 Notes and Precautions

.......................

.............

...........

5 2 5 Pow er D escent 14...................................................................

...............

..................

14 2 6 M ode 2 D escent .............................................................................................................

41 2 7 R eactor Shutdow n .............................................................................................................

48 2 8 R ecords R eview ..................................................

.................................................................

54 4 9 Support D ocum ents ..........................................................................................................

55 2 Addendum 1, RCS Degassification.ri...

...... ...........................

................................

56 3 Addendum 2, Main Turbine Operating Guidelines

..........................

58 1 Addendum 3, Post ReactorTrip Guideline

...................................................................

59 1 Addendum 4, Fast Reactor. Shutdown .........

............................................................

60 3 Addendum 5, Percefit Power vs Program RCS Tavg ...............................................

70 2 Addendum 6, Faileddor leaking Steam Dump Valves ..............................................

71 2 Addendum 7, Transferring SG feed to the AFW nozzles ........................................

74 2 Addendumi.'

8, Transferring feed from MFRV to LPRV ..........................................

77 2 Addendum 9 Feedwater, Isolation Signal Reset and Establishing SG Feed .............

81 1 Addendum lOiCV-0218 Boration While Inserting Rods ........................................

87/I * **: Addend4m 11 IInhibit Reactor Trip Log ..................................................................

89 I A*," 'ddendum 12 AFD Penalty Point Evaluation

..........................................................

90 1 Addendum l3 Low Power Feedwater Regulating Valves Exercise .........................

93 1 I Addendum 14 Deaerator Aux Steam Supply Valve (PV-7401)

Exercise .................

96/ , /.. .. .1 .Addendum 15 (Unit 2, 2RE16 Only) Turbine Shutdown without Bearing Lift Pump 97 I ": Lineup 1, Turbine Generator Systems Cooling Water Lineup ....................................

101 Usaoe I IN-HAND 2 -IN HAND CONTROLLING STATION 3 -REFERENCED 4 -AVAILABLE OPOP03-ZG-0006 Rev. 54 Page 2 of 102 Plant Shutdown From 100% to Hot Standby 1.0 Purpose 1.1 To provide instructions for plant shutdown from 100% power to Hot Standby (Mode 1 to Mode 3).1.2 To provide guidance for Post Reactor Trip plant stabilization.

2.0 References

2.1 Westinghouse

Precautions, Limitations and Setpoints, 5ZO1OZS1 101 2.2 Turbine Generator Instruction Book Volume 3 Operating Procedures (VTD-W 120-0003).

2.3 Westinghouse

Instruction Book, Steam Generator Feed Pump Drive Turbines, (VTD-W 120-0029).

2.4 Procedures

2.4.1 OPOP02-FW-0002, S.G.F.P Turbine 2.4.2 OPOP02-FW-0001, Main Feedwater 2.4.3 OPOP02-MS-0001, Main Steam System 2.4.4 OPOP02-HV-0001, Feedwater Heater Drains and Vents 2.4.5 OPOP02-CD-0001, Condensate System 2.4.6 OPOP02-AS-0001, Auxiliary Steam System 2.4.7 OPOP02-AF-0001, Auxiliary Feedwater 2.4.8 OPOP02-RS-0001, Rod Control 2.4.9 OPOP02-SB-0001, Steam Generator Blowdown System 2.4.10 OPSP03-FW-0001, Feedwater System Valve Operability Test 2.4.11 OPSP1O-ZG-0003, Shutdown Margin Verification Modes 3, 4 and 5 2.4.12 OPOP03-ZG-0007, Plant Cooldown 2.4.13 OPOP02-WG-0001, Gaseous Waste Processing System Operations 2.4.14 OPGP03-ZE-0033, RCS Pressure Boundary Inspection for Boric Acid Leaks 2.4.15 OPSP1O-DM-0003, Automatic Multiple Rod Drop Time Measurement 2.4.16 OPGP03-ZO-0042, Reactivity Management Program This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 3 of 1021 Plant Shutdown From 100% to Hot Standby 2.4.17 OPOPOI-ZA-0021, AC Electrical Notes and Precautions 2.4.18 OPOP02-EH-0001, Main Turbine Electro-Hydraul ic Control System 2.4.19 OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION 2.4.20 0PEP07-TM-0001, Main Turbine Mechanical Overspeed Test 2.4.21 OPGP03-ZA-0010, Performing and Verifying Station Activities 2.4.22 OPOP04-CV-0003, Emergency Boration 2.4.23 OPOP02-CV-0001, Makeup to the Reactor Coolant System 2.4.24 OPOP03-ZG-0008, Power Operations 2.4.25 OPEP02-ZX-0010, Reload Initial Start-Up Testing 2.5 ST-WT-YB-0044, ST-YB-HS-0584, Emergency DC Lube Oil Pump Service Time of 75 Minutes 2.6 MATS Item 8500034-866 (ST-UB-HL-047), Axial Flux Distribution 2.7 ST-HS-HS-2 1262, CVCS Pump Operation 2.8 MATS Item 9200505-936 (SPR 920128), Excessive Cooldown Due to Secondary Heat Loads.2.9 Technical Specification 3.1.1.4, 3.4.8 (Table 4.4-4 Item 4b), 4.3.1.1.2.a.2, 3.3.5.1, 3.7.1.6.2.10 Design Basis Documents:

2.10.1 5R149MB1027., Reactor Coolant System 2.10.2 5Z529ZB1003, NSSS Controls System 2.10.3 9GOI9MBO 117, Turbine Generator System 2.10.4 5S139MB0120, Feedwater System 2.10.5 5S109MB1026, Main Steam System 2.11 Response to SOER 90-003 2.12 CR 98-498, hydraulic transients were noticed in the condensate system downstream of feed water heater 25C 2.13 CR 98-11491, Observed Main Turbine speed coasting down at approx 120 RPM after opening generator output breaker This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 4 of 102 !Plant Shutdown From 100% to Hot Standby 2.14 ST-W2-NOC-000718, South Texas Delta 94 RSG Feedring Design 2.15 Technical Specification 3.3.5.1, 3.7.1.6 2.16 UFSAR 6.3.2.5 2.17 CREE 05-5888-20, Determine If The Maximum Allowable Continuous Speed For The SGFPs Can Be Increased.

2.18 CR 06-14683-3, Update MED description forNI(2)MDHS6173 to read "WHEN MAIN TURBINE IS NOT LATCHED, THE TURBINE DRAIN VALVES CANNOT BE STROKED CLOSED" 2.19 SPR 941157, Generator Exciter Cooling Was Isolated Causing Exciter To Overheat.2.20 CR 07-15570, While researching electrical circuits for electrical bus outage procedures two questions arose regarding the feedwater isolation valves (FWIV) energize-to-actuate modification.

2.21 OPOP02-CF-0004, Operation of the TGB Polymer Dispersant Injection System.2.22 CREE 11-5655-1, FWH 15/16(25/26) strings have the potential to isolate during plant shutdown 2.23 CREE 11-19352-2, Determine the maximum possible flow rate from one Boric Acid Transfer Pump through CV-0218 to support a faster emergency boration flow.2.24 Temporary Modification T2-13-5155-6, Unit 2 Main Turbine Bearing Lift pressure switch jumper.3.0 Prerequisites 3.1 IF reducing Reactor Power to less than 30%, THEN ENSURE two (2) CARS vacuum pumps available for operation per OPOP02-CR-0001, Main Condenser Air Removal.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 5 of 102 Plant Shutdown From 100% to Hot Standby 4.0 Notes and Precautions 4.1 If entering this procedure from an unplanned Reactor Trip Condition refer to Addendum 3, Post Reactor Trip Guideline for guidance.4.2 Operating PROCEDURES are written based on a defined set of plant conditions and equipment availability.

PROCEDURE changes are NOT required to document alternate performance based on conditions different from those assumed if the PROCEDURE can be performed safely. The decision to proceed lies with the Unit Supervisor/Shift Manager and is based on knowledge of system design and operation and the impact of omitting or re-sequencing steps.4.2.1 The Unit Supervisor/Shift Manager may authorize alternate performance for operating PROCEDURE sequence, including omitting steps, based on plant operating conditions.

The Unit Supervisor/Shift Manager ensures such an alternate performance does NOT adversely impact the safety of personnel or equipment, and documents the alternate method in the appropriate PROCEDURE or logbook. See OPGP03-ZA-0010, Performing and Verifying Station Activities for specific details.4.2.2 The Unit Supervisor/Shift Manager may authorize early start of procedure steps to enhance plant performance, WHEN the early start is of no safety impact for current plant conditions.

Documentation is NOT required for a early start as long as the step is completed before moving past this step in the overall sequence.4.2.3 Steps within this procedure SHALL be performed in order listed or in order provided in a authorized early start (Step 4.2.2) or alternate performance (Step 4.2.1). Steps that are authorized to be omitted SHALL be designated by placing "N/A" in the signoff or initial blanks. See OPGPO3-ZA-00 10, Performing and Verifying Station Activities for specific details.4.3 The Unit Supervisor/Shift Manager SHALL signoff or initial all steps unless otherwise designated within this procedure.

4.4 USE caution when making changes to plant fluid systems to minimize the potential for hydraulic transients.

4.5 This procedure contains Non-Intrusive Check valve testing. System Engineering should be notified prior to reaching Step 5.39 to allow for equipment set-up for these tests.4.6 This procedure provides provisions for transition to OPEP07-TM-0001, Main Turbine Mechanical Overspeed Test. IF the Main Turbine Mechanical Overspeed Test is to be performed THEN adequate personnel should be available to perform testing. Personnel should be staged and ready PRIOR to securing the Main Generator.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 7 Rev. 54 7 Page 6 of102 Plant Shutdown From 100% to Hot Standby 4.7 The following forms provide guidance for aid in the operation of the Main Turbine/Generator and Reactor: (These Forms may be used in Mode 1 and 2, as applicable.)

• OPOP03-ZG-0008, Form 1, Turbine Load Changes using the Limiter* OPOP03-ZG-0008, Form 2, Turbine Load Changes using the Setpoint Controls* OPOP03-ZG-0008, Form 3, (Automatic/Normal Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment* OPOP03-ZG-0008, Form 4, (Manual/Test Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

• OPOP03-ZG-0008, Form 5, Routine at Power Minor Rod Movements* OPOP02-CV-0001, Form 1, Modes 1-2 RCS Boration Checklist* OPOP02-CV-0001, Form 2, Modes 1-2 RCS Dilution Checklist* OPOP02-CV-0001, Form 3, Modes 1-2 RCS Alt Dilution Checklist* OPOP02-CV-0001, Form 4, Modes 1-2 Automatic Operation Checklist 4.8 If Tavg lowers to < 561'F, Comply with Technical Specification 3.1.1.4. IF Tavg cannot be restored to GREATER THAN 561°F within 15 minutes, THEN manually trip the reactor and GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

4.9 IF Reactor Power changes of greater than 15% Rated Thermal Power are performed in any one hour period, THEN NOTIFY Chemistry to perform an isotopic analysis of the RCS for Iodine to satisfy Technical Specification

3.4.8 Table

4.4-4 item 4.b.4.10 Each SG PORV SHALL remain OPERABLE, in AUTO with a nominal setpoint of 1225 +/- 7 psig. (MODES 1, 2) (REF 2.15)4.11 IF SG PORVs are required to be controlled in manual operation, OR in automatic operation with reduced setpoints, THEN an OAS entry is required to ENSURE compliance with Technical Specifications 3.3.5.1, and 3.7.1.6. (REF 2.15)4.12 Notification to NRC is NOT required for the manual actuation of the reactor protection system (RPS) for a pre-planned sequence during testing or reactor operation.

{NUREG-1022, Rev. 2 page 45, 1OCFR50.72(b)

(3)(iv)(A), IOCFR50.72 (b)(2)( iv)(B), I0CFR50.73(a)(2)(iv) (A)(I)}4.13 WHEN steady state operation at less than 85% of full power for greater than two weeks is anticipated, THEN NOTIFY Reactor Engineering and request a recommended Axial Flux Distribution (AFD) control strategy. (Reference 2.6)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 7 of 102 Plant Shutdown From 100% to Hot Standby 4.14 Degassing of the RCS is typically performed by one or a combination of both of the methods below: Mechanical degassing

-where the RCS is degassed by spraying and venting of the Pressurizer and VCT (Ref: Addendum 1, RCS Degassification)

Chemical degassing

-where the RCS is degassed by the addition of chemicals that react with the dissolved gases in the RCS.4.15 IF an RCS boundary opening is planned. THEN RCS Degassing should be performed as determined by the Shift Manager and Chemistry Management.

4.16 Rod Withdrawal Limits SHALL be observed in Modes I and 2.4.17 During power operation, only one CCP should be running. Two CCPs may be operated when required to maintain pressurizer water level, during the switching of one CCP to the other or surveillance testing. (Reference 2.7)4.18 WHEN reactor power is reduced below the Point of Adding Heat AND an unexpected or uncontrolled cooldown occurs, THEN the cooldown SHALL be stopped by isolating secondary heat loads, NOT by diluting and/or withdrawing control rods.4.19 RCS/PZR Pressure Control during steady state and transients, SHOULD be performed by the following: (WHEN controls are NOT responding as expected, THEN Manual intervention for pressure control is AUTHORIZED)

• Automatic control of Pressurizer Pressure Controller RC-PK-0655A (PZR Press Master Controller) to Control the output of the PZR CONTROL HTRS* Automatic control of Pressurizer Spray Valve Controller "PRZR SPR PCV-06551B" and "PRZR SPR PCV-0655C" to Control RCS/PZR Pressure* Backup Heaters may be cycled as necessary to aid in PZR turnover flow or pressure control.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 8 of 102 Plant Shutdown From 100% to Hot Standby 4.20 The following is the PZR Press Master Controller outputs in Manual Operation:

Function Controller Output VDC Controller Output % Signal Direction PCV-0655A OPENS 8.75 87.5 INC (+)PCV-0655A CLOSES 7.50 75.0 DEC (+)PZR PRESS DEV HI 7.19 72.0 INC (4)ALARM SPRAY FULL OPEN 7.19 72.0 INC (4)PRES PRESS DEV HI 6.99 70.0 DEC (0)ALARM RESETS SPRAY FULL 4.06 40.5 DEC (40)CLOSED CONT HTRS 0% PWR 3.44 34.5 INC (4)CONT HTRS 50% 2.50 25.0 N/A PWR CONTROL HTRS 1.56 15.5 DEC (4)100% PWR PZR PRESS DEV LO 0.94 9.5 DEC (+)ALARM & BU HTRS ON 4.21 WHEN drawing a bubble in the pressurizer and prior to starting a RCP, care should be taken to avoid rising pressurizer level. A significant rise in pressurizer level will allow cold water to cover the pressurizer liquid temperature element indicating a dramatic drop in pressurizer temperature.

Then the pressurizer heaters will heat this cool water, indicating a dramatic heating up of the pressurizer. (CR 05-5402)4.22 During Coastdown Operations or plant conditions where calculated Tref is in error (e.g., instrument failure, turbine offline, etc), Program Tavg should be utilized for Tavg vs. Tref comparisons. (REFER TO Addendum 5, Percent Power vs Program RCS Tavg.)4.23 Placing Excess Letdown in service will cause Plant Computer point UI 118 to become inaccurate in the NON-Conservative direction due to bypass flow causing approximately 2 MWth to bypass instrumentation.

This bypass flow and heat is NOT accounted for in the calorimetric.

4.24 In this procedure, WHEN the Main Turbine is online and the Tref instrumentation is functional, THEN the terms "Tref' and "Program Tavg" and are to be considered synonymous.

The indicated "Tref" may be substituted for "Program Tavg". WHEN using the indicated "Tref", THEN referral to Addendum 5, Percent Power vs Program RCS Tavg is NOT mandatory.

This procedure, when completed, SHALL be retained.

IIOPOP03-ZG-0006 Rev. 54 Page 9 of 102 Plant Shutdown From 100% to Hot Standby 4.24.1 "Tref' and "Program Tavg" are only truly equal at 100% power as "Tref' is not linear with "Program Tavg" once you go below 100% power ("Program Tavg" is linear from 0 to 100% power and "Tref' is NOT linear from 0 to 100%power). When stabilizing at a power level less than 100%, there may be some degree of temperature mismatch which needs to be evaluated prior to returning control rods to 'auto'. Past history has shown that automatic control rod motion may occur with as little as 0.3°F mismatch between "Tref" and "Program Tavg".4.25 Operate the Main Generator within the limits of the Main Generator Capability Curve in the Plant Curve Book.4.26 As plant conditions permit and with Shift Manager concurrence MAINTAIN SG blowdown flow rates as recommended by Chemistry.

4.27 IF auxiliary steam from the opposite unit is NOT available, THEN the auxiliary boiler may be started per 0POP02-AS-000L1 Auxiliary Steam System. At least three hours SHALL be allowed to warmup the auxiliary boiler from cold conditions.

4.28 Main Turbine load may be raised and lowered in either OPERATOR AUTO or MANUAL control at the Unit Supervisor/Shift Manager discretion.

Refer to OPOP01-TM-0001, Main Turbine/Generator Operations Guidelines.

4.29 The Main Turbine SHALL NOT be operated with Main Turbine Exhaust Pressure (Condenser Pressure "inHG") in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines.

IF the Main Turbine is operating in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines, THEN the following SHALL be performed within 5 minutes: (Reference 2.2)Main Turbine Exhaust Pressure (Condenser Pressure "inHG") SHALL be returned to a value outside the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines OR Main Turbine load (MWe) SHALL be lowered to a value outside the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines 4.30 IF operation outside the "Restrictive Zone" of Addendum 2, Main Turbine Operating Guidelines, can NOT be established within 5 minutes, THEN the Main Turbine SHALL be removed from service (placed on the jacking gear) in a safe and expeditious manner.(1-5% per minute) and remain out of service until an Engineering Evaluation can be performed.

This procedure, when completed, SHALL be retained.

E OPOP03-ZG-0006 Rev. 54 Page 10 of 102 Plant Shutdown From 100% to Hot Standby 4.3 I Main Turbine Exhaust Pressure should be at the lowest value attainable (i.e., Best Vacuum) during Turbine Startup/Shutdown and Low Load Operation to prevent the overheating of the LP turbine blading and excessive thermal expansion. (Reference 2.2)4.32 During plant shutdown, Gland Steam spillover pressure should be monitored using PI-6154. (CP009)4.33 During plant shutdown, CL-ACW pressure should be monitored using PI-6809. (CP009)4.34 SGs SHALL NOT be fed through the main feed nozzle when reactor power is less than 4%. This is to prevent thermal stratification when feedwater flows are less than 125,000 Ibm/hr. (Reference 2.14)4.35 IF it becomes necessary to operate Turbine Controls in Manual Mode: THEN all Turbine Control procedural steps referencing "Auto" Control Modes SHOULD be interpreted as "Manual" Control Mode.Since the burden on the operator is greatly increased during turbine manual operation, the operator SHOULD NOT start (rollup) the turbine in the Manual Mode unless it is unavoidable and approved by the Unit Supervisor/Shift Manager.Manual Operation of Turbine Controls including transitioning between "Manual" and "Auto" Control Modes is considered "Skill of the Craft". Refer to OPOPO 1 -TM-000 1, Main Turbine/Generator Operations Guidelines.

4.36 IF the Turbine Controls automatically transfers from "Auto" to "Manual" Control Modes, THEN Controls SHALL NOT be returned to "Auto" until the cause of the transfer has been evaluated and the Unit Supervisor/Shift Manager authorized returning to "Auto" control. Refer to OPOPOI-TM-0001, Main Turbine/Generator Operations Guidelines.

4.37 1(2)-FW-MOV-0108, FWH 11A(21A)/1 1B(21B) BYPASS ISOLATION MOV, may be 5% OPEN when the red light goes out and the green light remains illuminated.

Full closure of 1(2)-FW-MOV-0108 is obtained by HOLDING the jog switch in the closed position.The torque switch will electrically shut the motor off when the valve is fully seated.4.38 IF the steam dump system does NOT operate properly to maintain Tavg, THEN the Unit Supervisor/Shift Manager MAY direct actions to control RCS Tavg until the malfunctioning steam dump(s) can be repaired or isolated.4.39 WHEN Steam Dump or the Steam Chest drains open, THEN CLOSE the Steam Dump or the Steam Chest drains, as desired to reduce secondary heat loads.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page I Iof 102 Plant Shutdown From 100% to Hot Standby 4.40 During NORMAL plant operations, do NOT transfer between the Main Turbine IMP PRESS FEEDBACK modes of operation, WHEN Reactor Power is greater than 98%.Transferring the Main Turbine to IMP PRESS FEEDBACK to "IN or "OUT" mode, may result in a momentary load swing. Refer to OPOPO1-TM-0001, Main Turbine/Generator Operations Guidelines. (CR 04-12715)4.41 IF available for operation, WHEN operating

< 30% Power, THEN run two (2) Condenser Air Removal (CARS) pumps to aid in maintaining Condenser vacuum above 27" hg. At lower power levels, the scope of equipment in the plant that is exposed to vacuum expands. The shells of most Feedwater heaters and parts of the MSR's are in a vacuum, thus most steam leaks in the secondary turn into air leaks. A single CARS pump may NOT be able to handle (by design) the volume of air in-leakage. (CR 05-12083)4.42 Minimize operation in the VARs IN, Leading (Underexcited) region to limit eddy-current heating of the turbine end stator core. Operation in this region, or below the drawn-in curve (Main Generator Capability Curve in the Plant Curve Book) may cause a rise in heating and subsequent damage proportional to the depth and duration of operation below the curve. (CREE 07-4972-2) 4.43 Differences in steam leakage from the HP turbine glands as load is raised or lowered may result in unbalanced HP turbine gland steam pressure.

Manual adjustments may be performed, as necessary, on the following components to balance HP turbine gland steam pressure to acceptable values IAW OPOP02-GS-0001, Turbine Gland Seal Steam System:* 1(2)-GS-0201, GLAND STEAM ALTERNATE SPILLOVER BYPASS VALVE" 1(2)-GS-PC-6153, TURBINE STEAM SEAL HIGH PRESSURE TURBINE GLANDS PRESSURE CONTROLLER 1 I(2)-GS-MOV-0079, GLAND STEAM SPILLOVER PV-6156 BYPASS MOV OPERATOR" 1(2)-GS-PC-6156, TURBINE STEAM SEAL SPILLOVER PRESSURE CONTROLLER 4.44 Addendum 9, Feedwater Isolation Signal Reset and Establishing SG Feed, may be used to reset a Feedwater Isolation Signal, as directed by the Unit Supervisor/Shift Manager.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 12 of 102 Plant Shutdown From 100% to Hot Standby 4.45 WHEN the Reactor Power or Turbine Load values/ranges provided within this procedure are NOT obtained, THEN the Unit Supervisor/Shift Manager may authorize expanded ranges for the associated procedure step(s) based upon the following rules: (Rx Pwr = all) Expanded values/ranges for Rx Pwr are NOT allowed when meeting Technical Specification required Power Levels (Example:

Power level to meet QPTR requirements)

• (Rx Pwr > 98%) Expanded values/ranges are NOT allowed* (Rx Pwr 10% -98%) Rx Pwr values/ranges may be expanded by + 2% Rx Pwr* (Rx Pwr 10% -98%) Turbine Load values/ranges are approximations and may vary as seasons and systems change. If specific Turbine Load values are required they will be specified without the approximation symbol.* (Rx Pwr < 10%) Expanded values/ranges are NOT allowed.4.46 With the Energize-to-Actuate design of the FWIVs using ONLY the Non-Safety handswitches to close the FWIVs leaves a potential on a loss of non-class power for the FWIVs to go open. (Ref. 2.20)4.47 PAA dispersant injection SHOULD be isolated before reducing Reactor Power below 30%.4.48 FWH 15/16(25/26) strings have the potential to isolate during plant shutdown. (Refer to CREE 11-5655-1) 4.49 The principles of OPGP03-ZO-0042, Reactivity Management Program are in effect at all times during Operations in this procedure.

4.50 Pressurizer boron concentration should be maintained within 50 ppm of RCS concentration.

Equalize the boron concentration, as necessary, by energizing at least two (2) sets of Pressurizer Heaters to force additional spray.4.51 WHEN power is stabilized at a desired plateau, THEN ENSURE rod control in manual and the governor valves on the limiter to minimize the potential for unexpected power excursions.

This requirement may be relaxed during controlled power changes and as directed by this procedure.

4.52 Reactor Power monitoring during a power reduction or escalation:

• WHEN changing Reactor Power, THEN RCS loop Delta-T power indication should be used until Reactor Power and secondary conditions have stabilized.

Diverse power indications should also be monitored as a backup to confirm RCS loop Delta-T power (e.g. NIs and U 1118).* WHEN steady state conditions are reached at any power level and U 1118 has been confirmed to be accurate, THEN UI 118 should be primary power indication used while also monitoring diverse indications.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 13 of 102 Plant Shutdown From 100% to Hot Standby 4.53 Changes in Tavg, axial and radial power distribution, and control bank position affects the accuracy of tile Power Range NI detectors. (Reference 2.11)* Alternate indications of Reactor Power should be compared to NI power. Alternate indications include average RCS loop Delta-T, plant computer calorimetric estimates, turbine impulse pressure, and turbine generator load.* IF the Power Range NI detectors and calorimetric power indicate a difference greater than 2%, THEN Reactor Power should be stabilized and the respective NI detectors calibrated, unless otherwise directed by the Shift Manager.(Technical Specification 4.3.1.1.2.a.2) 4.54 Control Rod movement guidelines of 0PGPO3-ZO-0042, Reactivity Management Program, should be observed.4.55 IF less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 2800 ppm. IF greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 3500 ppm. (Ref. 2.16)4.56 Time in core life will affect RCS temperature control and control rod movement.

Reactor Engineering provides cycle specific reactivity parameters for the entire core cycle. This information takes into account MTC affects on RCS average temperature changes. Unit Supervisor should review Nuclear Design Report (NDR) as well for the effects of MTC at current time in core life.4.57 Zinc injection SHALL be secured as soon as practical following a reactor trip.4.58 Zinc injection SHALL be secured prior to a planned shutdown.

Zinc Injection may be secured at least one day prior to a planned refueling outage shutdown or as time allows prior to a forced outage to minimize the zinc transient due to zinc return. Consult outage plans or Chemistry for additional guidance as needed.4.59 IF a power reduction is planned for a duration of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, THEN zinc injection rate may be reduced to minimize the impact of zinc return based on plant experience.

Consult with Chemistry for additional guidance.4.60 Following an unplanned power reduction, sample and adjust zinc injection as needed to maintain zinc within the operating band.4.61 IF zinc concentrations rise to outside of the operating band due to zinc return during a downpower or shutdown, THEN zinc injection should be secured and not restarted until zinc is within the operating band and decreasing.

4.62 IF the Chemical Volume and Control System (CVCS) Mix Bed Demineralizer is out of service for greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, THEN reducing or securing zinc injection should be evaluated with Chemistry.

4.63 Significant changes in CVCS letdown flow will impact zinc concentrations in the RCS. IF CVCS letdown flow is reduced or rise by 20% or greater, THEN contact Chemistry for additional guidance.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 14 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.0 Power Descent 5.1 RECORD the Unit, Date and Time this procedure was entered.Unit Date Time 5.2 REVIEW OPGP03-ZO-0042, Reactivity Management Program.NOTE The following Step is a CONTINUOUS ACTION Step, the applicable Forms may be implemented as many times as needed or as directed by the Unit Supervisor/Shift Manager.5.3 REVIEW and IMPLEMENT the following Forms as applicable for guidance in the operation of the Main Turbine/Generator and Reactor: " OPOPO3-ZG-0008, Form 1, Turbine Load Changes using the Limiter" OPOPO3-ZG-0008, Form 2, Turbine Load Changes using the Setpoint Controls* 0POP03-ZG-0008, Form 3, (Automatic/Normal Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments

& Null Meter Alignment* OPOP03-ZG-0008, Form 4, (Manual/Test Mode ONLY) Online Main Generator Voltage/Reactive Load Adjustments" OPOP03-ZG-0008, Form 5, Routine at Power Minor Rod Movements* OPOP02-CV-0001, Form 1, Modes 1-2 RCS Boration Checklist* OPOP02-CV-0001, Form 2, Modes 1-2 RCS Dilution Checklist* OPOP02-CV-0001, Form 3, Modes 1-2 RCS Alt Dilution Checklist* OPOP02-CV-0001, Form 4, Modes 1-2 Automatic Operation Checklist This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 I Rev. 54 Page 15 of 102 Plant Shutdown From 100% to Hot Standby I Initials NOTE" Pressurizer boron concentration should be maintained within 50 ppm of RCS concentration.

Equalize the boron concentration, as necessary, by energizing at least two (2) sets of Pressurizer Backup Heaters to force additional spray. IF Shutdown is for refueling outage using CV-0218 boration, THEN all available Pressurizer Backup Heaters may be required.* Efforts to maintain Pressurizer boron concentration within 50 ppm of RCS concentration are secondary to safe Pressurizer Pressure control.* Do NOT place the Main Turbine in IMP PRESS FEEDBACK "IN" WHEN Reactor Power is greater than 98%. (CR 04-12715)5.4 IF directed by the Unit Supervisor/Shift Manager AND Plant conditions can support it, THEN the Backup Heaters may be energized as necessary to aid in PZR turnover flow.5.5 MAINTAIN Main Generator cold gas temperature greater than or equal to 90'F while reducing Main Generator load.5.6 PERFORM Addendum 13, Low Power Feedwater Regulating Valves Exercise, prior to placing the Low Power Feedwater Regulating Valves in-service.

5.7 PERFORM

Addendum 14, Deaerator Aux Steam Supply Valve (PV-7401)Exercise, prior to placing "AUX STEAM TO DEAERATOR PV-7401" in-service.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 I Rev. 54 Page 16 of 102 Plant Shutdown From 100% to Hot Standby initials CAUTION* The Main Turbine SHALL NOT be operated with Main Turbine Exhaust Pressure (Condenser Pressure "inHG") in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines.

IF the Main Turbine is operating in the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines, THEN the following SHALL be performed within 5 minutes: (Reference 2.2)Main Turbine Exhaust Pressure (Condenser Pressure "inHG") SHALL be returned to a value outside the "Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines OR Main Turbine load (MWe) SHALL be- lowered to a value outside the"Restrictive Zone" of the Exhaust Pressure Limitation Curve in Addendum 2, Main Turbine Operating Guidelines

• IF operation outside the "Restrictive Zone" of Addendum 2, Main Turbine Operating Guidelines can NOT be established within 5 minutes, THEN the Main Turbine SHALL be removed from service (placed on the jacking gear) in a safe and expeditious manner.(1-5% per minute) and remain out of service until an Engineering Evaluation can be performed.

NOTE Main Turbine Exhaust Pressure should be at the lowest value attainable (i.e., Best Vacuum)during Turbine Startup and Low Load Operation to prevent the overheating of the LP turbine blading and excessive thermal expansion. (Reference 2.2)5.8 ENSURE the Main Turbine Exhaust Pressure is within the limits of Addendum 2, Main Turbine Operating Guidelines.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 j Rev. 54 Page 17 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE Control rods may also be used for aiding in Tavg control provided Delta I is maintained within the limits of Technical Specifications.

If Tavg lowers to < 561'F, Comply with Technical Specification 3.1.1.4. IF Tavg cannot be restored to GREATER THAN 561OF within 15 minutes, THEN manually trip the reactor and GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

5.9 COMMENCE

RCS boration to establish the desired Tavg ramp rate. {CP004}5.10 COMMENCE turbine load reduction at the desired ramp rate. {CP007}5.11 ADJUST turbine load reduction or RCS boron concentration to maintain Tavg within 1.5 degrees of Program RCS Tavg (REFER TO Addendum 5, Percent Power vs Program RCS Tavg).5.12 WHEN Reactor Power goes below 98%, THEN PERFORM the following:

5.12.1 ARM the Modulate Signal for the Main Steam to DA valves by performing the following: (A single handswitch controls both valves)* PLACE 1(2)-MS-PV-7174 and 1(2)-MS-PV-7174A handswitch to the "MOD" position and return to "AUTO".CAUTION PLACING the Main Turbine in the IMP PRESS FEEDBACK "IN" mode, may result in a momentary load swing. REFER TO OPOPOI-TM-0001, Main Turbine/Generator Operations Guidelines.

5.12.2 IF directed by the Unit Supervisor/Shift Manager, THEN PLACE Main Turbine in the IMP-IN mode by depressing the IMP PRESS FEEDBACK "IN" push-button.

{CP007}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 18 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE 1(2)-MOV-0 108 may be 5% OPEN when the red light goes out and the green light remains.illuminated.

Full closure of 1(2)-MOV-0108 is obtained by HOLDING the jog switch in the closed position.

The torque switch will electrically shut the motor off when the valve is fully seated.5.13 {Mark N/A if feedwater heater 11 (21)A/B removed for maintenance)

WHEN Reactor Power goes below 95%, THEN ENSURE CLOSED 1(2)-MOV-0 108, FEEDWATER HEATER I IA/I IB(21A/21B)

BYPASS ISOLATION MOV.5.14 COMMENCE a Reactor Power reduction to less than = 65% Rx Pwr.5.15 IF Reactor Power is to remain below z 65% for an extended period, THEN one (1) SGFP may be secured as follows: 5.15.1 IF S/U SGFP 14(24) is NOT operating, THEN PLACE "S/U SGFP 14(24)" handswitch in the PULL TO LOCK position.{CP006}NOTE* WHEN the SGFP being secured is at its minimum speed, THEN the margin left oil the"SGFP MASTER SPEED" controller should be greater than or equal to 20% for unexpected transients.

{CP006}* IF SGFP testing is to be performed, THEN it may be desirable to maintain the S/U SGFP in PULL TO LOCK position.5.15.2 SECURE one SGFP per OPOP02-FW-0002, S.G.F.P. Turbine.5.15.3 WHEN SGFP testing is complete THEN ENSURE "S/U SGFP 14(24)" handswitch is in the AUTO position.

{CP006}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 19 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE IF a Plant Cooldown is planned, THEN it is recommended that the transfer of auxiliary steam loads occur as soon as possible to minimize time delays at low power levels.Closing the Deaerator Steam valves "SPLY VLVs PV-7174" and "PV-7174A"{CP008} in Step 6.12, should be delayed until < 20% Rx Pwr.5.16 IF desired at this time to warm up the Auxiliary Steam header or transfer auxiliary steam loads, THEN PERFORM the applicable portions of Step 6.3.5.17 IF Addendum 4, Fast Reactor Shutdown is to be performed, THEN Addendum 4, Fast Reactor Shutdown, Steps 1.1 through 1.3 MAY be performed as time permits.5.18 IF a Cooldown is planned, THEN CALCULATE the boron addition that will be necessary to obtain an RCS boration to Cold Shutdown (RCS at 68 0 F, Xenon-Free) Shutdown Margin concentration.

5.19 ESTABLISH the conditions (i.e., tools, hoses, power cords, etc) to enable the closure of the SG main feedwater regulating valve(s) isolation valves "FW REG VLV ISOL VLV(s)" {Do NOT CLOSE the valves until instructed by this procedure}.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 20 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.20 CONTINUE lowering Reactor Power and Turbine Load to between 14 and 16%Rx Pwr, WHILE performing the following steps at the indicated loads: 5.20.1 WHEN Power is at orjust above 50% Rx Pwr, CHECK AFD is within its target band near the AFD Penalty Point transition power level of 50% Reactor Power.5.20.1.1 IF AFD is outside of its target band near the AFD Penalty Point transition power levels (15% and 50%), THEN COMPLETE Addendum 12, AFD Penalty Point Evaluation, to ensure operability of the Technical Specification AFD Monitor Alarm and correct AFD Penalty Points if required.5.20.2 WHEN Power goes below 50% Rx Pwr, THEN VERIFY P9 status"P9 RX/TURB TRIP BLOCKED" PERM LAMPBOX illuminated.

{5M024}NOTE Condensate flow SHALL be limited to 11,000 gpm per pump to prevent Condensate Pump runout. Condensate flow should be approximately 7500 gpm at 40% Rx Pwr.5.20.3 IF Reactor Power is to remaini less than z 40% Rx Pwr for an extended period, THEN PERFORM the following, as directed by the Shift Manager: 5.20.3.1 SECURE one (1) condensate pump per OPOP02-CD-0001, Condensate System.5.20.3.2 SECURE one (1) feedwater booster pump, leaving one (1) feedwater booster pump in service and at least one (1) in standby per OPOP02-FW-0001, Main Feedwater.

5.20.3.3 ENSURE SGBD Temperature is less than 125°F using Plant Computer Point "T4405 SG BLWDN HX OUTLET TEMP". (References 2.4.9 and 2.12)5.20.3.4 ENSURE Condensate temperature exiting the blowdown heat exchanger is less than 265°F using Plant Computer Point "T4407 SG BLWDN HX COND OUTLET TEMP". (Reference 2.12)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 21 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.20.4 WHEN Power is less than 40% Rx Pwr, THEN VERIFY "P8 THREE LOOP OPERATION PERMITTED" PERM LAMPBOX illuminated.

{5M024}NOTE"C-20 AMSAC BLOCKED" permissive should come in with impulse pressure less than 224 psig.5.20.5 WHEN Reactor Power is less than 40% during lowering of power, THEN VERIFY C-20 AMSAC BLOCKED permissive is illuminated on 5M024.NOTE Coordinate with Chemistry as to when PAA dispersant injection should be secured. However, PAA dispersant injection SHOULD be secured prior to lowering below 30% Reactor Power.5.20.6 WHEN Reactor Power is being reduced to less than 3 o0%, THEN PRIOR TO reaching Reactor Power of 30%, ENSURE PAA dispersant injection secured per OPOP02-CF-0004, Operation of the TGB Polymer Dispersant Injection System.NOTE* WHEN Reactor Power is below z 34% Rx Pwr, THEN only one (1) SGFP is required to be operating.

IF SGFP testing is to be performed, THEN it may be desirable to maintain the S/U SGFP in PULL TO LOCK position.5.20.7 IF Reactor Power is to remain below z 34% for an extended period, THEN one (1) SGFP may be secured as directed by Shift Manager as follows: 5.20.7.1 WHEN the SGFP being secured is at minimum speed, THEN VERIFY greater than or equal to 20% margin left on the "SGFP MASTER SPEED" controller for unexpected transients.

{CP006}_5.20.7.2 IF "S/U SGFP 14(24)" is NOT operating, THEN PLACE "S/U SGFP 14(24)" handswitch in the PULL TO LOCK position.

{CP006}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 22 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.20.7.3 SECURE the unnecessary SGFP(s) per 0POP02-FW-0002, S.G.F.P. Turbine.5.20.7.4 WHEN SGFP testing is complete THEN ENSURE"S/U SGFP 14(24)" handswitch is in the AUTO position.

{CP006}5.20.7.5 PLACE "SGFP MASTER SPEED" controller in the MAN position.

{CP006}5.20.7.6 Slowly RAISE SGFP pump speed to approximately 5200 rpm. {CP006}5.20.8 PLACE the following MSR 11(21) and 12(22) steam vent valves in the "COND" (TO CNDSR) position:

{CP007}* "MSR 11(21) N STM VENT" (FWH 11(21)A)* "MSR 11(21) S STM VENT" (FWH I 1(21)A)0 "MSR 12(22) N STM VENT" (FWH 11 (21 )B)0 "MSR 12(22) S STM VENT" (FWH 11 (21)B)NOTE WHEN operating

< 30% Rx Pwr, THEN run two (2) Condenser Air Removal (CARS) pumps to aid in maintaining Condenser vacuum above 27" hg. At lower power levels, the scope of equipment in the plant that is exposed to vacuum expands. The shells of most Feedwater heaters and parts of the MSR's are in a vacuum, thus most steam leaks in the secondary turn into air leaks. A single CARS pump may NOT be able to handle (by design) the volume of air in-leakage. (CR 05-12083)5.20.9 IF available, THEN ENSURE two (2) CARS vacuum pumps in operation per 0POP02-CR-0001, Main Condenser Air Removal.5.20.10 ENSURE Turbine(s)

Gland Seal Steam Pressure(s)

AND Gland Steam spillover pressure is being maintained IAW OPOP02-GS-0001, Turbine Gland Seal Steam System.This procedure.

when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 23 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.20.11 WHEN Reactor Power is being reduced to less than = 20% Rx Pwr, THEN PRIOR TO reaching Reactor Power of 20% Rx Pwr, REMOVE the FW Heaters 1 1A(21A) and 1 1B(21B) from service by performing the following:

5.20.11.1 CLOSE the following valves:* "1(2)-HV-0003 FW HEATERS 1 IA(2 IA)""CONTINUOUS VENT VLV TO""CONDENSER

1. 1 3(23)" (55 ft TGB S End of FWH 11A(21A)}"1(2)-HV-0007 FW HEATERS I IB(21B)""CONTINUOUS VENT VALVE TO""CONDENSER

1. 13(23)"{55 ft TGB E End ofFWH 1 IB(21 B)}5.20.11.2 JOG CLOSED FW Heater I IA(21A) and 11 B(21B)"EXTR STM BLOCK" valves. {CP008}* "HTR 1 1A(21A) MOV-0063"* "HTR 1 1B(21 B) MOV-0067" 5.20.11.3 CLOSE the following valves:{55 ft TGB E of EHC Skid)"1(2)-MD-0223 ES TO FWH #1 lA(#21A)""DRIP LEG DRAIN LV-7926 HIGHSIDE""ISOL""1(2)-MD-0238 ES TO FWH # 11B(#21B)""DRIP LEG DRAIN LV-7928 HIGHSIDE""ISOL" 5.20.11.4 PERFORM the System Shutdown Valve Lineup per OPOP02-HV-0001, Feedwater Heater Drains and Vents.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 24 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.20.12 WHEN Reactor Power is between 20% to 25%, THEN ENSURE intermediate range high flux level trip and the power range low power trip "BISTABLE STATUS MONITORING" lights are OFF for the following:

{5M005}* "INTERMEDIATE RANGE HI CH 1"* "INTERMEDIATE RANGE HI CH 2"* "POWER RANGE LO CH 1"* "POWER RANGE LO CH 2"* "POWER RANGE LO CH 3" 0 "POWER RANGE LO CH 4" 5.20.13 WHEN Reactor Power is reduced below 20% Rx Pwr, THEN PERFORM the following:

5.20.13.1 SECURE "LP HTR DRIP PUMP(s)".

{CP008}i "11(21)" 1 "12(22)"1* "13(23)" 5.20.13.2 SECURE "MSR DRIP PUMP(s)".

{CP008}" "IlA(21A)"" "1IB(21IB)""12A(22A)""12B(22B)" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 25 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.20.14 ENSURE the "STM LN DRAHN"(s) for the running SGFP Turbine(s) are OPEN, N/A Non-running SGFP(s): (CP006)SGFP 11(21)"LV-7954/7955/7973" SGFP 12(22) "LV-7952/

7953/ 7974" SGFP 13(23) "LV-7950/

7951/ 7975" NOTE* The Steam Dump drains will open when turbine load reaches z 20%.* WHEN Steam Dump or the Steam Chest drains are opened, THEN, as necessary to aid in controlling secondary heat loads, cycle CLOSED the Steam Dump and the Steam Chest drains.0 Steam Dump and the Steam Chest drains should remain open at low power levels, when possible, to remove excess moisture.* When the Main Turbine is NOT latched the HP Main Turbine Drains and MST Crossunder Drains can NOT be closed. (Ref. 2.18)5.20.14.1 ENSURE the HP Main Turbine and MST Crossunder Drain Handswitch HS6173, "TURB STM LN DRN VLV" is in the OPEN position.

{CP007}5.21 WHEN Power is at orjust above 15% Rx Pwr, CHECK AFD is within its target band near the AFD Penalty Point transition power level of 15% Reactor Power.5.21.1 IF AFD is outside of its target band near the AFD Penalty Point transition power levels (15% and 50%), THEN COMPLETE Addendum 12, AFD Penalty Point Evaluation, to ensure operability of the Technical Specification AFD Monitor Alarm and correct AFD Penalty Points if required.This procedure, when completed, SHALL be retained.

7 OPOP03-ZG-0006 Rev. 54 Page 26 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE IF SU SGFP 14(24) is unavailable, OR it is desired to use the last SGFP to shutdown, THEN the lowering load may continue and the SGFP will be secured after AFW flow is established.

To minimize contaminated waste water discharges the SU SGFP is the Preferred Option when shutdown is associated with SG Tube Leakage.5.22 WHEN Reactor Power is between 14% and 20%., THEN PERFORM the following:

5.22.1 STABILIZE the Reactor Power, THEN MAINTAIN Reactor Power between 14% and 20%. {CP007}CAUTION Opening SU SGFP 14(24) discharge valve will lead to SG level and pressure oscillations due to the SGFP pressure being less than SU SGFP 14(24) discharge pressure.Oscillations can be minimized by ensuring the discharge pressures for both pumps are as close to equal as possible.5.22.2 ENSURE SU SGFP 14(24) "DISCH ISOL MOV-0518" closed.{CPO06}5.22.3 START SU SGFP 14(24) per OPOP02-FW-0001, Main Feedwater.

5.22.4 PLACE "SGFP MASTER SPEED" controller in the MAN position.{ CP006)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 27 of 102 Plant Shutdown From 100% to Hot Standby Initials CAUTION* SGFP bearing oil temperatures SHALL be monitored closely WHEN speed is > 5300 rpm." SGFP bearing oil temperatures may rise rapidly with speed > 5300 rpm.* Operation of the SGFP in the range of 5400-5600 RPM is acceptable with the following restrictions:

• 5500-5600 RPM:* May be operated in this range for one hour maximumrn.

• Individual SGFP flow SHALL be greater than 11000 gpm.* 5400-5500 RPM:* May be operated in this range indefinitely.

• Individual SGFP flow SHALL be greater than 8250 gpmn.5.22.5 IF the actions in Steps 5.22.6 through 5.22.9 were performed in OPOP02-FW-0001, THEN N/A Steps 5.22.6 through 5.22.9 AND GO TO Step 5.22.10.5.22.6 Slowly RAISE SGFP speed until:* Differential pressure across Startup SGFP discharge valve is less than or equal to 300 psid as indicated on S/U SGFP "PRESS PI-7187" and SG Inlet "PRESS PI-0558".

{CP006}, or* SGFP speed is approx 5400 rpm, or* SGFP bearing oil temperature reaches an operational limit, or Directed by the Unit Supervisor/Shift Manager to STOP the speed rise.5.22.7 ENSURE OPEN SU SGFP 14(24) "DISCH ISOL MOV-0518".

{CP006}5.22.8 Slowly LOWER SGFP speed and OBSERVE that the SU SGFP 14(24) starts supplying SG feedwater.

{CP006}5.22.9 WHEN SU SGFP 14(24) is supplying SG feedwater, THEN SECURE the SGFP per OPOP02-FW-0002, S.G.F.P. Turbine.5.22.10 ENSURE Turbine(s)

Gland Seal Steam Pressure(s)

AND Gland Steam spillover pressure is being maintained lAW OPOP02-GS-0001, Turbine Gland Seal Steam System.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 28 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.22.11 PERFORM the following to prepare the shift crew for removing the Main Generator from the Grid: 5.22.11.1 REVIEW Addendum 6, Failed or leaking Steam Dump Valves.5.22.11.2 ENSURE a pre-job briefing performed JAW Conduct of Operations Chapter 2, addressing expected Primary and Secondary responses to Steam Dump Operation.

5.22.11 .3 ENSURE dedicated Operator(s) available in TGB 55ft area to evaluate and isolate leaking or failed steam dumps, as needed.NOTE IF the steam dumps are NOT responding to plant conditions as required, THEN entry into an Emergency Operating (EOP) OR Abnormal Operating (AOP) procedure is NOT required as long as the excessive steam demand is controllable and addressed by Addendum 6, Failed or leaking Steam Dump Valves.5.22.11.4 MONITOR to ENSURE the steam dumps are responding to plant conditions as required (i.e., NO STUCK/FAILED OPEN VALVES).5.22.11.5 IF the steam dumps are NOT responding to plant conditions, as required, THEN as directed by the Unit Supervisor/Shift Manager, PERFORM Addendum 6, Failed or leaking Steam Dump Valves.5.22.12 IF directed by the Shift Manager to remove the Main Turbine or Generator from service, WHILE maintaining Reactor Power between 14% and 20%, THEN GO TO OPOP03-ZG-0005 "Plant Startup to 100%" Addendum 9, Extended Steam Dump Operations.

This procedure, when completed, SHALL be retained.

OPOP3-ZG0006Rev.

54 Pae 29 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.23 ENSURE a pre-job briefing on Transferring feed from MFRV to LPRV performed LAW Conduct of Operations Chapter 2. {REFER to Addendum 8, Transferring feed from MFRV to LPRV.}5.24 ENSURE Addendum 13, Low Power Feedwater Regulating Valves Exercise, complete.5.25 TRANSFER feed from the main feedwater regulating valves (MFRV) to the low power feedwater regulating valves (LPRV). {REFER to Addendum 8, Transferring feed from MFRV to LPRV.}CAUTION* Actual Reactor Power may be greater than the Reactor Power indicated by the nuclear instrumentation.

The RCS Delta-T will be more indicative of actual Reactor Power.I 1F Tavg lowers to < 5617F, THEN COMPLY with Technical Specification 3.1.1.4.* IF Tavg cannot be restored to GREATER THAN 5617F within 15 minutes, THEN manually trip the Reactor AND GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

NOTE IF a Reactor Shutdown is planned or SG main feedwater regulating valve(s) leak by seat, THEN isolating the SG main feedwater regulating valves (Step 5.40) MAY be PERFORMED in parallel with Main Turbine load reduction/shutdown, as directed by the Unit Supervisor/Shift Manager.5.26 CONTINUE Main Turbine load reduction.

{CP007}5.27 IF 0PEP07-TM-0001, Main Turbine Mechanical Overspeed Test, is to be performed, THEN REFER TO that Test to ensure heat soak requirements are met prior to lowering below 10% Turbine load.5.28 WHEN Turbine Load lowers below z 208 MWe, THEN ENSURE"ROD BANK SELECTOR SW" in MANUAL AND CONTROL Tavg within I°F of Program RCS Tavg (REFER TO Addendum 5, Percent Power vs Program RCS Tavg). {CP005}5.29 WHEN Main Turbine load lowers below- 138 MWe, THEN ENSURE P-1 3 PERM LAMPBOX "P13 TURB LOAD LESS THAN 10 PRCT" is illuminated.

{5M024}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 30 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.30 WHEN Rx Pwr lowers to between 6% and 10%, THEN PERFORM the following:

5.30.1 ENSURE "P10 MAN BLOCK INT/LO PR RX TRP PERM" PERM LAMPBOX off. {5M024}5.30.2 ENSURE "P7 POWER OPER RX TRIPS BLKD" PERM LAMPBOX illtuminated indicating the block for the following trips:{5M024}* Pressurizer Low Pressure (1870 psig)0 Pressurizer High Level (92%)* Two-loop Loss of Flow (91.8%)* RCP Undervoltage (10,014 volts)* RCP Underfrequency (57.2 Hz)5.30.3 ENSURE intermediate range high flux level trip unblocked as indicated by the following STATUS LAMPBOX lights being off:{5M023}* "INT RANIF. RYZ TRIP" "RI flZIkI:V1 TRAIN R"* "INT RANGE RX TRIP" "BLOCKED TRAIN R" 5.30.4 ENSURE power range low power trip unblocked as indicated by the following STATUS LAMPBOX lights being off: {5M023}* "LO PWR RANGE RX TRIP" "BLOCKED TRAIN R"* "LO PWR RANGE RX TRIP" "BLOCKED TRAIN S" 5.31 STOP lowering Main Turbine load between 6% and 8% Reactor Power.{CP007}5.32 ENSURE Turbine(s)

Gland Seal Steam Pressure(s)

AND Gland Steam spillover pressure is being maintained IAW OPOP02-GS-0001, Turbine Gland Seal Steam System.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 31 of 102 Plant Shutdown From 100% to Hot Standby I Initials NOTE* SGs SHALL NOT be fed through the main feed nozzle when Rx Pwr is less than 4%.This is to prevent thermal stratification when feedwater flows are less than 125,000 Ibm/hr.(Reference 2.14)* Rod pulls or dilutions may be necessary to hold Rx Pwr between 6% and 8%.5.33 IF Addendum 4, Fast Reactor Shutdown is to be performed AND it is desired to NOT transfer SG feed to the AFW nozzle (Step 5.39), THEN MAINTAIN Rx Pwr between 6% and 8%.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 32 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.34 IF steam dumps are in T-AVG position AND NOT Armed, THEN TRANSFER steam dumps to pressure control mode as follows: {CP007}CAUTION Placing the Steam Dumps in the Pressure Control mode may result in a Reactor Power rising due to possible controller drift resulting in unwanted valve movement.5.34.1 PLACE one "STEAM DUMP INTLK SEL" switch in OFF/RESET.(CP007)5.34.2 ENSURE "HDR PRESS CONT PK-0557" in Manual. (CP007)5.34.3 PLACE steam dump "MODE SEL" switch in STEAM PRES (PR)position. (CP007)5.34.4 ENSURE "DEMAND UI-0555" indicates 0% and all Steam Dump valves are closed. IF demand is greater than 0%, THEN slowly adjust "HDR PRESS CONT PK-0557" potentiometer. (CP007)5.34.5 ENSURE both "STEAM DUMP INTLK SEL" switches are in ON: (CPO07)* TRAIN A* TRAIN B 5.34.6 VERIFY "STM DUMP UNBLOCK AVAILABLE" light is ON.(CPoo7)5.34.7 ADJUST "HDR PRESS CONT PK-0557" setpoint to equal the current "MAIN STEAM HEADER PRESSURE PI-0557". (CP007)(Setpoint

= PI-0557 value / 1400 x 10)5.34.8 ENSURE "HDR PRESS CONT PK-0557" in AUTO. (CP007)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 33 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE Adjustment may take several changes to obtain a stable indication of Steam Dump Demand on UI-0555. Header Pressure Controller PK-0557 potentiometer should be adjusted in Step 5.34.9 until a stable 1% indication is set.5.34.9 Slowly adjust "HDR PRESS CONT PK-0557" potentiometer until"DEMAND UI-0555" indicates 1%. (CP007)5.34.10 ENSURE all steam dump valves closed by adjusting "HDR PRESS CONT PK-0557" setpoint as necessary.

5.35 IF steam dumps are in T-AVG position AND ARE Armed, THEN TRANSFER steam dumps to pressure control mode as follows: {CP007}5.35.1 PLACE the Steam Dump Controller "HDR PRESS CONT" PK-557 in Manual.5.35.2 ENSURE the Steam Dump Controller "HDR PRESS CONT" PK-557 at Minimum Demand.5.35.3 PLACE the Steam Dump MODE SEL SWITCH to STM PRESSURE Mode.5.35.4 MAINTAIN RCS Temperature within 3' of Program RCS Tavg (REFER TO Addendum 5, Percent Power vs Program RCS Tavg) by controlling Steam Dumps in Manual.5.35.5 ADJUST "HDR PRESS CONT PK-0557" setpoint to correspond to desired steam header pressure.5.35.6 ENSURE "HDR PRESS CONT PK-0557" in AUTO.5.36 IF the Main Turbine is being secured, THEN SECURE the MSRs per OPOP02-MS-0001, Main Steam System.This procedure, when completed, SHALL be retained.

0PO P03-Z G -0 0 0 6 R ev .54 Page 34 of 102 Plant Shutdown From 100% to Hot Standby Initials CAUTION I IF the steam dumps do NOT operate properly when the Main Turbine is tripped and steam flow lowers with the main feedwater regulating valves unisolated, THEN a P-14 FW isolation may occur.* If Tavg lowers to: 561'F, Comply with Technical Specification 3.1.1.4. IF Tavg cannot be restored to GREATER THAN 561'F within 15 minutes, THEN manually trip the reactor and GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

NOTE The steam dumps will open to maintain Reactor Power relatively constant during IF Turbine Controls are NOT in "Operator Auto", THEN Turbine Speed will rapidly lower when the GEN BKR is OPENED. (Reference 2.13)Main Turbine Generator Torsional alarms may be received when the main generator breaker is opened.5.37 SECURE the Main Generator as follows: 5.37.1 REDUCE Main Generator load to;- 63 MWe as indicated on the"GENERATOR" megawatt recorder.

{CP007}5.37.2 MAINTAIN Main Generator reactive load between 20 to 50 MVARs in the lagging direction (MVARs Out) as indicated on the GENERATOR "MVARS" meter. {CP007}5.3 7.3 PLACE "GEN BKR" in the PULL TO LOCK position.

{CP007}5.37.4 PLACE Main Generator Exciter Field Breaker "FLD BKR" in the PULL TO LOCK position.

{CP007}5.37.5 PLACE the "VOLT REG CONT" switch in the OFF position.{CP007}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 35 of 102 Plant Shutdown From 100% to Hot Standby Initials 5.37.6 ENSURE the following controls are positioned as indicated:{CP007}"BASE ADJUSTER":

full counterclockwise position (at ; 0%)."VOLTAGE ADJUSTER":

full counterclockwise position (at z 0%).NOTE Step 5.37.7 is based on the 250 VDC battery system design capacity with NO battery charger in service. (Reference 2.5)5.37.7 IF the Emergency Lube Oil Pump is required to be used during Main Turbine cooldown, THEN the Main Turbine SHALL be verified stopped and the Emergency Lube Oil Pump SHALL be stopped'within 75 minutes.5.37.8 START "L.O./SEAL OIL BACKUP PUMP". {CP007}NOTE Tripping the Main Turbine may cause the Standby EHC pump to AUTO- START.The Vendor recommends NO more than 15 rmin. in parallel EHC Pump operations.

5.37.9

• IF directed by the Unit Supervisor/Shift Manager, THEN START the STBY EHC Pump.* "EHC SPLY PUMP 11(21)"* "EHC SPLY PUMP 12(22)" 5.37.10 IF Main Turbine Mechanical Overspeed Testing is to be performed, THEN PERFORM OPEP07-TM-0001, Main Turbine Mechanical Overspeed Test.5.37.11 IF Main Turbine Mechanical Overspeed Testing will NOT be performed, THEN TRIP the Main Turbine. {CP007}5.37.12 ENSURE steam dump demand rises to maintain Reactor Power relatively constant.

{CP007}This procedure, when completed, SHALL be retained.

O 0POP03-ZG-0006 Rev. 54 Page 36 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE IF the steam dumps are NOT responding to plant conditions as required, THEN entry into an Emergency Operating (EOP) OR Abnormal Operating (AOP) procedure is NOT required, as long as, the excessive steam demand is controllable.

5.37.13 ENSURE the steam dumps are responding to plant conditions as required (i.e., NO STUCK/FAILED OPEN VALVES).5.37.14 IF the steam dumps are NOT responding to plant conditions as required, THEN as directed by the Unit Supervisor/Shift Manager, PERFORM Addendum 6, Failed or leaking Steam Dump Valves.5.37.15 VERIFY the Main Turbine Throttle and Governor Valves closed.{CP007}5.37.16 VERIFY all Reheat Stop and Intercept Valves closed. {CP007}5.37.17 ENSURE Turbine(s)

Gland Seal Steam Pressure(s)

AND Gland Steam spillover pressure is being maintained lAW OPOP02-GS-0001, Turbine Gland Seal Steam System.5.37.18 VERIFY the "UNIT TRIP" light is LIT. {CP007}5.37.19 DEPRESS the "GOV VLV LIMIT CONT" (lower) "\f" pushbutton until the valve position limit indication is 0%. {CP007}5.37.20 IF the Secondary is NOT being Shutdown, THEN COMMENCE Main Turbine Parameter Logsheet per 000I01-OL-0002 until the Main Generator output breaker is closed.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 37 of 102 Plant Shutdown From 100% to Hot Standby I Initials NOTE Reactor Shutdown MAY continue while waiting for turbine to coastdown in Steps 5.37.21 -Step 5.37.29.5.37.21 Determine EHC pump requirements and SECURE the un-necessary EHC pumps as follows: CAUTION WHEN the Main Turbine and Main Feedpurnp Turbines are NOT latched, THEN the EHC fluid does NOT return through the EHC coolers. Elevated temperatures are expected and allowed up to 140'F. Time in this configuration should be minimized.

5.37.21.1 IF EHC is NO longer required (i.e., Main Turbine and Main Feedpump Turbines secured and NOT to be re-started soon, as determined by Unit Supervisor/Shift Manager), THEN REMOVE the EHC System From Service per OPOP02-EH-0001 Main Turbine Electro-Hydraulic Control System, "Removing EHC System From Service".5.37.21.2 IF two (2) EHC pumps are running, THEN STOP one (1) EHC pump AND PLACE control switch in "PTL".{CP007}a. "EHC SPLY PUMP 11(21)" b. "EHC SPLY PUMP 12(22)" This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0006

-Rev. 54 Page 38 of12 I Plant Shutdown From 100% to Hot Standby Initials NOTE (Unit 2, 2RE16 Only)CR 13-5155 has identified and issue with Turbine Bearing Lift Pump which may prevent bearing lift being available for Main Turbine rotation on tile turning gear.Temporary Modification T2-13-5155-6 lowers the Bearing Lift pressure permissive to Main Turbine turning gear from 850 psig to 600 psig (N2LTPSH6211 "TURBINE GENERATOR BEARING LIFT PUMP DISCHARGE HIGH PRESSURE SWITCH").Temporary Modification T2-13-5155-6 installs a temporary gauge at test TEE for N2LTPSH6211 "TURBINE GENERATOR BEARING LIFT PUMP DISCHARGE HIGH PRESSURE SWITCH" (TGB 83ft Between Bearing 5 and 6 near the door where terminal box N2TIMBTNLO121 is located is underneath the catwalk).5.3 7.22 ENSURE the "TURB BRG LIFT PUMP" STARTS as Main Turbine speed reaches 600 rpm. {CP007}5.37.23 (Unit 2, 2RE16 Only) ENSURE temporary gauge isolation at test TEE for N2LTPSH62 11 "TURBINE GENERATOR BEARING LIFT PUMP DISCHARGE HIGH PRESSURE SWITCH" is OPEN.5.37.24 (Unit 2, 2RE16 Only) CHECK Bearing Lift pressure is greater than 600 psig on temporary gauge at test TEE for N2LTPSH62 11"TURBINE GENERATOR BEARING LIFT PUMP DISCHARGE HIGH PRESSURE SWITCH".5.37.25 (Unit 2, 2RE16 Only) MONITOR Bearing Lift pressure on temporary gauge at test TEE for N2LTPSH62 1I until Engineering is satisfied that Bearing Lift pressure will maintain above 600 psig.5.37.26 (Unit 2, 2RE16 Only) IF "TURB BRG LIFT PUMP" OR Main Turbine turning gear, is NOT available (or becomes unavailable), THEN PERFORM Addendum 15, (Unit 2, 2RE16 Only) Turbine Shutdown without Bearing Lift Pump.5.37.27 ENSURE Main Turbine turning gear in operation after achieving zero speed indication.

{CP007}5.37.27.1 ENSURE Main Turbine turning gear ENGAGES AND STARTS automatically.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 39 of 102: Plant Shutdown From 100% to Hot Standby I Initials NOTE Lineup 1 isolates cooling water to the exciter while shutdown to prevent condensation from forming. Lineup 1, Turbine Generator Systems Cooling Water Lineup, may be omitted with Engineering concurrence.

5.37.28 WHEN the Main Turbine is on the turning gear, THEN VERIFY"TV-6000", temperature control valve for CL-ACW to the exciter, operates correctly (i.e., goes closed when temperature is below setpoint).

{55 ft TGB SE Corner of WTA Regulating Cubicles}5.37.29 WHEN the Main Turbine is on the turning gear, THEN PERFORM Lineup 1, Turbine Generator Systems Cooling Water Lineup.(Reference 2.19)5.38 ENSURE a pre-job briefing on Transferring SG feed to the AFW nozzles performed IAW Conduct of Operations Chapter 2. {REFER TO Addendum 7, Transferring SG feed to the AFW nozzles.}5.39 WHILE maintaining Reactor Power between 6% and 8%, TRANSFER the IA(2A), 1B(2B), IC(2C) and ID(2D) SG feed from the Full Power Feed Nozzles to the AFW Nozzles. {REFER TO Addendum 7, Transferring SG feed to the AFW nozzles.}5.40 ENSURE SG MFRV(s) manual isolation valves are CLOSED: "1(2) FW-0068 SG 1A(2A) FW REG VLV ISOL VLV"{45 ft TGB S Side}"1(2) FW-0042 SG 1B(2B) FW REG VLV ISOL VLV"{45 ft TGB SE Corner}" "1(2) FW-0093 SG 1C(2C) FW REG VLV ISOL VLV"{45 ft TGB SE Comer}"1(2) FW-0109 SG ID(2D) FW REG VLV ISOL VLV"{45 ft TGB S Side}5.41 IF desired at this time to warm up the Auxiliary Steam header or transfer auxiliary steam loads, THEN PERFORM the applicable portions of Step 6.3.5.42 IF the Auxiliary Steam header or transfer of auxiliary steam was performed early, THEN ENSURE the Deaerator Steam valves handswitch "SPLY VLVs PV-7174" and "PV-7174A" on CP008 is in the Close position per Step 6.12.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 40 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE Coordinate with Chemistry prior to changing SGBD flowrates.

5.43 MINIMIZE secondary system heat loads (i.e. throttle steam flow to Deaerator, isolate drain lines, and minimize SGBD etc.). (Reference 2.8)5.44 IF directed by Operations Management to perform a Fast Reactor Shutdown, THEN PERFORM Addendum 4, Fast Reactor Shutdown.NOTE Mode 2 is entered when 2 of the 4 RCS loop Delta-T indications are LESS THAN OR EQUAL to 5%.5.45 LOWER Rx Pwr to between 4 and 5% with Control Rods in MANUAL.{CP005}5.46 RECORD the Unit, time and date the plant entered Mode 2.Unit: Time: Date: This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 j Rev. 54 Page 41 of 102 Plant Shutdown From 100% to Hot Standby Initials 6.0 Mode 2 Descent CAUTION WHEN Rx Pwr is reduced below 2% AND an unexpected or uncontrolled cooldown occurs, THEN an attempt SHALL be made to stop the cooldown by isolating secondary heat loads, and NOT by diluting and/or withdrawing control rods.WHEN Rx Pwr is between 2% -5% AND an unexpected or uncontrolled cooldown occurs, THEN an attempt SHOULD be made to stop the cooldown by isolating secondary heat loads, limited small reactivity changes (diluting or withdrawing control rods) permitted as directed by the Unit Supervisor/Shift Manager. NO large positive reactivity chan2es permitted.(Modes I and 2) IF Tavg lowers to < 561'F, THEN COMPLY with Technical Specification 3.1.1.4.IF Tavg cannot be restored to greater than 561'F within 15 minutes, THEN manually TRIP the reactor and GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

6.1 WHEN Rx Pwr is reduced to Mode 2, THEN ADJUST steam dump pressure controller "HDR PRESS CONT PK-0557" to maintain the RCS at no-load Tavg (567°F), a pot setting of approximately 8.46. {CP007}6.2 LOWER Rx Pwr to between 3% and 4% with control rods in MANUAL.{CP005}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 42 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE The performance of Step 6.3 may be performed early (at any Rx Power), delayed or deleted as directed by the Unit Supervisor/Shift Manager, dependent on plant conditions (e.g. high decay heat loads or rapid return to power).6.3 TRANSFER auxiliary steam loads as follows: CAUTION 0 Transferring auxiliary steam loads from the unit performing the shutdown to the operating unit will cause a slight power rise in the operating unit due to the additional steam loads being transferred.

Care SHALL be exercised in warming and pressurizing cold steam piping to prevent hydraulic transients.

NOTE* Step 6.3.1 is written for establishing auxiliary steam with a cold header. IF the auxiliary steam header is already warmed up, THEN only those steps necessary to supply auxiliary steam need to be performed and the other steps may be marked N/A.* Closing the Deaerator Steam valves "SPLY VLVs PV-7174" and "PV-7174A"{CP008} in Step 6.12, should be delayed until < 20% Rx Pwr.6.3.1 IF Unit 2(1) will supply Unit 1(2) auxiliary steam, THEN PERFORM the following:

6.3.1.1 OBTAIN permission for Unit 2(1) Control Room to align the Auxiliary Steam System to supply Unit 1(2).6.3.1.2 ENSURE "0-AS-MOV-0202 AS ISOL TO UNITS I AND 2" closed. {N of AUX Boilers SE Yard}6.3.1.3 Slowly OPEN "2-AS-0396 (1-AS-0395)

UNIT 2(1)AUX STEAM DISTRIBUTION BLOCK VLV BYPASS" to warm up the header. {SW of LC 12L}6.3.1.4 WHEN the auxiliary steam header has been warmed, THEN OPEN "2-AS-MOV-0301 (1-AS-MOV-0128)

UNIT 2(1) AUX STEAM DISTRIBUTION BLOCK VLV". {SW ofLC 12L}This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0006 Rev. 54 Page 43 of 102 Plant Shutdown From 100% to Hot Standby Initials 6.3.1.5 Slowly OPEN "I-AS-0395 MOV-0128 (2-AS-0396 MOV-0301)

BYPASS". {SW of LC 12L}_6.3.1.6 WHEN the auxiliary steam header has been warmed, THEN OPEN "I-AS-MOV-0128 (2-AS-MOV-0301)

UNIT 1(2) AUX STEAM DISTRIBUTION BLOCK VLV". {SW ofLC 12L}6.3.1.7 CLOSE "1-AS-0395 (2-AS-0396 MOV-0301)BYPASS". {SW of LC 12L}6.3.1.8 CLOSE "2-AS-0396(1-AS-0395)

UNIT 2(1) AUX STEAM DISTRIBUTION BLOCK VALVE BYPASS".{SW of LC 12L}6.3.1.9 IF it is desired to return auxiliary steam condensate to Unit 2(1), THEN ALIGN the Condensate System per 0POP02-CD-0001, Condensate System.6.3.2 IF the auxiliary boilers will supply Unit 1(2) auxiliary steam, THEN PERFORM the following:

6.3.2.1 ENSURE auxiliary boilers are available per 0POP02-AS-0001, Auxiliary Steam System.6.3.2.2 ENSURE "0-AS-MOV-0202 AS ISOL TO UNITS 1 and 2" open. {N of AUX Boilers SE Yard}6.3.2.3 Slowly OPEN "1-AS-0395 MOV-0128 (2-AS-0396 MOV-0301)

BYPASS" to warm up the auxiliary steam header. {SW of LC 12L}6.3.2.4 WHEN the auxiliary steam header has been warmed, THEN OPEN "1 -AS-MOV-0 128 (2-AS-MOV-030 1)UNIT 1(2) AUX STEAM DISTRIBUTION BLOCK VLV". {SW ofLC 12L}6.3.2.5 CLOSE "1-AS-0395 MOV-0128(2-AS-0396 MOV-0301)

BYPASS". {SW ofLC 12L}6.3.2.6 OPEN "0-AS-MOV-0205 RETURN COND TO AUX BLR DEAER BLOCK VLV". {N of AUX Boiler}6.4 CLOSE "MS SUPPLY TO AS" reducer outlet valves.{59 ft TGB SW Corner}* "1 (2)-MS-MOV-0 186"* "1(2)-MS-MOV-0304" This procedure, when completed, SHALL be retained.

I IPOP03-ZG-0006 Rev. 54 Page 44 of 102 Plant Shutdown From 100% to Hot Standby Initials 6.5 PLACE "STM SEAL SUPPLY SEL MOV-0316/0542 in the AUX position.{CP008}6.6 ENSURE Addendum 14, Deaerator Aux Steam Supply Valve (PV-7401)Exercise, complete.6.7 PLACE "AUX STEAM TO DEAERATOR PV-7401" controller PC-7401 in the MANUAL position.

{TGB Deaerator Stand}6.8 CLOSE "AUX STEAM TO DEAERATOR PV-7401" using controller PC-7401. {TGB Deaerator Stand}6.9 PLACE "AUX STEAM TO DEAERATOR(AUX STM SPLY) PV-7401" in the MODULATE(MOD) position.

{CP008}__6.10 Slowly RAISE the setpoint on the "AUX STEAM TO DEAERATOR PV-7401" controller PC-7401 to the desired setpoint for plant condition.

{TGB Deaerator Stand}6.11 OBSERVE "AUX STEAM TO DEAERATOR PV-7401" begins supplying steam to the deaerator.

{TGB Deaerator Stand}NOTE Closing the Deaerator Steam valves "SPLY VLVs PV-7174" and "PV-7174A" {CP008} in next Step, should be delayed until < 20% Rx Pwr.6.12 CLOSE the Deaerator Steam "SPLY VLVs PV-7174" and "PV-7174A"{CP008}CAUTION During low power operations a cooldown transient may occur due to Secondary System heat loads requiring more energy than is being produced. (Reference 2.8)6.13 LOWER Rx Pwr between 1% and 2% with Control Rods in MANUAL while performing Step 6.15, if applicable, during lowering of power. {CP005}6.14 IF directed by Operations Management to perform a Fast Reactor Shutdown, THEN PERFORM Addendum 4, Fast Reactor Shutdown.This procedure., when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 45 of 102 Plant Shutdown From 100% to Hot Standby Initials 6.15 IF ANY low power feedwater regulating valve is NOT controlling SG level properly, THEN PERFORM the following:

6.15.1 PLACE each affected valve in the MANUAL position.

{CP006}0 SG IA(2A)* SG 1B(2B)* SG I C(2C)* SG ID(2D)"LOW PWR FV-7151""LOW PWR FV-7152""LOW PWR FV-7153""LOW PWR FV-7154" 6.15.2 MONITOR SG levels (both wide range and narrow range) closely and MAINTAIN SG levels about the normal control band using AUTO/MANUAL control of the feedwater regulating valves.{CP006}NOTE IF Shutdown is due to SG Tube Leakage, THEN the use of AFW for cooldown will create more contaminated water waste.Use the Startup Feed Pump when it becomes available instead of AFW to limit the amount of contaminated water produced.6.16 IF feeding the SGs with Auxiliary Feedwater is desired, THEN PERFORM the following:

6.16.1 PLACE a minimum of two (2) AFW Pumps in service per OPOP02-AF-0001, Auxiliary Feedwater.

6.16.2 MONITOR SG levels closely until flow has been transferred onl all four SGs. {CP006}6.16.3 Perform Steps 6.16.4 and 6.16.5 simultaneously on one (1) SG at a time for a smooth transition to AFW.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 46 of 102 Plant Shutdown From 100% to Hot Standby Initials 6.16.4 REDUCE main feedwater flow to the SG by slowly closing the low power feedwater regulating valve. {CP006}* SG IA(2A) "LOW PWR FV-7151"* SG 1B(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153"* SG 1D(2D) "LOW PWR FV-7154" 6.16.5 ESTABLISH AFW flow to the SG per OPOP02-AF-0001, Auxiliary Feedwater.

0 SG 1A(2A)0 SG IB(2B)0 SG IC(2C)* SG ID(2D)6.16.6 PERFORM Steps 6.16.4 and 6.16.5 for the other three SGs.6.16.7 CLOSE preheater bypass valves. {CP006}* SG 1A(2A) "PREHTR BYPASS FV-7189"* SG IB(2B) "PREHTR BYPASS FV-7190"* SG IC(2C) "PREHTR BYPASS FV-7191" 0 SG ID(2D) "PREHTR BYPASS FV-7192" 6.16.8 IF the "S/U SGFP 14(24)" is NOT operating, THEN PLACE "S/U SGFP 14(24)" handswitch in the PULL TO LOCK position.{CP006}6.16.9 IF the "S.G.F.P Turbine" is operating, THEN SECURE the running S.G.F.P. Turbine per OPOP02-FW-0002, S.G.F.P. Turbine.6.16.10 IF the "S/U SGFP 14(24)" is operating, THEN SECURE the "S/U SGFP 14(24)" per OPOP02-FW-0001, Main Feedwater.

This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0006 Rev. 54 Page 47 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE* The preferred method for controlling SG steaming rates while feeding with AFW is with the SG PORVs.* (Mode 2) In order to support Startup and Shutdown activities the SG PORVs MAY be operated in MANUAL OR in automatic, to Maintain the secondary side pressure at or below an indicated steam generator pressure of 1225 psig.* IF SG PORVs are required to be controlled in MANUAL operation, OR in automatic operation with reduced setpoints, THEN an OAS entry is required to ENSURE compliance with Technical Specifications 3.3.5.1, and 3.7.1.6.6.16.11 CONTROL RCS temperature in desired band with SG PORVs.{CPO06)* SG 1A(2A)* SG 1B(2B)* SG IC(2C)* SG ID(2D)"PORV PV-7411""PORV PV-742 1""PORV PV-743 1""PORV PV-7441 " NOTE SG steaming rates may be controlled by manually operating SG PORVs.The preferred method for closing the MSIVs is to utilize the associated A and B train Safety Grade Handswitch(s).

{CP006}6.17 IF a secondary plant cooldown is planned without cooling down the primary OR it is desired to cooldown the primary using SG PORVs, THEN CLOSE MSIVs.* SG 1A(2A)* SG IB(2B)* SG I C(2C)* SG 1D(2D)"MSIV FSV-7414""MSIV FSV-7424""MSIV FSV-7434""MSIV FSV-7444" 6.17.1 MONITOR Main Steam Header Pressure using "PI-0557, MS HDR PRESSURE" indicator to ensure MSIVs close fully.{CP007)This procedure.

when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 48 of 102 Plant Shutdown From 100% to Hot Standby ____Initials NOTE Secondary plant shutdown will normally be completed per OPOPO3-ZG-0007, Plant Cooldown, after the Reactor has been shutdown.

IF a secondary plant cooldown is desired with the Reactor critical, THEN the Secondary Plant Shutdown section of OPOPO3-ZG-0007, Plant Cooldown, may be performed with the exception of opening the Reactor Trip Breakers.Adjust Source Range audio multiplier, as required, to monitor Source Range audible count rate.7.0 Reactor Shutdown 7.1 NOTIFY Chemistry to sample the RCS for boron concentration and record the results of the sample.RCS boron ppm 7.2 NOTIFY I&C to be on standby to perform PM for Intermediate Range NI Compensating Voltage Adjustment.

NOTE* Pressurizer boron concentration should be maintained within 50 ppm of RCS concentration.

• Equalize the boron concentration, as necessary, by energizing at least two (2) sets of Pressurizer Heaters to force additional spray.* Efforts to maintain Pressurizer boron concentration within 50 ppm of RCS concentration are secondary to safe Pressurizer Pressure control.7.3 IF directed by the Unit Supervisor/Shift Manager and Plant conditions can support it, THEN the Backup Heaters may be energized as necessary to aid in PZR turnover flow.7.4 IF Tavg lowers below 5617F at any time during the Reactor Shutdown, THEN TRIP the Reactor and GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

{CP005}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Pe 49 of 102 Plant Shutdown From 100% to Hot Standby Initials 7.5 IF Shutting down to Mode 5 OR 6, THEN CV-0218 Boration per Addendum 10, CV-0218 Boration While Inserting Rods, may be used to borate the RCS once Control Rod Insertion has commenced at the discretion of the Shift Manager.7.6 INSERT Control Bank Rods until both intermediate range readings go below 1.0 E-10 amps. {CP005}7.7 VERIFY P-6 interlock clears as indicated by the following lights being off:* "P6 SOURCE RANGE RX TRIP BLOCK PERM" PERM LAMPBOX {5M024}0 "SOURCE RANGE RX TRIP BLOCKED TRAIN R" STATUS LAMPBOX {5M023}* "SOURCE RANGE RX TRIP BLOCKED TRAIN S" STATUS LAMPBOX {5M023}* INTERMEDIATE RANGE CHI "P6" BISTABLE STATUS MONITORING

{5M005}* INTERMEDIATE RANGE CH2 "P6" BISTABLE STATUS MONITORING

{5M005}7.8 IF P-6 does NOT clear, THEN PERFORM the following:

7.8.1 LOWER

Reactor Power to less than 5.0 E-1 I amps.7.8.2 RESET both SR channels by momentarily placing each of the following switches to the UNBLOCKED position:

{CP005}* "SR TRN R"* "SR TRN S" 7.9 WHEN Source Range Nis are energized, THEN ENSURE audible Count Rate available.

7.10 IF audible Count Rate NOT available from Source Range Nis, THEN PERFORM recommendations per OPGP03-ZO-0042, Reactivity Management Program. "Addendum 3, Reactivity Management Elements -Reactor Shutdown".

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 m Rev. 54 Page 50 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE Shutdown Margin Verification may be performed in parallel with Steps 7.11 through 7.16.2.3 Shutdown Bank Insertion.

7.11 INSERT all Control Banks until Control Bank Rods are at zero (0) steps.{CP005}7.12 MAINTAIN at least the minimum shutdown margin as calculated per OPSP1O-ZG-0003, Shutdown Margin Verification Modes 3, 4 and 5.7.13 RECORD the Unit, Time and Date plant entered Mode 3.Unit: Time: Date: 7.14 NOTIFY System Engineering Department to perform OPGP03-ZE-0033, RCS Pressure Boundary Inspection for Boric Acid Leaks, as required by the procedure.

7.15 IF applicable, THEN PERFORM OPSPIO-DM-0003, Automatic Multiple Rod Drop Time Measurement, AND N/A Steps 7.16 through 7.16.2.3.This procedure, when completed, SHALL be retained.

oPOP03-ZG-0006 Rev. 54 Page 51 of 102 Plant Shutdown From 100% to Hot Standby Initials NOTE Rod Control Urgent Alarm and associated control room annunciator "ROD CONT URGENT ALARM" (5M03-B5) will be received when the reactor trip breakers are opened.O t t CAUTION SOpening tie Reactor Trip Breakers will cause a Feedwater Isolation 7.16 INSERT Shutdown Banks by one (1) of the following methods, RECORD N/A for the method NOT selected:

{CP005}7.16.1 CRDM Method: 7.16.1.1 PLACE the Control "ROD BANK SELECTOR SW" in the SB-A position.7.16.1.2 INSERT Shutdown Bank A Rods to zero (0) steps.7.16.1.3 PLACE the Control "ROD BANK SELECTOR SW" in the SB-B position.7.16.1.4 INSERT Shutdown Bank B Rods to zero (0) steps.7.16.1.5 PLACE the Control "ROD BANK SELECTOR SW" in the SB-C position.7.16.1.6 INSERT Shutdown Bank C Rods to zero (0) steps.7.16.1.7 PLACE the Control "ROD BANK SELECTOR SW" in the SB-D position.7.16.1.8 INSERT Shutdown Bank D Rods to zero (0) steps.7.16.1.9 PLACE the Control "ROD BANK SELECTOR SW" in the SB-E position.7.16.1.10 INSERT Shutdown Bank E Rods to zero (0) steps.7.16.1.11 ENSURE "Reactor Trip Log" is in INHIBIT per Addendum 11, Inhibit Reactor Trip Log.7.16.1.12 IF it is desired, THEN OPEN Reactor Trip Breakers.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 52 of 102 Plant Shutdown From 100% to Hot Standby Initials 7.16.2 Reactor Trip Breaker Method: 7.16.2.1 ENSURE "Reactor Trip Log" is in INHIBIT per Addendum 11, Inhibit Reactor Trip Log.7.16.2.2 OPEN the Reactor Trip Breakers.7.16.2.3 VERIFY ALL rods on the bottom.7.17 IF it is desired to reset the Feedwater Isolation Signal, THEN PERFORM Addendum 9, Feedwater Isolation Signal Reset and Establishing SG Feed.7.18 MAINTAIN SGs narrow range levels between 55% and 75% as follows: 7.18.1 IF AFW is the SG feedwater source, THEN CONTINUE operation per OPOP02-AF-0001, Auxiliary Feedwater.

7.18.2 IF SU SGFP 14(24) is the SG feedwater source, THEN: 7.18.2.1 ENSURE low power feedwater regulating valves in MANUAL. {CP006}* SG IA(2A) "LOW PWR FV-715 1"* SG 1B(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153"* SG 1D(2D) "LOW PWR FV-7154" 7.18.2..2 ENSURE preheater bypass valves are OPEN. {CP006}0 SG 1 A(2A) "PREHTR BYPASS FV-7189"* SG IB(2B) "PREHTR BYPASS FV-7190"* SG IC(2C) "PREHTR BYPASS FV-7191"* SG 1D(2D) "PREHTR BYPASS FV-7192" 7.18.2.3 THROTTLE low power feedwater regulating valves to maintain SG levels. {CP006}* SG IA(2A) "LOW PWR FV-715 1"* SG 1B(2B) "LOW PWR FV-7152"* SG IC(2C) "LOW PWR FV-7153"* SG 1D (2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 53 of 102 Plant Shutdown From 100% to Hot Standby Initials CAUTION Reactor Trip Breakers SHALL be OPEN prior to Securing the Rod Drive MG Sets. Securing Rod Drive MG Sets with Reactor Trip Breakers Closed will Trip Open the Reactor Trip Breakers.

Opening the Reactor Trip Breakers will cause a Feedwater Isolation.

NOTE Rod Drive Motor Generator MAY be required to be in operation if a transition to OPOP03-ZG-0007, Plant Cooldown is planned.7.19 IF desired by the Unit Supervisor/Shift Manager, THEN REMOVE the Rod Drive Motor Generator sets from operation per OPOP02-RS-0001, Rod Control.7.20 IF the Secondary is NOT being Shutdown, THEN COMMENCE Main Turbine Parameter Logsheet per 000101-OL-0002 until the Main Generator output breaker is closed.7.21 IF a transition to OPOP03-ZG-0007, Plant Cooldown is planned, THEN N/A the following steps, OTHERWISE PERFORM the following:

7.2 1.1 WHEN source range counts have been stabilized, THEN NOTIFY i&C to calibrate the "HI FLUX AT SHUTDOWN" alarm to five times the stabilized count rate.7.21.2 WHEN I&C has completed calibration, THEN PLACE the "HIGH FLUX AT SHUTDOWN" alarm switch in normal.7.22 IF the CVCS Demineralizers are Bypassed, THEN EVALUATE placing them in service per OPOP02-CV-0004, Chemical and Volume Control System Subsystem.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 54 of 102 Plant Shutdown From 100% to Hot Standby I 8.0 Records Review 8.1 Comments: 8.2 REVIEW procedure package to ensure all applicable sections are completed as required.Unit Supervisor/Shift Manager Date This procedure, when completed, SHALL be retained.

0 OP03-ZG-0006 Rev. 54 Page 55 of 102 Plant Shutdown From 100% to Hot Standby 9.0 Support Documents 9.1 Addendum 1, RCS Degassification

9.2 Addendum

2, Main Turbine Operating Guidelines

9.3 Addendum

3, Post Reactor Trip Guideline 9.4 Addendum 4, Fast Reactor Shutdown 9.5 Addendum 5, Percent Power vs Program RCS Tavg 9.6 Addendum 6, Failed or leaking Stearn Dump Valves 9.7 Addendum 7, Transferring SG feed to the AFW nozzles 9.8 Addendum 8, Transferring feed from MFRV to LPRV 9.9 Addendum 9, Feedwater Isolation Signal Reset and Establishing SG Feed 9.10 Addendum 10, CV-0218 Boration While Inserting Rods 9.11 Addendum 11, Inhibit Reactor Trip Log 9.12 Addendum 12, AFD Penalty Point Evaluation 9.13 Addendum 13, Low Power Feedwater Regulating Valves Exercise 9.14 Addendum 14, Deaerator Aux Steam Supply Valve (PV-7401)

Exercise 9.15 Addendum 15, (Unit 2, 2RE16 Only) Turbine Shutdown without Bearing Lift Pump 9.16 Lineup 1, Turbine Generator Systems Cooling Water This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 56 of 102 Plant Shutdown From 100% to Hot Standby Addendum 1 RCS Degassification Page 1 of 2 Initials NOTE IF Chemical degassing, THEN perform this Addendum only when recommended by Chemistry 1.0 ESTABLISH nitrogen cover gas on the Volume Control Tank (VCT) as follows:{CVCS Chemical Mixing Tank Room}1.1 TAG CLOSED "1(2)-CV-0 178 HYDROGEN SUPPLY TO VCT ISOL" with an Equipment Clearance Order.1.2 OPEN "1(2)-CV-0181 NITROGEN SUPPLY TO VCT ISOL".1.3 OPEN "1 (2)-NL-0033 UNIT 1(2) VOLUME CONTROL TANK N2 SUPPLY VLV". {51 ft Cubicle 335 Access from 41 ft}2.0 INITIATE RCS degassing as follows: 2.1 NOTIFY Radwaste Operator(s) of pending degassification so they can raise the gas flow as allowed per OPOP02-WG-0001, Gaseous Waste Processing System Operations.

2.2 NOTIFY

Health Physics of pending degassification to the Gaseous Waste Processing System.NOTE The effluent from the primary sample sink is directed back to VCT during pressurizer vapor space degassing.

2.3 NOTIFY

Chemistry to initiate degassing the pressurizer vapor space.2.4 OPEN "VCT VENT PCV-01 15". {CP004}2.5 PLACE "DIVERT LCV-0I 12A" in the VCT position.

{CP004}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 57 of 102 Plant Shutdown From 100% to Hot Standby Addendum 1I RCS Degassification Page 2 of 2 Initials CAUTION* To prevent lifting the VCT relief valve, VCT pressure SHALL NOT be allowed to exceed 65 psig during fill.To prevent RCS boron dilution, RCS makeup flow boron concentration SHALL be at greater than or equal to RCS boron concentration.

2.6 RAISE

VCT level to between 90 and 95% using "RC M/U CONT". {CP004}2.7 VERIFY VCT level between 90 and 95% on "VCT LEVEL LI-01 12".(CP004}2.8 MAINTAIN VCT pressure between 15 and 30 psig while raising level.2.9 VERIFY VCT pressure between 15 and 30 psig on "VCT PRESS PI-0115".2.10 PLACE "RC M/U CONT" in the STOP position.

{CP004}2.11 ALLOW VCT pressure to decay to the new minimum value (approximately 15 to 20 psig) while maintaining VCT level between 90 and 95%. {CP004_NOTE IF the LWPS does NOT have the capacity to receive the water from letdown or it is desired to minimize water usage, THEN it is permissible to adjust charging and letdown to obtain the desired VCT levels.2.12 PLACE "DIVERT LCV-01 12A" in the AUTO position.

{CP004}2.13 VERIFY "1(2)-CV-PV-3 111 NITROGEN TO VCT PV" maintains VCT pressure greater than or equal to 15 psig. {CP004}2.14 LOWER VCT level to 30% by placing "DIVERT LCV-01 12A" in the RHT position.

{CP004}2.15 PLACE "DIVERT LCV-01 12A" in the AUTO position.

{CP004}2.16 IF additional RCS degassification is required, THEN RETURN TO Step 2.5 of this addendum and perform Steps 2.5 through 2.15 as necessary to complete RCS degassification.

2.17 RETURN "RC M/U CONT" to the configuration desired by the Unit Supervisor.

{CP004}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 58 of 102 Plant Shutdown From 100% to Hot Standby j Addendum 2 Main Turbine Operating Guidelines Page 1 of I Exhaust Pressure Limitations Maximum Permissible Condensing Pressure 20 21 22 (L.0~0 23 24 25 25.ý26 27 28 29 30 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 Load -MW CD14035 U1(2/2/12)

This procedure, when completed, SHALL be retained.

OPOPo3-ZG-0006 Rev. 54 Page 59 of 102 Plant Shutdown From 100% to Hot Standby IAddendu 3 Post Reactor Trip Guideline Page I of I Add m 3 P NOTE The following is an aid to the Unit Supervisor/Shift Manager in determining the necessary steps to be performed following a normal reactor trip. See Notes and Precautions Step 4.2 for additional guidance.Initials 1.0 The following Steps are recommended to be performed during a Normal Post Reactor Trip plant stabilization:

• Step 5.1* Step 5.4* Step 5.13* Step 5.20.2* Step 5.20.4* Step 5.20.6* Step 5.20.10 through Step 5.20.14.1* Step 5.28 through Step 5.30* Step 5.34 through Step 5.36* Step 5.37.3 through Step 5.37.6* Step 5.37.13 through Step 5.37.29* Step 6.3.1* Step 6.4 through Step 6.12* Step 7.1* Step 7.7* Step 7.9 through Step 7.10* Step 7.13 through Step 7.14* Step 7.16.1.11 or 7.16.2.1* Step 7.18 through end of procedure as applicable This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 60 of 102 Plant Shutdown From 100% to Hot Standby E Addendum 4 FsRector Shutdown Page I of 10 Initials 1.0 To conduct a Fast Reactor Shutdown from MODE 1 or 2 to HOT STANDBY (MODE 3) PERFORM the following:

1.1 OBTAIN

permission to conduct Fast Reactor Shutdown.Shift Manager Operations Manager NOTE* Notification to NRC is NOT required for the manual actuation of the reactor protection system (RPS) for a pre-planned sequence during testing or reactor operation.

{NUREG-1022, Rev. 2 page 45, 1OCFR50.72( b)(3)( iv)(A), IOCFR50.72 (b)(2)( iv)(B), 1 OCFR50.73(a)(2)(iv) (A)(1)}* IF unexpected responses or evolutions occur after the reactor trip breakers are opened, THEN the Unit Supervisor/Shift Manager will have to determine if NRC notification is required for the unexpected conditions.

1.2 ENSURE

pre-job briefing performed prior IAW Conduct of Operations Chapter 2.Talking Points:* Expected RCS temperature response* Expected PZR Level response* Expected PZR Pressure response* Emergency Boration criteria* Entry into OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION* Personnel Responsibilities

• Notifications to NRC* Just In Time Training 1.3 ENSURE the transfer of auxiliary steam loads has been completed lAW the applicable portions of this procedure main body, Step 6.3.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 61 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 4 Fast Reactor Shutdown Page 2 of 10 NOTE* Secondary plant shutdown will normally be completed per OPOP03-ZG-0007, Plant Cooldown, after the Reactor has been shutdown.* Deenergizing the CRDM Coils, will prevent thermal lockup of the RCCAs.* IF an unexpected response is received, THEN GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

• Rod Control Urgent Alarm and associated control room annunciator "ROD CONT URGENT ALARM" (5M03-B5) will be received when the reactor trip breakers are opened.CAUTION* Emergency boration per OPOP04-CV-0003, Emergency Boration.

must be initiated, IF 2 or more control rods are NOT fully inserted.

Full insertion by DRPI position and "Rod Bottom" lights must be verified Opening the Reactor Trip Breakers will cause a Feedwater isolation (Low Tave with P-4 FW Isol).S/G levels must be monitored closely.* Opening the Reactor Trip breakers as described in the following step does NOT require entry into OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION unless the Unit Supervisor/Shift Manager determines the results of opening the Reactor Trip breakers is NOT as expected.* IF Tavg lowers below 561'F at any time during this Section (Fast Reactor Shutdown), THEN ENSURE the Reactor is TRIPPED and GO TO OPOP05-EO-EOOO, REACTOR TRIP OR SAFETY INJECTION.

{CP005}Initials 1.4 OPEN the Reactor Trip Breakers to insert all Control and Shutdown rods.1.5 VERIFY the following:

• All Control Bank and Shutdown Bank rod at bottom lights are lit." All Reactor Trip and Bypass breakers are OPEN." Neutron flux is Lowering.This procedure, when completed, SHALL be retained.

7 OPOP03-ZG-0006 Rev. 54 Page 62 of 102 Plant Shutdown From 100% to Hot Standby Addendum 4 Fast Reactor Shutdown Page 3 of 10 Initials 1.6 ADJUST steam dump pressure controller "HDR PRESS CONT PK-0557" to maintain the RCS at no-load Tavg (567 'F), a pot setting of approximately 8.46.{CP007}1.7 IF it is desired to feed the S/G's with Main Feedwater, THEN RESET the Feedwater Isolation Signal by PERFORMING the following:

1.7.1 ENSURE

both "LOW PWR" and "NORM" feedwater regulating valve controllers in manual with a zero demand. {CP006}* SG IA(2A)* SG 1A(2A)* SG S B(2B)* SGc 1B(2B)S SG I C(2C)SG S C(2C)SG S D(2D)SG I D(2D)"LOW PWR FV-7151" (manual & "0" demand)"NORM FCV-055 1" (manual & "0" demand)"LOW PWR FV-7152" (manual & "0" demand)"NORM FCV-0552" (manual & "0" demand)"LOW PWR FV-7153" (manual & "0" demand)"NORM FCV-0553" (manual & "0" demand)"LOW PWR FV-7154" (manual & "0" demand)"NORM FCV-0554" (manual & "0" demand)This procedure, when completed, SHALL be retained.

oPoPo3-ZG-oo06 Rev. 54 Page 63 of 102]Plant Shutdown From 100% to Hot Standby E Addendum 4 Fast Reactor Shutdown Page 4 of 10 Initials 1.7.2 ENSURE the following valves are CLOSED and their associated handswitch are in the CLOSE position:

{CP006}: SG 1A(2A)SG IA(2A)SG 1A(2A)SG 1B(2B)SG 1B(2B)SG 1B(2B)SG 1C(2C)SG 1C(2C)Sc 1C(2C)S 1D(2D)GSO 1D(2D)SG 1D(2D)"PREHTR BYPASS FV-7189" (HS in CLOSE & CLOSED)"FWIV FV-7141" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7148A" (HS in CLOSE & CLOSED)"PREHTR BYPASS FV-7190" (HS in CLOSE & CLOSED)"FWIV FV-7142" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7147A" (HS in CLOSE & CLOSED)"PREHTR BYPASS FV-7191" (HS in CLOSE & CLOSED)"FWIV FV-7143" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7146A" (HS in CLOSE & CLOSED)"PREHTR BYPASS FV-7192" (HS in CLOSE & CLOSED)"FWIV FV-7144" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7145A" (HS in CLOSE & CLOSED)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 I Rev. 54 Page 64 of 102 Plant Shutdown From 100% to Hot Standby Addendum 4 Fast Reactor Shutdown Page 5 of 10 Initials 1.7.3 RESET the following "FW ISOL RESET TRAIN A" and "TRAIN B" applicable signals by depressing both train pushbuttons:

{CP006}* SI/SG HI-HI LEVEL "RESET" pushbuttons.

• TRAIN A* TRAIN B* LO TAVE "RESET (BLOCK)" pushbuttons.

• TRAIN A* TRAIN B NOTE Performance of the next step will release the "LOW PWR" feedwater regulating valves for normal operation.

• IF the Reactor Trip Breakers are closed, THEN the "NORM" feedwater regulating valves will also be released for normal operation.

0 IF the Reactor Trip Breakers are open, THEN the "NOT RESET" white lights will remain illuminated after the next step is performed.

1.7.4 DEPRESS

both TRAIN A and TRAIN B FW CONT/BYP VLVS"RESET" pushbuttons.

{CP006}* TRAIN A* TRAIN B 1.7.5 IF the Reactor Trip Breakers are closed, THEN VERIFY both TRAIN A and TRAIN B FW CONT/BYP VLVS "NOT RESET" white lights extinguished.

{CP006}* TRAIN A* TRAIN B This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 65 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 4 Fast Reactor Shutdown Page 6 of 10 Initials 1.7.6 RESET the following safety grade solenoid valves by momentary placing each SG's "FW ISOL RESET TRAIN A" and "TRAIN B" to OPEN: {CP006}0 SG lA(2A)0 SG 1A(2A)* SG 1B(2B)0 SG 1B(2B)* SG 1C(2C)* SG 1 C(2C)0 SG ID(2D)0 SG ID(2D)"FW ISOL RESET TRAIN A""FW ISOL RESET TRAIN B""FW ISOL RESET TRAIN A""FW ISOL RESET TRAIN B""FW ISOL RESET TRAIN A""FW ISOL RESET TRAIN B""FW ISOL RESET TRAIN A""FW ISOL RESET TRAIN B" 1.8 IF feeding the SGs with Auxiliary Feedwater is desired, THEN PERFORM the following:

1.8.1 ENSURE

all SG LO-LO Level AFW actuation's are reset.1.8.2 ENSURE a minimum of two (2) AFW Pumps are in service per OPOP02-AF-0001, Auxiliary Feedwater.

1.8.3 ESTABLISH

AFW flow to the SG per OPOP02-AF-0001, Auxiliary Feedwater.

• SG IA(2A)* SG S B(2B)* SG 1 C(2C)* SG ID(2D)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 66 of 102 Plant Shutdown From 100% to Hot Standby Addendum 4 Fast Reactor Shutdown Page 7 of 10 Initials 1.9 MAINTAIN SGs narrow range levels between 55 and 75% as follows: 1.9.1 IF AFW is the SG feedwater source, THEN CONTINUE operation per OPOP02-AF-0001, Auxiliary Feedwater.

1.9.2 IF SU SGFP 14(24) is the SG feedwater source, THEN ESTABLISH flow as follows: 1.9.2.1 ENSURE the low power feedwater regulating valves in the MANUAL position fully closed. {CP006}* SG 1 A(2A) "LOW PWR FV-7151"* SG IB(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153"* SG ID(2D) "LOW PWR FV-7154" 1.9.2.2 OPEN the preheater bypass valves. {CP006}* SG 1A(2A) "PREHTR BYPASS FV-7189" S SG S B(2B) "PREHTR BYPASS FV-7190"* SG IC(2C) "PREHTR BYPASS FV-7191"* SG ID(2D) "PREHTR BYPASS FV-7192" 1.9.2.3 THROTTLE low power feedwater regulating valves to maintain SG levels. {CP006}0 SG IA(2A) "LOW PWR FV-7151" 0 SG I B(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153" S SG S D (2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 67 of 102 Plant Shutdown From 100% to Hot Standby Addendum 4 Fast Reactor Shutdown Page 8 of 10 Initials 1.10 VERIFY P-6 interlock clears as indicated by the following lights being OFF:* "P6 SOURCE RANGE RX TRIP BLOCK PERM" PERM LAMPBOX {5M024}0 "SOURCE RANGE RX TRIP BLOCKED TRAIN R" STATUS LAMPBOX {5M023}* "SOURCE RANGE RX TRIP BLOCKED TRAIN S" STATUS LAMPBOX {5M023}* INTERMEDIATE RANGE CHI "P6" BISTABLE STATUS MONITORING

{5M005}* INTERMEDIATE RANGE CH2 "P6" BISTABLE STATUS MONITORING

{5M005}1.11 IF P-6 does NOT clear, THEN PERFORM the following:

1.11.1 RESET both SR channels by momentarily placing each of the following switches to the UNBLOCKED position:

{CP005}* "SR TRN R"* "SR TRN S" 1.12 ENSURE SG MFRV(s) manual isolation valves are CLOSED: (Continue with this Addendum while performing this step)* "1(2) FW-0068 SG 1A(2A) FW REG VLV ISOL VLV"{45 ft TGB S Side}* "1(2) FW-0042 SG I B(2B) FW REG VLV ISOL VLV"{45 ft TGB SE Comer}* "1(2) FW-0093 SG 1C(2C) FW REG VLV ISOL VLV" (45 ft TGB SE Comer}* "1(2) FW-0109 SG 1D(2D) FW REG VLV ISOL VLV" (45 ft TGB S Side}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 I Rev. 54 Page 68 of 102 1 Plant Shutdown From 100% to Hot Standby Addendum 4 Fast Reactor Shutdown Page 9 of 10 Initials 1.13 WHEN Source Range NIs are energized, THEN ENSURE audible Count Rate available.

1.14 IF audible Count Rate NOT available from Source Range NIs, THEN PERFORM recommendations per OPGP03-ZO-0042, Reactivity Management Program,"Addendum 3, Reactivity Management Guideline

-Reactor Shutdown".

1.15 NOTIFY Chemistry to sample the RCS for boron concentration and record the results of the sample.RCS boron ppm NOTE* Pressurizer boron concentration should be maintained within 50 ppm of RCS concentration.

• Equalize the boron concentration, as necessary, by energizing at least two (2) sets of Pressurizer Heaters to force additional spray." Efforts to maintain Pressurizer boron concentration within 50 ppm of RCS concentration are secondary to safe Pressurizer Pressure control.1.16 IF directed by the Unit Supervisor/Shift Manager and Plant conditions can support it, THEN the Backup Heaters may be energized as necessary to aid in PZR turnover flow.1.17 MAINTAIN at least the minimum shutdown margin as calculated per OPSP1O-ZG-0003, Shutdown Margin Verification Modes 3, 4 and 5.1.18 RECORD the Unit, time and date plant entered Mode 3.Unit: Time: Date: 1.19 IF a Cooldown is planned and directed by the Unit Supervisor/Shift Manager, THEN commence an RCS boration to Cold Shutdown (RCS at 68'F, Xenon-Free)

Shutdown Margin concentration.

This procedure, when colnpleted, SHALL be retained.

IOPOP03-ZG-0006 Rev. 54 Page 69 of 102 Plant Shutdown From 100% to Hot Standby E Addendum 4 Fast Reactor Shutdown Page 10 of 10 CAUTION Reactor Trip Breakers SHALL be OPEN prior to Securing the Rod Drive MG Sets. Securing Rod Drive MG Sets with Reactor Trip Breakers Closed will Trip Open the Reactor Trip Breakers.

Opening the Reactor Trip Breakers will cause a Feedwater Isolation.

NOTE Rod Drive Motor Generator MAY be required to be in operation if a transition to OPOP03-ZG-0007, Plant Cooldown is planned.Initials 1.20 IF desired by the Unit Supervisor/Shifi Manager, THEN REMOVE the Rod Drive Motor Generator sets from operation per OPOP02-RS-0001, Rod Control.1.21 IF the Secondary is NOT being Shutdown, THEN COMMENCE Main Turbine Parameter Logsheet per 00010 1-OL-0002 until the Main Generator output breaker is closed.1.22 IF a transition to OPOP03-ZG-0007, Plant Cooldown is planned, THEN N/A the following steps, OTHERWISE PERFORM the following:

1.22.1 WHEN source range counts have been stabilized, THEN NOTIFY I&C to calibrate the "HI FLUX AT SHUTDOWN" alarm to five times the stabilized count rate.1.22.2 WHEN I&C has completed calibration, THEN PLACE the "HIGH FLUX AT SHUTDOWN" alarm switch in normal.1.23 IF the CVCS Demineralizers are Bypassed, THEN EVALUATE placing them in service per 0POP02-CV-0004, Chemical and Volume Control System Subsystem.

1.24 NOTIFY System Engineering Department to perform OPGP03-ZE-0033, RCS Pressure Boundary Inspection for Boric Acid Leaks, as required by the procedure.

1.25 ENSURE "Reactor Trip Log" is in INHIBIT per Addendum 11, Inhibit Reactor Trip Log.1.26 RETURN to this procedure main body, Section 8.This procedure, when completed, SHALL be retained.

O POP03-ZG-0006 Rev. 54 [ Page 70 of 102 Plant Shutdown From 100% to Hot Standby Addendum 5 Percent Power vs Program RCS Tavg Page 1 of I Auctioneered High RCS Tavg = 592'F Percent Program] Percent Program Percent Program Percent Program Power Tavg Power Tavg Power Tavg Power Tavg 1 567.25 26 573.50 51 579.75 76 586.00 2 567.50 27 573.75 52 580.00 77 586.25 3 567.75 28 574.00 53 580.25 78 586.50 4 568.00 29 574.25 54 580.50 79 586.75 5 568.25 30 574.50 55 580.75 80 587.00 6 568.50 31 574.75 56 581.00 81 587.25 7 568.75 32 575.00 57 581.25 82 587.50 8 569.00 33 575.25 58 581.50 83 587.75 9 569.25 34 575.50 59 581.75 84 588.00 10 569.50 35 575.75 60 582.00 85 588.25 11 569.75 36 576.00 61 582.25 86 588.50 12 570.00 37 576.25 62 582.50 87 588.75 13 570.25 38 576.50 63 582.75 88 589.00 14 570.50 39 576.75 64 583.00 89 589.25 15 570.75 40 577.00 65 583.25 90 589.50 16 571.00 41 577.25 66 583.50 91 589.75 17 571.25 42 577.50 67 583.75 92 590.00 18 571.50 43 577.75 68 584.00 93 590.25 19 571.75 44 578.00 69 584.25 94 590.50 20 572.00 45 578.25 70 584.50 95 590.75 21 572.25 46 578.50 71 584.75 96 591.00 22 572.50 47 578.75 72 585.00 97 591.25 23 572.75 48 579.00 73 585.25 98 591.50 24 573.00 49 579.25 74 585.50 99 591.75 25 573.25 50 579.50 75 585.75 100 592.00 This procedure, when completed, SHALL be retained.

0POP03-ZG-0006 Rev. 54 Page 71 of 102 Plant Shutdown From 100% to Hot Standby Addendum 6 Failed or leaking Steam Dump Valves Page 1 of 3 1.0 IF Steam Dumps are operating improperly in AUTO for low power in Steam Pressure Mode, THEN PERFORM the following{CP007}:

1.1 PLACE

Steam Dump "HDR PRESS CONT" PK-0557 in MANUAL.1.2 Manually ADJUST Steam Dumps to maintain RCS Tavg within 3YF of Tref.2.0 IF a failure prevents control of ALL Steam Dumps, THEN SELECT "OFF/ RESET" on the following switches: 2.1 SELECT "OFF/ RESET" on the following switches {CP007}:* "STEAM DUMP TRAIN A -INTLK SEL"* "STEAM DUMP TRAIN B -INTLK SEL" 2.2 TRANSFER RCS cooling to the SG PORVs (Refer To Steps 6.16 and 6.17).3.0 DISPATCH operator to close at least one (1) Steam Dump isolation valve for each failed or leaking Steam Dump Valve (N/A NOT failed or leaking Steam Dump Valves): " MS-PV-7485 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 1 {TGB 55 ft East Side Of Condenser 11(21)1 o MS-0136 STM DUMP PV-7485 ISOLATION o MS-0138 STM DUMP PV-7485 ISOLATION" MS-PV-7486 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 3 {TGB 55 ft East Side Of Condenser 11(21)}o MS-0135 STM DUMP PV-7486 ISOLATION o MS-0137 STM DUMP PV-7486 ISOLATION* MS-PV-7487 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 4 {TGB 55 ft West Side Of Condenser 11(21)}o MS-0132 STM DUMP PV-7487 ISOLATION o MS-0134 STM DUMP PV-7487 ISOLATION" MS-PV-7488 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 2 {TGB 55 ft West Side Of Condenser 11(21)1 o MS-0131 STM DUMP PV-7488 ISOLATION o MS-0133 STM DUMP PV-7488 ISOLATION This procedure, when completed., SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 72 of 1021 Plant Shutdown From 100% to Hot Standby Addendum 6 Failed or leaking Steam Dump Valves Page 2 of 3* MS-PV-7489 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group I {TGB 55 ft East Side Of Condenser 12(22))o MS-0128 STM DUMP PV-7489 ISOLATION o MS-0130 STM DUMP PV-7489 ISOLATION" MS-PV-7490 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 3 {TGB 55 ft East Side Of Condenser 12(22))o MS-0127 STM DUMP PV-7490 ISOLATION o MS-0129 STM DUMP PV-7490 ISOLATION* MS-PV-7491 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 4 {TGB 55 ft West Side Of Condenser 12(22))o MS-0124 STM DUMP PV-7491 ISOLATION o MS-0126 STM DUMP PV-7491 ISOLATION" MS-PV-7492 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 2 {TGB 55 ft West Side Of Condenser 12(22))o MS-0123 STM DUMP PV-7492 ISOLATION o MS-0125 STM DUMP PV-7492 ISOLATION* MS-PV-7493 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group I {TGB 55 ft East Side Of Condenser 13(23)1 o MS-0120 STM DUMP PV-7493 ISOLATION o MS-0 122 STM DUMP PV-7493 ISOLATION" MS-PV-7494 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 3 {TGB 55 ft East Side Of Condenser 13(23))o MS-0119 STM DUMP PV-7494 ISOLATION o MS-0121 STM DUMP PV-7494 ISOLATION This procedure, when completed, SHALL be retained.

I OPOPo3-ZG-0006 Rev. 54 Page 73 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 6 Failed or leaking Steam Dump Valves Page 3 of 3* MS-PV-7495 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 4 {TGB 55 ft West Side Of Condenser 13(23)}o MS-01 16 STM DUMP PV-7495 ISOLATION o MS-01 18 STM DUMP PV-7495 ISOLATION" MS-PV-7496 MAIN STEAM DUMP TO CONDENSER PRESSURE CONTROL VALVE Group 2 {TGB 55 ft West Side Of Condenser 13(23)}o MS-0115 STM DUMP PV-7496 ISOLATION o MS-0117 STM DUMP PV-7496 ISOLATION 4.0 WHEN the failed or leaking Steam Dump Valve(s) isolated, THEN RETURN the following switches to ON {CP007}:* "STEAM DUMP TRAIN A -INTLK SEL"* "STEAM DUMP TRAIN B -INTLK SEL" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 74 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 7 Transferring SG feed to the AFW nozzles Page 1 of 3 NOTE Non-Intrusive Check Valve Testing for AF system valves SHALL be performed as determined by the Unit Supervisor/Shift Manager and System Engineer per OPEP07-ZE-0008.

• IF Addendum 4, Fast Reactor Shutdown is to be performed AND Reactor Power is maintained between 6 and 8% Reactor Power, THEN this Addendum may be marked N/A (the establishment of feedwater after the Rx Trip Breakers are opened will align feed to the AFW nozzles).* The transfer of SG feed in the following Steps should be performed for one (1) SG at a time per Operator.

Additional Operators may parallel the process by transferring their assigned SGs, if directed by the Unit Supervisor/Shift Manager.Initials 1.0 TRANSFER the SG feed to the AFW nozzle as follows: 1.1 OBSERVE the SG level(s).

{CP006}* SG IA(2A)* SG IB(2B)* SG 1 C(2C)* SG 1D(2D)This procedure., when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 75 of 102 Plant Shutdown From 100% to Hot Standby Addendum 7 Transferring SG feed to the AFW nozzles Page 2 of 3 Initials 1.2 IF SG level(s) vary by greater than 3% during the transfer, THEN PERFORM the following on the affected SG: 1.2.1 PLACE the low power feedwater regulating valve in the MAN position.

{CP006}* SG 1A(2A) "LOW PWR FV-7151"* SG 1B(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153"* SG 1D(2D) "LOW PWR FV-7154" 1.2.2 ADJUST the low power feedwater regulating valve to return the S/G level to program SG level. {CP006}S SG 1 A(2A) "LOW PWR FV-7151"* SG 1B(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153" S SG 1 D(2D) "LOW PWR FV-7154" 1.2.3 PLACE the low power feedwater regulating valve in the AUTO position.

{CP006}SG I1A(2A) "LOW PWR FV-7151" SG 1B(2B) "LOW PWR FV-7152" SG S C(2C) "LOW PWR FV-7153" SG ID(2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 76 of 102 Plant Shutdown From 100% to Hot Standby Addendum 7 Transferring SG feed to the AFW nozzles Page 3 of 3 Initials 1.3 OPEN the preheater bypass valve(s).

{CPO06}* SG 1A(2A) "PREHTR BYPASS FV-7189"* SG IB(2B) "PREHTR BYPASS FV-7190"* SG IC(2C) "PREHTR BYPASS FV-7191"* SG 1D(2D) "PREHTR BYPASS FV-7192" 1.4 IF the Feedwater Isolation Valve OPERABILITY test is required to be performed, THEN PERFORM OPSP03-FW-0001, Feedwater System Valve Operability Test, on the feedwater isolation valves. {CP006}* SG IA(2A) "FWIV FV-7141"* SG 1B(2B) "FWIV FV-7142"* SG IC(2C) "FWLV FV-7143"* SG 1D(2D) "FWIV FV-7144" CAUTION USE the Non-Safety grade switches to CLOSE the valves in the following Step. Using the Safety Grade Switches will cause a Total Feedwater Isolation.

1.5 USING

the Non-Safety grade switches, CLOSE the feedwater isolation valves.{CP006}* SG IA(2A) "FWIV FV-7141"* SG IB(2B) "FWIV FV-7142"* SG 1C(2C) "FWIV FV-7143"* SG ID(2D) "FWIV FV-7144" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 I Rev. 54 Page 77 of 102 Plant Shutdown From 100% to Hot Standby P :Addendum 8 Transferring feed from MFRV to LPRV Page 1 of 4]Initials CAUTION Iniil Failure to closely monitor SG levels may result in a Reactor Trip.1.0 ENSURE Addendum 13, Low Power Feedwater Regulating Valves Exercise, complete.2.0 IF SU SGFP 14(24) is NOT operating, THEN PERFORM the following:

2.1 ENSURE

"SGFP MASTER SPEED" controller in the MAN position.

{CP006}2.2 ENSURE SGFP pump speed approximately 5200 rpm. {CP006}NOTE Plant maneuvers SHOULD normally NOT be performed during the transfer of feedwater regulating valves.* IF a Reactor Shutdown is planned OR SG main feedwater regulating valve(s) leak by seat, THEN isolating the SG main feedwater regulating valves MAY be PERFORMED in parallel with Main Turbine load reduction/shutdown, as directed by the Unit Supervisor/Shift Manager.3.0 ENSURE Main Turbine load stabilized AND the SG narrow range level stabilized within the normal control band. {CP006}* SG 1A(2A)* SG IB(2B)* SG I C(2C)* SG ID(2D)This procedure, when completed, SHALL be retained.

0 OPOP03-ZG-0006 Rev. 54 Page 78 of 102 Plant Shutdown From 100% to Hot Standby Addendum 8 Transferring feed from MFRV to LPRV Page 2 of 4 Initials NOTE The transfer of SG feed in the following Steps should be performed for one (1) SG at a time per Operator.

Additional Operators may parallel the process by transferTing their assigned SGs, if directed by the Unit Supervisor/Shift Manager.4.0 ENSURE the selected SG low power feedwater regulating valve (LPRV) in the MAN position.

{CP006}* SG l A(2A)0 SG IB(2B)0 SG I C(2C)* SG ID(2D)"LOW PWR FV-7 I 51""LOW PWR FV-7152""LOW PWR FV-7153""LOW PWR FV-7154" 5.0 PLACE the selected SG main feedwater regulating valve (MFRV) in the MAN position.{CP006}* SG IA(2A)* SG IB(2B)0 SG I C(2C)* SG ID(2D)"NORM FCV-055 1""NORM FCV-0552""NORM FCV-0553""NORM FCV-0554" 6.0 VERIFY the selected SG feedwater flow at steady state values. {CP006}SG IA(2A)SGC 1B(2B)SG S C(2C)SG 1D(2D)flow flow_ flow flow This procedure, when completed, SHALL be retained.

0POP03-ZG-0006 Rev. 54 Page 79 of 102 Plant Shutdown From 100% to Hot Standby Addendum 8 Transferring feed from MFRV to LPRV Page 3 of 4 I Initials CAUTION WHEN transferring the selected SG feedwater regulating valves, THEN MAINTAIN the selected SG feedwater flow at approximately the steady state value observed in the previous Step. {CP006}7.0 Slowly THROTTLE OPEN the selected SG low power feedwater regulating valve (LPRV) z 1% OR until a rise is noticed in the feedwater flow. {CP006}SG 1A(2A)SG 1B(2B)SG IC(2C)SG 1D(2D)"LOW PWR FV-715 1""LOW PWR FV-7152""LOW PWR FV-7153""LOW PWR FV-7154" NOTE IF the selected SG main feedwater regulating valve (MFRV) seat leakage is excessive, THEN the Unit Supervisor/Shift Manager may direct the selected FW REG VLV ISOL VLV(s) to be CLOSED while performing the transfer.8.0 Slowly THROTTLE CLOSED the selected SG main feedwater regulating valve until SG feedwater flow has returned to the steady state values observed earlier. {CP006}* SG IA(2A)0 SG IB(2B)* SG I C(2C)* SG ID(2D)"NORM FCV-055 1""NORM FCV-0552""NORM FCV-0553""NORM FCV-0554" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 80 of 102 Plant Shutdown From 100% to Hot Standby Addendum 8 Transferring feed from MFRV to LPRV Page 4 of 4 Initials 9.0 PERFORM Steps 7 and 8, UNTIL the selected SG main feedwater regulating valve (MFRV) is CLOSED. {CP006}* SG 1A(2A)* SG IB(2B)* SG 1 C(2C)* SG ID(2D)10.0 PLACE the selected SG low power feedwater regulating valve (LPRV) in the AUTO position.

{CP006}* SG IA(2A) "LOW PWR FV-715 1"& SG IB(2B) "LOW PWR FV-7152"* SG IC(2C) "LOW PWR FV-7153"* SG I D(2D) "LOW PWR FV-7154" 11.0 MONITOR the selected SG low power feedwater regulating valve(s) for proper operation.

{CP006}* SG 1A(2A)* SG 1B(2B)* ScG C(2C)S SG 11D(2D)12.0 IF required, THEN RETURN to Addendum LPRVs."LOW PWR FV-7151""LOW PWR FV-7152""LOW PWR FV-7153""LOW PWR FV-7154" Step 3 to transfer additional MFRVs to This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 81 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 9 Feedwater Isolation Signal Reset and Establishing SG Page 1 of 6 Feed NOTE This Addendum may be used to reset a Feedwater Isolation Signal, as directed by the Unit Supervisor/Shift Manager.1.0 IF it is desired to reset the Feedwater Isolation Signal, THEN PERFORM the following:

1.1 ENSURE

both "LOW PWR" and "NORM" feedwater regulating valve controllers in manual with a zero (0) demand. {CP006}* SG IA(2A)* SG IA(2A)0 SG lB(2B)0 SG IB(2B)* SG S C(2C)* SG IC(2C)SG ID(2D)SG 1D(2D)"LOW PWR FV-715 1" (manual & "0" demand)"NORM FCV-0551 " (manual & "0" demand)"LOW PWR FV-7152" (manual & "0" demand)"NORM FCV-0552" (manual & "0" demand)"LOW PWR FV-7153" (manual & "0" demand)"NORM FCV-0553" (manual & "0" demand)"LOW PWR FV-7154" (manual & "0" demand)"NORM FCV-0554" (manual & "0" demand)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 82 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 9 Feedwater Isolation Signal Reset and Establishing SG Page 2 of 6 Feed 1.2 ENSURE the following valves are CLOSED and their associated handswitch are in the CLOSE position:

{CP006}* SG 1A(2A)* SG lA(2A)* SG 1A(2A)* SG 1B(2B)* SG 1B(2B)* SG 1B(2B)* SG 1C(2C)0 SG IC(2C)SG S C(2C)SG ID(2D)SG S D(2D)SG S D(2D)"PREHTR BYPASS FV-7189" (HS in CLOSE & CLOSED)"FWIV FV-7141" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7148A" (HS in CLOSE & CLOSED)"PREHTR BYPASS FV-7190" (HS in CLOSE & CLOSED)"FWIV FV-7142" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7147A" (HS in CLOSE & CLOSED)"PREHTR BYPASS FV-7191" (HS in CLOSE & CLOSED)"FWIV FV-7143" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7146A" (HS in CLOSE & CLOSED)"PREHTR BYPASS FV-7192" (HS in CLOSE & CLOSED)"FWIV FV-7144" (HS in CLOSE & CLOSED)"FWIV BYPASS FV-7145A" (HS in CLOSE & CLOSED)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 83 of 102 Plant Shutdown From 100% to Hot Standby I Addendumn 9 Feedwater Isolation Signal Reset and Establishing SG Page 3 of 6 Feed NOTE* IF an actual SG HI-HI LEVEL signal is present (or was present from wet layup condition), THEN the only way to reset FWI from the control boards is to ensure level is restored and clear the P4 signal by closing the Rx Trip Breaker before depressing the reset push buttons.* The SG HI-HI LEVEL bistable signal must be clear for successful performance of the following steps.1.3 IF a SG HI-HI LEVEL actuation signal has NOT been RESET, THEN PERFORM the following:

1.3.1 ENSURE

that a Reactor Trip signal is NOT present.1.3.2 CLOSE the Rx Trip Breakers 1.4 RESET the following "FW ISOL RESET TRAIN A" and "TRAIN B" applicable signals by depressing both train pushbuttons:

{CP006}* SI/SG HI-HI LEVEL "RESET" pushbuttons.

• TRAIN A* TRAIN B* LO TAVE "RESET (BLOCK)" pushbuttons.

• TRAIN A* TRAIN B 1.5 VERIFY both red FW ISOL TRAIN A and FW ISOL TRAIN B lights on ESF Monitoring Panel 6M21 are extinguished.

This procedure, when completed, SHALL be retained.

0POP03-ZG-0006 Rev. 54 Page 84 of 102 Plant Shutdown From 100% to Hot Standby Addendum 9 Feedwater Isolation Signal Reset and Establishing SG Page 4 of 6 Feed NOTE Performance of the next step will release the "LOW PWR" feedwater regulating valves for normal operation.

IF the Reactor Trip Breakers are closed, THEN the "NORM" feedwater regulating valves will also be released for normal operation.

IF the Reactor Trip Breakers are open, THEN the "NOT RESET" white lights will remain illuminated after the next step is performed.

1.6 DEPRESS

both TRAIN A and TRAIN B FW CONT/BYP VLVS "RESET" pushbuttons.

{CP006}* TRAIN A* TRAIN B 1.7 IF the Reactor Trip Breakers are closed, THEN VERIFY both TRAIN A and TRAIN B FW CONT/BYP VLVS "NOT RESET" white lights extinguished.

{CP006}* TRAIN A* TRAIN B 1.8 IF the Reactor Trip Breakers are opened after resetting FWI, THEN RESET the following "FW ISOL RESET" "TRAIN A" and "TRAIN B" by depressing both trains pushbuttons:

{CP006}* LO TAVE "RESET (BLOCK)" pushbuttons.

• TRAIN A* TRAIN B FW CONT/BYP VLVS "RESET"* TRAIN A* TRAIN B This procedure, when completed, SHALL be retained.

oPOP03-ZG-0006 Rev. 54 Page 85 of 102 Plant Shutdown From 100% to Hot Standby Addendum 9 Feedwater Isolation Signal Reset and Establishing SG Page 5 of 6 Feed 1.9 RESET the following safety grade solenoid valves by momentary placing each SG's "FEEDWATER ISOL" .TRAIN A" and "TRAIN B" to OPEN: {CP006}* SG I A(2A) "FEEDWATER ISOL TRAIN A"* SG 1A(2A) "FEEDWATER ISOL TRAIN B"* SG IB(2B) "FEEDWATER ISOL TRAIN A"* SG I B(2B) "FEEDWATER ISOL TRAIN B"* SG 1C(2C) "FEEDWATER ISOL TRAIN A"* SG 1 C(2C) "FEEDWATER ISOL TRAIN B" 0 SG 1D(2D) "FEEDWATER ISOL TRAIN A" 0 SG I D(2D) "FEEDWATER ISOL TRAIN B" 2.0 MAINTAIN SGs narrow range levels between 55 and 75% as follows: 2.1 IF feeding the SGs with Auxiliary Feedwater is desired. THEN PERFORM the following:

2.1.1 ENSURE

all SG LO-LO Level AFW actuation's are RESET.2.1.2 ENSURE a minimum of two (2) AFW Pumps are in service per OPOP02-AF-0001, Auxiliary Feedwater.

2.1.3 ESTABLISH

AFW flow to the SG per OPOP02-AF-0001, Auxiliary Feedwater.

0 SG IA(2A)0 SG IB(2B)0 SG S C(2C)* SG ID(2D)This procedure, when completed, SHALL be retained.

0POP03-ZG-0006 Rev. 54 Page 86 of 102 Plant Shutdown From 100% to Hot Standby Addendum 9 Feedwater Isolation Signal Reset and Establishing SG Page 6 of 6 Feed 2.2 IF SU SGFP 14(24) is the SG feedwater source, THEN ESTABLISH flow as follows: 2.2.1 ENSURE low power feedwater regulating valves (LPRV) in the MANUAL position filly closed. {CP006}SG 1A(2A) "LOW PWR FV-7151" SG IB(2B) "LOW PWR FV-7152" SG 1C(2C) "LOW PWR FV-7153" SG 1D(2D) "LOW PWR FV-7154" 2.2.2 OPEN the preheater bypass valves. {CP006}* SG 1A(2A) "PREHTR BYPASS FV-7 189" S SG I 1B(2B) "PREHTR BYPASS FV-7190"* SG 1C(2C) "PREHTR BYPASS FV-7191"* SG 1D(2D) "PREHTR BYPASS FV-7192" 2.2.3 THROTTLE OPEN low power feedwater regulating valves (LPRV) to maintain SG levels. {CPO06}* SG 1A(2A) "LOW PWR FV-715 1"* SG S B(2B) "LOW PWR FV-7152"* SG S C(2C) "LOW PWR FV-7153" S SG S D (2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 87 of 102 Plant Shutdown From 100% to Hot Standby Addendum 10 CV-0218 Boration While Inserting Rods Page I of 2 CAUTION IF less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 2800 ppm. IF greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 3500 ppm. (Ref. 2.16)NOTE* ALT BORATION ISOL MOV-0218 may be operated as required to maintain system inventory; taking note of times when boration flow is secured or restarted.

• All indications and controls for this Addendum are located on Control Panel CP004.* Flow rate through ALT BORATION ISOL MOV-0218 with one BA pump in service should be limited to 170 gpm on FI-0120A to prevent BA pump runout (CREE 11-19352-2).

1.0 RECORD

the amount of boric acid required to raise the RCS boric acid concentration to 68°F Xenon-free conditions with "all rods are fully inserted" per Figure 5.5 of the Plant Curve Book.(Refer to Plant Curve Figure 3.1, 3.2A or 3.2B, Boron Addition, as necessary)

BA gallons 2.0 ENSURE the following valves are closed: " "BA FLOW CONT VLV FCV-110A"* "TO VCT OUTL FCV-0110B"" "RMW FLOW CONT VLV FCV-01I1A"* "FILL FCV-0111 B" 3.0 IF RCP seal differential pressure is low, THEN it may be desirable to realign seal return to the VCT to maintain RCP seal DP by performing the following:

3.1 OPEN "I(2)-CV-0171 CVCS SEAL RETURN VCT ISOL". (19 ft MAB High Energy Valve Room 80)3.2 CLOSE "1(2)-CV-0 170 CVCS SEAL RETURN TO CHARGING PUMP SUCTION ISOL". (19 ft MAB High Energy Valve Room 80)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 88 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 10 CV-0218 Boration While Inserting Rods Page 2 of 2 4.0 ENSURE a charging pump is running and discharging to the RCS.5.0 ENSURE charging flow indicated on FI-0205A is greater than 50 GPM.6.0 START a BA pump.7.0 OPEN "ALT BORATION ISOL MOV-0218." 8.0 RECORD BA flow rate on "ALT BORATE FLOW FI-0120A." BA Flow Rate GPM 9.0 DETERMINE the time required to borate the RCS to the 68'F Xenon-free conditions with"all rods are fully inserted" per Figure 5.5 of the Plant Curve Book: BA (Step 1)Boration Time = -Minutes BA Flow Rate (Step 8)10.0 Coordinate with Chemistry to periodically monitor RCS boric acid concentration, as required.11.0 WHEN the CV-0218 boration is no longer required, THEN PEFORM the following:

11.1 CLOSE "ALT BORATION ISOL MOV-0218." 11.2 STOP the BA pump.11.3 RESTORE charging and letdown lineup as directed by the Unit Supervisor or Shift Manager.12.0 IF RCP seal return was aligned to the VCT, THEN PERFORM one of the following, OTHERS IWE N/A: 12.1 IF desirable to realign seal returm to the CCP Suction, THEN PERFORM the following:

12.1.1 OPEN "1(2)-CV-0170 CVCS SEAL RETURN TO CHARGING PUMP SUCTION ISOL". (19 ft MAB High Energy Valve Room 80)12.1.2 CLOSE "1 (2)-CV-0171 CVCS SEAL RETURN VCT ISOL".(19 ft MAB High Energy Valve Room 80)12.2 IF desirable to continue Operation with the RCP seal return aligned to the VCT, THEN Caution Tag the CCP hand switch on Control Panel CP004 to identify the off-normal alignment.

Caution Tag ECO#This procedure, when completed, SHALL be retained.

POP03-ZG-0006 Rev. 54 Page 89 of 102 Plant Shutdown From 100% to Hot Standby I Addendum I 1 Inhibit Reactor Trip Log Page 1 of 1 Initials NOTE* Unit 1 ICS drops 101, 103, 115, 121, 123 used to access TRIP LOG ARM/INHIBIT page.* Unit 2 ICS drops 100, 102, 114, 120, 122 used to access TRIP LOG ARM/INHIBIT page.* Reactor Trip Log can be manually printed, CONTACT ICS System Engineer for support.1.0 This addendum SHALL be performed in Modes 3 through 6 or Core Off Loaded to SFP.Record current plant Mode: Mode 2.0 At one of the listed ICS drops SELECT "MENU".3.0 SELECT "DATA ANALYSIS AND MAINTENANCE".

4.0 ENSURE

"TOP LEVEL MENU" is selected.5.0 SELECT "OPERATOR STATION PROGRAMS".

6.0 SELECT

"Trip Log Report Alarm/Inhibit." NOTE Step 7 will cause ICS Alarm point SPZCTRIPINH to alarm.7.0 SELECT "CLICK TO INHIBIT" 8.0 RECORD "TRIP LOG REPORT" indication.(check appropriate block)INHIBIT El ARM I]9.0 IF "TRIP LOG REPORT" indicates ARM, THEN CONTACT UNIT SUPERVISOR/SHIFT MANAGER and SUBMIT a Condition Report.10.0 RECORD ICS alarm page point SPZCTRIPINH "RX/TURBINE TRIP LOGS" indication.(check appropriate block)INHIBIT El ARM l_11.0 IF ICS point SPZCTRIPINH displays ARM, THEN CONTACT UNIT SUPERVISOR/SHIFT MANAGER and SUBMIT a Condition Report.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 90 of 102 Plant Shutdown From 100% to Hot Standby Addendum 12 AFD Penalty Point Evaluation Page 1 of 3 1.0 Notes and Precautions For AFD Penalty Point Evaluation 1.1 Use this addendum to evaluate and adjust AFD Penalty Points as required.1.2 The indicated AFD SHALL be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of: 1.2.1 One minute penalty deviation for each 1 minute of power operation outside of the target band at thermal power levels equal to or above 50% of Rated Thermal Power, and 1.2.2 One-half minute penalty deviation for each 1 minute of power operation outside of the target band at thermal power levels between 15% and 50% of Rated Thermal Power.1.3 The Technical Specification AFD Monitor Alarm function is performed by the Delta Flux Nuclear Application Program (NAP) on the plant computer.1.4 The Delta Flux Nuclear Application Program uses Power Range Nuclear Instrumentation as the input for Percent Rated Thermal Power.1.4.1 Power Range Nuclear Instrumentation is NOT always an accurate indication of Percent Rated Thermal Power. The Power Range Nuclear Instrumentation may be de-calibrated during plant transients that include temperature changes or rod motion.1.4.2 AFD Penalty Point accumulation with power near the AFD transition points (15%and 50%) may not be accurate if the Power Range Nuclear Instrumentation does not accurately reflect actual Percent Rated Thermal Power.1.5 WHEN AFD is outside of its target band near the AFD transition points, THEN AFD Penalty Point accumulation should be monitored to ensure accurate accumulation of Penalty Points.1.5.1 As an example, IF AFD is outside of its target band and actual Percent Rated Thermal Power is 48%, THEN AFD Penalty Points would be inaccurately accumulated if Power Range Nuclear Instrumentation indicated above 50%.1.5.2 As an example, IF AFD is outside of its target band and actual Percent Rated Thermal Power is 18%, THEN AFD Penalty Points would be inaccurately accumulated if Power Range Nuclear Instrumentation indicated below 15%.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 91 of 102 Plant Shutdown From 100% to Hot Standby Addendum 12 AFD Penalty Point Evaluation Page 2 of 3 1.6 IF a de-calibration of the Power Range Nuclear Instrumentation has caused or is causing Penalty Points to be accumulated incorrectly, THEN the AFD Monitor Alarm should be declared INOPERABLE.

1.6.1 IF AFD Penalty Points are conservative (i.e. more Penalty Points than required have been collected or are being collected), THEN the AFD Monitor Alarm may be considered OPERABLE at the discretion of the Shift Manager.1.6.2 The AFD Monitor Alarm should remain INOPERABLE until Penalty Points are being accumulated correctly AND the historical Penalty Points are corrected in the plant computer Delta Flux Nuclear Application Program.1.6.3 Performance of OPSP03-NI-0001 (Power Range NI Calibration) should be used to make Power Range Nuclear Instrumentation match actual Percent Rated Thermal Power.1.6.4 Step 2 of this Addendum should be used to review and adjust historical Penalty Points in the plant computer Delta Flux Nuclear Application Program.2.0 Review and adjust Delta Flux Nuclear Application Program Penalty Points 2.1 Review power (NI, UI 118, Delta T), Delta Flux and target band limits.2.2 Most accurate or conservative power indication should be used.2.3 Obtain Shift Managers approval to adjust Delta Flux Nuclear Application Program Penalty Points.2.4 From the Nuclear Applications Program, select the DF: Delta Flux Menu (i.e. click on the"M" next to "DF: Delta Flux").2.5 From the DF: Delta Flux Menu, select "DF0300 -DF Penalty Point 24 Hr History.2.6 Select the line in the DF0300 window for the hour to review.2.7 Review the Penalty Point history for each minute of the selected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 92 of 102 Plant Shutdown From 100% to Hot Standby Addendum 12 AFD Penalty Point Evaluation Page 3 of 3 2.8 IF the Penalty Points for any minute in the selected one hour period needs to be corrected, THEN perform the following:

2.8.1 Select

the gray box for the minute to be corrected.

The selected minute number (0 to 59) will be displayed in the upper left corner of the window as MIN.2.8.2 Select the correct Penalty Point gray box (0.0, 0.5, or 1.0) for the selected minute.2.8.3 Verify the corrected Penalty Point value is displayed for the selected minute. The plant computer will add an "S" quality tag for any manually entered (substituted)

Penalty Point minutes.2.8.4 Repeat Steps 2.8.1 to 2.8.3 to correct Penalty Points for each minute during the selected hour as required.2.8.5 Close the DF Penalty History for the selected hour. (Double click the minus sign in the upper left corner of the window).2.9 Repeat Steps 2.6 through 2.8.5 to correct Penalty Points for any other hour period.2.10 Obtain a peer check that Penalty Points have been correctly adjusted.This procedure, when completed, SHALL be retained.

0POP03-ZG-0006 Rev. 54 Page 93 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 13 Low Power Feedwater Regulating Valves Exercise Page 1 of 3 j Initials NOTE* This addendum is used to stroke the Low Power Feedwater Regulating Valves (LPRV)prior to placing in-service on a shutdown.* A Plant Operator SHALL be stationed at the SG Low Power Feedwater Regulating Valves during valve stroking to verify smooth operation.

1.0 ENSURE

all four Low Power Feedwater Regulating Valves are CLOSED: {CP006}" SG IA(2A) "LOW PWRFV-7151"* SG 1B(2B) "LOW PWR FV-7152"* SG IC(2C) "LOW PWR FV-7153"* SG I D(2D) "LOW PWR FV-7154" 3.0 CLOSE the isolation valves for the Low Power Feedwater Regulating Valves: {29 ft TGB S Side}* "1(2)-FW-0193 SG 1A(2A) LOW POWER REG VALVE (FV-7151)

INLET ISOL"* "l(2)-FW-0191 SG IB(2B) LOW POWER REG VALVE (FV-7152)

INLET ISOL"* "1(2)-FW-0189 SG IC(2C) LOW POWER REG VALVE (FV-7153)

INLET ISOL"* "1(2)-FW-0187 SG 1D(2D) LOW POWER REG VALVE (FV-7154)

INLET ISOL" 4.0 CYCLE the Low Power Feedwater Regulating Valves full OPEN and CLOSED to ensure smooth operation:

{CP006}* SG IA(2A) "LOW PWR FV-7151"* SG 1B(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153"* SG 1D(2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 94 of 102 Plant Shutdown From 100% to Hot Standby Addendum 13 Low Power Feedwater Regulating Valves Exercise Page 2 of 3 Initials 5.0 VERIFY locally the Low Power Feedwater Regulating Valves CLOSED: {29 ft TGB S Side}* "I(2)-FW-FV-7151 SG IA(2A) LOW POWER FEED REG VALVE"* "1(2)-FW-FV-7152 SG 1B(2B) LOW POWER FEED REG VALVE"* "1(2)-FW-FV-7153 SG IC(2C) LOW POWER FEED REG VALVE" S' 1(2)-FW-FV-7154 SG 1D(2D) LOW POWER FEED REG VALVE" 6.0 ENSURE all four Low Power Feedwater Regulating Valves are CLOSED: {CP006}* SG IA(2A) "LOW PWR FV-7151"* SG 1B(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153"_* SG ID(2D) "LOW PWR FV-7154" 7.0 OPEN the isolation valves for the Low Power Feedwater Regulating Valves: {29 ft TGB S Side}* "1(2)-FW-0 193 SG IA(2A) LOW POWER REG VALVE (FV-7151)

INLET ISOL"* "1(2)-FW-0 191 SG IB(2B) LOW POWER REG VALVE (FV-7152)

INLET ISOL"" "I(2)-FW-0 189 SG 1C(2C) LOW POWER REG VALVE (FV-7153)

INLET ISOL" S"1 (2)-FW-0 187 SG 1D(2D) LOW POWER REG VALVE (FV-7154)

INLET ISOL".This procedure, when completed, SHALL be retained.

IOPOP03-ZG-0006 Rev. 54 Page 95 of 102 Plant Shutdown From 100% to Hot Standby I Addendum 13 Low Power Feedwater Regulating Valves Exercise Page 3 of 3 Initials 8.0 INDEPENDENTLY VERIFY OPEN the isolation valves for the Low Power Feedwater Regulating Valves: {29 ft TGB S Side}* "l(2)-FW-0193 SG 1A(2A) LOW POWER REG VALVE (FV-7151)

INLET ISOL"* "l(2)-FW-0191 SG 1B(2B) LOW POWER REG VALVE (FV-7152)

INLET ISOL" _* "1(2)-FW-0189 SG 1C(2C) LOW POWER REG VALVE (FV-7153)

INLET ISOL"* "1(2)-FW-0187 SG 1D(2D) LOW POWER REG VALVE (FV-7154)

INLET ISOL" This procedure, when completed, SHALL be retained.

0 oPOP03-ZG-0006 Rev. 54 Page 96 of 102 Plant Shutdown From 100% to Hot Standby Addendum 14 Deaerator Aux Steam Supply Valve (PV-7401)

Exercise Page 1 of I Initials NOTE" This addendum is used to stroke "AUX STEAM TO DEAERATOR PV-7401" prior to placing in-service on a shutdown.* A Plant Operator SHALL be stationed at "AUX STEAM TO DEAERATOR PV-7401" during valve stroking to verify smooth operation.

1.0 ENSURE

"AUX STEAM TO DEAERATOR PV-7401" controller PC-7401 in the MANUAL position AND CLOSED. {TGB Deaerator Stand)2.0 CLOSE one of the following isolations valves to "AUX STEAM TO DEAERATOR PV-7401":

{TGB Deaerator Stand)* 1(2)-AS-0036, "MAIN DEAERATOR AUXILIARY STEAM SUPPLY FV-7401 INLET ISOLATION VALVE".* 1(2)-AS-0298, "MAIN DEAERATOR AUXILIARY STEAM SUPPLY PV-7401 OUTLET ISOLATION VALVE".3.0 PLACE "AUX STEAM TO DEAERATOR(AUX STM SPLY) PV-740 1" in the MODULATE(MOD) position.

{CP008}4.0 CYCLE "AUX STEAM TO DEAERATOR PV-7401" using controller PC-7401 full OPEN and CLOSED to ensure smooth operation.

{TGB Deaerator Stand)5.0 ENSURE "AUX STEAM TO DEAERATOR PV-7401" is CLOSED. {TGB Deaerator Stand}6.0 ENSURE "AUX STEAM TO DEAERATOR PV-7401" controller PC-7401 in the MANUAL position AND CLOSED. {TGB Deaerator Stand)7.0 MOMENTARILY PLACE "AUX STEAM TO DEAERATOR(AUX STM SPLY)PV-7401" in the CLOSE position.

{CP008}8.0 ENSURE both isolations valves to "AUX STEAM TO DEAERATOR PV-7401" OPEN: {TGB Deaerator Stand)* 1(2)-AS-0036, "MAIN DEAERATOR AUXILIARY STEAM SUPPLY FV-7401 INLET ISOLATION VALVE".* 1(2)-AS-0298, "MAIN DEAERATOR AUXILIARY STEAM SUPPLY PV-7401 OUTLET ISOLATION VALVE".This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 97 of 102 Plant Shutdown From 100% to Hot Standby E Addendum 15 (Unit 2, 2RE16 Only) Turbine Shutdown without Page 1 of 4 I Bearing Lift Pump I NOTE CR 13-5155 has identified an issue with Turbine Bearing Lift Pump which may prevent bearing lift being available for Main Turbine rotation on the turning gear.Temporary Modification T2-13-5155-6 lowers the Bearing Lift pressure permissive to Main Turbine turning gear from 850 psig to 600 psig (N2LTPSH6211 "TURBINE GENERATOR BEARING LIFT PUMP DISCHARGE HIGH PRESSURE SWITCH").Temporary Modification T2-13-5155-6 installs a temporary gauge at test TEE for N2LTPSH621I1 "TURBINE GENERATOR BEARING LIFT PUMP DISCHARGE HIGH PRESSURE SWITCH" (TGB 83 ft Between Bearing 4 and 5 near the door where terminal box N2TMBTNLO121 is located is underneath the catwalk).This Addendum will rotate the Main Turbine 180 degrees every 15 minutes even if Bearing Lift is NOT available.

• A Plant Operator SHALL be stationed to monitor the Turbine operations without bearing lift.The Main Turbine SHOULD be rotated during cool down until metal temperatures are LESS THAN 300'F.The Turbine Lube Oil System SHALL remain in operation until Main Turbine shell temperatures are LESS THAN 300'F.* The Main Turbine SHOULD be rotated whenever Gland Sealing Steam is applied.The Main Turbine MAY be taken off the turning gear with Gland Sealing Steam applied:* For up to 15 minutes if turbine metal temperatures are GREATER THAN or EQUAL to 300'F.* For tIp to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if Turbine Lube Oil is operating and turbine metal temperatures are LESS THAN 300'F.* The Main Turbine SHALL NOT be rotated by steam until proven to be free to rotate by use of the Turning Gear.* Main Turbine metal temperatures are indicated on the following ICS points (Point Group 9):* T6014, HP TURB MIDSPAN BASE TEMP* T6015, HP TURB MIDSPAN COVER TEMP* T6016, HP TURB GOV END BASE TEMP* T6017, HP TURB GOV END COVER TEMP* T6018, HP TURB GEN END BASE TEMP* T6019, HP TURB GEN END COVER TEMP This procedure, when completed, SHALL be retained.

0POP03-ZG-0006 Rev. 54 Page 98 of 102 Plant Shutdown From 100% to Hot Standby Addendumn 15 (Unit 2, 2RE16 Only) Turbine Shutdown without Page 2 of 4 Bearing Lift Pump Initials NOTE When the Main Turbine shaft is rotated without Bearing Lift oil pressure a condition called"slip stick" is a concern. Slip stick would be noticeable as an unusual vibration or shaking as the Main Turbine shaft is rotated. Main Turbine shaft rotation should be stopped immediately and Control Room notified if slip stick occurs.1. ENSURE Pressure switch N2LTPSH6211 (TURBINE GENERATOR BEARING LIFT PUMP DISCHARGE HIGH PRESSURE SWITCH) jumper toggle switch in bypass (ON)per Temporary Modification T2-13-5155-6.

2. ENSURE Turbine Lube Oil System in operation per OPOP02-LT-0001, Turbine Lube Oil System.3. IF Bearing Lift Pump is available, THEN ENSURE Bearing Lift Pump is in service.4. Manually ROTATE Main Turbine using Turning Gear in manual as follows: a) ENSURE the "TURN GEAR" switch is in MANUAL. (CP007)b) ENSURE Turning Gear is engaged using the manual engagement lever. (At Turning Gear Motor)c) IF the Turning Gear DOES NOT engage, THEN PERFORM the following:

i) RELEASE the manual engagement lever. (At Turning Gear Motor)ii) PLACE the "TURBINE TURNING GEAR MOTOR JOG SWITCH" in "JOG".("MN TURBINE TURNING GEAR" panel)iii) Momentarily DEPRESS the "START" pushbutton to jog motor.("MN TURBINE TURNING GEAR" panel)iv) ENGAGE the Turning Gear with the manual engagement lever.(At Turning Gear Motor)d) IF desired to use the JOG function to rotate the Main Turbine shaft, THEN GO TO Step 4k) of this Addendum.e) PLACE "TURBINE TURNING GEAR MOTOR JOG SWITCH" in "RUN".(TGB 83' "MN TURBINE TURNING GEAR" panel)f) DEPRESS "START" pushbutton to start turning gear motor. ("MN TURBINE TURNING GEAR" panel)This procedure, when completed, S-tALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 99 of 102 Plant Shutdown From 100% to Hot Standby Addendum 15 (Unit 2, 2RE16 Only) Turbine Shutdown without Page 3 of 4 Bearing Lift Pump Initials g) VERIFY Main Turbine shaft is rotated 180 degrees.h) DEPRESS the "STOP" pushbutton to stop turning gear motor. (At "MN TURBINE TURNING GEAR" panel)i) Perform this Addendum every 15 minutes until Main Turbine metal temperatures are LESS THAN 300 0 F.j) GO TO Step 4n) of this Addendum k) ENSURE the "TURBINE TURNING GEAR MOTOR JOG SWITCH" in "JOG".("MN TURBINE TURNING GEAR" panel)1) Momentarily DEPRESS the "START" pushbutton to jog motor to rotate the Main Turbine shaft 180 degrees. ("MN TURBINE TURNING GEAR" panel)m) Perform this Addendum every 15 minutes until Main Turbine metal temperatures are LESS THAN 300 0 F.n) WHEN ready to secure Turning Gear operations, THEN PERFORM the following:

i) IF the Turning Gear is operating in MANUAL, THEN DEPRESS the "STOP" pushbutton. (At "MN TURBINE TURNING GEAR" panel)ii) VERIFY Turbine Turning Gear disengages.(I) IF Turning Gear Motor is engaged, THEN PROCEED as follows: 1. PLACE the power supply for the Turning Gear Motor on MCC 1G5(2G5)/F4 in "OFF".2. REMOVE motor shaft protective cover from top of Turning Gear Motor. (2 nuts)3. ROTATE Turning Gear Motor shaft CLOCKWISE until Turning Gear disengages from main turbine shaft.4. REPLACE motor shaft protective cover on top of Turning Gear Motor.5. PLACE the power supply for the Turning Gear Motor on MCC 1G5(2G5)/F4 to "ON".This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 100 of 102 Plant Shutdown From 100% to Hot Standby I Addendumn 15 (Unit 2, 2RE16 Only) Turbine Shutdown without Page 4 of 4 1 Bearing Lift Pump I Initials NOTE IF Turning Gear is NOT available or it is recommended by Engineering that it is unsafe to rotate the Main Turbine, THEN Step 5 should be performed as soon as possible to minimize Main Turbine Rotor bow.5. IF unable to rotate the Main Turbine, THEN PERFORM the following:

a) ENSURE RCS cooling is transferred to SG PORVs per procedure Steps 6.16 and 6.17.b) ENSURE "COND POLISH BYPASS MOV-0132" is open. {CP008}c) DISPATCH operator to place all non-running "CONDENSER VACUUM PUMP LOCAL HANDSWITCH" in the "OFF" position.

{29 ft TGB}d) ENSURE Condenser Vacuum Breakers are open. {CP009}e) DISPATCH operator to secure running Condenser Vacuum Pumps per OPOP02-CR-0001, Main Condenser Air Removal. {29 ft TGB I t) ENSURE turbine exhaust hood spray valves in the MOD position.

{CP007}g) SECURE Turbine Gland Seal System per OPOP02-GS-0001, Turbine Gland Seal Steam System.h) DISPATCH operator to secure auxiliary steam per OPOP02-AS-0001, Auxiliary Steam System.6. ENSURE Engineering evaluation of Main Turbine Rotor prior to rotating Main Turbine.This procedure, when completed, SHALL be retained.

IOPOP03-ZG-0006 Rev. 54 Page 101 of 102 Plant Shutdown From 100% to Hot Standby Lineup I Turbine Generator Systems Cooling Water Lineup Page I of 2 UNIT 1 (Circle Unit Performing Lineup) UNIT 2 EXCEPTIONS DEVICE COMPONENT NOUN N UMBER DESCRIPTION REMARKS Personnel participating in device manipulation:

Name Initials Device lineup completed by: Operator Date Time Lineup I Reviewed: Unit Supervisor Date This procedure, when completed, SHALL be retained.

OPOP03-ZG-0006 Rev. 54 Page 102 of 102 Plant Shutdown From 100% to Hot Standby Lineup I Turbine Generator Systems Cooling Water Lineup Page 2 of 2 DEVICE COMPONENT NOUN LOCATION POSITION ALIGNED NEW TAG NUMBER ] DESCRIPTION REQUIRED BY NEEDED CL-ACW OUTLET FOR EXCITER COOLER TGB 55' SW Corner, WTA Regulator I (2)-AC-0062 I A(2A) Cubicles CLOSED CL-ACW OUTLET FOR EXCITER COOLER TGB 55' SW Comer, WTA Regulator 1(2)-AC-0063 IA(2A) Cubicles CLOSED CL-ACW OUTLET FOR EXCITER COOLER TGB 55' SW Comer of WTA Regulator 1I(2)-AC-0060 1B(2B) OUTLET Cubicles CLOSED CL-ACW OUTLET FOR EXCITER COOLER TGB 55' SW Comer of WTA Regulator 1I(2)-AC-0061 1B(2B) OUTLET Cubicles CLOSED This Form when completed, has NO retention.

SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION D0527 STI 33I7718960 0P P 3Z -0 7Rev. 71 Pae I of21 OPOP03-ZG-0007 7~Pg~f1 Plant Cooldown Quality Safety-Related Usage: IN HAND Effective Date: 11/06/2013 Controlling Station D. Rohan NA Crew IC Operations PREPARER TECHNICAL USER COGNIZANT DEPT.Usge Table of Contents , 4 1.0 Purpose and Scope....................................................

~ ... 3..........

4 2 .0 R eferences 3..... .......................................................................

................................

3 2 3. 0 Notes and Precautions

............................

/..........................

.............

9 2 4.0 P rerequisites

..................................................................................................................

18 2 5.0 M ode 3 C ooldow n ......................................................

20... ... .... ........ .........................

9 2 6.0 M ode 4 C ooldow n 45..................................................................................................

45 2 7.0 M ode 5 C ooldow n ........................................

5.............5.

.............

..............................

5 2 8.0 R CS D epressurization

............................

76............................

...............

.........................

76 2 9.0 R C S D rain dow n or V entin 8......................9.......................

........................................

89 2 10.0 R Seco d ary Pai n t S udow n o e t ..................

I:-.........1

...............

.......................................

1 89 2 10.0 Secondary Plant Shutdown 108 2.11.0 R ecords R eview 115............................

.................

115 2 11.0 Recor s Revi w ........................ .........

...... .* ........................................

1 4 12.0 Support D ocum ents ..............

/ .........

.....................................................................

1.o 116 3 Addendum 1, RCS Cooldown Lim itations ..............................................................

117 3 Addendum 2, RCS/RJ-R Simplified Elevati-6n Diagram .........................................

118 3 Addendum 3, Determination of Rt 'S Volume to be Drained ..................................

119 3 Addendum 4, RC-LG 3662 RCS LE, VEL SIGHTGLASS (SLINKY) Upper C o n e t o .s e t l D i ...: Connection A ssembly D ,agram ......................................................................

120 Ad e d m 5 ......

2 Addendum 5,RCS.D egassification

..............................................................

121 2 Addendum 6>Degassing-the RCDT and PRT ..........................................................

124 2 Addendum 7,Conditions for Steam Generator Decay Heat Removal ....................

127 2 A ddendum 8, Gripper Releasing

..............................................................

128 2 KAddendum 9, Plant Cooldown with the PZR Water Solid ......................................

132 2 Addendum 10, MODE 5 Cooldown with MSIVs OPEN ........................................

140 4 jjc' Addendum41H, RCS/PZR Pressure Operations Guideline

.......................................

161 7,> 44 OPOP03-ZG-0007 Rev. 71 Page 2 of 216 Plant Cooldown U sage Table of Contents ..............................................................................................

Pa. ge 2 Addendum 12, Solid Plant Operations Entry Checklist

...........................................

166 2 Addendum 13, Manual Blowdown of Main Steam lines upstream of MS1Vs ........ 175 1 Addendum 14, MOV-0016A, B & C Emergency Operations Guideline

................

181 2 Addendum 15, Rx Head Venting and RCS level Instruments Disagreements E valuation G uideline .......................................................................................

182 2 Addendum 16, Throttling 1(2)-CV-0198 RMW ISOL for Technical Specification 3 .4 .1.3 ..............................................................................................................

18 8 2 Addendum 17, Moving Control Rods with High Reactor Coolant Turbidity

.........

192 3 Addendum 18, Indicated Pressurizer Level When Solid vs. Pressurizer Temperature

.........................................................................................................................

19 4 1 Addendum 19, Controlling RCS Inventory at or above Elv 39 ft. 4.9 in ................

195 1 Addendum 20, Venting Reactor Vessel Head Using Head Vent Throttle Valve(s) 205 2 Addendum 21, Closure of Personnel Air Lock Doors .............................................

207 2 Data Sheet 1, RCS and Pressurizer Cooldown Rates ...............................................

209 1 Lineup 1, RV to PZR Equalizing Line Lineup ........................................................

213 Form 1, CVCS Line Boration Tracking Form ...........................

215 Usage I -IN HAND 2 -IN HAND CONTROLLING STATION 3 -REFERENCED 4 -AVAILABLE This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 3 of 216 Plant Cooldown 1.0 Purpose and Scope 1.1 This procedure provides the instructions for a plant cooldown from Hot Standby to Cold Shutdown.

Included are instructions for maintaining Shutdown Margin and proper temperature/pressure relationships.

1.2 This procedure SHALL be used to satisfy Technical Specification Surveillance Requirement 4.4.9.1.1, TRM 4.4.9.2 (partially).

2.0 References

2.1 Technical

Specifications (TS) 3.4.1.3, 3.4.1.4.2, 3.4.9.3, 3.5.3.1, 4.4.9.1.1, 4.4.9.3.4, 3.4.1.4.1, 3.4.9.1, 3.6.1.1, 3.6.1.3.2.2 Technical Requirements Manual (TRM) 3.1.2.1, 3.4.9.2, 4.4.9.2 2.3 UFSAR Section 9.3.4.1.2.6, 15.4.6.2, 5.2.2.11.3 2.4 MATS Items 8600865-866, 8600863-866 and 8401177-866 (IEC 78-05)2.5 OPCP03-ZC-0005, Chemical Addition to the Reactor Coolant System 2.6 OPOP02-WG-0001, Gaseous Waste Processing System Operations 2.7 OPGPO3-ZE-0033, RCS Pressure Boundary Inspection For Boric Acid Leaks 2.8 OPGP03-ZO-0012, Plant Systems Chemistry Control 2.9 OPOP02-AF-0001, Auxiliary Feedwater 2.10 OPOP02-CR-000 1, Main Condenser Air Removal 2.11 OPOP02-CV-000I, Makeup to the Reactor Coolant System 2.12 OPOP02-CV-0004, Chemical and Volume Control System Subsystem 2.13 0POP02-RC-0004, Operation of Reactor Coolant Pump 2.14 OPOP02-RH-0001, Residual Heat Removal System Operation 2.15 OPOP02-SB-0002, Steam Generator Wet Layup Recirc 2.16 OPOP02-SP-0001, Solid State Protection System 2.17 OPOP03-ZG-0009, Mid-Loop Operation This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 4 of 216 Plant Cooldown 2.18 0POP03-ZG-0010, Refueling Operations 2.19 OPSP02-RC-0403, RCS COMS T Hot Set 2 ACOT (P-0403, T-0413, T-0423, T-0433, T-0443)2.20 0PSP02-RC-0404, RCS COMS T Cold Set 3 ACOT (P-0404, T-0414, T-0424,T-0434, T-0444)2.21 OPSP03-MS-0002, Main Steam System Cold Shutdown Valve Operability Test 2.22 OPOP02-CD-0001, Condensate System 2.23 0POP02-FW-0001, Main Feedwater 2.24 0POP02-RS-0001, Rod Control 2.25 OPOP02-GS-0001, Turbine Gland Seal Steam System 2.26 OPOP02-RC-0003, Filling and Venting the Reactor Coolant System 2.27 0POP03-ZG-0003, Secondary Plant Startup 2.28 OPOP03-ZG-00 11, Secondary Plant Cold Startup 2.29 OPSP03-CV-001 1, Chemical and Volume Control System Valve Operability Test (Cold Shutdown)2.30 0PSP03-CV-0014, CVCS Equipment Verification 2.31 OPSP03-FW-0002, Feedwater System Valve Operability Test (Cold Shutdown)2.32 OPSP03-HC-0004, Reactor Containment Building Normal Purge System Valve Operability Test (Cold Shutdown)2.33 0PSP03-RH-0007, Residual Heat Removal System Valve Operability Test (Cold Shutdown)2.34 OPGP03-ZO-0042, Reactivity Management Program 2.35 OPMP07-DM-0003, Rapid Refueling Rod Holdout Operation 2.36 OPSP03-CV-0009, Monthly Boration Flow Path Verification 2.37 OPSPIO-ZG-0003, Shutdown Margin Verification

-Modes 3, 4 and 5 2.38 MATS Item 8500092-866 (SER 84-079)2.39 MATS Items 8500141-866, 8500088-866 (SER 82-07)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 5 of 216 Plant Cooldown 2.40 MATS Item 8500184-866 (UFSAR Section 3.9.1.1.6.9) 2.41 MATS Item 8600015-866 (OMR 85-283)2.42 MATS Item 8501477-866 (SER 81-091)2.43 MATS Item 8501603-860 (IEB 80-12)2.44 MATS Item 8600210-866 (UFSAR Section 5.4.7.2.3) 2.45 MATS Items 8601358-866, 8500310-866, 8601060-866, 8601230-866 2.46 MATS Item 8700678-860 (SER 87-013), Rx Vessel Stud Corrosion from Primary Coolant Leak 2.47 MATS Item 8802104-936 (ST-HL-AE-3398) 2.48 MATS Item 9000929-936, (GNL 90-006)2.49 MATS Item 9001090-872, Incorporate FCs 2.50 MATS Item 9001263-936, (SPR 900465) Uniform Heatup and Cooldown Rates 2.51 MATS Item 9200042-936, (IEN 91-73)2.52 MATS Item 9200129-936, (OTH 92-001)2.53 Preliminary Response to Generic Letter 870912, ST-HL-AE-2356 2.54 SPR 920540, Letdown Isolation While Bypassing MSIVs 2.55 SPR 920607, Possible Reactor Vessel Head Vacuum Affecting Level Indication 2.56 SPR 930426, Administrative Cooldown Limit Exceeded 2.57 Boron Dilution Analysis, ST-HL-AE-1765, 1988 2.58 Pressurizer Differential Temperature, IEB 88-11 2.59 Thermal Stratification of the Pressurizer Surge Line, ST-HL-AE-2992 2.60 Westinghouse Precautions, Limitations and Setpoints, 5Z00OZS1 101 2.61 Westinghouse Steam Generator Precautions, Limitations and Setpoints ST-WN-YB-2958 2.62 WCAP 12067, Pressurizer Surge Line and Residual Heat Removal Line Stratification Section 1.4.7, Dec. 1988.2.63 Westinghouse Technical Bulletin No. 77-14, CREE 97-14236-1, CREE 00-3288-1, CREE 01-15527-1 2.64 Westinghouse Technical Bulletin No. 92-05 This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 6 of 216 Plant Cooldown 2.65 Design Basis Documents:

2.65.1 5R124MB1027, Reactor Coolant System 2.65.2 5Z529ZB 1003, NSSS Controls 2.65.3 9GOI9N4B10 117, Turbine Generator System 2.65.4 5S139MB0120, Feedwater System 2.65.5 5Sl09MB1026, Main Steam System 2.66 OPSP03-RC-00 10, Pressurizer Power Operated Relief Valve Operability Test 2.67 ST-HL-AE-4549, Technical Specification 4.4.4.1 Action b. Operating PORV Through One Cycle of Full Travel 2.68 ST-HS-HS-26623, Operability and Reportability Review of Station Problem Report 932774 for Unit I and 2 2.69 CR-95-604, Degraded Standby Decay Heat Removal Capability Via Natural Circulation 2.70 USQE 95-0011, Justification for RCS Cooldown Limits 2.71 USQE 96-0014, Justification for Plant Cooldown with Rods Partially Withdrawn 2.72 HL&P LOG# A41010-00438-AVB, SDM Reference 2.73 CR 96-5716., Pressurizer Temperature Cooldown Limit Exceeded 2.74 CR 97-14212-2, CREE "Control Rod H2 Would Not Withdraw" 2.75 Reg Guide 1.22 and Reg Guide 1.118 2.76 UFSAR 7.1.2.5 2.77 CREE 97-14388-1 2.78 ST-UB-NOC-1847, Unit 2, Cycle 7 Modes 3, 4, and 5 RSAC Confirmation (CR 98-16471)2.79 CR 96-15748-5, Incorporate Recommendations of SER 98-01 2.80 TSC-218, RCS Overpressure Protection Systems 2.81 USQE 01-9518-6, UFSAR change to add analysis of CVCS malfunctions in Mode 3 that increases Reactor Coolant Inventory (Section 15.5.2).2.82 CREE 02-5478, MC CALC 6500, remove the limitation that closes the MSIV's when RCS temperature is less than 245°F.2.83 Westinghouse Nuclear Safety Advisory Letter NSAL-02-14, Steam Line Break During Mode 3 2.84 CR 02-3367, Erratic indication on TI-0607 Pressurizer Vapor Temperature indication.

2.85 ST-HL-AE-2498, Limits on Hydraulic Transients This procedure, when completed, SHALL be retained.

2.86 CR 03-3694, Safety Injection due to moisture in Main Steam Lines 2.87 CR 03-4704, PZR PORV actuation during Solid Plant Conditions 2.88 Calc 04-RC-0001 Rev 1, RCS Dilution from the Pressurizer 2.89 OPOP07-RC-0001, RC Vent Rig/Sightglass Installation and Removal 2.90 OPMIPO8-CV-0135 "Letdown I-X Outlet Pressure Calibration".

2.91 CR 05-3071, (LER 2-05-003)

SSPS ECO to support FWIV energize to actuate MOD in 2RE10 caused COMS to be inoperable during solid plant operations in violation of TS 3.4.9.3.2.92 CAQ-S 06-4091-5, U2 'A' Train RHR Header Pressure Rose to the RHR Relief Setpoint 2.93 UFSAR 6.3.2.5 2.94 CREE 08-5240-2, Evaluation to document the risks associated with not performing steps to prevent control rod thermal binding.2.95 5ZO1OZSI 101, Precautions, Limitations and Setpoints 2.96 CREE 08-13703-1, Evaluate the current requirements in OPOP03-ZG-0007 prohibiting rod movement with RCS turbidity levels above 1. 1 NTU during plant cooldown to determine if the limit can be raised or an alternate method. can be used to ensure CRDM performance is not degraded.2.97 CREE 08-14659-2, Evaluate any procedural changes/notes that need to be made to prevent CCP 2B high vibration alarm under high flow conditions from being written up in the future.2.98 CR 00-4057, At Low RCS Pressure and No RCP Running RCS Flow Indicators May Indicate 20-25% Flow.2.99 CR 98-14576, Uncertainty may be as high as 10 gpm for the indicated value. This could result in 135 gpm actual flow through CV0198.2.100 CR 09-15273, During swapping of steam reducers in Unit 2 from the large reducer to the small reducer the steam relief valve in Unit 1 lifted (PSV-8795) due to high header pressure.2.101 Calculation ZC-7040, Loop uncertainty calculation for RCS Wide Range pressure monitoring instrumentation.

2.102 OE 27762, Unintended Reduction in Pressurizer Inventory (SEN 278).2.103 CR 09-13289, SOER 09-1 RECOMMENDATION

1. 5 QUESTION 5.9.2.104 CREE 02-1259-4, Indicated Level When Solid This procedure, when completed, SHALL be retained.

7POP03-ZG-0007 Rev. 71 Page 8 of 216 Plant Cooldown 2.105 CR 10-6519, While performing RCS drain down, received rad monitor alarms for unit vent (80 101).2.106 OPCP01-ZA-0038, Plant Chemistry Specifications 2.107 OPMP04-ZG-00 12, Equipment Hatch Removal and Installation 2.108 Calculation NC-07090, Evaluation of Boron Dilution Flow Paths in Modes 5b and 6 2.109 SCAQ CR 11-7747, LER 1-1 1-001 -During the review of the evolution to pump the IB RHUT to the VCT a possible dilution source may have been identified that is not covered by OPSP03-CV-0014, CVCS Equipment Verification.

2.110 CR 11-31378-47, Operation of TV-4494 With RCS Cooled and Vented Below 115 'F.2.111 CR 12-28399, Entered OPOP04-CR-0001 Due To Lowering Condenser Vacuum.2.112 CREE 13-4092-7, Determine Selected Volumes CVCS Piping to borate for compliance with T.S. 3.4.1.4.2, and 3.9.1.2.113 OPGPO3-ZM-0028, Erection and Use of Temporary Scaffolding This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page9of216 E : Plant Cooldown 3.0 Notes and Precautions

3.1 Steps

in this procedure that begin with *** are considered non-time critical and the cooldown SHOULD continue without delays, WHEN it is being performed to meet a Technical Specification LCO ACTION or using POP09s, POP04s, POP05s (EOPs) and time does NOT permit step completion.

Actions to complete *** Steps SHOULD be addressed as soon as time/manpower permits.3.2 Operating PROCEDURES are written based on a defined set of plant conditions and equipment availability.

PROCEDURE changes are NOT required to document alternate performance based on conditions different from those assumed if the PROCEDURE can be performed safely. The decision to proceed lies with the Shift Manager/Unit Supervisor and is based on knowledge of system design and operation and the impact of omitting or re-sequencing steps.3.2.1 The Shift Manager/Unit Supervisor may authorize alternate performance for operating PROCEDURE sequence, including omitting steps, based on plant operating conditions.

The Shift Manager/Unit Supervisor ensures such an alternate perfornance does NOT adversely impact the safety of personnel or equipment, and documents the alternate method in the appropriate procedure or logbook. REFER TO OPGP03-ZA-00 10, Performing and Verifying Station Activities for specific details.3.2.2 The Shift Manager/Unit Supervisor may authorize early start of procedure steps to enhance plant performance, WHEN the early start is of NO safety impact for current plant conditions.

Documentation is NOT required for an early start as long as the step is completed before moving past this step in the overall sequence.3.2.3 Steps within this procedure SHALL be performed in order listed or in order provided in an authorized early start (Step 3.2.2) or alternate perfonnance (Step 3.2.1). Steps that are authorized to be omitted SHALL be designated by placing "N/A" in the signoff or initial blanks. REFER TO OPGP03-ZA-00 10, Performing and Verifying Station Activities for specific details.3.2.4 Shift Manager/Unit Supervisor may authorize holding RCS Temperature and/or Pressure at selected levels to support outage/maintenance conditions.

The Shift Manager/Unit Supervisor ensures holding RCS Temperature and/or Pressure has NO safety impact for current plant conditions, and such performance does NOT adversely impact the safety of personnel or equipment, and documents the reason in the appropriate procedure or logbook.3.2.5 WHEN the conditions requiring a Plant Cooldown NO longer exist, the Shift Manager/Unit Supervisor may authorize exiting this procedure to perform OPOP03-ZG-0001, Plant Heatup.3.3 The Unit Supervisor SHALL signoff or initial all steps unless otherwise designated within this procedure.

This procedure, when completed, SHALL be retained.

3.4 *** Prior to commencing Plant Cooldown, any valves that were backseated with ECO tags SHALL be removed from their backseat.

This is preferred but NOT required if the cooldown is required by Technical Specifications.

3.5 *** Prior to initiating cooldown., careful consideration should be given to the consequences and effects on other plant parameters. (Ref 2.54)3.6 IF the cooldown is due to an unisolable RCS leak, THEN a prompt depressurization and cooldown of the RCS is required.* The 80°F/hr administrative cooldown limit does NOT apply. (Ref 2.70)* The temperature differential between the Pressurizer water space and the RCS SHALL be maintained less than or equal to 250'F. (Ref 2.62)3.7 Caution SHALL be used during changes in plant status to minimize the potential for hydraulic transients.

3.8 WHEN Mode 5 is entered, THEN it is permissible to open both doors of the personnel airlock simultaneously.

3.8.1 IF both doors are to be left opened, THEN the key switch SHALL be placed in the OFF position to de-energize the hydraulic system solenoids.

3.9 The following are examples of precautionary measures that may be used to minimize cooldown induced events: (Ref 2.54)* Isolating known steam demands prior to equalizing around the Main Steam Isolation Valves. (e.g. turbine drains)* Raising Pressurizer level in anticipation of the level shrink due to a cooldown.* Promptly isolating Main Steam in the event of an unanticipated excessive steam demand.* Taking manual control of the charging flow control valve as needed to raise pressurizer level.This procedure, when completed, SHALL be retained.

II OPOP03-ZG-0007 Rev. 71 Page I1Iof 216 Plant Cooldown 3.10 Shift Manager or designee perfornms a thorough review of the outage schedule to ensure safety of the plant during Solid Plant Operations.

This review is performed by comparing the outage schedule to the guidelines on Addendum 12. Shift Manager or designee ensures compensatory measures or other remedial actions address Solid Plant risk concerns.3.10.1 All applicable issues will be analyzed to determine if they meet the Solid Plant Operations guidelines identified in this procedure.

3.10.2 IF the guidelines are NOT met, THEN: 3.10.2.1 Operations SHALL document those issues in the report to the Outage Manager. Operations identifies any HIGHER RISK EVOLUTIONS which require additional reviews.3.10.2.2 Corrective actions and/or contingencies SHALL be developed and implemented to minimize risk (i.e., schedule changes, deletion from outage, compensatory measures, training, etc.).3.10.3 All corrective actions should be evaluated for the following:

3.10.3.1 Effectiveness to ensure actions taken, especially schedule revisions, have NOT affected shutdown issues elsewhere.

3.10.3.2 To ensure safety of the plant and plant personnel have been maintained.

3.11 Inoperable Instrument substitution:

WHEN a instrument (i.e., flow, pressure, temperature, etc) specified in this procedure is inoperable and there is a Functional instrument that performs the equivalent function or provides an alternate method to read the desired variable (i.e., QDPS. ICS, etc), THEN the Shift Manager/Unit Supervisor may authorize the use of the Alternate instrument in place of the instrument specified by this procedure.

The Shift Manager/Unit Supervisor ENSURES that the use of an alternate instrument does NOT adversely impact the safety of personnel or equipment, and documents the alternate method in the appropriate PROCEDURE or logbook. This Step does NOT apply when performing Surveillance testing. Instruments used during the performance of a Surveillance Test SHALL be IAW OPGP03-ZM-0016, Installed Plant Instrumentation Calibration Verification Program. (See OPGP03-ZA-0010, Performing and Verifying Station Activities, for specific details.)3.12 Technical Specification 3.4.1.4.1, Mode 5 with reactor coolant loops filled, is satisfied by one RCP running for a period of time such that any non-condensable gas is removed from the SG tubes AND when RCS pressure is maintained above the VCT pressure. (Ref 2.78)3.13 Technical Specification 3.4.1.4.2, Mode 5 with reactor coolant loops NOT filled (Also referred to as Mode 5b), is applicable when ALL RCPs are stopped AND RCS pressure is or has been less than VCT pressure. (Ref 2.78)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 12 of 216 Plant Cooldown 3.14 The following are required to maintain COMS Operable (Ref. TS 3.4.9.3, TS Bases Table 3.8-1): " SSPS Actuation Cabinets "A" and "B" and Logic Cabinets "R" and "S" must be energized. (Ref. Step 3.15)* Pressurizer PORVs must be energized.

• MRDS Switches may be in "DEFEAT ALL" or "DEFEAT/CVI AVAIL".* SSPS Actuation Trains may be in TEST or OPERATE.* Electrical distribution Channels I and II (Train A) & III (Train B) must be operable.One train must be powered by its associated operable Class 1E power system (batteries, inverters and chargers).

• QDPS (APC-BI and APC-DI) and 7300 Cabinet (Channel 2-ZRRO16 and Channel 3-ZRRO 18) portions for the following:

• BI(2)RC-P-0404 DI(2)RC-P-0403

• B 1 (2)RC-T-0414 D 1 (2)RC-T-0413

• BI(2)RC-T-0424 Dl(2)RC-T-0423

• BI(2)RC-T-0434 DI(2)RC-T-0433

• B 1(2)RC-T-0444 DI(2)RC-T-0443 3.15 SSPS Actuation Cabinets "A" and "B" and Logic Cabinets "R" and "S" must be energized for both COMS to be OPERABLE.

Logic Cabinet "R" and Actuation Cabinet "A" required for PCV-0655A "Pressurizer PORV Train A". Logic Cabinet "S" and Actuation Cabinet "B" required for PCV-0656A "Pressurizer PORV Train B". {LER 2-05-003, SSPS ECO to support FWIV energize to actuate MOD in 2RE10 caused COMS to be inoperable during solid plant operations in violation of TS 3.4.9.3. (CR 05-5960, CR 05-3071)}.

3.16 Plant Cooldown should be a smooth evolution spread over the entire 60 minute period.A short time accelerated cooldown SHALL NOT be performed. (Ref 2.50)3.17 The administrative limit for cooldown of the RCS, excluding the Pressurizer, is 80'F in any one hour period. This limit may be exceeded at the Shift Manager's discretion. (Ref 2.56 and Ref 2.70)3.18 The administrative limit for cooldown of the Pressurizer is 160'F in any one hour period.3.19 Upper head subcooled.

margin SHALL be checked every 30 minutes using Reactor Vessel head thermocouple to ensure the upper head remains subcooled during depressurization to prevent void formation.

This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 13 of 216 Plant Cooldown 3.20 RCS and Pressurizer temperature changes SHALL be monitored per Data Sheet I during all cooldown cycles.3.21

• Pressurizer boron concentration SHALL be maintained within 50 ppm of RCS concentration prior to securing all RCPs. Additional Pressurizer heaters SHALL be energized to equalize the boron concentration.

3.22 WHEN altering RCS pressure, THEN RCP seal injection flows SHALL be maintained between 8 and 13 gpm to each RCP.3.23 During RCS cooldown it may be required to place "PRESS CONT PCV-0135" in manual to control letdown flow.3.24 To minimize the effects of surge line thermal stratification, the temperature differential between the Pressurizer liquid and the Reactor Coolant SHALL NOT exceed 320'F.(Ref 2.59 and 2.62)3.25 Auxiliary spray SHALL NOT be initiated with a temperature differential greater than 621°F. (TRM 3.4.9.2.c) 3.26 IF less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 2800 ppm. (Ref. 2.93)3.27 IF greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 3500 ppm. (Ref. 2.93)3.28 Evolutions such as equalizing around and opening a Main Steam Isolation Valve may cause a primary plant cooldown and SHALL be performed in a controlled manner.(Ref 2.54)3.29 Following a RCFC Inlet Temperature HI Alarm, Control Room indication associated with RCS Pressure loops 403 and 404 should NOT be used for the purposes of complying with the Technical Specification Pressure-Temperature Heatup/Cooldown curves until the transmitters have been recalibrated.

Pressure loops 403 and 404 would remain operable for purposes of COMS input. Pressure Loops 405, 406, and 407 are NOT affected. (Ref 2.101)3.30 In Mode 5 and 6 (Head on Rx Vessel), the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" SHALL be manned at all times when remote level indication is NOT available. (Remote level indication MAY be Video Monitors in the Control Room)3.31 In Mode 5 and 6 (Head on Rx Vessel) with RCS Depressurized and maintaining RCS Inventory at a fixed value, periodic venting of the Reactor Vessel Head may be required due to gas buildup. Monitor RVWL Sensors 1, HJTC Train "A" or Train "C" Computer Points IITE2004 and IITE3004 respectively, if available.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 14 of216 Plant Cooldown 3.32 In Mode 5 and 6 (Head on Rx Vessel) with RCS Depressurized and maintaining RCS Inventory at a fixed value, IF unexpected VCT diversions are required to maintain PZR and VCT levels, THEN Monitor Reactor Vessel Head for voiding as gasses come out of solution in the Reactor or the Steam Generators displacing the water.3.33 (In Mode 5 and 6 with RCS Depressurized)

WHEN Reactor Vessel Head voiding is indicated, THEN vent the Head per Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

3.34 (UNIT 2 ONLY) Under high flow (> 200 gpm) and low RCS pressure conditions, the high vibration alarm may come in for the inboard bearing on CCP 2B (7H and 7V POSITIONS).

This alarm is expected because of a shaft surface characteristic and is not indicative of a pump vibration problemr.

CCP 2B should still be checked out to ensure proper operations however a Condition Report is not required if the only issue is the high vibration on 7H and 7V POSITIONS. (Reference 2.97)3.35 Static pressure zero and span shifts on the RCS flow transmitters will affect flow indication when the RCS is NOT at normal operating pressure.

When the RCS is at low pressure with no RCP running, there may be RCS flow indication up to 25% flow even when there is no actual flow through the loop. This is an expected indication that occurs due to static pressure effect on the flow transmitters measuring differential pressure across elbow taps on the RCS piping. (Reference 2.98)3.36 (In Mode 5 and 6 and the RCS is pressurized)

WHEN Reactor Vessel Head voiding is indicated by Pressurizer level rising with constant or rising VCT level and RVWL Sensors 1, HJTC Train "A" or Train "C" Computer Points IITE2004 and IITE3004 respectively temperature rising, THEN Addendum 20, Venting Reactor Vessel Head Using Head Vent Throttle Valve(s), provides the steps for venting the Reactor Vessel Head to PRT.3.37 WHEN RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" is in service, THEN ENSURE "OPOP07-RC-0001, RC Vent Rig/Sightglass Installation and Removal", LINEUP 1, "RC-LG-3662 RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup PERFORMED daily and DOCUMENTED in a temporary log.3.38 IF differences in indications exist between the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)", RVWL and pressurizer Cold Calibrated channel, THEN STOP any draining of the RCS and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 15 of 216 Plant Cooldown 3.39 One (1) of the following Pressurizer vent paths should be established prior to drain down: (Ref 2.52)The Pressurizer spray line vent valves RC-0502 and RC-0503 open to atmosphere.(Preferred Path).* A minimum of one Pressurizer Code Safety Valve removed.3.40 During coupling and uncoupling of the RCP shaft, RCS leakage along the pump shaft and through the seal housing may occur.3.41 WHEN RCS temperature is less than 150'F, THEN the reactor vessel head vent rig may be connected per OPOP02-RC-0003, Filling and Venting the Reactor Coolant System, Addendum 1.3.42 IF control rods have been withdrawn per Addendum 8 AND the reactor trip breakers open unexpectedly, THEN Addendum 8 may be re-performed, if desired, to withdraw control rods prior to recommencing a plant cooldown.3.43 PZR indicated temperature should be less than 2007F above RCS Loop temperature prior to filling the Pressurizer to avoid exceeding a cooldown limit, or Normal Spray/Aux Spray flow should be established.

3.44 The PZR Surge Line Temperature should be monitored during plant cooldown. (Ref 2.73)3.45 When draining the RCS, THEN Addendum 3 MAY be referred to VERIFY actual volume drained correlates with expected volume drained. (Ref 2.79)3.46 Degassing of the RCS is typically performed by one or a combination of both of the methods below: Mechanical degassing

-where the RCS is degassed by spraying and venting of the Pressurizer and VCT.Chemical degassing

-where the RCS is degassed by the addition of chemicals that react with the dissolved gases in the RCS.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 16 of 216 Plant Cooldown 3.47 WHEN Pressurizer vapor space temperature TI-0607 is NOT functional, THEN use the associated functional Pressurizer water space temperature TI-0608 for all Pressurizer temperature indications called out in this procedure.

Use of the liquid temperature element alone is more conservative

[will provide higher indicated change for a given actual system change] and better represents actual metal temperature.

Use of the liquid temperature indication alone will provide assurance that cooldown limits will NOT be exceeded.(CREE 02-3367)Example: Pressurizer vapor space temperature TI-0607 is non-functional, THEN substitute Pressurizer water space temperature TI-0608, for Pressurizer vapor space temperature TI-0607 in this procedure.

3.48 Four (4) RCP operation NOT permitted WHEN RCS average temperature is less than 140'F. Do NOT run all 4 RCPs UNTIL RCS average temperature reaches greater than 140'F. (3 pumps or less restriction RCS < 140'F, This limitation is required to demonstrate acceptable fuel assembly top nozzle hold down spring forces in the Cold Zero Power lift force calculation.)

3.49 REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown.

3.50 During Shutdown conditions, WHEN an ECO is on SI-MOV-0016A, B, C AND its associated LHSI train is required to be functional for LOCA injection/recirculation, THEN ESTABLISH a "dedicated" Operator watch position in accordance with Addendum 14, MOV-0016A, B & C Emergency Operations Guideline.

3.51 REFER TO OPOP07-RC-0001 "RC Vent Rig/Sightglass Installation and Removal" as required, for instructions for the installation, removal, operation and monitoring of the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" or the "Additional (Tygon) RCS Level Sightglass".

3.52 WHEN the RCS is Depressurized, THEN for Rx Head venting or for RCS Instruments Disagreements, REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline, as required.3.53 WHEN Borated water drips/spills onto Carbon steel components, THEN ENSURE it is immediately cleaned up or a CR written to document the condition.

3.54 Whenever the reactor coolant temperature is above 160'F, at least one Reactor Coolant Pump should be in operation. (Ref. 2.60)3.55 During a normal cooldown, at least one Reactor Coolant Pump shall be operated to ensure that the temperature difference between the loops does not exceed 25°F. (Ref. 2.60)3.56 IF Hydrogen Peroxide addition is required, THEN dilution water amount used in the chemical addition SHALL be accounted for to maintain required shutdown margin.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 17 of 216 Plant Cooldown 3.57 Minimize the time at lowered RCS inventory (fuel in the reactor with level at or below the reactor vessel flange). Controls for Infrequently Performed Evolution per OPGP03-ZA-0506, Tests or Evolutions Requiring Additional Controls, and OPGPO3-ZO-0049, Conduct of Tests or Evolutions Requiring Additional Controls, SHALL be in place prior to lowering RCS level below 0% Pressurizer Cold Calibration Level elevation (elevation 52 ft 2 in) at Step 9.30.3.58 WHEN Steam Generator (SG) temperature is lowered, THEN SG narrow range level indication will indicate higher than actual level.3.59 Addendum 7 contains a list of conditions that should be met prior to taking credit for using the Steam Generators as a decay heat removal means while in Mode 5.3.60 During plant cooldown, all SGs will normally be connected to the steam header to assure a uniform cooldown of the RCS. (UFSAR 5.2.2.11.3) 3.61 The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by performing Addendum 13. MONITOR the following "MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points for indications of moisture buildup in the Main Steam Lines: 0 LD7900, S/G IA(2A) MS LN DRN FROM MS-2001* LD7901. S/G I B(2B) MS LN DRN FROM MS-2002* LD7902, S/G I C(2C) MS LN DRN FROM MS-2003* LD7903, S/G ID(2D) MS LN DRN FROM MS-2004 3.62 Deaerator Storage Tank Level SHALL be maintained in normal band of 60% to 80% when condenser vacuum is established.

Going below 60% level may affect condenser vacuum.(Ref. 2.111)3.63 The principles of OPGP03-ZO-0042, Reactivity Management Program, are in effect at all times during Operations in this procedure.

3.64 Shutdown margin SHALL be verified adequate based on the RCS boron concentration.

3.65 IF planned to place the RCS in MODE 5 with reactor coolant loops NOT filled or MODE 6 AND planned to swap the CVCS Bed Demineralizers in service during RCS in MODE 5 with reactor coolant loops NOT filled or MODE 6 THEN FLUSH the oncoming Demineralizers per OPOP02-CV-0004, Chemical and Volume Control System Subsystem PRIOR TO entering RCS in MODE 5 with reactor coolant loops NOT filled and MODE 6 conditions. (Ref 2.57)3.66 IF Personnel Air Lock (PAL) doors are open in Mode 5, THEN Addendum 21, Closure of Personnel Air Lock Doors, is available to establish containment closure.This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 18 of 216 Plant Cooldown Initials 4.0 Prerequisites NOTE This section may be performed concurrently with Step 3.0 through Step 5.28. Step 4.1 may be marked N/A if the following conditions exists: The cooldown is being performed to meet a Technical Specification LCO ACTION or using a POP09, POP04 or POP05 (EOP), and time does NOT permit step completion.

OR Plant General Manager permission is obtained AND the cooldown is for a Non-Rapid Refueling Outage or non-refueling outage where heatup to Mode 3 is to be performed prior to attempting to withdraw any Control or Shutdown Rod.OR Plant General Manager permission is obtained, AND the cooldown is for a Rapid Refueling Outage AND a cycle specific evaluation (CREE) on the risks of cooling down without performing steps to prevent control rod thermal binding is performed.

4.1 *** To prevent Control and Shutdown Rods thermal binding the Movable Rod Gripper Paws must be disengaged from the Drive Shafts, this is accomplished by performing one of the following steps prior to commencing a plant cooldown: (Ref 2.64 and 2.74)4.1.1 OPEN Reactor Trip Breakers (RTBs) and VERIFY all Control and Shutdown Rods tripped greater than 2 steps (NO Rods manually inserted).

4.1.2 Completion

of Addendum 8 for any Bank with Control or Shutdown Rods which were manually inserted.4.1.3 Completion of OPSP0O-DM-0003, Automatic Multiple Rod Drop Time Measurement.

4.2 *** To prevent Control and Shutdown Rods thermal binding the Stationary Rod Gripper Paws must be disengaged from the Drive Shafts, this is accomplished prior to commencing a plant cooldown by ensuring the Reactor Trip Breaker are open and tagged with an Equipment Clearance.

ECO #ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.This procedure., when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 19 of 216 Plant CooldownI Initials 4.3 WHEN the RH or SI systems are subject to high pressure in-leakage (i.e.Accumulator or RCS press > 250 psi), THEN PRIOR TO starting Low Head Safety Injection (LHSJ), High Head Safety Injection (HHSI), or Residual Heat Removal (RHR) pump(s), ENSURE that a remote vent path is available to vent the RHR system or a plan developed to promptly establish a remote vent path during the performance of the activity. (Ref. 2.92)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 20 of 216 Plant CooldownI Initials 5.0 Mode 3 Cooldown NOTE Boration to begin the establishment of Cold Shutdown (RCS at 681F, Xenon-Free)

Shutdown Margin concentration MAY commence without delay after Mode 3 entry, as directed by the Shift Manager/Unit Supervisor.

5.1 RECORD

the Unit, date and time this procedure was entered.Unit: Date: Time: Hrs.5.2 IF an RCS boundary opening is planned, THEN CONTACT Operations and Chemistry Management for recommendations on the RCS degas method(s) to be used. (Mechanical, Chemical or combination of both)5.3 *** IF the "HI FLUX AT SHUTDOWN" alarm has NOT been previously calibrated, THEN PERFORM the following:

5.3.1 WHEN source range counts have been stabilized, THEN NOTIFY I&C to calibrate the "HI FLUX AT SHUTDOWN" alarm to five times the stabilized count rate.5.3.2 WHEN I&C has completed calibration, THEN PLACE the "HIGH FLUX AT SHUTDOWN" alarm switch in normal.5.4 *** TEST RCB Evacuation Alarm and VERIFY audible in the RCB.NOTE IF AMSAC is bypassed, THEN "C-20 AMSAC BLOCKED" permissive is unreliable and will NOT change states. (i.e., whatever state this permissive is in when AMSAC is bypassed will be locked-in.)

5.5 PLACE

the AMSAC SYSTEM BYPASS switch in the "BYPASS" position.(ZRR054 in Relay Room Rm# 202)This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 21 of 216 Plant Cooldown I Initials NOTE IF conditions exist which prevent a Pressurizer boron sample from being taken, THEN energizing all Pressurizer heaters for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following boration should establish Pressurizer boron concentration within 50 ppm of RCS boron concentration.

5.6 NOTIFY

Chemistry to sample the RCS and Pressurizer and RECORD boron concentrations.

RCS/Time PZR/Time NOTE Normal RCS Pressure/Temperature in Mode 3 is 2235 psig/565-571°F.

5.7 ESTABLISH

conditions for a RCS boration to Cold Shutdown (RCS at 68°F, Xenon-Free)

Shutdown Margin concentration by PERFORMING the following:

5.7.1 DETERMINE

required Shutdown Margin (SDM) boron concentration for RCS temperature of 68°F Xenon-free conditions with "all rods are fully inserted" per Figure 5.5 of the Plant Curve Book.5.7.2 RECORD SDM Boron Concentration (Cb) for the RCS temperature from Step 5.7.1 (OR the selected Temperatures from Step 5.18.1. 1), THEN add 25 ppm (OPGP03-ZO-0042, Addendum 3).ppm + 25 ppm = ppm (Step 5.7.1) (+25 Cb)5.7.3 IF any rod(s) is NOT fully inserted, THEN RAISE the required SDM boron concentration in Step 5.7.2, by the following: " N/A = add zero ppm* for each rod stuck less than OR equal to 18 steps withdrawn add 60 ppm* for each rod stuck greater than 18 steps withdrawn add 228 ppm.5.7.4 Sum the additional SDM Boron Concentration due to "rod(s) NOT fully inserted" determined in Step 5.7.3 with "+ 25 Cb" from Step 5.7.2.ppm + _ ppm = ppm (Step 5.7.2) (Step 5.7.3) (+ Rods Cb)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 22 of 216 Plant Cooldown Initials 5.7.5 IF the Pressurizer boron concentration is less than (<) SDM boron concentration in Step 5.7.4 AND It is desired to begin an RCS cooldown without waiting on boron concentration to be greater than (>) SDM boron concentration in Step 5.7.4, THEN RAISE the required SDM boron concentration in Step 5.7.4, by the Additional Required Boron from the table below for the applicable Pressurizer conditions: (N/A = add zero ppm)Required SDM Boron. Current PRZR Level Additional RCS tm Boron Required (ppm) I(%) (ppm)0-500 0-30 44.8 500- 1000 0-30 89.7 1000- 1500 0-30 134.5 1500-2000 0- 30 179.3 2000-2500 1 0-30 224.2 0-500 30-60 78.5 500- 1000 30-60 157.0 1000- 1500 30-60 235.5 1500-2000 30-60 314.0 2000 -2500 30 -60 392.5 0- 500 60 -Solid 129.9 500 -1000 60 -Solid 259.8 1000- 1500 60 -Solid 389.6 1500-2000 60 -Solid 519.5 2000 -2500 60 -Solid 649.4 5.7.6 Sum the additional determined in Step SDM Boron Concentration due to "PZR" 5.7.5 with "Cb" from Step 5.7.4.(Step 5.7.5)ppm + ppm (Step 5.7.4)_ ppm (SDM Cb)This procedure, when completed, SHALL be retained.

NOTE" IF less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 2800 ppm. (Ref. 2.93)* IF greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since Reactor Shutdown, THEN RCS Boron Concentration SHALL be limited to less than 3500 ppm. (Ref. 2.93)* To ensure Pressurizer level does not hinder cooldown REFER TO Step 5.29 for Pressurizer level control.5.7.7 COMMENCE the BORATION of the RCS to a Boron Concentration greater than the RCS SDM Boron Concentration (Cb) recorded in Step 5.7.6.5.7.8 WHEN the RCS SDM Boron Concentration (Cb) recorded in Step 5.7.6 is obtained, THEN record boron concentrations in Step 5.18.2.5.8 INITIATE Pressurizer spray by placing at least two groups of Pressurizer backup heaters in the ON position. (Ref 2.40)NOTE" IF conditions exist which prevent a Pressurizer boron sample from being taken, THEN energizing all Pressurizer heaters for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following boration should establish Pressurizer boron concentration within 50 ppm of RCS boron concentration." (Good Operational practice)

Once per 30 minutes recommended sample frequency until RCS SDM Boron Concentration (Cb) recorded in Step 5.7.6 is obtained (Frequency of sampling at Shift Manager/Unit Supervisor's discretion).

• (Good Operational practice)

Once per hour recommended sample frequency during RCS cooldown if boric acid additions are being performed using manual makeup to VCT (Frequency of sampling at Shift Manager/Unit Supervisor's discretion).

5.9 NOTIFY

Chemistry to begin periodic sampling of the RCS and Pressurizer for boron concentrations.

5.10 *** ENSURE the CVCS demineralizer selected by Chemistry in service, per OPOP02-CV-0004, Chemical and Volume Control System Subsystem.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 24 of'216 Plant Cooldown Initials NOTE Additional Letdown orifices MAY be placed in service during the cooldown to maintain letdown flow.5.11 ESTABLISH maximum letdown flow per OPOP02-CV-0004, Chemical and Volume Control System Subsystem.

NOTE IF evaluation of SG hideout return is necessary, THEN SG blowdown should remain established as long as possible on the targeted SG(s).5.12 *** NOTIFY Chemistry to evaluate for SG hideout return.5.13

• NOTIFY SED that the plant is in Mode 3 to allow SED to determine if OPGP03-ZE-0033, RCS Pressure Boundary Inspection For Boric Acid Leaks, is required to be performed. (Ref 2.7)NOTE* Performance of a Cooldown MAY commence while degassing is in progress as long as the RCS SDM Boron Concentration (Cb) requirements are satisfied in accordance with Step 5.18.Mechanical degassing of RCS should be done with RCS pressure as high as allowable to promote inventory turn over.IF RCS, PRT or RCDT degassing is delayed, THEN critical path timeline for RCS depressurization may be impacted while waiting on the degassing to be completed.

Steps 5.14 and 5.15 may be performed concurrently.

CAUTION IF an RCS boundary opening is planned, THEN ENSURE the RCS is degassed as necessary to lower the RCS to less than 4 cc/kg Hydrogen and VCT Hydrogen to less than 4%.5.14

• IF performing a Mechanical degas (Step 3.46), THEN COMMENCE degassing of the RCS per Addendum 5.5.15 *** COMMENCE degassing the RCDT and PRT per Addendum 6.This procedure, when completed, SHALL be retained.I OPOP03-ZG-0007 Rev. 71 Page 25 of 216 Plant Cooldown Initials 5.16 *** IF a RCS boundary opening is planned, THEN NOTIFY Chemistry to sample for H 2 Concentration of the RCS/VCT as required.5.17 IF Analog Channel Operational Testing has NOT been completed within the previous 31 days, THEN NOTIFY I&C to perform the following Analog Channel Operational Testing procedures for the PORV channels: OPSPO2-RC-0403, RCS COMS T HOT Set 2 ACOT (P-0403, T-0413, T-0423, T-0433, T-0443) Performed Date / Time OPSP02-RC-0404, RCS COMS T COLD Set 3 ACOT (P-0404, T-0414, T-0424, T-0434, T-0444) Performed Date / Time This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 26 of 216 Plant Cooldown Initials CAUTION Cooldown SHALL NOT Commence until RCS SDM Boron Concentration (Cb)requirements are satisfied in accordance with the following Step.*Enhancing an OUTAGE schedule is NOT a justification for "time does NOT permit obtaining the RCS SDM Boron Concentration".

5.18 ENSURE the RCS SDM Boron Concentration (Cb) requirements to begin a Cooldown are completed as follows: 5.18.1 IF performing a TS LCO directed Cooldown AND time does NOT permit obtaining the Cold Shutdown (RCS at 68°F, Xenon-Free)

Shutdown Margin Boron Concentration of Step 5.7.6, THEN PERFORM the following:

5.18.1.1 DETERMINE and RECORD required SDM boron concentration for RCS Temperature and SDM Boron Concentration that will be used for each Temperature plateau, THEN add 25 ppm.RCS Tempi __RCS Temp 2 RCS Temp 3 __RCS Temp 4 __RCS Temp 5__SDM Cb ppm + 25 ppm = _ ppm SDM Cb ppm + 25 ppm = ppm SDM Cb ppm + 25 ppm = -ppm SDM Cb ppm + 25 ppm = -ppm SDM Cb ppm + 25 ppm = ppm 5.18.1.2 Using Step 5.7.3 as a guide, add applicable boron for"rod(s) NOT fully inserted".

RCS Temp, __RCS Temp 2 __RCS Temp 3 __RCS Temp 4 RCS Temp 5 Cb ppm + Rods ppm = ppm Cb ppm + Rods ppm = ppm Cb ppm + Rods ppm = ppm Cb ppm + Rods ppm = ppm Cb ppm + Rods ppm = ppm 5.18.1.3 Using Step 5.7.5 as a guide, add applicable boron due to "PZR".RCS Temp, __RCS Temp 2 __RCS Temp 3 __RCS Temp 4 RCS Temp 5 Cb ppm + PZR ppm = ppm Cb ppm + PZR ppm = ppm Cb ppm + PZR ppm = ppm Cb ppm + PZR ppm = ppm Cb ppm + PZR ppm = .ppm This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 27 of 216 Plant Cooldown I Initials NOTE The following Step 5.18.1.4 MAY be completed out of the numbered sequence and as specified in the Step.WHEN the RCS is BORATED to greater than the SDM Cb of Step 5.7.6, THEN complete Step 5.18.2 and mark the remaining Steps of 5.18.1.1 -5.18.1.4 N/A.5.18.1.4 ENSURE the RCS is BORATED to greater than the Shutdown Margin boron concentration for the RCS Temperatures recorded in Step 5.18.1.3, prior to a Cooldown to the associated Temperature plateau.RCS Tempi __RCS Cb > SDM Cb Time RCS Temp 2__RCS Cb > SDM Cb Time RCS Temp 3__RCS Cb> SDM Cb Time Perform Ind. Verif Perform Ind. Verif Perform Ind. Verif Perform Ind. Verif Perform Ind. Verif RCS Temp 4 __RCS Temp 5__RCS Cb > SDM Cb RCS Cb > SDM Cb Time Time This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 28 of 216 Plant Cooldown Initials NOTE Step 5.18.2 SHALL NOT be marked N/A AND SHALL always be completed, as soon as possible and within the limitation specified in Step 5.18.1.4. (Ref 2.83)5.18.2 ENSURE the RCS Shutdown Margin Concentration (Cb) requirements of Step 5.7.6 are completed, THEN RECORD boron concentrations.

RCS/Time PZR/Time 5.18.3 IF opening of the RCS is planned AND a Hydrogen Peroxide addition is required, THEN CONTINUE to borate the RCS an additional 10 ppm to ensure sufficient margin is available in RCS Shutdown Margin Concentration (Cb) requirements in addition to Step 5.7.6 to account for the dilution water used to perform Hydrogen Peroxide addition.ppm + 10 ppm-(Step 5.7.6) (ppm Required RCS Cb)5.18.4 IF Rapid Refueling is planned, THEN CONTINUE to borate the RCS to SHUTDOWN MARGIN LIMIT CURVE Plant Curve Book Figure 5.5, Mode 5 Cb, ARO, 687F to 2007F, K <= 0.95. (Ref 2.35)Required RCS Cb ppm 5.18.5 IF Rapid Refueling is planned AND a Hydrogen Peroxide addition is required, THEN CONTINUE to borate the RCS an additional 10 ppm to ensure sufficient margin is available in addition to SHUTDOWN MARGIN LIMIT CURVE Plant Curve Book Figure 5.5, Mode 5 Cb, ARO, 68°F to 200'F, K <= 0.95 to account for the dilution water used to perform Hydrogen Peroxide addition. (Ref 2.35)ppm + 10 ppm =(Step 5.18.4)__ ppm (Required RCS Cb)This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 29 of216 Plant Cooldown Initials NOTE If an RHT will be used as an RCS fill source per OPOP02-RC-0003, Filling and Venting the Reactor Coolant System, or OPOP03-RC-0100, RCS Vacuum Fill, calculation NC-07090 requires VCT (50% level) to have Cb greater than 2875 ppm to account for any dead legs of unknown boron concentration before RHT of greater than 2800 ppm is aligned to the VCT.Borating the RCS to greater than 2875 ppm on shutdown will ensure RCS and VCT are ready to receive RHT water upon refill. (Refs. 2.108, 2.109 & 3.13)5.18.6 IF the following is planned, THEN CONTINUE to borate the RCS to Boron Concentration (Cb) of greater than 2875 ppm (Refs. 2.108, 2.109 & 3.13):* Entry into Mode 5b (Mode 5 with reactor coolant loops NOT filled) OR Mode 6" RHT(s) will be used as an RCS fill source in OPOP02-RC-0003, Filling and Venting the Reactor Coolant System or 0POP03-RC-0 100, RCS Vacuum Fill.5.18.7 IF an RCP will remain inservice

[no planned entry into Mode 5b (Mode 5 with reactor coolant loops NOT filled) OR Mode 6], THEN PLACE Caution Tags on all four RCP handswitches stating Boron Concentration for entry in Mode 5b (Mode 5 with reactor coolant loops NOT filled) OR Mode 6 should be confirmed prior to securing RCPs.ECO #5.18.7.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.5.18.8 Commence Form 1, CVCS Line Boration Tracking Form.I This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 30 of 216 Plant Cooldown Initials CAUTION (Normal Cooldown or TS LCO directed Cooldown with time) RCS temperature SHALL NOT be allowed to lower to 540'F prior to blocking Safety Injection to prevent a Safety Injection Actuation.(Normal Cooldown or TS LCO directed Cooldown with time) RCS boration to RCS Shutdown Margin Concentration (Cb) requirements of Step 5.7.6 SHALL be completed prior to blocking Safety Injection.

• IF any Steam Generator is "dry" or expected to steam "dry", THEN a Normal Cooldown or TS LCO directed Cooldown SHALL NOT commence until the RCS Cold Shutdown (RCS at 68°F, Xenon-Free)

Shutdown Margin concentration requirements are satisfied.

REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown 5.19 PLACE Pressurizer Master pressure controller "PRESS CONT PK-0655A" in the MANUAL position.5.20 Using the Pressurizer Spray Valve Controller(s) "PRZR SPR PCV-0655B and/or PCV-0655C", REDUCE RCS pressure to between 1900 and 1950 psig.5.21 MONITOR RCS and Pressurizer temperature and pressure per Data Sheet 1.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 31 of 216 Plant Cooldown Initials 5.22 WHEN RCS pressure is less than 1985 psig (P-I 1), THEN PERFORM the following:

5.22.1 ENSURE RCS Shutdown Margin Concentration (Cb) requirements are met by performing the following:

5.22.1.1 ENSURE one of the following:

a. (Normal Cooldown or TS LCO directed Cooldown with time) ENSURE the RCS boration to RCS Shutdown Margin Concentration (Cb) requirements of Step 5.7.6 concentration completed. (Ref 2.83), OR b. (TS LCO directed Cooldown AND time does NOT permit obtaining the RCS SDM Boron Concentration of Step 5.22.1. 1) ENSURE the RCS boration to RCS Shutdown Margin Concentration (Cb) requirements of Step 5.7.6 are completed, as soon as possible AND the RCS Boron Concentration (Cb) meets the requirements of Step 5.18.1.4.5.22.1.2 PERFORM OPSP I O-ZG-0003, Shutdown Margin Verification

-Modes 3, 4 and 5, for RCS 68°F, xenon free conditions.

5.22.2 BLOCK Pressurizer pressure safety injection signal by taking both Train R and Train S block switches to the BLOCK position.

{CP005}"PRZR PRESS SI TRAIN R""PRZR PRESS S1 TRAIN S" 5.22.3' VERIFY both Train R and Train S Pressurizer pressure safety injection block lamp boxes illuminated.

{CP005}"PRZR PRESS SI BLOCKED TRAIN R" ("PRZR SI BLOCKED TRAIN R")* "PRZR PRESS SI BLOCKED TRAIN S" ("PRZR S1 BLOCKED TRAIN S")This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 32 of216 Plant Cooldown Initials 5.22.4 BLOCK low compensated steam line pressure safety injection signal by taking both Train R and Train S block switches to the BLOCK position.

{CP005}* "LO STM LN PRESS SI TRAIN R""LO STM LN PRESS SI TRAIN S" 5.22.5 VERIFY both Train R and Train S low compensated steam line pressure injection block lamp boxes illuminated.

{CP005}"LO STM LN PRESS SI BLOCKED TRAIN R"* "LO STM LN PRESS SI BLOCKED TRAIN S" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 33 of 216 Plant Cooldown I Initials NOTE* The RCP configuration recommended for RCS Cooldown conditions are as follows:* (Preferred)

RCP 1D(2D) [RCP I D(2D) is the only pump with adequate spray flow in single pump configuration]

• (Alt 1) RCP 1D(2D) and RCP 1A(2A)* (Alt 2) RCP 1 D(2D) and RCP 1 B(2B) or RCP 1 C(2C)* (Alt 3) RCP 1 A(2A) and RCP 1 B(2B) or RCP 1 C(2C)* (Alt 4) RCP 1B(2B) and RCP 1C(2C)* MAXIMIZING the Reactor Coolant System flowrate during CRUD Cleanup may reduce DOSE rates for the remainder of the Outage due to faster cleanup time.* MAXIMIZING the Reactor Coolant System flowrate during cooldown may raise DOSE rates for the remainder of the Outage due to large CRUD burst.* For Planned Outages, the RCP configuration recommended by the Outage Planning Team should be considered; the Shift Manager/Unit Supervisor has final selection judgment.* For Un-Planned Outages, the RCP configuration recommended by Health Physics and Chemistry should be considered; the Shift Manager/Unit Supervisor has final selection judgment.* IF NO RCP configuration recommendation is provided, THEN the default RCP configuration is RCP ID(2D) running and 3 RCPs secured, the Shift Manager/Unit Supervisor has final selection judgment.* The Shift Manager/Unit Supervisor may authorize changes to the operating RCP configuration during the cooldown if conditions warrant (i.e. large CRUD burst, pump problems, cleanup issues, etc)." REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown.

5.23 ESTABLISH the operating RCP(s) configuration for the cooldown, by PERFORMING the following:

5.23.1 IF Planned Outage, THEN CONTACT Outage Planning Team for RCP configuration for cooldown recommendation.

5.23.2 CONTACT Chemistry for RCP configuration for cooldown recommendation and to verify that RCS activity level (crud) is acceptable and full flow clean-tip NOT required.5.23.3 CONTACT Health Physics for RCP configuration for cooldown recommendation.

This procedure, when completed, SHALL be retained.

NOTE IF an RCP is secured, THEN ensure that the RCP is allowed to coast down prior to start or stop of another RCP.5.23.4 SECURE or START the applicable RCPs per OPOP02-RC-0004, Operation of Reactor Coolant Pump as determined by the Shift Manager/Unit Supervisor.

{CP005}5.23.5 NOTIFY Health Physics to monitor for rising area radiation due to CRUD burst during cooldown.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 35 of 216 Plant Cooldown I Initials NOTE The cooldown should be spaced over the entire 60-minute period vice a short time, high rate cooldown. (Ref 2.50)WHEN RCS temperature is greater than 340'F, THEN the preferred method of feeding SGs is with S/U SGFP 14(24) per 0POP02-FW-0001, Main Feedwater.

Whenever possible, steam dumps should be used to steam the SGs.SGs SHALL NOT be fed with the Main Feedwater System until RCS temperature is greater than 340'F. (Ref 2.85)AFW can be used as an alternative to S/U SGFP 14(24). IF AFW is being used, THEN SG PORVs should be the method used for cooling whenever possible.Using AFW and steam dumps WILL require periodic monitoring of hotwell level to ensure proper level control operation.

Use of S/U SGFP 14(24) while steaming through the SG PORVs will require periodic monitoring of hotwell level to ensure proper level control operation.

The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by performing Addendum 13.MONITOR the following "MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points for indications of moisture buildup in the Main Steam Lines:* LD7900, S/G 1A(2A) MS LN DRN FROM MS-2001* LD7901, S/G 1B(2B) MS LN DRN FROM MS-2002* LD7902, S/G IC(2C) MS LN DRN FROM MS-2003* LD7903, S/G 1D(2D) MS LN DRN FROM MS-2004 5.24 *** ENSURE backseated valves are removed from backseated position AND ECO tags removed prior to plant cooldown.

This is preferred but NOT required if the cooldown is required by Technical Specifications.

5,25 *** IF cooldown is desired using Steam Dumps AND the MSIVs are closed, THEN INITIATE warm up of Main Steam lines per OPOP03-ZG-0003, Secondary Plant Startup or OPOP03-ZG-00 11, Secondary Plant Cold Startup.(Ref 2.54)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 36 of 216 Plant Cooldown I Initials NOTE Thermal binding of the closed CVCS CHARGING TO RCS LOOP COLD LEG MOV during the cooldown is probable and valve motor has NO protective thermal overloads.

5.26 *** PERFORM the following to the NOT In-Service CVCS CHARGING TO RCS LOOP COLD LEG MOV OPERATOR, (the closed valve)CV-MOV-0006 or CV-MOV-0003:

5.26.1 De-Energize the MOV and hang ECO on supply breaker to prevent Valve Operator Motor Damage. ECO #_5.26.2 CAUTION tag Control Room handswitch to identify "Local Manual operation is required prior to re-energizing the Valve Operator to prevent Valve Motor damage from thermal binding of valve".ECO #5.26.3 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.5.27 *** PLACE a Caution Tag on the following valves reading "Ensure the differential pressure between the charging header and the RCS is less than 2500 psid prior to closing this MOV OR Secure the charging pump OR Close CV-MOV-0025 prior to closing CV-MOV-0003 or CV-MOV-0006." (Ref 2.68)ECO #* "LOOP A ISOL MOV-0003" (NORM)* "LOOP C ISOL MOV-0006" (ALT)* ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.5.28 VERIFY Prerequisites Section 4.0 is complete.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 37 of 216 Plant Cooldown Initials CAUTION* IF Pressurizer level exceeds 55% (Hot Cal), THEN IMMEDIATE Operator action required to reduce level to Less Than 55% (Hot Cal) is authorized.

• IF performing Pressurizer Level control using manual FCV-0205 control (Step 5.29, 3 rd bullet), THEN Charging Pump Operation is restricted to one (1) Charging Pump while level is above program level.NOTE Step 5.29 may be performed as directed by the Unit Supervisor at any time during the cooldown to control Pressurizer level.* Manual control of Pressurizer Level Requires heightened Operator attention to prevent RCS Pressure/Level transients.

A Reactor Operator SHALL be assigned responsibility of controlling Primary Plant Parameters.

• Establish pressurizer level between 45-55% at the start of the cooldown and the level may be allowed to lower to program over the duration of the cooldown.

Once the level reaches program a second charging pump may be started to maintain pressurizer level to support the continued cooldown of the RCS until Mode 4 in which only one charging pump is allowed to be in-service.

• REFER TO Step 3.9.5.29 ENSURE Pressurizer level is being maintained by one of the following:

• CVCS is in AUTO and maintaining level at or near Program Level* CVCS is in MANUAL and Operator is MAINTAINING level at or near Program Level* IF instructed by the Unit Supervisor, THEN PLACE "CHG FLOW CONT VLV FCV-0205" in MANUAL and CONTROL Pressurizer level between 45% and 55% (Hot Cal).This procedure, when completed, SHALL be retained.

I 0OPOP03-ZG-0007 Rev. 71 Page 38 of216 Plant Cooldown Initials CAUTION* Rapid over cooling of the RCS will cause a RAPID lowering in PZR level. Cooldown rates should be established slowly while ENSURING that the PZR level is controllable." The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by performing Addendum 13. MONITOR the following "MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points for indications of moisture buildup in the Main Steam Lines:* LD7900, S/G IA(2A) MS LN DRN FROM MS-2001* LD7901, S/G IB(2B) MS LN DRN FROM MS-2002* LD7902, S/G IC(2C) MS LN DRN FROM MS-2003* LD7903, S/G 1D(2D) MS LN DRN FROM MS-2004 5.30 ESTABLISH cooldown using the steam dumps {CP007} or SG PORVs.{CP006}5.31 MAINTAIN SG Narrow Range levels between 55 and 75%.CAUTION Pressurizer pressure SHALL NOT be allowed to drop to less than 1900 psig until the Pressurizer pressure safety injection signal has been blocked per Step 5.22.1-5.32 ENSURE Safety Injection blocked per Step 5.22.5.33 ENSURE 50 to 100°F subcooling is maintained continuously while maintaining RCS pressure between 1000 and 1900 psig during RCS cooldown.5.34 WHEN RCS Tavg is 5637F (P-12) AND the steam dumps are being used for RCS cooldown, THEN momentarily PLACE steam dump "TNTLK SEL" switches in the BYPASS INTERLCK position.

{CP007}* Train A "INTLK SEL"* Train B "JNTLK SEL" 5.35 ADJUST cooldown rate as required to maintain a differential temperature of less than 25°F between the RCS loops.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 39 of 216 Plant Cooldown I Initials 5.36 MAINTAIN RCP seal injection flow to each RCP between 8 and 13 gpm.NOTE REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown.

Static pressure zero and span shifts on the RCS flow transmitters will affect flow indication when the RCS is NOT at normal operating pressure.

When the RCS is at low pressure with no RCP running, there may be RCS flow indication up to 25% flow even when there is no actual flow through the loop. This is an expected indication that occurs due to static pressure effect on the flow transmitters measuring differential pressure across elbow taps on the RCS piping. (Reference 2.98)5.37 WHEN RCS temperature is between 440 and 450'F, THEN LOWER RCS pressure to between 900 and 1000 psig.5.38 CLOSE safety injection accumulator discharge valves as follows: 5.38.1 CLOSE accumulator discharge valve breakers using the power lockout switches.

{CPOO1 }* "ACC 1A(2A) PWR LOCKOUT MOV-0039A"* "ACC 1B(2B) PWR LOCKOUT MOV-0039B"* "ACC IC(2C) PWR LOCKOUT MOV-0039C" 5.38.2 CLOSE accumulator discharge valves. {CP001 }* "ACC IA(2A) DISCH ISOL MOV-0039A"* "ACC 1B(2B) DISCH ISOL MOV-0039B"* "ACC 1C(2C) DISCH ISOL MOV-0039C" 5.38.3 OPEN accumulator discharge valve breakers using the power lockout switches.

{CP001 }* "ACC 1A(2A) PWR LOCKOUT MOV-0039A"* "ACC 1B(2B) PWR LOCKOUT MOV-0039B1"* "ACC 1C(2C) PWR LOCKOUT MOV-0039C" 5.39 MAINTAIN RCP seal injection flow to each RCP between 8 and 13 gpm.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 40 of 216 E Plant Cooldown Initials NOTE OPOP02-RC-0004, Operation of Reactor Coolant Pump, Addendum I may be referenced to ensure that Number 1 seal leak off values for the RCPs are within acceptable limits.REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown 5.40 WHEN Steps 5.37 to 5.39 are complete, THEN LOWER RCS pressure to between 575 and 625 psig.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 41 of216 Plant Cooldown Initials NOTE* Step 5.41 may be performed concurrently with the remainder of this procedure as determined by the Unit Supervisor, but must be performed prior to Mode 5 entry.* Step 5.41 provides the following: " If RCP seal differential pressure is low it may be desirable to realign seal return to the VCT to maintain RCP seal DP." Required by Calculation NC-07090, Evaluation of Boron Dilution Flow Paths in Mode 5b (Mode 5 with reactor coolant loops NOT filled) and Mode 6, to prevent low boron water from being introduced into the RCS. Required to flush seal return line with 58 gallons of RCS Cb concentration water, however not required to be maintained once line is flushed. (Refs. 2.108, 2.109 & 3.13)o To ensure Seal Return line is flushed adequately ensure that seal return is lined up to VCT based on the following:

1. Add RCP Seal Leakoffflowrate from all four RCPs to get a total Seal Leakoffflowrate (example FT-0156 = 1.5 gpm, FT-0157 = 1.5 gpm, FT-0158 = 1.5 gpm, FT-0159 =1.5 gpm for a total Seal Leakoff flowrate of 6 gpm).2. 58 gallons divided by total Seal Leakoffflowrate to give time needed to flush 58 gallons (example 58 gallons / 6 gpm = 9.66 minutes)3. Double time of flush for conservatism (9.66 x 2 = 19.32 minutes).5.41 ALIGN 2.109): seal return to the VCT by performing the following (Refs. 2.108 &5.41.1 IF Entry into Mode 5b (Mode 5 with reactor coolant loops NOT filled) OR Mode 6 is planned, THEN ENSURE RCS Boron Concentration (Cb) greater than 2875 ppm. (Ref. Step 5.18.6)5.41.2 OPEN "1(2)-CV-0171 CVCS SEAL RETURN VCT ISOL".(19 ft MAB High Energy Valve Room 80)5.41.3 CLOSE "1(2)-CV-0170 CVCS SEAL RETURN TO CHARGING PUMP SUCTION ISOL". (19 ft MAB High Energy Valve Room 80)5.41.4 IF desirable to continue Operation with the RCP seal return aligned to the VCT, THEN Caution Tag the CCP hand switch on Control Panel CP004 to identify the alignment.

ECO#5.41.4.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 42 of216 Plant Cooldown Initials 5.41.5 IF desirable to ALIGN Seal Return to Charging Pump Suction, THEN PERFORM the following:

5.41.5.1 ENSURE adequate time has been allowed to flush 58 gallons through Seal Return to VCT.5.41.5.2 OPEN "1(2)-CV-0170 CVCS SEAL RETURN TO CHARGING PUMP SUCTION ISOL".(19 ft MAB High Energy Valve Room 80)5.41.5.3 CLOSE "1(2)-CV-0171 CVCS SEAL RETURN VCT ISOL". (19 ft MAB High Energy Valve Room 80)NOTE The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by perfonring Addendum 13. MONITOR the following-MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points for indications of moisture buildup in the Main Steam Lines:* LD7900, S/G 1A(2A) MS LN DRN FROM MS-2001* LD7901, S/G IB(2B) MS LN DRN FROM MS-2002* LD7902, S/G 1C(2C) MS LN DRN FROM MS-2003* LD7903, S/G I D(2D) MS LN DRN FROM MS-2004 5.42 LOWER RCS temperature to between 350'F and 3557F using the steam dump valves {CP007} or SG PORVs. {CP006}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page43of216 Plant Cooldown Initials NOTE REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown 5.43 WHEN RCS temperature is between 3507F and 3557F, THEN PERFORM the following:

5.43.1 LOWER RCS pressure 325 psig to 425 psig.5.43.2 MAINTAIN RCP seal differential pressure greater than or equal to 230 psid.* "RCP 1A(2A) SEAL I DP PI-0152"* "RCP 1B(2B) SEAL I DP P1-0153"* "RCP 1C(2C) SEAL I DP PI-0154"* "RCP 1D(2D) SEAL 1 DP PI-0 155" NOTE IF a RCS boundary opening is planned, THEN RCS H, concentration should be reduced to less than 4 cc/kg prior to opening the RCS.5.44 IF a RCS boundary opening is planned, THEN NOTIFY Chemistry to sample the RCS for H 2 concentration.

5.45 ENSURE Analog Channel Operational Testing completed for the PORV channels within the previous 31 days.OPSP02-RC-0403, RCS COMS T HOT Set 2 ACOT (P-0403, T-0413, T-0423, T-0433, T-0443) Performed Date/Ti me OPSP02-RC-0404, RCS COMS T COLD Set 3 ACOT (P-0404, T-0414, T-0424,T-0434, T-0444) Performed Date/Time 5.46 ENSURE OPSP03-RC-0010, Pressurizer Power Operated Relief Valve Operability Test, is current (i.e., has been performed within the past 3 months).(Ref 2.70) Performed Date/Time This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page44of216 Plant Cooldown Initials 5.47 NOTIFY Chemistry to sample the RCS and Pressurizer and RECORD boron concentrations.

RCS/Time PZR/Time 5.48 ENSURE the Shutdown Margin requirements of Step 5.7.6 {RCS boration to Cold Shutdown (RCS at 68°F, Xenon-Free) concentration) completed. (Ref.83)5.49 RECORD SDM Boron Concentration that will be used from Step 5.47: ppm This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 45 of 216 Plant Cooldown Initials CAUTION PRIOR to entry into Mode 4 the RCS SHALL be borated to the Cold Shutdown (RCS at 68°F, Xenon-free) condition.

6.0 Mode 4 Cooldown 6.1 *** NOTIFY Security to establish the Shutdown Modes Protective Strategy for Unit 1(2).6.2 *** ENSURE the following RCP seal "STANDPIPE FILL" valves are LOCKED CLOSED: {RCB RHR HX Rooms}* "1 (2)-RH-0171"* "I (2)-RH-0172"* "l(2)-RH-0173" 6.3 *** ENSURE Equipment Clearance Order hung for "STANDPIPE FILL" valves in Step 6.2. ECO #6.3.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.NOTE SSPS Actuation Cabinets "A" and "B" and Logic Cabinets "R" and "S" must be energized for both COMS to be OPERABLE.

Logic Cabinet "R" and Actuation Cabinet "A" required for PCV 0655A "Pressurizer PORV Train A". Logic Cabinet "S" and Actuation Cabinet "B" required for PCV 0656A "Pressurizer PORV Train B". {LER 2-05-003, SSPS ECO to support FWIV energize to actuate MOD in 2RE10 caused COMS to be inoperable during solid plant operations in violation of TS 3.4.9.3 (CR 05-5960, CR 05-3071)).

6.4 ARM the Cold Overpressure Mitigation System.(Technical Specification 3.4.9.3) {CP004}* "OVERPRESS MIT" for PCV-0655A* "OVERPRESS MIT" for PCV-0656A This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 46 of 216 Plant Cooldown I Initials NOTE Perform Non-Intrusive Check Valve Testing for AF system valves as determined by the Shift Manager/Unit Supervisor and System Engineer per OPEP07-ZE-0008.

• SGs SHALL NOT be fed with the Main Feedwater System, WHEN the RCS temperature is less than 340'F. (Ref 2.85)* The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by performing Addendum 13. MONITOR the following"MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points for indications of moisture buildup in the Main Steam Lines:* LD7900, S/G 1A(2A) MS LN DRN FROM MS-2001* LD7901, S/G 1B(2B) MS LN DRN FROM MS-2002* LD7902, S/G 1C(2C) MS LN DRN FROM MS-2003* LD7903, S/G 1D(2D) MS LN DRN FROM MS-2004 6.5 IF SG levels are being maintained by S/U SGFP 14(24), THEN TRANSFER feed to AFW as follows: 6.5.1 ESTABLISH feed to SGs per OPOP02-AF-0001, Auxiliary Feedwater.

6.5.2 USE SG PORVs (preferred) and/or Steam Dumps to steam the SGs.{CP006}6.5.3 CLOSE low power feedwater regulation valves to isolate S/U SGFP 14(24) flow to the SGs. {CP006}* SG 1A(2A) "LOW PWR FV-715 1"* SG 1B(2B) "LOW PWR FV-7152"* SG IC(2C) "LOW PWR FV-7153"* SG ID(2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 47 of 216 Plant Cooldown Initials 6.5.4 CLOSE preheater bypass valves. {CP006}* SG 1A(2A) "PREHTR BYPASS FV-7189"* SG I1B(2B) "PREHTR BYPASS FV-7190"* SG IC(2C) "PREHTR BYPASS FV-7191"* SG 1D(2D) "PREHTR BYPASS FV-7192" 6.5.5 MAINTAIN SG Narrow Range levels between 55 and 75%.{CP006}CAUTION Steps 6.6 through 6.11 SHALL be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering Mode 4 and prior to one or more RCS cold leg temperatures lowering below 325°F.(Technical Specifications 3.5.3.1 and 3.4.9.3)NOTE Tile Administrative Cooldown Rate is 80°F/hr. (Ref 2.70)The cooldown should be spaced over the entire 60 minute period vice a short time high rate cooldown. (Ref 2.50)Mode 4 is entered when 2 of the 4 RCS average temperature indications are LESS THAN 350 0 F.6.6 LOWER RCS temperature to between 3301F and 349°F.6.7 RECORD the Unit, date and time the plant entered Mode 4.Unit: Date: Time: Hrs.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 48 of216 Plant Cooldown I___Initials NOTE Steps 6.8 through 6.12 may be performed concurrently.

6.8 DISABLE

positive displacement charging pump by opening and tagging breaker "POS DISP CHG PMP I G8(2G8)/N2" with an Equipment Clearance Order. (Technical Specification 3.4.9.3)ECO #6.8.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.6.9 DISABLE one "CENT CHARGING PUMP" by racking out and tagging the associated pump breaker with an Equipment Clearance Order.(Technical Specification 3.4.9.3)ECO #* "IA(2A) E1C(E2C)/4"* "IB(2B) E1A(E2A)/9" ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.6.10 DISABLE one "HIGH HEAD SAFETY INJECTION PUMP" by racking out and tagging the associated pump breaker with an Equipment Clearance Order.(Technical Specification 3.5.3.1) ECO #* "IA(2A) E1 A(E2A)/5" S "I B(2B) E1B(E2B)/5" S "1 C(2C) E 1 C(E2C)/5" ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 49 of216 Plant Cooldown Initials 6.11 PLACE a second "HIGH HEAD SAFETY INJECTION PUMP" in Standby by racking out and tagging the associated pump breaker with a Caution Tag.(Technical Specification 3.5.3.1) ECO #* "IA(2A) E1A(E2A)/5""IB(2B) E1B(E2B)/5""1C(2C) E1C(E2C)/5" ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.6.12 PERFORM Addendum 16, Throttling "I(2)-CV-0198 RMW ISOL" for Technical Specification 3.4.1.3. ECO #6.12.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 50 of 216 Plant Cooldown Initials CAUTION Two trains of LHSI SHALL remain operable in Mode 4. It is permissible to consider a LHSI train OPERABLE, if the only impact to its operability is the associated RHR train in service.(TS 3.5.3.1)I NOTE* Plant cooldown may continue using the steam dumps or SG PORVs while performing Steps 6.13 through 6.16. Step 6.5 may be performed concurrent with Steps 6.13 through 6.16.* The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by performing Addendum 13. MONITOR the following "MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points for indications of moisture buildup in the Main Steam Lines:* LD7900, S/G 1A(2A) MS LN DRN FROM MS-2001* LD7901, S/G IB(2B) MS LN DRN FROM MS-2002* LD7902, S/G 1C(2C) MS LN DRN FROM MS-2003* LD7903, S/G ID(2D) MS LN DRN FROM MS-2004* REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown 6.13 WHEN RCS temperature is between 330'F and 349°F, THEN PERFORM the following:

6.13.1 LOWER RCS pressure 325 psig to 350 psig.6.13.2 MAINTAIN RCP seal differential pressure greater than or equal to 230 psid.* "RCP 1A(2A) SEAL I DP P1-0152" 0 "RCP 1B(2B) SEAL 1 DP PI-0153"* "RCP 1C(2C) SEAL 1 DP PI-0154"* "RCP ID(2D) SEAL I DP PI-0155" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 51 of 216 Plant Cooldown I Initials NOTE Perform Non-Intrusive Check Valve Testing for RH and SI system valves as determined by the Shift Manager/Unit Supervisor and System Engineer per OPEP07-ZE-0008.

Step 6.12 SHALL be completed prior to performance of Step 6.14.6.14 PLACE RHR Train A or B in service per OPOP02-RH-0001, Residual Heat Removal System Operation.

• RHR Train A* RHR Train B 6.15 PLACE low pressure letdown in service per OPOP02-CV-0004, Chemical and Volume Control System Subsystem.

6.16 IF any idle RHR train boron concentration has NOT been equalized with RCS boron concentration, THEN PLACE idle RHR trains in service per OPOP02-RH-0001, Residual Heat Removal System Operation, to equalize boron concentration.

• RHR Train A* RHR Train B* RHR Train C NOTE REFER TO Addendum 11, RCS/PZR Pressure Operations Guideline for general techniques for control of RCS/PZR Pressure during heatup/cooldown 6.17 WHEN available RHR train suction valves have been opened, THEN MAINTAIN RCS Pressure between 325 psig and 400 psig.This procedure, when completed, SHALL be retained.

I 7POP03-ZG-0007 Rev. 71 Page 52 of 216 Plant Cooldown Initials CAUTION Use of a pressurizer steam bubble during periods of low pressure, low temperature operation is preferred.

This steam bubble will dampen the plant's response to potential transient generating inputs, providing easier pressure control with slower response rates. (UFSAR 5.2.2.11.3)

NOTE Pressurizer level (Cold Cal) should be used as the primary Pressurizer level indicator through out the remaining cooldown.

Other indications (RHT level rising, Pressurizer hot calibrated level tracking) should also be used so as not to rely on only one instrument. (Ref. 2.102)Addendum 3, Determination of RCS Volume to be Drained, MAY be referred to VERIFY actual volume drained correlates with expected volume drained. (Ref 2.79)I 6.18 ENSURE the Pressurizer program level indication selector handswitch "PZR PROG LEVEL" selected to the COLD CAL LI-0675 position.6.19 IF directed by Operations Management to take the PZR Water Solid at this point in the cooldown, THEN PERFORM Addendum 9 "Plant Cooldown with the PZR Water Solid" while continuing with this procedure.

NOTE* IF Addendum 10 is started and the MSIVs/MSIBs become closed for reasons other than instructed in Addendum 10, THEN the Shift Manager/Unit Supervisor SHALL determine the steps between 6.21 through 7.35 required to be performed.

• IF performing Addendum 10, THEN Steps 6.21 through 7.35 may be marked NA.6.20 IF directed by Operations Management to perform the plant cooldown with the MSIVs or MSIBs open into MODE 5 conditions, THEN GO TO Addendum 10,"MODE 5 Cooldown with MSIV's OPEN".NOTE Secondary plant shutdown SHALL be completed through Step 10.15 prior to placing the SGs in wet layup per Step 7.15.2.6.21 INITIATE secondary plant shutdown per Section 10.0 at the discretion of the Shift Manager.This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 53 of 216 Plant Cooldown Initials NOTE* Performance of plant cooldown may continue while degassing is in progress.Mechanical degassing of RCS should be done with RCS pressure as high as allowable to promote inventory turn over.Decay Heat removal is accomplished by a combination of RHR and SG PORVs. When shifting to RHR only, two trains of RHR may be required.6.22 CONTINUE RCS cooldown to between 240'F and 250'F by adjusting RHR HX flow WHILE performing the following:

6.22.1 Maintain RCS pressure between 325 psig and 400 psig. {CP011 6.22.2 Maintain RCP seal differential pressure greater than or equal to 230 psid.* "RCP 1A(2A) SEAL 1 DP PI-0152"* "RCP 1B(2B) SEAL I DP PI-0153"* "RCP IC(2C) SEAL 1 DP P1-0154"* "RCP ID(2D) SEAL 1 DP PI-0155" This procedure, when completed, SHALL be retained.

o0POP03-ZG-0007 Rev. 71 Page 54 of 216 Plant Cooldown Initials 6.23 WHEN RCS temperature is less than 245°F, THEN PERFORM the following:

6.23.1 ENSURE Steam Dumps are CLOSED {CP007}6.23.2 CLOSE MSIVs {CP006}* SG 1A(2A) "MSIV FSV-7414"* SG 1B(2B) "MSIV FSV-7424"* SG IC(2C) "MSIV FSV-7434"* SG ID(2D) "MSIV FSV-7444" 6.24 CONTINUE RCS cooldown by adjusting RHR HX flow until RCS temperature is between 215'F and 220'F. {CP001 }6.25 THROTTLE OPEN SG PORVs to vent the SGs. {CP006}* SG 1 A(2A) "PORV PV-7411"* SG IB(2B) "PORV PV-742 1"* SG 1 C(2C) "PORV PV-743 1"* SG 1D(2D) "PORV PV-7441" 6.26 ENSURE Step 5.41 has been performed.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 1 Page 55 of 216 Initials 7.0 Mode 5 Cooldown CAUTION Step 7.2 places the unit in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Technical Specification LCO Action Statement.

All three HHSI pumps SHALL be inoperable within the allotted time.NOTE The administrative cooldown rate is 80°F/hr. (Ref. 2.70)The cooldown should be spaced over the entire 60 minute period vice a short time high rate cooldown. (Ref. 2.50)REFER TO Shutdown Risk Assessment Report if it is desired to take credit for the Steam Generators for decay heat removal purposes in Mode 5.7.1 IF it is desired to take credit for the Steam Generators for decay heat removal purposes in Mode 5, THEN ENSURE the conditions listed in Addendum 7, Conditions for Steam Generator Decay Heat Removal, are met.NOTE Mode 5 is entered when 2 of the 4 RCS average temperature indications are LESS THAN OR EQUAL TO 200°F.7.2 CONTINUE RCS cooldown to between 1907F and 1997F WHILE performing the following:

7.2.1 MAINTAIN

RCS pressure 325 psig and 400 psig.7.2.2 MAINTAIN RCP seal differential pressure greater than or equal to 230 psid.* "RCP 1A(2A) SEAL I DP PI-0 152"* "RCP IB(2B) SEAL I DP PI-0 153"* "RCP IC(2C) SEAL 1 DP PI-0154"* "RCP 1D(2D) SEAL 1 DP PI-0155" 7.3 RECORD the Unit, date and time the plant entered Mode 5.Unit: Date: Time: Hrs.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 I Rev. 71 Page 56 of 216 Plant Cooldown Initials 7.4 DISABLE the operable and standby "HIGH HEAD SAFETY INJECTION PUMPS" by racking out and tagging the respective pump breakers with an Equipment Clearance Order. ECO #" "IA(2A) EIA(E2A)/5"" "IB(2B) EI B(E2B)/5" S"I C(2C) E 1 C(E2C)/5" ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.7.5 CLOSE HHSI cold leg injection valves. {CP00I }* "LOOP A Tc INJ MOV-0006A"* "LOOP B Tc [NJ MOV-0006B"* "LOOP C Tc INJ MOV-0006C" NOTE Plant cooldown may continue while performing Steps 7.6 through 7.15.5.* During Shutdown conditions, WHEN an ECO is on SI-MOV-0016A, B, C AND its associated LHSI train is required to be functional for LOCA injection/recirculation, THEN ESTABLISH a "dedicated" Operator watch position in accordance with Addendum 14, MOV-0016A, B & C Emergency Operations Guideline.

7.6 PLACE

LHSI pump hand switches in the PULL TO LOCK position.

{CPOO1}* "LHSI PUMP 1A(2A)"* "LHSI PUMP 1B(2B)"* "LHSI PUMP 1 C(2C)" 7.7 PLACE all three containment spray pumps hand switches in the PULL TO LOCK position.

{CP002} (Ref 2.18, Ref 2.41)* "CSS PUMP IA(2A)"* "CSS PUMP I B(2B)"* "CSS PUMP IC(2C)" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 57 of216 Plant CooldownI Initials 7.8 IF the control rods are NOT fully inserted, THEN manually INSERT all control rods by selecting the appropriate bank on the "ROD BANK SEL" switch.NOTE* Performance of Step 7.9 will generate a feedwater isolation signal due to low Tavg.* Rod Control Urgent Alarm and associated control room annunciator "ROD CONT URGENT ALARM" (5M03-B5) will be received when the reactor trip breakers are opened 7.9 IF the reactor trip breakers are closed, THEN OPEN reactor trip breakers by placing the "REACTOR TRIP" switch in the TRIP position.

{CP005}NOTE The "MASTER RELAY DEFEAT SWITCH" (MRDS) must be pulled out to move it from RX TRIP to DEFEAT/CVI AVAIL.7.10 WHEN the reactor trip breakers are open, THEN PLACE the "MASTER RELAY DEFEAT SWITCH" in each SSPS Logic Cabinet in either the DEFEAT/CVI AVAIL or DEFEAT ALL position as directed by the Shift Manager/Unit Supervisor.

7.10.1 "PROTECTION SYSTEM LOGIC TRAIN R, LOGIC CABINET, LOGIC TEST PANEL" (SSPS) (ZRROO)7.10.2 "PROTECTION SYSTEM LOGIC TRAIN S, LOGIC CABINET, LOGIC TEST PANEL" (SSPS) (ZRR008)7.11 Steps 7.27, 7.30, 7.31, 7.32 and 7.36 have been identified as candidates for early start. Early starts of these steps have been evaluated as having NO safety impact for current plant conditions.

The Shift Manager/Unit Supervisor may authorize early start of procedure steps to enhance plant performance.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 58 of 216 IPlant CooldownI Initials NOTE" Step 7.12 is required by Calculation NC-07090, Evaluation of Boron Dilution Flow Paths in Mode 5b (Mode 5 with reactor coolant loops NOT filled) and Mode 6, to prevent low boron water from being introduced into the RCS. (Refs. 2.108, 2.109 & 3.13)" Step 7.12 will perform one of the following to comply with Calculation NC-07090: " The CVCS Loop A or C charging line already isolated will remain isolated by ECO throughout Mode 5b (Mode 5 with reactor coolant loops NOT filled) and Mode 6.OR* Swap CVCS Loop A and C charging lines so that both flowpaths will be borated with shutdown boron concentration water prior to establishing 0PSP03-CV-0014, CVCS Equipment Verification.

• IF Plant Cooldown with the PZR Water Solid is planned, THEN Step 7.12 SHOULD be completed PRIOR TO Water Solid conditions.

7.12 PERFORM one of the following to ensure CVCS to RCS Loops A and C isolation lines are not a source of RCS dilution in Mode 5b and Mode 6 (Refs.2.108, 2.109 & 3.13): 7.12.1 (Preferred Method) Swapping 1(2)-CV-MOV-0003 and 1(2)-CV-MOV-0006 to borate prior to Mode 5B. (Reference 2.112)7.12.1.1 ENSURE the following:

a. RCS Boron Concentration Greater Than 2800 ppm.b. Charging flow 50 gpn OR greater.c. Flow through the inservice CVCS line for 6 minutes OR greater.* "LOOP C ISOL MOV-0006"* "LOOP A ISOL MOV-0003" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 59 of 216 Plant Cooldown Initials 7.12.1.2 ENSURE one of the following (Ref 2.68): a. IF the differential pressure between the charging header and the RCS is less than 2500 psid, THEN perform Step 7.12.1.3 or 7.12.1.4.OR b. SECURE the CVCS charging pump per OPOP02-CV-0004, Chemical and Volume Control System Subsystem OR c. ENSURE charging flow to RCS is secured as follows: 1. ENSURE miniflow valve OPEN for inservice Centrifugal Charging Pump: CCP 1A(2A) "RECIRC FCV-0201" CCP 1B(2B) "RECIRC FCV-0202" 2. ENSURE "CHG FLOW CONT FK-0205" in MAN and CLOSED.7.12.1.3 IF "LOOP A ISOL MOV-0003" is in service, THEN PERFORM the following:

a. OPEN "LOOP C ISOL MOV-0006" b. CLOSE "LOOP A ISOL MOV-0003" c. GO TO Step 7.12.1.5 7.12.1.4 IF "LOOP C ISOL MOV-0006" is in service. THEN PERFORM the following:

a. OPEN "LOOP A ISOL MOV-0003" b. CLOSE "LOOP C ISOL MOV-0006" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 60 of 216 Plant Cooldown Initials 7.12.1.5 IF charging to RCS was secured in Step 7.12.1.2, THEN PERFORM one of the following (Ref 2.68): a. PLACE IN SERVICE a charging pump per OPOP02-CV-0004, Chemical and Volume Control System Subsystem, OR b. RESTORE charging flow to RCS is as follows: 1. ADJUST CV-FCV-0205 to obtain Charging flow of 50 gpm or greater.2. CLOSE inservice Centrifugal Charging Pump Recirc valve: CCP lA(2A) "RECIRC FCV-0201" CCP 1B(2B) "RECIRC FCV-0202" 7.12.1.6 PERFORM the following to ensure inservice charging line to RCS is borated prior to entering mode 5B: a. Charging flow 50 gpm OR greater.b. Flow through through the inservice charging line for 6 minutes OR greater.* "LOOP C ISOL MOV-0006"* "LOOP A ISOL MOV-0003" 7.12.1.7 UPDATE Form 1, CVCS Line Boration Tracking Form.7.12.1.8 GO TO Step 7.13.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 61 of 216 E: ~ Plant CooldownI Initials 7.12.2 (Alternate Method) IF the NOT In-Service CVCS CHARGING TO RCS LOOP COLD LEG MOV OPERATOR, (the closed valve)CV-MOV-0006 or CV-MOV-0003 will remain isolated throughout Mode 5b and Mode 6, THEN PERFORM the following for the NOT In-Service valve, OTHERWISE GO TO Step 7.12.1: 7.12.2.1 REQUEST Unit Operations Manager permission to perform this Section.7.12.2.2 De-Energize the MOV and hang ECO on supply breaker.ECO #7.12.2.3 CAUTION tag Control Room handswitch to identify"Potential dilution source must remain isolated in Mode 5b and Mode 6".ECO #7.12.2.4 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.7.12.2.5 UPDATE Form 1, CVCS Line Boration Tracking Form.I CAUTION Use of a pressurizer steam bubble during periods of low pressure, low temperature operation is preferred.

This steam bubble will dampen the plant's response to potential transient generating inputs, providing easier pressure control with slower response rates. (UFSAR 5.2.2.11.3) 7.13 IF directed by Operations Management to take the PZR Water Solid at this point in the cooldown, THEN PERFORM Addendum 9 "Plant Cooldown with the PZR Water Solid" while continuing with this procedure.

7.14 IF establishing a SG Nitrogen blanket, THEN PERFORM the following:

I This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 62 of 216 Plant Cooldown Initials NOTE A set of hoses is normally stored in the OUTAGE BOX. IF performing this in a Forced Outage, hoses may need to be fabricated.

7.14.1 INSTALL bulk Nitrogen supply hoses to each SG connection:

"1(2)-MS-0510 N2 BLANKET ISOL"{57 ft IVC Loop A}S "1(2)-MS-0508 N2 BLANKET ISOL"{57 ft IVC Loop B}"1 (2)-MS-0506 N2 BLANKET ISOL"{57 ft IVC Loop C}"1(2)-MS-0504 N2 BLANKET ISOL"{57 ft IVC Loop D}This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 63 of 216 Plant Cooldown Initials 7.14.2 ENSURE OPEN the following nitrogen valves: 1 (2)-NL-0080

'LOW PRESSURE NITROGEN TO STEAM GENERATOR BLANKETING ISOLATION VALVE" (YDB Ely 29' between Unit 1(2) TGB and IVC on the South Side, West End)* I (2)-NL-0 186 "LOW PRESSURE NITROGEN TO S/G ID(2D) BLANKETING ISOLATION VALVE" {IVC Rm 404 Elv 51', behind FSV-7444 between MS Line and Wall}* 1(2)-NL-0 188 "LOW PRESSURE NITROGEN TO S/G 1C(2C) BLANKETING ISOLATION VALVE" {IVC Rm 401 Elv 51', behind FSV-7434 between MS Line and Wall}* 1 (2)-NL-0 190 "LOW PRESSURE NITROGEN TO S/G 1B(2B) BLANKETING ISOLATION VALVE" {IVC Rm 402 Elv 51', behind FSV-7424 between MS Line and Wall}* 1(2)-NL-0 192 "LOW PRESSURE NITROGEN TO S/G IA(2A) BLANKETING ISOLATION VALVE" {IVC Rm 403 Elv 51', behind FSV-7414 between MS Line and Wall}7.14.3 ENSURE SG PORVs CLOSED. {CP006}* SG IA(2A) "PORV PV-741 1"* SG 1B(2B) "PORV PV-7421"* SG I C(2C) "PORV PV-743 1"* SG I D(2D) "PORV PV-744 1" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 64 of 216 Plant Cooldown Initials 7.14.4 OPEN "N2 BLANKET ISOL" valves. {IVC 57 ft}* SG I A(2A) "1 (2)-MS-05 10 N2 BLANKET ISOL"* SG 1A(2A) "1(2)-MS-0509 N2 BLANKET ISOL"* SG 1B(2B) "1 (2)-MS-0508 N2 BLANKET ISOL"* SG IB(2B) "1(2)-MS-0507 N2 BLANKET ISOL"* SG 1 C(2C) "1 (2)-MS-0506 N2 BLANKET ISOL"* SG 1 C(2C) "1 (2)-MS-0505 N2 BLANKET ISOL"* SG ID(2D) "1(2)-MS-0504 N2 BLANKET ISOL"* SG ID(2D) "1 (2)-MS-0503 N2 BLANKET ISOL" 7.14.5 USING Control Room pressure instrumentation for reference, ADJUST each regulator to maintain a positive pressure less than 2.0 psig on the Steam Generators.

• 1(2)-MS-PCV-8393 "LOW PRESSURE NITROGEN TO S/G 1(2)D BLANKETING PRESSURE CONTROL VALVE" {57 ft IVC Loop D}* 1(2)-MS-PCV-8394 "LOW PRESSURE NITROGEN TO S/G 1(2)C BLANKETING PRESSURE CONTROL VALVE" {57 ft IVC Loop C)* 1(2)-MS-PCV-8395 "LOW PRESSURE NITROGEN TO S/G 1(2)B BLANKETING PRESSURE CONTROL VALVE" {57 ft IVC Loop B I* "1 (2)-MS-PCV-8396 "LOW PRESSURE NITROGEN TO S/G 1(2)A BLANKETING PRESSURE CONTROL VALVE{57 ft IVC Loop A}This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 65 of 216 Plant Cooldown Initials CAUTION In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.7.15 IF placing SGs in wet layup, THEN PERFORM the following:

7.15.1 VERIFY secondary plant shutdown completed through Step 10.15.(Step 7.15.2 may generate a feedwater isolation signal depending on position of MRDS switch.)7.15.2 PLACE SGs in wet lay up by feeding each SG per OPOP02-AF-0001, Auxiliary Feedwater, until Wide Range level indicates between 98%and 100%.* SG 1A(2A)* SG IB(2B)* SG 1C(2C)* SG ID(2D)7.15.3 SECURE AFW operation per OPOP02-AF-0001, Auxiliary Feedwater.

7.15.4 PLACE SGs in recirculation per OPOP02-SB-0002, Steam Generator Wet Layup Recirc.7.15.5 ESTABLISH SG wet lay up chemistry as required per OPCPOI-ZA-0038, Plant Chemistry Specifications.

7.16 ENSURE AFW operation is secured per OPOP02-AF-0001, Auxiliary Feedwater.

This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 66 of'216 Plant Cooldown Initials CAUTION Maintain RCS Temp 190-195 until RCS can be depressurized to prevent exceeding the maximum Delta Temp on the PZR Surge Line per WCAP analysis.7.17 IF the RCS cooldown is being performed due to an unisolable RCS leak, THEN GO TO Section 8.0, RCS Depressurization. (Ref 2.59)7.18 Commence an RCS cooldown to between 180'F and 185°F, MAINTAIN RCS pressure between 325 psig and 400 psig.NOTE I&C PM to place Core Exit T/C Alarm System in-service is required prior to RCS Tave less than 160'F.7.19 IF RCS drain down to reduced inventory will be performed, THEN NOTIFY I&C to place Core Exit T/C Alarm System in service (PMs Unit 1-95003780, Unit 2-95003781).

WAN #7.20 NOTIFY Chemistry to sample the VCT for H2 concentration.

This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 67 of216 Plant Cooldown Initials 7.21 IF RCS boundary opening is NOT planned AND all RCPs will be secured, THEN PERFORM the following:

7.21.1 Establish a control band of Hydrogen on the RCS of 5-15 cc/kg using Addendum 5.7.21.2 IF Hydrogen lowers below 5 cc/kg OR per Chemistry request, THEN PERFORM the following:

7.21.2.1 CONTACT Chemistry to determine if Hydrogen should be aligned to the VCT.7.21.2.2 IF Hydrogen is to be aligned to VCT, THEN PERFORM the following:

a. ENSURE ECO for Hydrogen Supply"I(2)-CV-0178 HYDROGEN SUPPLY TO VCT ISOL" is released.b. VERIFY Gaseous Waste Processing System (GWPS) is in service per OPOP02-WG-0001, Gaseous Waste Processing System Operations.

c. VERIFY H2 System aligned to supply the VCT.d. ENSURE "VCT VENT PCV-01 15" open.e. CLOSE "1(2)-CV-0181 NITROGEN SUPPLY TO VCT ISOL". (CVCS Chemical Mixing Tank Room)f. CLOSE "1(2)-NL-0033 UNIT 1(2) VOLUME CONTROL TANK N2 SUPPLY VLV". (51 ft Cubicle 335 Access from 41 ft)g. OPEN "1(2)-CV-0178 HYDROGEN SUPPLY TO VCT ISOL". (CVCS Chemical Mixing Tank Room)NOTE The "DIVERT LCV-01 12A" valve may be placed in the "RHT" position as necessary to control VCT Level.h. PLACE "DIVERT LCV-01 12A" in the VCT position.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 68 of 216 Plant Cooldown____

Initials CAUTION VCT pressure SHALL NOT be allowed to exceed 65 psig during fill.i. RAISE VCT level to between 90 and 95% using"RC M/U CONT". {CP004}j. MAINTAIN VCT pressure less than 60 psig while raising level. {CP004}k. PLACE "RC M/U CONT" in the STOP position.1. ALLOW VCT pressure to decay to the new minimum value (approximately 15 to 20 psig)while maintaining VCT level between 90 and 95%. {CP004}NOTE IF the LWPS does NOT have the capacity to receive the water from letdown or it is desired to minimize water usage, THEN it is permissible to adjust charging and letdown to obtain the desired VCT levels.in. PLACE "DIVERT LCV-01 12A" in the AUTO position.

{CP004}n1. VERIFY "H2 SUPPLY REGULATOR PV-3 110" maintains greater than or equal to 15 psig on the VCT.o. LOWER VCT level to 30% by placing"DIVERT LCV-01 12A" in the RHT position.p. PLACE "DIVERT LCV-01 12A" in the AUTO position.q. IF additional hydrogen cover gas is required on the VCT, THEN REPEAT Steps 7.21.2.2.h through 7.21.2.2.p This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 69 of 216 Plant Cooldown Initials r. WHEN Chemistry reports Hydrogen concentration is satisfactory in the RCS (5-15 cc/kg), THEN PERFORM the following:

1. TAG CLOSE "I(2)-CV-0178 HYDROGEN SUPPLY TO VCT ISOLATION VALVE" with an Equipment Clearance Order. (CVCS Chemical Mixing Tank Room)2. OPEN "1(2)-CV-0181 NITROGEN SUPPLY TO VCT ISOLATION VALVE". (CVCS Chemical Mixing Tank Room)3. OPEN "1(2)-NL-0033 LOW PRESSURE NITROGEN TO VOLUME CONTROL TANK ISOLATION VALVE". (51 ft Cubicle 335 Access friom 41 fi)4. RECORD ECO#5. ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 70 of 216 Plant Cooldown Initials CAUTION The stepping or tripping of CRDMs during periods when coolant crud levels are high should be kept to a minimum by applying the following:

• Reactor Trip Breakers should be open prior to Hydrogen Peroxide addition to minimize the attracting of crud particles in the CRDMs coils. (Ref 2.63 & 2.96)Reactor Trip Breakers should remain open after a Hydrogen Peroxide addition until coolant turbidity levels have returned to 1.1 NTU or less. (Ref 2.63 & 2.96)* Control Rods should NOT be moved until coolant turbidity levels have returned to 1.1 NTU or less. (Ref 2.63 & 2.96)0 IF coolant turbidity levels are greater than 1.1 NTU AND Control Rod movement is to be performed, THEN PERFORM Addendum 17, Moving Control Rods with High Reactor Coolant Turbidity. (Ref. 2.96)* IF a RCS boundary opening is planned, THEN RCS H2 Concentration should be less than 4 cc/kg prior to opening the RCS.IF Hydrogen Peroxide addition is required, THEN dilution water amount used in the chemical addition SHALL be accounted for to maintain required shutdown margin.(Refer to Steps 5.18.3 and 5.18.5)NOTE IF OPSPO3-CV-0014, CVCS Equipment Verification, has been performed AND Hydrogen Peroxide addition is desired, THEN ENSURE Hydrogen Peroxide add completed prior to entering Mode 5b (Mode 5 with reactor coolant loops NOT filled) (Refs. 2.108, 2.109 & 3.13).7.22 IF opening of the RCS is planned AND a Hydrogen Peroxide addition is required, THEN PERFORM the following:

7.22.1 IF OPSP03-CV-00 14, CVCS Equipment Verification, has been performed, THEN ENSURE OPSP03-CV-0014 is set to allow Hydrogen Peroxide addition.7.22.2 ENSURE RCS boron concentration greater than or equal to that recorded in Step 5.18.3.RCS Cb/Time PZR Cb/Time This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 71 of 216 Plant Cooldown Initials 7.22.3 IF Rapid Refueling is planned, THEN ENSURE RCS boron concentration is greater than or equal to that recorded in Step 5.18.5.(Ref 2.35)RCS Cb/Time PZR Cb/Time 7.22.4 NOTIFY Chemistry to add Hydrogen Peroxide to the RCS per OPOP02-CV-0001, Makeup to Reactor Coolant System, and OPCPO3-ZC-0005, Chemical Addition to the Reactor Coolant System.7.23 NOTIFY Chemistry to sample the RCS for fission products and H2 concentration.

7.24 OBTAIN concurrence from Health Physics and Chemistry to continue the RCS cooldown below 180'F.7.25 WHEN concurrence from Health Physics and Chemistry is obtained, THEN COMMENCE RCS Cooldown to 150'F, MAINTAIN RCS pressure between 325 psig and 400 psig.7.26 IF Core Exit T/C Alarm System was placed in service, THEN PERFORM the following:

7.26.1 WHEN RCS temperature is less than 1607F, THEN VERIFY "RC MID LOOP CORE EXIT TEMP HI" Lamp box IM02 E-1 alarm clear. {CPOO I)7.26.2 WHEN RCS temperature reaches 150'F, THEN CONTINUE with Steps 7.27 through 7.30.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 72 of 216 Plant Cooldown Initials NOTE Plant cooldown may continue while performning Step 7.27.7.27 ENSURE an inspection of the RCS for boron deposits and other evidence of primary system leakage is performed per OPGP03-ZE-0033, RCS Pressure Boundary Inspection For Boric Acid Leaks, as determined necessary by SED.(Ref 2.46)CAUTION In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.7.28 ENSURE idle RHR trains have been placed in service per OPOP02-RH-0001, Residual Heat Removal System Operation, to equalize boron concentrations with the RCS.* RHR Train A* RHR Train B* RHR Train C NOTE RHR Train A or B should remain in operation to supply low pressure letdown.7.29 SECURE selected RHR trains per OPOP02-RH-0001, Residual Heat Removal System Operation, as directed by the Shift Manager/Unit Supervisor.

• RHR Train A* RHR Train B* RHR Train C 7.30 MAINTAIN RCS pressure between 325 and 400 psig.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 73 of 216 Plant Cooldown Initials 7.31 WHEN Rapid Refueling rod holdout operations are desired, THEN PERFORM the following:

CAUTION* During rod holdout operations, RCS temperature should be stable when moving control rods to prevent control rod thermal binding and two positive reactivity changes during rod motion.Reactor trip breakers must be open during a RCS cooldown when any control rods are on the bottom.OPSP03-CV-0014, CVCS Equipment Verification MAY isolate the letdown flowpath (Example:

closing the Detain outlets before opening the Demin Bypass), ENSURE the letdown flowpath restrictions are planned for PRIOR TO performing 0PSP03-CV-00 14, CVCS Equipment Verification.

7.3 1. 1 ENSURE RCS is borated to SHUTDOWN MARGIN LIMIT CURVE Plant Curve Book Figure 5.5, Mode 5 Cb, ARO, 68°F to 200'F, K <= 0.95. (Ref 2.35)RCS Cb/Time PZR Cb/Time 7.3 1.2 PERFORM OPSP03-CV-0014, CVCS Equipment Verification, to isolate dilution paths while rods are in the Rapid Refueling position.An Equipment Clearance Order SHALL be hung to ensure dilution paths remain isolated.

ECO #7.31.2.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.7.31.3 IF Control Rods are to be locked out for GREATER THAN 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with Reactor Vessel Head on the flange and fuel is in the vessel, THEN REFER TO OPOP03-ZG-0012, Operation with Rods in the Rapid Refueling Position.7.31.4 IF RCS turbidity levels are greater than 1.1 NTU AND Control Rod holdout is to be performed, THEN PERFORM Addendum 17, Moving Control Rods with High Reactor Coolant Turbidity. (Ref.2.96)7.31.5 PERFORM the appropriate sections of OPMP07-DM-0003, Rapid Refueling Rod Holdout Operation.

7.31.6 WHEN the Rapid Refueling rod holdout operations are complete, THEN ENSURE the Reactor Trip Breakers are open.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007

[Rev. 71 Page 74 of216 Plant Cooldown Initials NOTE Valve exercising during cold shutdown SHALL commence within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of achieving cold shutdown and continue until all testing is complete or the plant is ready to return to operation at power. For extended outages, testing need not be commenced in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided all valves required to be tested during cold shutdown will be tested before plant startup.However, it is not the intent to keep the plant in cold shutdown to complete cold shutdown testing.Cold Shutdown Testing reference:

IST Plan Rev. 14, Section 5.4.2, Valve Exercise Test OM-2004 CODE, ISTC-3521 (g)7.32 INITIATE Cold Shutdown testing within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of Mode 5 entry. Cold Shutdown testing consists of perfonnance of the following surveillances:

OPSP03-CV-001 1, Chemical and Volume Control System Valve Operability Test (Cold Shutdown)OPSP03-FW-0002, Feedwater System Valve Operability Test (Cold Shutdown)OPSPO3-HC-0004, Reactor Containment Building Normal Purge System Valve Operability Test (Cold Shutdown)0PSP03-MS-0002, Main Steam System Cold Shutdown Valve Operability Test OPSP03-RC-0010, Pressurizer Power Operated Relief Valve Operability Test OPSP03-RH-0007, Residual Heat Removal System Valve Operability Test (Cold Shutdown)* OPSP03-RH-00

10. RHR System Valve Leak Tests* OPSPO3-SI-0028, SIS Accumulator Valves Operability Test (Cold Shutdown)7.33 NOTIFY the Plant Operations Surveillance Coordinator to identify the specific procedures required to be performed prior to restart.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 75 of216 Plant Cooldown Initials CAUTION LHSI cold leg injection valves SHALL remain open on operating RHR trains.WHEN RCS temperature reaches 140'F, THEN ENSURE less than 4 RCPs in operation.

7.34 WHEN RCS temperature reaches 150'F, THEN CLOSE LHSI cold leg injection valves on all idle RHR trains. {CPOO }* "LOOP A Tc INJ MOV-003 IA"* "LOOP B Tc INJ MOV-003 IB"* "LOOP C Tc INJ MOV-003 IC" 7.35 IF necessary to control ambient temperature in the RCB to between 65°F and 70 0 F, THEN ADJUST the number of RCFCs in operation per OPOP02-HC-0001, Containment HVAC.7.36 IF desired to access reactor cavity during shutdown, THEN NOTIFY appropriate personnel to perform the required inspections of the safety stairway going down the east side of the cavity per OPGP03-ZM-0028, Erection and Use of Temporary Scaffolding.

7.37 IF desired to cooldown below 150TF. THEN GO TO Section 8.0, OTHERWISE Maintain RCS temperature 150'F-180'F and RCS pressure between 325 and 400 psig.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 76 of 216 j Plant Cooldown Initials 8.0 RCS Depressurization CAUTION* Cold Calibrated Pressurizer Level must be kept greater than 30% to keep the Pressurizer heaters covered.* WHEN RCS temperature reaches 140'F, THEN ENSURE less than four (4) RCPs in operation.

• Whenever the reactor coolant temperature is above 160 0 F, at least one Reactor Coolant Pump should be in operation. (Ref. 2.60)* During a normal cooldown, at least one (1) Reactor Coolant Pump shall be operated to ensure that the temperature difference between the loops does NOT exceed 25°F. (Ref.2.60)8.1 IF desired to perform OPMP07-DM-0003, Rapid Refueling Rod Holdout Operation, THEN MAINTAIN RCS pressure as required by 0PMP07-DM-0003.

8.2 ENSURE

Pressurizer "HTR CONT GRP 1C(2C)" control switch in the PTL position.

{CP004}8.3 MONITOR RCP seal differential pressure to maintain greater than 230 psid.{CP004}0 "RCP IA(2A) SEAL I DP PI-0152"* "RCP 1B(2B) SEAL 1 DP PI-0 153"* "RCP 1 C(2C) SEAL I DP PI-0 154"* "RCP ID(2D) SEAL 1 DP PI-0155" 8.4 IF PZR is NOT Solid (has bubble) and it is desired to take the PZR Solid, THEN PERFORM the following:

8.4.1 PERFORM

"Addendum 12, Solid Plant Operations Entry Checklist" 8.4.2 OBTAIN permission to conduct Solid Plant Operations.

Shift Manager Operations Manager This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 77 of 216 Plant Cooldown Initials 8.5 REDUCE RCS pressure to approximately 310 psig by: IF PZR Solid, by ADJUSTING "PRESS CONT PCV-0135" as necessary.

IF PZR NOT Solid, ADJUSTING Pressurizer spray and backup heater groups as necessary.

{CP004} (Ref 2.62)CAUTION IF the RCS depressurization is due to an unisolable RCS leak, THEN the temperature differential between Pressurizer liquid and the Reactor Coolant SHALL NOT exceed 2507F. (Ref 2.62)IF the RCS depressurization is NOT due to an unisolable RCS leak, THEN the temperature differential between Pressurizer liquid and the Reactor Coolant SHALL NOT exceed 320'F.WHEN RCS Temperature is less than 140'F, THEN ENSURE less than four (4) RCPs in operation.

(3 pumps or less restriction, This limitation was required to demonstrate acceptable hold down forces in the Cold Zero Power lift force calculation.)

NOTE The administrative cooldown rate is 80°F/hr. (Ref 2.70)The administrative cooldown limit for the Pressurizer is 160°F/hr.The cooldown should be spaced over the entire 60 minute period vice a short time high rate cooldown. (Ref 2.50)8.6 IF RCS cooldown is due to an unisolable RCS leak, THEN PERFORM the following:

8.6.1 ADJUST

RHR flow as necessary to cooldown the RCS to between 180'F and 185°F.8.6.2 MAINTAIN RCS temperature greater than or equal to 180°F by adjusting RIR flow as necessary, WHILE reducing Pressurizer liquid temperature and pressure to less than 350'F/125 psig by using Pressurizer spray.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 78 of 216 Plant Cooldown Initials NOTE IF Step 8.6, is applicable, THEN it should be completed prior to performing Step 8.7.8.7 ADJUST RHR flow as necessary to commence a cooldown of the RCS to 105 0 F.8.8 WHEN RCS temperature reaches 140'F, THEN ENSURE less than four (4)RCPs are in operation.

8.9 IF PZR is Solid, THEN REDUCE Pressurizer liquid temperature to approximately RCS Temperature by ADJUSTING Pressurizer spray, WHILE maintaining cooldown limits., N/A if PZR is NOT Solid.CAUTION OPSP03-CV-0014,.

CVCS Equipment Verification MAY isolate the letdown flowpath (Example:

closing the Demin outlets before opening the Detrin Bypass), ENSURE the letdown flowpath restrictions are planned for PRIOR TO performing OPSP03-CV-0014, CVCS Equipment Verification.

8.10 IF the RCS is to be depressurized to less than VCT pressure OR drained, THEN ENSURE OPSP03-CV-0014, CVCS Equipment Verification, is performed before securing RCPs. An Equipment Clearance Order SHALL be hung to ensure dilution paths remain isolated. (Ref2.78)ECO #8.10.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.8.11 COMMENCE Non-Intrusive Check Valve Testing for CC & CV system valves as determined by the Shift Manager/Unit Supervisor and System Engineer per OPEP07-ZE-0008, OTHERWISE N/A.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 79 of 216 E :Plant Cooldowniti Initials NOTE IF NO RCS temperature recommendation where all RCP(s) are to be shutdown is provided, THEN the default RCP(s) Shutdown temperature recommendation is 140'F, the Shift Manager/Unit Supervisor has final selection judgment.The Shift Manager/Unit Supervisor may authorize changes to the operating RCP configuration during the cooldown if conditions warrant (i.e. large CRUD burst, pump problems, cleanup issues, etc).8.12 ESTABLISH the RCP(s) Shutdown temperature criteria for the cooldown, by PERFORMING the following:

8.12.1 CONTACT Outage Planning Team for RCP(s) Shutdown temperature recommendation.

8.12.2 CONTACT Chemistry for RCP(s) Shutdown temperature recommendation and to verify that RCS activity level (crud) is acceptable and full flow clean-up NOT required.8.12.3 CONTACT Health Physics for RCP(s) Shutdown temperature recommendation.

This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 80 Plant CooldownI Initials CAUTION* At least one (1) RCP SHALL be maintained in service UNTIL RCS boron concentration is greater than or equal to Cold Shutdown Xenon Free Concentration and all significant Boron changes are completed. (Ref. 2.60)* At least one (1) RCP SHALL be maintained in service UNTIL RCS forced oxygenation and degassification are completed.

• In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.* WHEN RCS Temperature is less than 140°F, THEN ENSURE less than four (4) RCPs in operation.

(3 pumps or less restriction, This limitation was required to demonstrate acceptable fuel assembly top nozzle hold down spring forces in the Cold Zero Power lift force calculation.)

• Whenever the reactor coolant temperature is above 160'F, at least one (1) Reactor Coolant Pump should be in operation. (Ref. 2.60)* During a normal cooldown, at least one (1) Reactor Coolant Pump shall be operated to ensure that the temperature difference between the loops does NOT exceed 25°F. (Ref. 2.60)* The conditions listed below applies to a "normal" cooldown ONLY, IF abnormal conditions exist, THEN the Shift Manager/Unit Supervisor may determine when to stop all RCPs.8.13 WHEN the following conditions have been satisfied:

• RCS temperature is < 160'F (Ref. 2.60)* RCS forced oxygenation and degassification are completed* RCS boron concentration is greater than or equal to Cold Shutdown Xenon Free Concentration (Ref. 2.60)* All significant Boron changes are completed (Ref. 2.60)* The temperature differential between the auxiliary spray water and the pressurizer is < 320'F (Ref. 2.60)* Normal Spray flow is NOT required for continued PZR cooling (i.e., Aux Spray is available) (Ref. 2.60)* The RCS temperature has been reduced to the temperature value determined by Step 8.12 THEN, ENSURE all RCPs are stopped per OPOP02-RC-0004, Operation of Reactor Coolant Pumps: {CP004}"RCP 1A(2A)""RCP IC(2C)""RCP 1B(2B)""RCP 1D(2D)" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 81 of 216 Plant Cooldown Initials 8.14 WHEN all RCPs stopped, THEN REDUCE RCS pressure to approximately 100 psig by performing the following:

IF PZR is Solid, ADJUSTING "PRESS CONT PCV-0135" as indicated on "OUTLET PRESS PI-0135", OTHERWISE N/A.IF PZR is NOT Solid, REDUCE Pressurizer liquid temperature and pressure by adjusting Pressurizer spray, Aux spray and backup heater groups as necessary, OTHERWISE N/A.a. ENSURE OPEN Pressurizer Spray valves, N/A if Aux Spray in service. {CP004}* "PRZR SPR PCV-0655B"* "PRZR SPR PCV-0655C" b. ENSURE OPEN Pressurizer "AUX SPRAY LV-3119" c. ENSURE CLOSE CVCS Loop A and C isolation valves* "LOOP A ISOL MOV-0003"* "LOOP C ISOL MOV-0006" NOTE Charging and Letdown flows may be adjusted to aid in AUX SPRAY FLOW.d. THROTTLE (CLOSED) Pressurizer Spray valves to reduce pressure in the RCS: {CP004}* "PRZR SPR PCV-06551B"* "PRZR SPR PCV-0655C" 8.15 WHEN RCS Pressure is LESS THAN 100 psig, THEN CLOSE RCP "SEAL LKF ISOL" valves. {CP004}* RCP IA(2A) "SEAL NO I LKF ISOL FV-3 154"* RCP IB(2B) "SEAL NO I LKF ISOL FV-3155"* RCP I C(2C) "SEAL NO 1 LKF ISOL FV-3156"* RCP ID(2D) "SEAL NO I LKF ISOL FV-3157" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 82 of216 Plant Cooldown Initials NOTE* The purpose of Step 8.16 is to cool the PZR vapor space to 250'F (PZR NOT solid) OR cool to approximately RCS temperature and less than 200'F (PZR solid).* IF Pressurizer vapor space temperature TI-0607 is NOT functional, THEN use Pressurizer water space temperature TI-0608 for temperature indications in this procedure. (Ref 2.84)8.16 CIRCULATE flow to cool the PZR as follows: {CP004}8.16.1 ENSURE OPEN Pressurizer Spray valves, N/A if Aux Spray in service. {CP004}* "PRZR SPR PCV-0655B"* "PRZR SPR PCV-0655C" 8.16.2 ENSURE OPEN Pressurizer "AUX SPRAY LV-3119" 8.16.3 ENSURE CLOSED CVCS Loop A and C isolation valves.* "LOOP A ISOL MOV-0003"* "LOOP C ISOL MOV-0006" NOTE Charging and Letdown flows may be adjusted to aid in AUX SPRAY FLOW.8.16.4 THROTTLE (CLOSED) Pressurizer Spray valves to reduce Temperature in the Pressurizer:

{CP004}* "PRZR SPR PCV-0655B"* "PRZR SPR PCV-0655C" This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 83 of 216 Plant Cooldown Initials 8.16.5 WHEN the PZR vapor space temperature has cooled to < 2507F (PZR NOT solid) OR to approximately RCS temperature and <200'F (PZR solid), THEN PERFORM the following:

8.16.5.1 ENSURE OPEN Pressurizer Spray valves. {CP004}* "PRZR SPR PCV-0655B"* ~"PRZR SPR PCV-0655C" 8.16.5.2 ENSURE OPEN CVCS Loop A or C isolation valve.* "LOOP A ISOL MOV-0003"* "LOOP C ISOL MOV-0006" 8.16.5.3 ENSURE CLOSE Pressurizer "AUX SPRAY LV-3119".8.17 ENSURE 0PSP03-RC-001 1, Reactor Vessel Head Vent Flow Verification and Valve Operability Test (Cold Shutdown) performed. (CR 03-5002)8.18 IF Pressurizer level will be lowered below 0% Cold Calibrated level, THEN INSTALL the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" per OPOP07-RC-0001, RC Vent Rig/Sightglass Installation and Removal.8.19 PERFORM the following to establish the conditions for placing Nitrogen on the PZR: 8.19.1 ENSURE Restoration of PRT/RCDT/GWPS from H 2 Degassing per OPOP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation.

8.19.2 ENSURE a Nitrogen supply hose is installed to the hose connection downstream of "1 (2)-RC-0166, PRZR PORV LINE VENT VALVE".{RCB Top of PRZR}8.19.3 CLOSE Pressurizer Relief Tank "1(2)-RC-0025, N2 SUPPLY ISOL".{RCB 6 ft E of PRT}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 84 of216 Plant Cooldown Initials CAUTION The Nitrogen supply is routed directly into the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" connection.

This can cause an indicated level difference of several inches due to the higher pressure at the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" connection.

RCS H 2 concentration should be reduced to less than 4 cc/kg prior to opening the RCS.8.19.4 ENSURE "1(2)-RC-0028, PRT GAS SAMPLE BOTTLE BYPASS VALVE" closed. { 10 ft NMAB Non-Radioactive Pipe Chase Room 064}8.19.5 ENSURE "1 (2)-NL-0039, PZR NITROGEN SUPPLY VALVE" open. {10 ft MAB Non-Radioactive Pipe Chase Room 064}8.19.6 ENSURE "1 (2)-RC-PV-3654, PRT N2 SUPPLY" maintaining between 2 and 6 psig. { 10 ft MAB Non-Radioactive Pipe Chase Room 064}8.19.7 OPEN PRTN2 "ICIV FV-3653".

{CP004}8.19.8 OPEN PRT N2 "OCIV FV-3652".

{CP004}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 85 of 216 Plant Cooldown Initials CAUTION ALL RCS dilution paths SHALL BE ISOLATED WHEN the RCS is in a "LOOPS NOT FILLED" condition, per T. S. 3.4.1.4.2 and 3.9.1. (CR 07-929)8.20 PRIOR TO placing the RCS in a "LOOPS NOT FILLED" condition (i.e., Depressurized), ALL RCS dilution paths SHALL BE ISOLATED by performing the following:

CAUTION 0PSP03-CV-0014, CVCS Equipment Verification MAY isolate the letdown flowpath (Example:

closing the Demin outlets before opening the Demin Bypass), ENSURE the letdown flowpath restrictions are planned for PRIOR TO performing OPSP03-CV-0014, CVCS Equipment Verification.

8.20.1 ENSURE 0PSP03-CV-0014, CVCS Equipment Verification, has been performed.

8.20.2 ENSURE an Equipment Clearance Order is hung to ensure dilution paths in the previous Step remain isolated.* ECO #* ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.8.21 PRIOR TO placing the RCS in a "LOOPS NOT FILLED" condition (i.e., Depressurized), ENSURE 0PSP03-CV-0009, Monthly Boration Flow Path Verification, is performed for Verification of Boration Flowpath.NOTE When RCS/Letdown temperature is LESS THAN 115 'F, TV-4494 operates with its disc in the closed position at zero demand. The resultant cavitation at the disc edge causes damage to the valve disc. During prolonged operation with the RCS cooled and vented, the controller should be placed in manual and demand set for 30% to 50% to minimize cavitation damage of the valve disc and optimize CCW flow through the Letdown Heat Exchanger. (Reference 2.110)8.22 WHEN RCS temperature is LESS THAN OR EQUAL to 115 0 F, THEN TRANSFER "CCW TO LTDN HX TEMP CONTR TV-4494" to "MANUAL" control with controller demand set between 30% and 50%.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page86of216 Plant Cooldown Initials CAUTION IF PZR is Solid, THEN ENSURE PZR temperature is less than 2007F before commencing Step 8.23.8.23 REDUCE RCS pressure to approximately 0 (zero) psig by performing the following:

8.23.1 IF PZR Solid, THEN ADJUSTING "PRESS CONT PCV-0135" as indicated on "OUTLET PRESS PI-0135".8.23.2 IF PRZR NOT Solid, THEN PERFORM the following:

8.23.2.1 NOTIFY MAB Watch of pending diversion of letdown flow to the RHT.NOTE Pressurizer level changes SHALL be performed slowly to allow metal soak and steady Nitrogen flow.IF RCS Chemistry gases limits has been met, THEN cooldown of PRZR vapor space may be done by using auxiliary spray as an alternate to Steps 8.23.2.3 and 8.23.2.4 after completion of one (1) full cycle of Pressurizer level.* Steps 8.23.2.3 and 8.23.2.4 may be repeated as required.* One (1) full cycle of Pressurizer level is required to ENSURE the upper Pressurizer metal temperature is reduced.8.23.2.2 VERIFY Chemistry sample of gases (i.e., Iodine, Xenon) in the RCS is within Chemistry limits.8.23.2.3 RAISE Pressurizer level to approximately 95% Cold Calibrated channel level as follows. {CP004}Manually RAISE charging flow using"CHG FLOW CONT VLV FCV-0205".

Manually LOWER letdown flow using"PRESS CONT PCV-0135".

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 87 of 216 Plant Cooldown Initials 8.23.2.4 WHEN Pressurizer level is approximately 95%, THEN LOWER Pressurizer level to less than 50% Cold Calibrated channel level as follows: {CP004}Manually LOWER charging flow using"CHG FLOW CONT VLV FCV-0205".

Manually RAISE letdown flow using"PRESS CONT PCV-0 135".8.24 WHEN RCS pressure is approximately 0 psig, THEN PERFORM the following:

8.24.1 VERIFY Pressurizer vapor space temperature is less than 200'F.8.24.2 OPEN "I(2)-RC-0100, N2 SUPPLY TO PZR ISOL".{RCB 6 ft E of PRT}8.24.3 OPEN "I (2)-RC-0 102, N2 SUPPLY ISOL VALVE TO PZR PORV".{RCB Top of PRZR}8.24.4 OPEN "1 (2)-RC-0 166., PRZR PORV LINE VENT".{RCB Top of PRZR}8.24.5 ADJUST Nitrogen regulator PV-3654 to maintain between 1 and 5 psig on the Pressurizer.

{ 10 ft MAB Non-Radioactive Pipe Chase Room 064}8.25 NOTIFY I&C to ADJUST the PCV-0135 controller gain to "Normal Ops" settings per OPMP08-CV-0135 "Letdown HX Outlet Pressure Calibration".

NOTE Placing the Reactor Coolant Purification Pump (RCPP) in-service will allow letdown flow rates of up to 250 gpm which will facilitate N2 injection to the U-Tubes.8.26 WHEN Pressurizer Cold Calibrated level is less than 80% AND IF RCS Purification is desired, THEN PLACE the Reactor Coolant Purification Pump (RCPP) In Service per OPOP02-CV-0004, Chemical and Volume Control System Subsystem. (Ref 2.63))This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 88 of216 Plant CooldownI Initials NOTE WHEN the Pressurizer and PRT are cross connected, the driving force for flow to vent the Head is from Pressurizer level. Experience shows that levels less than 45% Cold Calibrated level are often insufficient to vent the Head and level occasionally needs to be raised to 50% or higher to allow for Head venting.8.27 ADJUST Charging and Letdown Flow to maintain Pressurizer level between 70% and 30% Cold Calibrated channel level.CAUTION Monitor Reactor Vessel Head for voiding as gasses come out of solution in the Reactor.Monitor RVWL Sensors 1 HJTC Train A or Train C computer points IITE2004 and IITE3004 respectively.

NOTE Venting of the Reactor Vessel Head MAY be PERFORMED as required using Step 3.33 or Step 3.36, as applicable.

8.28 OPEN Pressurizer Relief Tank "1(2)-RC-0025 N2 SUPPLY ISOL".{RCB at PRT}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 89 of 216 Plant Cooldown Initials 9.0 RCS Drain down or Venting 9.1 ENSURE RCS make up plan is in place prior to lowering RCS level below the flange (Ref. OPOP02-SI-0003 for RCS inventory values) (Ref. 2.103).9.2 ENSURE Mid-Loop and RCB Coordinators are available to manage and monitor work activities prior to entry into a reduced RCS inventory condition.

9.3 Draining

the RCS below 0% Pressurizer Cold Calibration Level (elevation 52 ft 2 in) at Step 9.30 is an Infrequently Performed Evolution per OPGP03-ZA-0506, Tests or Evolutions Requiring Additional Controls.ENSURE OPGP03-ZO-0049, Conduct of Tests or Evolutions Requiring Additional Controls, has been reviewed and is being implemented.

CAUTION ALL RCS dilution paths SHALL BE ISOLATED WHEN the RCS in a "LOOPS NOT FILLED" condition, per T. S. 3.4.1.4.2 and 3.9.1. (CR 07-929)9.4 PRIOR TO placing the RCS in a "LOOPS NOT FILLED" condition (i.e., Depressurized), ALL RCS dilution paths SHALL BE ISOLATED by performing the following:

CAUTION OPSP03-CV-0014, CVCS Equipment Verification MAY isolate the letdown flowpath (Example:

closing the Demin outlets before opening the Demin Bypass), ENSURE the letdown flowpath restrictions are planned for PRIOR TO performing OPSP03-CV-0014, CVCS Equipment Verification.

9.4.1 ENSURE

0PSP03-CV-0014, CVCS Equipment Verification, has been performed.

9.4.2 ENSURE

an Equipment Clearance Order is hung to ensure dilution paths in the previous Step remain isolated." ECO #* ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.9.5 IF it is desired to remove water from the SG U-Tubes using nitrogen, THEN PERFORM OPOP02-RC-00 10, Removal of Water From SG U-Tubes Using Nitrogen.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 90 of 216 Plant Cooldown Initials 9.6 IF MID-LOOP or REDUCED INVENTORY Operation is planned, THEN the following MAY be PERFORMED WHILE continuing this procedure:

9.6.1 PERFORM

Prerequisites of OPOP03-ZG-0009, Mid-Loop Operation.

9.6.2 REVIEW

the following to ensure plant status is acceptable for reduced RCS inventory operation:

• Plant Operations Technical Specifications Surveillance Status SM* Night Orders SM* Operability Assessment System SM* Equipment Clearance Orders SM* Temporary Modifications SM* Locked Valve List and Deviation Log SM This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 91 of 216 Plant CooldownI Initials CAUTION The I&C PMs for calibration of the hot leg level indication will result in leaving the instrument valved out at the associated instrurnent blocks to prevent over ranging the instrument.

9.7 IF this is the first Mid-Loop entry in an outage, THEN ENSURE I&C Maintenance has performed the following:

COMPLETED OPMP08-RC-3660, RCS Loop A Level For Mid-Loop Operation Calibration (L-3660) in the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Date/Time COMPLETED OPMPO8-RC-3661., RCS Loop C Level For Mid-Loop Operation Calibration (L-3661) in the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Date/Time CAUTION* Prior to initiating an RCS drain down, the Shift Manager SHALL review the impact of on-going or scheduled work activities on systems affecting RHR cooling capabilities.

• Any activity determined to have a potential for RCS perturbation SHALL be stopped, rescheduled, or completed prior to beginning an RCS drain down. (Ref 2.51)WHEN draining the RCS, THEN Addendum 3 MAY be referred to VERIFY actual volume drained correlates with expected volume drained. (Ref 2.79)9.8 IF the reactor vessel head vent rig has NOT been installed, THEN CONTACT maintenance to install the reactor vessel head vent rig per OPOP07-RC-0001, RC Vent Rig/Sightglass Installation and Removal.9.9 ENSURE low pressure letdown in service with flow to the VCT/RHT per OPOP02-CV-0004, Chemical and Volume Control System Subsystem, Placing Low Pressure Letdown In Service section.9.10 CLOSE "LETDN ORIF HDR ISOL FV-00l1'".

{CP004}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 92 of 216 Plant Cooldown Initials 9.11 CLOSE "ORIF ISOL VLV"s. {CP004}"120-150 GPM""85-100 GPM""25-30 GPM""1(2)-CV-FV-0012""1(2)-CV-FV-0013""1 (2)-CV-MOV-00 14" 9.12 CLOSE letdown isolation valves. {CP004}9.12.1 "LETDN ISOL LCV-0465" 9.12.2 "LETDN ISOL LCV-0468" CAUTION Seal injection SHOULD be maintained in service any time RCS level is LOWERED through the RCP seal package. Seal injection SHALL be maintained in service any time RCS level is RAISED through the seal package for all coupled (non-backseated)

RCPs. Seal injection is NOT required to be in service for uncoupled (backseated)

RCPs. This prevents infiltration of crud from the RCS to the seal package. The bottom of the RCP seal package is at elevation 35 ft. 4 1/2 in. The top of the RCP seal package is at elevation 37 ft. 6 1/2 in. (Ref. CR 09-102-22)

RCS up shaft leakage can occur during the coupling or uncoupling of the RCP shaft.9.13 ISOLATE charging flow by closing "CHG FLOW CONT VLV FCV-0205".

{CP004}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 93 of 216 Plant CooldownI Initials NOTE Perforn Non-Intrusive Check Valve Testing for CC system valves as determined by the Shift Manager/Unit Supervisor and System Engineer per OPEP07-ZE-0008.

9.14 IF the Pressurizer vent path to be established by Step 9.22 is to be open to atmosphere, THEN ENSURE the following valves are CLOSED:* "1(2)-RC-0102 N2 SUPPLY ISOL VALVE TO PRZR PORV". {RCB Top of PRZR}* "I(2)-RC-0100, N2 SUPPLY TO PZR ISOL".{RCB 6 ft E of PRT}* PRT "I (2)-RC-0025 N2 SUPPLY ISOL".{RCB 6 ft E of PRT}9.15 ENSURE braided tubing for both upper and lower RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" connections are connected.

9.16 ENSURE RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" is placed in service per OPOP07-RC-0001., LINEUP 1, "RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup". REFER TO Addendum 4.9.17 COMMENCE temporary logging the perfonnance of OPOP07-RC-0001, LINEUP 1, "RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup" as required. (Step 3.37)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 94 of 216 Plant Cooldown Initials CAUTION Prior to draining the RCS, a vent path (Such as described in Step 3.39) SHALL be established to prevent drawing a vacuum. (Ref 2.52)NOTE IF the RCS is vented to atmosphere, THEN the Steam Generators SHALL NOT be considered a heat sink. (Technical Specification 3.4.1.4.1) 9.18 NOTIFY Health Physics of intention to VENT the Pressurizer to atmosphere using the vent valve on the head vent rig. (Ref OPOP02-RC-0003, Addendum 1) {Follow radiological instructions provided by Health Physics)9.19 ENSURE the equipment hatch emergency closure team and equipment in place per OPMP04-ZG-00 12, Equipment Hatch Removal and Installation, prior to venting the RCS with the equipment hatch open.9.20 ENSURE CLOSED and TAG CLOSED CVCS Auxiliary Spray Level Control Valve "AUX SPRAY LV-3119" with an Equipment Clearance Order (to prevent expelling water from PZR & Head vents).* ECO#* ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.9.21 PLACE the Reactor Vessel head to pressurizer equalizing line in service as follows: 9.21.1 ENSURE the reactor vessel head venting manifold is connected per"OPOP02-RC-0003, Addendum 1, Filling and Venting the RCS".{RCB on RV Head)9.21.2 PERFORM Lineup 1, RV to PZR Equalizing Line Lineup.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 95 of 216 Plant Cooldown Initials CAUTION The following step will breach the RCS. IF unexpected results occur, THEN STOP venting and CONTACT Unit Supervisor for additional instructions.

9.21.3 SLOWLY OPEN "1(2)-RC-0507 RX VESSEL HEAD VENTING MANIFOLD VENT" valve. {RCB on RV Head_9.22 WHEN the Pressurizer has equalized with the atmosphere, THEN ESTABLISH a vent path for the Pressurizer using one of the following paths. (Designate which below)* The Pressurizer spray line vent valves 1(2)-RC-0502 and 1(2)-RC-0503 open. (Preferred Path)* PZR Safety valve removed.9.23 DOCUMENT Vent Path.(Describe) 9.24 RECORD ECO used for Vent Path. ECO#9.24.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.9.25 ENSURE the following valves are closed on the Reactor Vessel Head Vent Rig: "I(2)-RC-0507 RX VESSEL HEAD""VENTING MANIFOLD VENT VALVE"{RCB On RV Head}S "1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN"{RCB On RV Head}* "1(2)-RC-0504 RV HEAD/PRZR""EQUALIZING LINE DRAIN VLV"{73 ft RCB Outside On SG IA(2A) N Shield Wall}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 96 of 216 Plant Cooldown Initials 9.26 ENSURE the following valves are OPEN: " "I(2)-RC-0 132 RX VESSEL HEAD""ATMOSPHERIC VENT VALVE"{RCB On RV Head}"1(2)-RC-0506 RX VESSEL HEAD VENTING""MANIFOLD PI-3636 ISOL" (with PI-3636 installed, OTHERWISE N/A) {RCB On RV Head}"1(2)-RC-0508 RX VESSEL HEAD VENTING"MANIFOLD PRZR EQUAL LINE ISOL"{RCB On RV Head}S "1(2)-RC-0501 RV HEAD/PRZR EQUALIZING"LINE ISOL VLV"{73 ft RCB Outside On SG IA(2A) Shield Wall}" "1(2)-RC-0 163 PZR SPRAY LINE VENT VALVE{RCB Top of PRZR}" "1(2)-RC-0 103 PZR SPRAY LINE VENT VALVE"{RCB Top of PRZR}9.27 ENSURE RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" is being monitored.

9.28 OPEN PZR Spray Valves* RC-PCV-655B

• RC-PCV-655C 9.29 ENSURE "DIVERT LCV-0I 12A" in the AUTO position.

{CP004}9.30 IF the pressurizer level is greater than 10%, THEN Using Low Pressure Letdown and at a RCS Drain Rate of less than 250 gpm, LOWER pressurizer level to 10% Cold Calibrated channel level (55 ft 6 inch elevation) as follows: 9.30.1 DETERMINE RCS volume to be drained to 55 ft 6 inch elevation using Addendum 3, Determination of RCS Volume to be Drained.9.30.2 PLACE "DIVERT LCV-0 I12A" to the RHT position.

{CP004}9.30.3 Manually RAISE letdown flow using "PRESS CONT PCV-0 135".{CP004 }This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 97 of 216 Plant Cooldown I Initials NOTE Each 1% of Pressurizer Cold Calibrated level is between 4 and 5 inches elevation change.Elevation 55 ft 6 inch 10% PZR Cold Calibrated level 9.30.4 As pressurizer level lowers to 10% AND before the PZR level goes off scale low, COMPARE Pressurizer level with RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" level indication.

9.30.5 IF pressurizer level AND RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indications do NOT agree within six inches of each other, THEN STOP the drain and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

9.30.6 IF pressurizer level will be maintained, THEN REFER TO Addendum 19, Controlling RCS Inventory at or above Elv 39 ft. 4.9 in.9.31 PLACE Reactor Vessel head temperature on trend display. (Plant Computer points (upper head) IITE2040 & IITE3040, (sensor #1) IITE2004 & IITE3004, (sensor #2) IITE2007 & IITE3007, (sensor #3) IITE2010 & IITE3010}9.32 Using Low Pressure Letdown and at a RCS Drain Rate of less than 250 gpm, DRAIN the RCS to the 48 ft elevation as indicated on RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)", OTHERWISE N/A 9.32.1 IF RCS level will be maintained, THEN REFER TO Addendum 19, Controlling RCS Inventory at or above Elv 39 ft. 4.9 in.This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 98 of 216 Plant Cooldown Initials CAUTION Draining the RCS has the potential to draw a vacuum in the Reactor Vessel Head.A vacuum in the Reactor Vessel Head will cause the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indication to be in error.Draining should be done at a slow controlled rate with all differences in level indication resolved prior to continued draining. (Ref 2.55)When draining the RCS, then Addendum 3 MAY be referred to VERIFY actual volume drained correlates with expected volume drained. (Ref 2.79)IF during any RCS draining process, fluctuations are observed in RHR pump flow, amps, OR discharge pressure, THEN any RCS drain in progress SHALL be stopped to allow RCS water level to stabilize and any RCS water level recovery SHALL be initiated as necessary to ensure RHR system operation.

Raising RCS level above the topof the RX Head will re-establish the water plug in the RX Head vent to atmosphere.

Periodic venting of the RX Head to remove gasses that come out of solution WILL be required to maintain the RX Head full. The RX Head vent to atmosphere requires opening 1 (2)-RC-0509 to remove the water plug (Loop Seal) during each venting cycle.Due to ALARA concerns from Dose when venting the Head, it is recommended that RCS level NOT be maintained above the top of the RX Head for long periods of time when the RCS is de-pressurized.

IF the RVWL system is available AND indicates less than 100% during Step 9.33, THEN STOP the drain down and INVESTIGATE the level differences.

IF during any RCS draining process, fluctuations are observed in RHIR pump flow, amps, OR discharge pressure, THEN any RCS drain in progress SHALL be stopped to allow RCS water level to stabilize and any RCS water level recovery SHALL be initiated as necessary to ensure RHR system operation.

9.33 At a RCS Drain Rate of less than 250 gpm, REDUCE RCS level to between 47 ft. 4 in. and 46 ft. 5 in. as indicated on the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)".

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 99 of 216 Plant Cooldown Initials NOTE Opening "1(2)-ED-0163 LOWER INTERNALS STORAGE AREA DRAIN" will provide a drain path for "1(2)-RC-0509 RX VESSEL HEAD VENTING MANIFOLD DRAIN".* 1(2)-ED-0163 will be closed in OPOP03-RC-0100, RCS Vacuum Fill or OPOP02-RC-0003, Filling and Venting the Reactor Coolant System.9.34 OPEN "1(2)-ED-0163 LOWER INTERNALS STORAGE AREA DRAIN" to drain the lower internals storage area to the containment normal sump. (19 ft RCB 105' AZ between C & D RCPs)9.35 ENSURE suitable drain hose is connected to "1(2)-RC-0509 RX VESSEL HEAD" "VENTING MANIFOLD DRAIN" {RCB On RV Head}.NOTE WHEN RVWL Sensor Point 1 has been uncovered, THEN indicated temperature will rise to about 750'F due to heating from the heated junction thermocouple.

9.36 VERIFY water in reactor vessel head less than 180'F as indicated by Plant Computer points (tipper head) IITE2040 and IITE3040.CAUTION Following drain of Reactor Head Vent line, containment gas activity could rise.(Ref. 2 .105), 9.37 WHEN RCS level is between 47 ft. 4 in. and 46 ft. 5 in. as indicated on RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)", THEN PERFORM the following:

9.37.1 NOTIFY Health Physics of Reactor Head Vent drain and possible rise in containment gas activity.9.37.2 IF Health Physics requires any compensatory measures for Reactor Head Vent drain, THEN ENSURE compensatory measures are implemented.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 100 of216 Plant Cooldown Initials 9.37.3 OPEN the following valves to remove water plug (Loop Seal) from Head and Pressurizer vent manifold and Rx Head Vent line:* "1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN"{RCB On RV Head}1("1(2)-RC-0507 RX VESSEL HEAD""VENTING MANIFOLD VENT VALVE"{RCB On RV Head}9.37.4 OPEN the Reactor Vessel head vent valves. {CP005}0 "ISOL HV-3657A"* "ISOL HV-3657B"* "ISOL HV-3658A"* "ISOL HV-3658B"* "HEAD VENT THROT VLV HCV-0601"* "HEAD VENT THROT VLV HCV-0602" 9.38 WHEN Head and Pressurizer vent manifold and Rx Head Vent line draining is complete, THEN NOTIFY Mechanical Maintenance that the head vent spool pieces have been drained and may be removed to support Rx Head removal.9.39 IF RCS level is to be raised above 47 ft. 4 in, THEN CLOSE the following valves: "1(2)-RC-0507 RX VESSEL HEAD""VENTING MANIFOLD VENT VALVE"{RCB On RV Head}"1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN"{RCB On RV Head}This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 101 of216 Plant Cooldown Initials 9.40 IF RCS level is to remain below 47 ft. 4 in, THEN PERFORM the following:

CLOSE "1(2)-RC-0507 RX VESSEL HEAD""VENTING MANIFOLD VENT VALVE"{RCB On RV Head}ENSURE "1(2)-RC-0509 RX VESSEL HEAD" "VENTING MANIFOLD DRAIN"{RCB On RV Head} remains OPEN to vent the HEAD.NOTE IF RCS water level reaches 39 ft 3 in. (i.e., Reactor Vessel Flange), THEN SGs can NO longer be considered a heat sink.9.41 IF RCS water level is to be drained down to less than or equal to 39 ft. 3 in.(i.e., Rx Vessel Flange), THEN ENSURE actions associated with Technical Specification LCO 3.4.1.4.2 are met.NOTE The temperature rise will occur when sensor is uncovered prior to RVWL point indicating dry.WHEN RVWL Sensor Point 1 has been uncovered, THEN indicated temperature will rise to about 750'F due to heating from the heated junction thermocouple.

CAUTION Inadequate head venting due to rapid draining can cause Reactor Vessel water level to remain higher than loop level.9.42 USING Low Pressure Letdown and at a RCS Drain Rate of less than 150 gpm, DRAIN the RCS to 45' 3.4" or (IF available) until RVWL Sensor Point 1 heated junction thermocouple temperature rises by 20'F. (Plant Computer Points I1TE2004, A-Train, IITE3004, C-Train)This procedure, when completed, SHALL be retained.

0OP0P3-ZG-0007 Rev. 71 Pg102 of 216 IPlant CooldownI Initials CAUTION Do NOT allow an excessive amount of time to elapse prior to re-establishing Seal Injection to the Coupled RCPs.9.43 IF required, THEN SECURE the following as needed to aid maintenance in back seating Reactor Coolant Pumps.* Seal Injection* Centrifugal Charging Pumps* RC Purification Pump 9.44 IF RVWL is available, THEN COMPARE RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indication with RVWL level.9.45 IF RVWL is available AND RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indications do NOT agree within six inches of each other, THEN STOP the drain and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

9.46 IF RCS level will be maintained, THEN REFER TO Addendum 19., Controlling RCS Inventory at or above Elv 39 ft. 4.9 in.9.47 REFER TO Addendum 3 to determine draindown volume.NOTE The temperature rise will occur when the sensor is uncovered prior to RVWL point indicating dry.9.48 Using Low Pressure Letdown and at a RCS Drain Rate of less than 100 gpm, DRAIN the RCS to 39' 4.9" or (IF available) until RVWL Sensor Point 2 (39'4.9") heated junction thermocouple temperature rises by 20'F. (Plant Computer points IITE2007, A-Train, IITE3007, C-Train)9.49 WHEN RCS level is at RVWL sensor 2 level (39 ft 5 in), IF RVWL is available, THEN COMPARE RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indication with RVWL level.9.50 IF RVWL is available AND RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indications do NOT agree within six inches of each other, THEN STOP the drain and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

This procedure, when completed, SHALL be retained.

7 OPOP03-ZG-0007 Rev. 71 Page 103 of 216 Plant CooldownI Initials 9.51 IF RCS level will be maintained, THEN REFER TO Addendum 19, Controlling RCS Inventory at or above Elv 39 ft. 4.9 in.NOTE Performing OPOPO7-RC-OOO1, RC Vent Rig/Sightglass Installation and Removal, Lineups for RCS Loop Narrow Range Hot Leg level Gauge will NOT result in the hot leg instrument being placed in service. Notification of I&C personnel to be on STANDBY will result in placing these instruments in service sooner.9.52 IF MID-LOOP or REDUCED INVENTORY Operation is planned, THEN the following MAY be PERFORMED WHILE continuing this procedure:

9.52.1 PERFORM Loop I and Loop 3 Narrow Range Hot Leg Level Instruments lineups lAW OPOPO7-RC-OOO1, RC Vent Rig/Sightglass Installation and Removal: " Lineup 2, Loop I Narrow Range Hot Leg Level Valve Lineup* Lineup 3, Loop 3 Narrow Range Hot Leg Level Valve Lineup 9.52.2 REQUEST I&C place the Narrow Range Hot Leg Level instruments in service:* LOOP I LEVEL TRANSMITTER LT-3660* LOOP 3 LEVEL TRANSMITTER LT-3661 I&C I&C 9.52.3 RECORD RVWL, narrow range hot leg level indications and fifth highest core exit temperature TR-001 on CPO18 OR QDPS on OPOPO3-ZG-0009, Mid-Loop Operation, Form 1, Mid-Loop Operation Logsheet, every four hours.9.52.4 RECORD narrow range hot leg sightglass level on OPOPO3-ZG-0009 Form 1, Mid-Loop Operation Logsheet, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.9.52.5 PERFORM OPOPO3-ZG-0009, Form 2, Mid-Loop Checklist, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during reduced RCS inventory operations.

9.52.6 ENSURE Maintenance and testing activities planned OR in progress will NOT cause RCS perturbations.

Mid-Loop Coordinator Date Time 9.53 ADJUST RHR HX flow as required to maintain RFIR HX inlet temperature less than 140'F.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 717 Page 104of216 Plant CooldowiiI Initials 9.54 CHECK status of RCPs: COUPLED UNCOUPLED RCP IA(2A)RCP IB(2B)RCP 1C(2C)RCP ID(2D)9.55 PLACE Seal Injection in service to all coupled RCPs per OPOP02-CV-0004, Chemical and Volume Control System Subsystem.

This procedure, when completed, SHALL be retained.

IOPOP03-ZG-0007 Rev. 71 Page 105 of 216 Plant Cooldown Initials CAUTION Seal injection SHOULD be maintained in service any time RCS level is LOWERED through the RCP seal package. Seal Injection SHALL be maintained in service any time RCS level is RAISED through the seal package for all coupled (non-back seated) RCPs. Seal Injection is NOT required to be in service for uncoupled (back seated) RCPs. This prevents infiltration of crud from RCS to the seal package. The bottom of the RCP seal package is at elevation 35 ft. 4 1/2 in. The top of the RCP seal package is at elevation 37 ft. 6 1/2 in. (Ref. CR 09-102-22)9.56 IF it is desired to LOWER RCS level through the Seal Package with RCPs uncoupled (back seated), THEN PERFORM Steps 9.56.1 through 9.56.3 for the RCPs which are uncoupled (back seated), OTHERWISE GO TO Step 9.57.9.56.1 ENSURE pumps are on their backseat.9.56.2 ROUTE a drain hose from the applicable "RCP SEAL INJ LINE DRAIN" valves to the Containment Sump or a Radiological Drain as appropriate:

{19' RCB IMB in overhead at respective RCP}* RCP "A" 1(2)-CV-0595A RCP "A" 1(2)-CV-0038A

• RCP "B" 1(2)-CV-0595B RCP "B" 1(2)-CV-0038B

• RCP "C" 1(2)-CV-0595C RCP "C" 1(2)-CV-0038C

• RCP "D" 1(2)-CV-0595D RCP "D" 1(2)-CV-0038D 9.36.3 OPEN the following valves as appropriate:

• RCP "A" 1(2)-CV-0595A RCP "A" 1(2)-CV-0038A

• RCP "B" 1(2)-CV-0595B RCP "B" 1(2)-CV-0038B

• RCP "C" 1(2)-CV-0595C RCP "C" 1(2)-CV-0038C

• RCP "D" 1(2)-CV-0595D RCP "D" 1(2)-CV-0038D I This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 106 of 216 Plant Cooldown Initials 9.57 IF MID-LOOP OR REDUCED INVENTORY Operation is planned, THEN PERFORM the following:

9.57.1 OBTAIN Shift Manager and Operations Manager approval to establish the Prerequisites conditions required for entry into MID-LOOP or REDUCED INVENTORY Operation. (REFER TO OPGP03-ZO-0035, Reduced RCS Inventory Operations and OPOP03-ZG-0009, Mid-Loop Operation)

Shift Manager Operations Manager Date Date Time Ti me 9.57.2 COMMENCE the establishment of the Prerequisites conditions required for entry into MID-LOOP or REDUCED INVENTORY Operation. (REFER TO OPGP03-ZO-0035, Reduced RCS Inventory Operations and OPOP03-ZG-0009, Mid-Loop Operation)

CAUTION OPSP03-CV-0014, CVCS Equipment Verification MAY isolate the letdown flowpath (Example:

closing the Detain outlets before opening the Demin Bypass), ENSURE the letdown flowpath restrictions are planned for PRIOR TO performing OPSP03-CV-0014, CVCS Equipment Verification.

9.58 ENSURE OPSP03-CV-0014, CVCS Equipment Verification, has been performed.

An Equipment Clearance Order shall be hung to ensure dilution paths remain isolated.

ECO#9.58.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.CAUTION Reducing RCS level to less than elevation 36 ft 9 in., with nuclear fuel in the Reactor Vessel, is NOT permitted using this procedure. (Ref 2.47)9.59 REDUCE RCS level to between 36 ft 9 in and 37 ft 9 in, as desired.This procedure, when completed, SHALL be retained.

Initials 9.60 IF RVWL is available AND RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indications do NOT agree within six inches of each other, THEN STOP the drain and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

9.61 MAINTAIN RCS level as follows: NOTE IF partial drain down is to accommodate RCP work (i.e., seal package or motor uncoupling), THEN RCS level should be maintained between 36 ft 9 in. and 37 ft 9 in.9.61.1 RAISE RCS level as necessary using gravity drain from the RWST through the LHSI pump and cold leg injection valves in the idle RHR train. (Preferred Method)9.61.2 RAISE RCS level as necessary using a CCP normal charging or seal injection flow paths. (Alternate Method)9.61.3 REDUCE RCS level as necessary using low pressure letdown.9.62 Closely MONITOR RHR pump paramneters.

IF Mode 6 operations are planned, THEN GO TO OPOP03-ZG-00 10, Refueling Operations.

NOTE Reduced inventory is defined as a Reactor Vessel water level less than 36 feet 3 inches.9.63 IF RCS drain down to less than 36 ft 9 in, with nuclear fuel in thle Reactor Vessel is planned, THEN EXIT this procedure and GO TO OPOP03-ZG-0009, Mid-Loop Operation.

9.64 WHEN RCS partial drain down is NO longer desired, THEN FILL RCS per 0POP02-RC-0003, Filling and Venting the Reactor Coolant System.9.65 IF Plant Heat uip is planned, THEN GO TO 0POP03-ZG-0001, Plant Heat up.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 108 of 216 Plant Cooldown Initials 10,0 Secondary Plant Shutdown NOTE IF S/U SGFP 14(24) is NOT required to feed the SGs, THEN applicable steps in this section may be performed in parallel with any other section of this procedure.

10.1 ENSURE the Condensate Polishers are bypassed prior to establishing Secondary Wet Lay up Chemistry controls.10.2 IF desired, THEN ESTABLISH wet lay up chemistry in the condenser hotwell per OPCP01-ZA-0038, Plant Chemistry Specifications.

10.3 VERIFY S/U SGFP 14(24) isolated from the SGs with the preheater bypass valves closed. {CP006}* SG IA(2A) "PREHTR BYPASS FV-7189"* SG 1B(2B) "PREHTR BYPASS FV-7190"& SG 1C(2C) "PREHTR BYPASS FV-7191"* SG ID(2D) "PREHTR BYPASS FV-7192" 10.4 VERIFY low power feedwater regulating valves closed. {CP006}* SG 1A(2A) "LOW PWR FV-7151"* SG IB(2B) "LOW PWR FV-7152"* SG 1C(2C) "LOW PWR FV-7153"* SG 1D(2D) "LOW PWR FV-7154" 10.5 PLACE "S/U SGFP 14(24)" handswitch in the PULL TO LOCK position.{CP006}10.6 OPEN "1(2)-FW-0476 SGFP HEADER BYPASS VLV".{29 ft TGB E Cond Polish}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 109 of216 Plant Cooldown Initials 10.7 CRACK OPEN low power feedwater regulating valves to pressurize the downstream piping. {CP006}* SG I A(2A) "LOW PWR FV-7151"* SG IB(2B) "LOW PWR FV-7152"* SG IC(2C) "LOW PWR FV-7153"* SG I D(2D) "LOW PWR FV-7154" 10.8 ESTABLISH feedwater warm-up flowpath by opening the following valves:{44 ft IVC}* "1(2)-FW-0473, SG I A(2A) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0474, SG 1A(2A) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0471, SG IB(2B) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0472, SG 1B(2B) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0469, SG 1C(2C) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0470, SG IC(2C) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0467, SG 1D(2D) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0468, SG 1D(2D) FW WARM-UP TO COND ISOL VLV".10.9 THROTTLE OPEN low power feedwater regulating valves to establish flow through the warm-up flowpath.

{CP006}SG IA(2A) "LOW PWR FV-7151" SG S B(2B) "LOW PWR FV-7152" SG IC(2C) "LOW PWR FV-7153" SG 1D(2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 110 of 2 16 Plant Cooldown Initials CAUTION If auxiliary steam supplied from the opposite Unit, then changing steam flow will affect Reactor power. Coordinate as necessary with opposite Unit Control Room when opposite Unit main steam supplying auxiliary steam header. Improper control of auxiliary steam can result in a Reactivity Event. (Ref. 2.100)10.10 Slowly COOL the Deaerator by adjusting "AUX STEAM TO DEAERATOR PV-7401" controller PC-7401 output as required to gradually lower steam pressure on the Deaerator.

{Deaerator Stand)10.11 WHEN Deaerator pressure is between 0.3 and 5.0 psig, THEN DISPATCH Operator to open FW-0621, Deaerator Vent Condenser Vent valve{TGB 83'j, to prevent pulling vacuum on DA.10.12 IF Unit Supervisor concurs AND it is desired to raise Deaerator turnover, THEN THROTTLE OPEN the following valves as required: "1 (2)-FW-0486 DEAERATOR STOR TK #2 DUMP""CONT VLV BYPASS VLV"{29 ft TGB S of CNDSR 13(23))"1(2)-FW-0487 DEAERATOR STOR TK #1 DUMP""CONT VLV BYPASS VLV"{29 ft TGB S of CNDSR 13(23))"1(2)-FW-0484 DEAERATOR STORAGE TANK""BLOWDOWN VLV" {55 ft TGB E of CNDSR 11(21)in Overhead by RHDT 1 1A(21A)}S "1(2)-FW-0485 DEAERATOR STORAGE TANK""BLOWDOWN VLV" (55 ft TGB E of CNDSR 13(23)in Overhead by RHDT I IB(21B)}10.13 IF it is desired to go to Intermediate Path Recirc to raise Deaerator turnover, THEN PERFORM the following.

10.13.1 CLOSE condensate "1(2)-CD-0569 RECIRC LINE ISOLATION VALVE". (29 ft TGB E CNDSR Pit in Overhead, Access From 55 ft TGB}10.13.2 OPEN "1(2)-FW-0227 FW CLEANUP ISOL" valve.(55 ft TGB SE Corner)10.13.3 OPEN "1(2)-FW-0228 FW CLEANUP ISOL" valve.155 ft TGB SE Corner}10.13.4 OPEN "FW RECIRC HDR BLOCK MOV-0332".

{CP008)}This procedure, when completed, S1HALL be retained.

OPOP03-ZG-0007 Rev. 71 Page IIIof216 Plant Cooldown Initials 10.13.5 THROTTLE OP EN condensate recirc valves local controllers as required to obtain desired flow rate: "COND RECIRC VALVE FV-7022A"{55 ft TGB NE of CNDSR 12(22)}* "COND RECIRC VALVE FV-7022B"{55 ft TGB NE of CNDSR 12(22)}10.14 WHEN Deaerator temperature reaches 150°F in each DA Storage tank (Plant Computer points T7187, T7188, T7189 and T7190 FW DEAER TEMP), THEN DISPATCH Operator to close the following feedwater warm-up isolation valves: {44 ft IVC}* "1(2)-FW-0473, SG I A(2A) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0474, SG IA(2A) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0471, SG I B(2B) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0472, SG 1B(2B) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0469, SG 1C(2C) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0470, SG 1C(2C) FW WARM-UP TO COND ISOL VLV".* "1(2)-FW-0467, SG 1D(2D) FW WARM-UP TO COND ISOL VLV".S "1 (2)-FW-0468, SG 1D(2D) FW WARM-UP TO COND ISOL VLV".10.15 CLOSE low power feedwater regulating valves. {CP006}S SG I A(2A) "LOW PWR FV-7151"* SG 1B(2B) "LOW PWR FV-7152"* SG IC(2C) "LOW PWR FV-7153"* SG 1D(2D) "LOW PWR FV-7154" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 112 of 216 Plant Cooldown I Initials NOTE Performance of Step 10.16 will generate a feedwater isolation signal due to low Tavg.Rod Control Urgent Alarm and associated control room annunciator "ROD CONT URGENT ALARM" (5M03-B5) will be received when the reactor trip breakers are opened 10.16 IF the reactor trip breakers are closed, THEN OPEN reactor trip breakers by placing the "REACTOR TRIP" switch in the TRIP position.

{CP005}10.17 IF Intermediate Path Recirc was established, AND it is desired to secure Intermediate Path Recirc, THEN PERFORM the following.

10.17.1 CLOSE condensate recirc valves using local controllers:

• "COND RECIRC VALVE FV-7022A"{55 ft TGB NE of CNDSR 12(22)}* "COND RECIRC VALVE FV-7022B"{55 ft TGB NE of CNDSR 12(22)1 10.17.2 CLOSE "FW RECIRC HDR BLOCK MOV-0332".

{CP008}10.17.3 CLOSE "1(2)-FW-0228 FW CLEANUP ISOL" valve.{55 ft TGB SE Comer)10.17.4 CLOSE "1(2)-FW-0227 FW CLEANUP ISOL" valve.{55 ft TGB SE Corner}10.17.5 OPEN "1(2)-CD-0570 RECIRC LINE ISOL BYPASS VALVE". {29 ft TGB E CNDSR Pit in Overhead, Access from 55 ft TGB}10.17.6 OPEN condensate "1(2)-CD-0569 RECIRC LINE ISOLATION VALVE".{29 ft TGB E CNDSR Pit in Overhead, Access From 55 ft TGB}10.17.7 CLOSE "l(2)-CD-0570 RECIRC LINE ISOL BYPASS VALVE".{29 ft TGB E CNDSR Pit in Overhead, Access from 55 ft TGB}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 113 of 216 Plant Cooldown Initials 10.18 ENSURE CLOSED "N2 PURGE TO CONDENSER" valves (NW Corner of CNDSR Pit):* 1(2)-NL-0253, "N2 PURGE TO CONDENSER 11(21)"* 1(2)-NL-0254, "N2 PURGE TO CONDENSER 12(22)"5* 1(2)-NL-0255, "N2 PURGE TO CONDENSER 13(23)" 10.19 ENSURE Steam Generator Blowdown is secured per 0POP02-SB-0001, Steam Generator Blowdown System.10.20 ENSURE Condensate Polishers are bypassed prior to breaking condenser vacuum.10.21 PERFORM the following to break condenser vacuum: 10.2 1.1 ENSURE the following MSIVs and MSIBs are CLOSED: {CP006}* SG IA(2A) "MSIV FSV-7414"* SG 1B(2B) "MSIV FSV-7424"* SG 1C(2C) "MSIV FSV-7434"* SG ID(2D) "MSIV FSV-7444"* SG IA(2A) "MSIB FV-7412"* SG 1B(2B) "MSIB FV-7422"* SG IC(2C) "MSIB FV-7432"* SG ID(2D) "MSIB FV-7442" 10.21.2 IF desired to depressurize the Main Steam Header, THEN SLOWLY OPEN Steam Dumps as needed to depressurize Main Steam Header.10.21.2.1 WHEN Main Steam Header is depressurized, THEN CLOSE Steam Dumps.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 114 of216 Plant Cooldown Initials 10.21.3 PLACE ALL the non-running "CONDENSER VACUUM PUMP LOCAL HANDSWITCH" in the "OFF" position.10.21.4 OPEN vacuum breakers and PLACE ALL condenser vacuum pumps in STOP per 0POP02-CR-0001, Main Condenser Air Removal.{CPO09}* "VAC BKR 11(21) MOV-0023"* "VAC BKR 12(22) MOV-0024"* "VAC BKR 13(23) MOV-0025"* CR Pump 11 (21)* CR Pump 12(22)* CR Pump 13(23)CAUTION If gland steam is being supplied from auxiliary steam from opposite Unit, then changing gland steam flow will affect Reactor power. Coordinate as necessary with opposite Unit Control Room when opposite Unit main steam supplying auxiliary steam header. Improper control of auxiliary steam can result in a Reactivity Event. (Ref. 2. 100)10.22 SECURE gland steam per 0POP02-GS-0001, Turbine Gland Seal Steam System.10.23 STOP all feedwater booster pumps. {CP008}* "FW BOOSTER PUMP 11(21)"* "FW BOOSTER PUMP 12(22)"* "FW BOOSTER PUMP 13(23)" 10.24 IF desired, THEN PLACE low pressure feedwater heaters in wet lay up per OPOP02-CD-0001, Condensate System.10.25 IF desired, THEN PLACE high pressure feedwater heaters in wet lay up per OPOP02-FW-0001, Main Feedwater.

I This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 115 of 216 Plant Cooldown Initials 10.26 IF Unit Supervisor authorized the following valves to be opened, AND it is desired, THEN CLOSE the following valves: "1(2)-FW-0486 DEAERATOR STOR TK #2 DUMP""CONT VLV BYPASS VLV"{29 ft TGB S of CNDSR 13(23)}S "1(2)-FW-0487 DEAERATOR STOR TK #1 DUMP""CONT VLV BYPASS VLV"{29 ft TGB S of CNDSR 13(23)}S "1(2)-FW-0484 DEAERATOR STORAGE TANK""BLOWDOWN VLV" {55 ft TGB E of CNDSR 11(21)in Overhead by RHDT I IA(21A)}S "1 (2)-FW-0485 DEAERATOR STORAGE TANK""BLOWDOWN VLV" {55 ft TGB E of CNDSR 13(23)in Overhead by RHDT 1 IB(21B)}10.27 STOP the Condensate Pumps. {CP009}* "COND PUMP 11(21)"* "COND PUMP 12(22)"* "COND PUMP 13(23)" 10.28 WHEN all Condensate Pumps are secured, THEN SWAP FWBP Seal Water Strainers per 0POP02-FW-0001, Main Feedwater.

11.0 Records Review 11.1 REVIEW completed procedure package to ensure all applicable sections are completed as required.Shift Manager/Unit Supervisor Date This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 116 of1216 Plant Cooldown 12.0 Support Documents 12.1 Addendum 1L RCS Cooldown Limitations 12.2 Addendum 2, RCS/RHR Simplified Elevation Diagram 12.3 Addendum 3, Determination of RCS Volume to be Drained 12.4 Addendum 4, RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" Upper Connection Assembly Diagram 12.5 Addendum 5, RCS Degassification 12.6 Addendum 6, Degassing the RCDT and PRT 12.7 Addendum 7, Conditions for Steam Generator Decay Heat Removal 12.8 Addendum 8, Movable Gripper Releasing 12.9 Addendum 9, Plant Cooldown with the PZR Water Solid 12.10 Addendum 10, MODE 5 Cooldown with MSIVs OPEN 12.11 Addendum 11, RCS/PZR Pressure Operations Guideline 12.12 Addendum 12, Solid Plant Operations Entry Checklist 12.13 Addendum 13, Manual Blowdown of Main Steam lines upstream of MSIVs 12.14 Addendum 14, MOV-0016A, B & C Emergency Operations Guideline 12.15 Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline 12.16 Addendum 16, Throttling "1(2)-CV-0198 RMW ISOL" for Technical Specification 3.4.1.3 12.17 Addendum 17, Moving Control Rods with High Reactor Coolant Turbidity 12.18 Addendum 18, Indicated Pressurizer Level When Solid vs. Pressurizer Temperature 12.19 Addendum 19, Controlling RCS Inventory at or above Elv 39 ft. 4.9 in.12.20 Addendum 20, Venting Reactor Vessel Head Using Head Vent Throttle Valve(s)12.21 Addendum 21, Closure of Personnel Air Lock Doors 12.22 Data Sheet 1, RCS and Pressurizer Cooldown Rates 12.23 Lineup 1, RV to PZR Equalizing Line Lineup 12.24 Form 1, CVCS Line Boration Tracking Form This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 117 of216 Plant Cooldown Addendum 1 I RCS Cooldown Limitations Page 1 of I RCS COOL-DOWN LIMIT CURVE MATERIAL PROPERTi' BASIS COv. TE'.L _L~r C ..M.TERi.AL

-- ",.INTERMEDIATE SHELL R-1606--3 COPPER CONTENT: CONSERVATIVELY

&SSUMNED AS 0.)0 WTO-R O NDT IT [ AL. 10:F RT NOT AFTER 32 EFFPY 1/4, 91' F 3/..4-T, 647F SINGLE CLR\VE 12,PPLICAELE FOR COOLDOWN RATES UP TO 100"F-/HR FOR THE SEPR.,'ICE PERIOD JP TO 32 EFP'i. AHD OJ'.ITAIKS MARCIKS OF 1]0F AND 60 PSIG FOR POSSIBLE INSTRUMENKT ERRORS.30O0 w 999 C.)w 9999 (99 92000 UN .ACCEPTABLE

_COOLDCWN PATE 'F."HR ACJCEPTABLE (6-2 1 P S110)25--I--50_ -SO___I (O (1 207)1000 0 1 0 0 300 INDICATED TEMPERATURE

('F)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 f Page 118 of216 Plant Cooldown Addendum 2 RCS/RHR Simplified Elevation Diagram Page 1 of 1 REACTOR COOLANT SYSTEM PRESSURIZER T '&"RHR (out)SECTION A-A HOT LEG STP D-0794 Rev 2 This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 119of216 Plant Cooldown S Addendum 3 Determination of RCS Volume to be Drained Page 1 of I RCS DRAIN RCS DRAIN REFERENCE PLANT DOWN RHT LEVEL DOWN RHT LEVEL ELEVATION VOLUME CHANGE VOLUME CHANGE WITHOUT N 2 (%)(1) WITH N 2 (%)(1)(gallons) (gallons)Pressurizer 50% Level 68 ft 10 in (3) --- (2) ---101455 94662 Pressurizer 10% Level 55 ft 6 in 96016 6.4 (2)2_____________77021 2.Reactor Vessel Flange 39 ft 3 ' (2)(RVWL Sensor Point 2) (.39 ft 5 in) 83564 14.6 5412726.9 4.5 ft below Vessel Flange 34 ft 10.1 in 75440 9.6 (2) 16.3 (RVWL Sensor Point 3) (34 ft 10.25 in) 40307 Top of Hot Leg Nozzle (RVWL Sensor Point 4) 33 ft 5.5 in 36345 46.0 36345 4.7 Narrow Range Hot Leg 33 ft 2 in Level +11 inches (33 ft 5.625 in) 35,468 1.0 35,468 1.0 Narrow Range Hot leg 32 ft 9 in 32,636 3.3 32,636 3.3 Level +6 inches (alarn)Hot Leg Centerline (RVWL Sensor Point 5) 32 ft 3 in 30,281 2.8 30,281 2.8 Bottom of Narrow Range 31 ft 11.5 in 29,079 1.4 29,079 1.4 Hot Leg Level Indication

_I I 1 (1) DRAIN DOWN -Level change is percent level rise from reducing RCS level from previous reference point.(2) NITROGEN INJECTION

-RCS volume and RHT level changes are variable and based on the AMOUNT AND RATE of Nitrogen injected into the Steam Generators.

These levels are also variable based upon the CVCS letdown rate and other plant conditions.

(3) VARIANCE -Nitrogen is assumed NOT injected at this point.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 120 of216 Plant Cooldown Addendum 4 RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" Page I of I Upper Connection Assembly Diagram RC-01 02 FLEX HOSE RC-0166 SIGHTGLASS LG-3662 RCS LOOP A INTERMEDIATE LEG RC-0057A RC-0200 RC-0518 RC-0058A SIGHTGLASS DRAIN VALVE CD101660(06/14/00)

DRAIN This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 121 of 216 Plant Cooldown Addendum 5 RCS Degassification Page 1 of 3 Initials NOTE RCS Degasification can be used for removal of Hydrogen gas and/or Fission Product Gasses as recommended by Chemistry.

Steps in this addendum may be re-sequenced and re-performed based on Chemistry recommendations.

1.0 IF recommended by Chemistry THEN, ESTABLISH nitrogen cover gas on the Volume Control Tank (VCT) as follows: {CVCS Chemical Mixing Tank Room 51 ft MAB}1.1 TAG CLOSED "1(2)-CV-0178 HYDROGEN SUPPLY TO VCT ISOLATION VALVE" with an Equipment Clearance Order.1.2 OPEN "1(2)-CV-0181 NITROGEN SUPPLY TO VCT ISOLATION VALVE".1.3 OPEN "1 (2)-NL-0033 LOW PRESSURE NITROGEN TO VOLUME CONTROL TANK ISOLATION VALVE".1.4 RECORD ECO#1.5 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.2.0 INITIATE Mechanical RCS degassing as follows: CAUTION During performance of this addendum the VCT gas flow to the GWPS may be raised. Do NOT exceed 2.5 psig on GWPS inlet header OR 5.8 SCFM system outlet flow. (Prevents GWPS from tripping)2.1 NOTIFY MAB Watch of pending degasification so they can raise gas flow as allowed per OPOP02-WG-0001, Gaseous Waste Processing System.2.2 NOTIFY Health Physics of pending degasification to the Gaseous Waste Processing System.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 122 of 2166 Plant Cooldown Addendum 5 RCS Degassification Page 2 of 3 Initials NOTE During Mechanical degas the effluent from the primary sample sink is directed back to VCT during pressurizer vapor space degassing.

2.3 IF performing a Mechanical degas, THEN NOTIFY Chemistry degassing of the pressurizer vapor space may commence.2.4 OPEN "VCT VENT PCV-01 15". {CP004}2.5 PLACE "DIVERT LCV-01 12A" in the VCT position.

{CP004}CAUTION To prevent lifting the VCT relief valve, VCT pressure SHALL NOT be allowed to exceed 65 psig during fill.T To prevent RCS boron dilution, RCS makeup flow boron concentration SHALL be at greater than or equal to RCS boron concentration.

2.6 RAISE

VCT level to between 90 and 95% using "RC M/U CONT" OR gravity feed from the RWST, while maintaining VCT pressure between 15 and 30 psig (PI-0115).

{CP004}2.7 VERIFY VCT level between 90 and 95% on "VCT LEVEL LI-01 12". {CP004}2.8 PLACE "RC M/U CONT" in the STOP position.

{CP004}2.9 ALLOW VCT pressure to decay to the new minimum value (approximately 15 to 20 psig) while maintaining VCT level between 90 and 95%. {CP004}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 123 of216 Plant Cooldown Addendum 5 RCS Degassification Page 3 of 3 Initials NOTE IF the LWPS does NOT have the capacity to receive the water from letdown or it is desired to minimize water usage, THEN it is permissible to adjust charging and letdown to obtain the desired VCT levels.It is permissible to raise Pressurizer level in anticipation of the level shrink due to a cooldown.

NA when PZR is Solid. (Ref 3.9)2.10 WHEN notified by Chemistry to place nitrogen cover gas on the Volume Control Tank (VCT) THEN, 2.10.1 ENSURE nitrogen cover gas on the Volume Control Tank (VCT) as follows:{CVCS Chemical Mixing Tank Room 51 ft MAB}2.10.1.1 TAG CLOSED "1(2)-CV-0178 HYDROGEN SUPPLY TO VCT ISOLATION VALVE" with an Equipment Clearance Order.2.10.1.2 OPEN "1(2)-CV-0181 NITROGEN SUPPLY TO VCT ISOLATION VALVE".2.10.1.3 OPEN "1 (2)-NL-0033 LOW PRESSURE NITROGEN TO VOLUME CONTROL TANK ISOLATION VALVE".2.10.2 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.2.10.3 ENSURE "1(2)-CV-PV-31 11 NITROGEN TO VCT PV" maintains VCT pressure greater than or equal to 15 psig. {CP004}2.11 IF performing a Chemical degas AND additional Mechanical degasification is NOT required, THEN maintain VCT level between 90 and 95% until notified by Chemistry that the level may be lowered in preparation for the Peroxide addition.2.12 LOWER VCT level to 30% by DIVERTING to the RHT OR by raising Pressurizer level for RCS Cooldown (Ref 3.9)2.13 IF additional RCS degasification is required, THEN RETURN TO Step 2.5 and perform Steps 2.5 through 2.11 as necessary to complete RCS degasification.

2.14 RETURN "RC M/U CONT" to the configuration desired by the Unit Supervisor.

{CP004}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 124 of'216 Plant Cooldown Addendum 6 Degassing the RCDT and PRT Page 1 of 3 1.0 Prerequisites 1.1 N2 supply system is available to the Reactor Coolant Drain Tank (RCDT) and Pressurizer Relief Tank (PRT).1.2 Gaseous Waste Processing System (GWPS) is in service per OPOP02-WG-0001, Gaseous Waste Processing System Operations.

2.0 Discussion

Degassing the PRT and RCDT will purge both tanks to the GWPS with nitrogen to remove Hydrogen and fission product gases. There are 2 flowpaths available for degassing the PRT and RCDT. H 2 Degassing RCDT 4 PRT 4 GWPS and H 2 Degassing PRT 4 RCDT 4 GWPS.Both methods may be used to ensure all Hydrogen is removed as recommended by Chemistry.

However, the flowpath of H 2 Degassing RCDT 4 PRT 4 GWPS provides the best purge flow indications because of 2 instruments in the purge line. The flowpath of H 2 Degassing PRT 4 RCDT 4 GWPS provides the best purge flowrate due to a large N 2 supply regulator.

Good Operational practice has shown that the flowpath of H 2 Degassing PRT 4 RCDT 4 GWPS should be established and allowed to continue until both the PRT and RCDT H 2 are less than 4 %and the RCS H 2 is less than 4 cc/kg. This can be modified by Chemistry or the Shift Manager/Unit Supervisor.

3.0 Notes

and Precautions

3.1 Performance

of this addendum will connect the PRT and the RCDT and allow for simultaneous degassification of both tanks.3.2 Performance of this addendum vents the PRT and RCDT to the GWPS.3.3 Degassification of the PRT and RCDT should be performed until the RCS H 2 is less than 4 cc/kg or as recommended by Chemistry.

3.4 Do NOT exceed 2.5 ps ig on GWPS inlet header OR 5.8 SCFM system outlet flow.(Prevents GWPS from tripping)3.5 Placing the PRT in recirc SHOULD facilitate a faster Degass.3.6 Valve cycling/testing during cooldown may add Hydrogen to the PRT.3.7 The explosive range of hydrogen gas concentration is from 4 to 70%.3.8 Based on experience, purging H 2 from the PRT and the RCDT takes approx. 9 to 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 125 of216 Plant Cooldown Addendum 6 Degassing the RCDT and PRT Page 2 of 3 Initials 4.0 Procedure 4.1 ENSURE pre-Job briefing performed prior IAW Conduct of Operations Chapter 2.4.2 IF desired, THEN PLACE the PRT is in RECIRC per OPOP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation.

4.3 DEGASS

the PRT and the RCDT by performing one the following, N/A option NOT performed: (Largest N 2 Regulator Flowpath)

PERFORM degassing the PRT and the RCDT per OPOP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation, Section "H 2 Degassing PRT 4 RCDT 4 GWPS".(Best Instrumented Flowpath)

PERFORM degassing the PRT and the RCDT per OPOP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation, Section "H 2 Degassing RCDT 4 PRT 4 GWPS".4.4 IF desired by Chemistry or the Shift Manager/Unit Supervisor, WHEN tile PRT/RCDT H 2 is less than 4 % AND the RCS H 2 is less than 4 cc/kg OR as requested by Chemistry or the Shift Manager/Unit Supervisor, THEN reverse the DEGASS flow of the PRT and the RCDT by performing the method NOT selected in previous Step, N/A NOT performed option: PERFORM degassing the PRT and the RCDT per OPOP02-RC-0001., Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation, Section "H 2 Degassing RCDT 4 PRT 4 GWPS".PERFORM degassing the PRT and the RCDT per OPOP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation, Section "H 2 Degassing PRT 4 RCDT 4 GWPS".This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 126 of 216 Plant Cooldown Addendum 6 Degassing the RCDT and PRT Page 3 of 3 Initials 4.5 IF desired, THEN STOP the PRT RECIRC per OPOP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation.

4.6 WHEN the PRT/RCDT H, is less than 4 % AND the RCS H2 is less than 4 cc/kg, THEN PERFORM OPOP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation, Restoration of PRT/RCDT/GWPS from H 2 Degassing.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 127 of 216 Plant Cooldown Addendum 7 Conditions for Steam Generator Decay Heat Removal Page 1 of I 1.0 ENSURE the following conditions exist for SGs credited for decay heat removal means in Initials Mode 5: 1.1 VERIFY Pressurizer heaters available to control RCS pressure.1.2 VERIFY AFW available to feed credited SGs.1.3 VERIFY SG PORVs functional for credited SGs.1.4 VERIFY that the RCS is NOT vented.1.5 VERIFY water level in the credited SGs GREATER THAN 10% narrow range.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 128 of 216 Plant Cooldown Addendum 8 Movable Gripper Releasing Page 1 of 4 Initials NOTE IF the reactor trip breakers open unexpectedly, THEN this addendum may be reperformed as necessary to withdraw control rods.Steps may be marked NA for Rod Banks that have NO potentially engaged Movable Grippers.Insertion attempts of sticking control rods should be completed prior to proceeding.

1.1 Document

in the Station Log which Control Rods are required to be released by this addendum.1.2 PLACE Rod Drive MG Set in service per OPOP02-RS-0001, Rod Control.1.3 ENSURE a reactor trip signal is NOT present.1.4 CLOSE Reactor Trip Breakers.1.5 TURN "ROD CONT SU" switch to RESET and VERIFY the following occurs: 1.5.1 All step counters reset to zero.1.5.2 "ROD CONT URGENT ALARM" resets as evident by viewing the annunciator extinguishes.

1.5.3 "ROD CONT NON URGENT ALARM" resets as evident by viewing the annunciator extinguishes.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 129 of 216]Plant Cooldown Addendum 8 Movable Gripper Releasing I Page 2 of 4 I Initials CAUTION MONITOR nuclear instrumentation continuously and anticipate criticality while Control Rods are being withdrawn.

1.6 PLACE

"ROD BANK SEL" switch in SB-A position.NOTE DRPI may NOT see a 2 Step Rod movement.1.7 PLACE Rod Control switch in OUT position to WITHDRAW Shutdown Bank A to 2 steps as indicated on Group step counters.1.8 OBSERVE Shutdown Bank A, Group 1 and 2 step counters indicate rod withdrawal.

1.9 PLACE

"ROD BANK SEL" switch in "SB-B" position.1.10 PLACE Rod Control switch in OUT position to WITHDRAW Shutdown Bank B to 2 steps as indicated on Group step counters.1.11 OBSERVE Shutdown Bank B, Group 1 and 2 step counters indicate rod withdrawal.

1.12 PLACE "ROD BANK SEL" switch in "SB-C" position.1.13 PLACE Rod Control switch in OUT position to WITHDRAW Shutdown Bank C to 2 steps as indicated on Group step counter.1.14 OBSERVE Shutdown Bank C step counter indicates rod withdrawal.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 130 of216 Plant Cooldown Addendum 8 Movable Gripper Releasing Page 3 of 4 Initials 1.15 PLACE "ROD BANK SEL" switch in "SB-D" position.1.16 PLACE Rod Control switch in OUT position to WITHDRAW Shutdown Bank D to 2 steps as indicated on Group step counter.1.17 OBSERVE Shutdown Bank D step counter indicates rod withdrawal.

1.18 PLACE "ROD BANK SEL" switch in "SB-E" position.1.19 PLACE Rod Control switch in OUT position to WITHDRAW Shutdown Bank E to 2 steps as indicated on Group step counter.1.20 OBSERVE Shutdown Bank E step counter indicates rod withdrawal.

1.21 PLACE "ROD BANK SEL" switch in "CB-A" position.1.22 PLACE Rod Control switch in OUT position to WITHDRAW Control Bank A to 2 steps as indicated on Group step counters.1.23 OBSERVE Control Bank A, Group I and 2 step counters indicate rod withdrawal.

1.24 PLACE "ROD BANK SEL" switch in "CB-B" position.1.25 PLACE Rod Control switch in OUT position to WITHDRAW Control Bank B to 2 steps as indicated on Group step counters.1.26 OBSERVE Control Bank B, Group I and 2 step counters indicate rod withdrawal.

1.27 PLACE "ROD BANK SEL" switch in "CB-C" position.1.28 PLACE Rod Control switch in OUT position to WITHDRAW Control Bank C to 2 steps as indicated on Group step counters.1.29 OBSERVE Control Bank C, Group 1 and 2 step counters indicate rod withdrawal.

This procedure, when completed, SHALL be retained.

7 -OPOP03-ZG-0007 Rev. 71 Page 131 of 216 Plant Cooldown Addendum 8 Movable Gripper Releasing Page 4 of 4 Initials 1.30 PLACE "ROD BANK SEL" switch in "CB-D" position.1.31 PLACE Rod Control switch in OUT position to WITHDRAW Control Bank D to 2 steps as indicated on Group step counters.1.32 OBSERVE Control Bank D, Group 1 and 2 step counters indicate rod withdrawal.

NOTE Rod Control Urgent Alarm and associated control room annunciator "ROD CONT URGENT ALARM" (5M03-B5) will be received when the reactor trip breakers are opened.CAUTION Do NOT GO TO OPOP04-RS-0001, Rod Control Malfunction, after performing Step 1.33 of this addendum.1.33 OPEN Reactor Trip Breakers.1.34 IF Control and Shutdown Rods are NOT inserted (Rod Bottom Lights on), THEN EVALUATE Shutdown Margin.This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 132 of 216 I I Plant Cooldown Addendum 9 Plant Cooldown with the PZR Water Solid Page 1 of 8 NOTE The administrative cooldown rate is 80°F/hr. (Ref 2.70)The administrative cooldown limit for the Pressurizer is 160°F/hr.The cooldown should be spaced over the entire 60 minute period vice a short time high rate cooldown. (Ref 2.50)* MONITOR RCS and Pressurizer temperature and pressure per Data Sheet 1.* PRESS CONT PCV-0135 may NOT respond fast enough to limit pressure transients, IMMEDIATE MANUAL Operation/Intervention to control pressure is AUTHORIZED.

Whenever the plant is water solid AND the reactor coolant pressure is being maintained by low pressure letdown, it is recommended to keep all three letdown orifices open. (Ref 2.65.1 and 2.95)If all RCPs have stopped for more than 5 minutes and the RCS temperature is greater than the charging and seal injection water temperature, DO NOT start the first RCP until a steam bubble is formed in the pressurizer. (Ref 2.65.1, UFSAR 5.2.2.11.3)

If all RCPs are stopped and the reactor coolant is being cooled down by the residual heat removal heat-exchangers, a non-uniform temperature distribution may occur in tile RCS.DO NOT attempt a start of a RCP unless a steam bubble exists in the pressurizer or an acceptable temperature profile can be demonstrated. (Ref 2.65.1, UFSAR 5.2.2.11.3)

When the reactor coolant pressure is being maintained by the low pressure letdown control valve during water solid operation, changes to the flow rate through the RHR loop by throttling of valves or starting and stopping the RHR pumps may cause a rise in reactor coolant pressure.For example, stopping the RHR pumps may cause a rise in the reactor coolant pressure of between 100 and 150 psig. (Ref 2.65.1)It is recommended where possible that the RHR train that is used for low pressure letdown NOT be used for RCS cooling. This will minimize pressure transients while adjusting RCS cooldown rates.During plant cooldown, all SGs will normally be connected to the steam header to assure a uniform cooldown of the RCS. (UFSAR 5.2.2.11.3)

WHEN RCS Temperature is less than 140'F, THEN ENSURE less than 4 RCPs in operation.

(3 pumps or less restriction, This limitation was required to demonstrate acceptable hold down forces in the Cold Zero Power lift force calculation.)

REFER TO Addendum 18, Indicated Pressurizer Level When Solid vs. Pressurizer Temperature as an aide in estimating Pressurizer Cold Calibrated level indications when Pressurizer is solid. (Ref. 2.104)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 133 of 216 Plant Cooldown Addendum 9 Plant Cooldown with the PZR Water Solid Page 2 of 8 CAUTION In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.* Control RCS Pressure using Spray and PZR Heaters until directed otherwise.

• WHEN available RHR train suction valves have been opened, THEN Maintain RCS Pressure between 325 psig and 400 psig.DO NOT Isolate all RHR Trains inlet lines from the reactor coolant loop unless there is a steam bubble in the pressurizer or the charging pumps are stopped. (Ref 2.65.1)Use of a pressurizer steam bubble during periods of low pressure, low temperature operation is preferred.

This steam bubble will dampen the plant's response to potential transient generating inputs, providing easier pressure control with slower response rates. (UFSAR 5.2.2.11.3)

Initials 1.0 PLACING the RCS in a SOLID Condition:

1.1 PERFORM

"Addendum 12, Solid Plant Operations Entry Checklist".

1.2 OBTAIN

permission to conduct Solid Plant Operations.

Shift Manager Operation Manager 1.3 NOTIFY I&C to calibrate PCV-0135 controller to "Solid Ops" settings per 0PMP08-CV-0 135, Letdown HX Outlet Pressure Calibration.

1.4 ENSURE

low pressure letdown in service per OPOP02-CV-0004, Chemical and Volume Control System Subsystem.

1.5 ENSURE

Residual Heat Removal (RHR) Train A or B in service per OPOP02-RH-0001, Residual Heat Removal.This procedure, when completed, SHALL be retained.

OPOP0 3-ZG-0007 Rev. 71 Page 134 of 216]Plant Cooldown Addendum 9 Plant Cooldown with the PZR Water Solid Page 3 of8 I NOTE The RCP configuration recommended for PZR Water Solid conditions are as follows: " (Preferred)

RCP ID(2D) [RCP 1D(2D) is the only pump with adequate spray flow in single pump configuration]

• (Alt 1) RCP 1D(2D) and RCP 1A(2A)* (Alt 2) RCP 1 D(2D) and RCP 1 B(2B) or RCP I C(2C)* (Alt 3) RCP 1 A(2A) and RCP I B(2B) or RCP 1 C(2C)" (Alt 4) RCP 1B(2B) and RCP 1C(2C) (Inadequate RCP spray flow this configuration, Aux Spray required)* For Planned Outages, the RCP configuration recommended by the Outage Planning Team should be considered; the Shift Manager/Unit Supervisor has final selection judgment.For Un-Planned Outages, the RCP configuration recommended by Health Physics and Chemistry should be considered; tile Shift Manager/Unit Supervisor has final selection judgment.IF NO RCP configuration recommendation is provided, THEN the default RCP configuration is RCP 1D(2D) running and 3 RCPs secured, the Shift Manager/Unit Supervisor has final selection judgment.* The Shift Manager/Unit Supervisor may authorize changes to the operating RCP configuration during the cooldown if conditions warrant (i.e. large CRUD burst, pump problems, cleanup issues, etc)." PRIOR to securing an RCP during the remainder of the cooldown, CONTACT Chemistry and Health Physics for input on CRUD Cleanup.* In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.The Shift Manager/Unit Supervisor may authorize the STOPPING of additional operating RCPs per OPOP02-RC-0004, Operation of Reactor Coolant Pumps, WHEN the cooldown rate lowers to undesirable levels or for outage support activities." Four (4) RCP operation NOT permitted WHEN RCS average temperature is less than 140'F.Do NOT run all 4 RCPs UNTIL RCS average temperature reaches greater than 140'F.(3 pumps or less restriction RCS < 1407F, This limitation is required to demonstrate acceptable fuel assembly top nozzle hold down spring forces in the Cold Zero Power lift force calculation.)

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 135 of 216 ,Plant Cooldown Addendum 9 Plant Cooldown with the PZR Water Solid Page 4 of 8 NOTE (Good Operational practice)

MAXIMIZE the Reactor Coolant System flowrate during CRUD Cleanup to reduce DOSE rates for the remainder of the Outage." (Good Operational practice)

Do NOT have a planned RCP start during PZR Water Solid conditions.

IF RCPs require starting, THEN they should be performed prior to the PZR Water Solid conditions.

Initials 1.6 ESTABLISH the operating RCP(s) configuration for PZR Water Solid operations, by PERFORMING the following:

1.6.1 IF Planned Outage, THEN CONTACT Outage Planning Team for RCP configuration for PZR Water Solid operations recommendation, the Shift Manager/Unit Supervisor has final selection judgment.1.6.2 CONTACT Chemistry for RCP configuration for PZR Water Solid operations recommendation and to verify that RCS activity level (crud) is acceptable AND continued full flow clean-tip is NOT required, the Shift Manager/Unit Supervisor has final selection judgment.1.6.3 CONTACT Health Physics for RCP configuration for PZR Water Solid operations recommendation, the Shift Manager/Unit Supervisor has final selection judgment.1.6.4 SECURE or START the applicable RCPs per OPOP02-RC-0004, Operation of Reactor Coolant Pump as determined by the Shift Manager/Unit Supervisor.

{CP005}" "RCP 1A(2A)"" "RCP 1B(2B)"* "RCP IC(2C)"" "RCP ID(2D)" (preferred running pump)1.6.5 NOTIFY Health Physics to monitor for rising area radiation due CRUD burst during cooldown.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 136 of 216 Plant Cooldown I E Addendumn 9 Plant Cooldown with the PZR Water Solid Page 5 of 8 I Initials NOTE* Temperature Stratification may occur in the PZR while going Solid. Cooldown Rates may suddenly be exceeded as the Liquid Space and Vapor Space thermocouples are exposed to these differences.

• SSPS Actuation Cabinets "A" and "B" and Logic Cabinets "R" and "S" must be energized for both COMS to be OPERABLE.

Logic Cabinets "R" and Actuation Cabinets "A" required for PCV 0655A "Pressurizer PORV Train A". Logic Cabinets "S" and Actuation Cabinets "B" required for PCV 0656A "Pressurizer PORV Train B". {LER 2-05-003, SSPS ECO to support FWIV energize to actuate MOD in 2RE10 caused COMS to be inoperable during solid plant operations in violation of TS 3.4.9.3. (CR 05-5960, CR 05-3071)}.

1.7 ENSURE

the Cold Overpressure Mitigation System is ARMED.(Technical Specification 3.4.9.3) {CP004}* "OVERPRESS MIT" for PCV-0655A* "OVERPRESS MIT" for PCV-0656A 1.8 PLACE "CHG FLOW CONT FK-0205" in MAN.NOTE* Letdown pressure and flow are affected by RHR Pump head and RCS pressure changes. During RCS pressure/temperature changes and RHR pump flow changes, letdown flow should be monitored on "LETDN HX OUTL FLOW FI-0 132" AND "PRESS CONT PCV-0135" AND should be Adjusted As Required to maintain desired letdown flow and Pressurizer level." The Recommended Letdown flowrate for chemical degas is approx. 220 gpm.* To prevent pump runout, Charging Pump flowrate at this pressure and temperature should NOT exceed 300 gpm.1.9 ADJUST "PRESS CONT PCV-0135" to obtain the 220 gpm or desired letdown flow (as indicated on "LETDN HX OUTL FLOW FI-0132")1.10 ENSURE RCS pressure between 325 psig and 400 psig. {PT-0403 or PT-0404}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 137 of]216 Plant Cooldown Addendum 9 Plant Cooldown with the PZR Water Solid Page 6 of 8 Initials 1.11 ADJUST "CHG FLOW CONT FK-0205" to maintain Total Charging flow (Charging

+ Seal Injection) as required for Pressurizer Level Control (Letdown flow + shrinkage).

1.12 PLACE "PRESS CONT PCV-0 135" in AUTO.1.13 ENSURE RCS pressure between 325 psig and 400 psig. {PT-0403 or PT-0404}CAUTION* Adjusting charging flow and/or letdown flow to change Pressurizer level at a rate of< 0.5%/min (Step 1.14) or < 0.25 %/min (Steps 1.15 and 1.21), will ensure a slow and controllable fill of the PZR. This is necessary to limit thermal stresses on the PZR materials by excessive cooldown rates, at the point of bubble collapse.* The Pressurizer MAY go water solid as early as 85% Cold Calibrated level indications.

• DURING the final approach to Water Solid, RCS Cooldown rates should be stable AND constant as possible.1.14 IF Pressurizer level is Less than 83 % (Cold Cal), THEN ADJUST "CHG FLOW CONT FK-0205" and "PRESS CONT PCV-0 135" to raise Pressurizer Level (Cold Cal) at a rate of< 0.5 %/min (approx +50 GPM over steady state flowrate + cooldown).

1.15 IF Pressurizer level is Greater than 83 % (Cold Cal), THEN ADJUST "CHG FLOW CONT FK-0205" and "PRESS CONT PCV-0 135" to raise Pressurizer Level (Cold Cal) at a rate of< 0.25 %/min (approx +25 GPM over steady state flowrate + cooldown).

1.16 ENSURE RCS pressure between 325 psig and 400 psig. {PT-0403 or PT-0404}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 138 of 216 J Plant Cooldown I Addendum 9 Plant Cooldown with the PZR Water Solid Page 7 of 8 j NOTE* The administrative cooldown rate is 80°F/hr. (Ref 2.70)* The administrative cooldown limit for the Pressurizer is 160°F/hr.* The cooldown should be spaced over the entire 60 minute period vice a short time high rate cooldown. (Ref 2.50)* MONITOR RCS and Pressurizer temperature and pressure per Data Sheet 1.CAUTION* The Pressurizer MAY go water solid as early as 85% Cold Calibrated level indications.

• Adjustments to PZR spray flow must be made in small increments, to prevent rapid collapse of bubble, causing excessive cooldown rates.Initials 1.17 MONITOR RCS and Pressurizer temperature and pressure per Data Sheet 1.1.18 ENERGIZE PZR Heaters AND CONTROL PZR Spray Valves as necessary to maintain RCS pressure CONSTANT, between 325 psig and 400 psig.1.19 MONITOR PZR Surge Line temperature to ensure outflow.1.20 CLOSELY MONITOR the following, during the PZR fill process:* PZR Surge Line Temperature.

• PZR Liquid Space Temperature.

• RCS pressure.1.21 WHEN Pressurizer level is Greater than 83 % (Cold Cal), THEN ADJUST"CHG FLOW CONT FK-0205"and "PRESS CONT PCV-0 135" to raise Pressurizer Level (Cold Cal) at a rate of < 0.25 %/min (approx +25 GPM over steady state flowrate + cooldown).

1.22 ENSURE RCS pressure between 325 psig and 400 psig. {PT-0403 or PT-0404}This procedure, when completed, SHALL be retained.

oPOP03-ZG-0007 Rev. 71 Page 139 of216 Plant Cooldown Addendum 9 Plant Cooldown with the PZR Water Solid Page 8 of 8 CAUTIONI ERCSpressure should be closely monitored after indicated Cold Calibrated level reaches 90 %.NOTE The PZR spray valves will still have an effect on pressure after the RCS is solid due to the colder spray water.* WHEN RCS pressure is stable with rising letdown flow, THEN the RCS is in a Solid condition.

Initials 1.23 ADJUST the PZR Spray Valve(s) AND Heaters, as necessary, to cooldown the PZR at less than or Equal to cooldown limits.1.24 WHEN PZR Spray valves NO longer influence pressure, THEN ADJUST Charging and Letdown to STABILIZE RCS pressure between 325 psig and 400 psig.1.25 WHEN PZR is in a SOLID Condition, THEN 1.25.1 ENSURE "PRESS CONT PCV-0135" is maintaining RCS pressure between 325 psig and 400 psig. {PT 403 or PT 404}.1.25.2 ENSURE all PZR heaters are OFF.1.25.3 HANG an Equipment Clearance Order to ENSURE PZR heaters remain off. ECO#1.25.4 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.1.26 CONTINUE RCP Operation and Spray flow through the PZR to aid in RCS degassification.

1.27 ENSURE that degassing of the VCT per Addendum 5 and Degassing the RCDT and PRT per Addendum 6 is in progress, complete or as directed by Chemistry Dept, OTHERWISE NA.1.28 NOTIFY the Chemistry Dept. to determine chemical degassification and hydrogen peroxide addition requirements.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 140 of216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 1 of 21 NOTE* Performance of plant cooldown may continue while degassing is in progress.Mechanical degassing of RCS should be done with RCS pressure as high as allowable to promote inventory turn over.0 Decay Heat removal is accomplished by a combination of RHR and Steam Flow (SG PORVs or Steam dumps). When shifting to RHR only, two trains of RHR may be required.* IF Addendum 10 is started and the MSIVs/MSIBs become closed for reasons other than instructed in Addendum 10, THEN the Shift Manager/Unit Supervisor SHALL determine the procedure main body steps between 6.21 through 7.35 required to be performed.

This Addendum step numbering begins at step 100.0 CAUTION Plant cooldown using this Addendum has 2 methods of heat removal. A suggested technique is to control RCS Temperature and cooldown rate with the RHR system, while setting the steam flow rate of the SG PORVs or Steam dumps to a minimum level and adjusting it only as necessary to maintain temperature limits.Initials 100.0 CONTINUE RCS cooldown to between 250°F and 205'F by adjusting RHR HX flow or Steam Flow, WHILE performing the following:

100.1 Maintain RCS pressure between 325 psig and 400 psig. {CPOO }100.2 Maintain RCP seal differential pressure greater than or equal to 230 psid.0 "RCP IA(2A) SEAL 1 DP PI-0 152"* "RCP 1B(2B) SEAL 1 DP PI-0153"* "RCP 1C(2C) SEAL I DP PI-0 154" 0 "RCP ID(2D) SEAL 1 DP PI-0155" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 141 of216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 2 of12 NOTE The following step should reduce the steam demand to an amount that will aid in cooling the SG metal while the main load of decay heat is removed via the RHR system.* The operator has continuous permission to adjust the position of the steam demand controlling components as required for temperature control.* Small incremental changes in valve position are recommended due to the time delay required to see the respondent temperature change.* Using AFW and steam dumps WILL require periodic monitoring of hotwell level to ensure proper level control operation.

• During this cooldown the SGs are expected to go into a vacuum condition.

Initials 100.3 MAINTAIN SG Narrow Range levels between 55 and 75% using AFW.{CP006}100.4 INITIATE secondary plant shutdown per Section 10.0, Step 10.0 through Step 10.16, at the discretion of the Shift Manager.100.5 THROTTLE steam flow, as necessary, from the SGs by performing the following:

100.5.1 ENSURE steam dump controller in manual 100.5.2 ADJUST steam dump controller to desired setting 100.5.3 CLOSE/OPEN steam line drains as necessary This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 142 of 216 Plant Cooldown I Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 3 of 2l1 Initials 101.0 Mode5 Cooldown CAUTION Step 101.2 places the unit in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Technical Specification LCO Action Statement.

All three HHSI pumps SHALL be inoperable within the allotted time.NOTE The administrative cooldown rate is 80°F/hr. (Ref 2.70)The cooldown should be spaced over the entire 60 minute period vice a short time high rate cooldown. (Ref 2.50)REFER TO Shutdown Risk Assessment Report if it is desired to take credit for the Steam Generators for decay heat removal purposes in Mode 5.101.1 IF it is desired to take credit for the Steam Generators for decay heat removal purposes in Mode 5, THEN ENSURE the conditions listed in Addendum 7 are met.101.2 CONTINUE RCS cooldown to between 190'F and 199°F WHILE performing the following:

101.2.1 MAINTAIN RCS pressure 325 psig and 400 psig.101.2.2 MAINTAIN RCP seal differential pressure greater than or equal to 230 psid.* "RCP 1A(2A) SEAL 1 DP PI-0152"* "RCP 1B(2B) SEAL 1 DP PI-0153"* "RCP IC(2C) SEAL 1 DP PI-0154" 0 "RCP ID(2D) SEAL 1 DP PI-0155" 101.3 RECORD the Unit, date and time the plant entered Mode 5.Unit: Date: Time: Hrs.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 143 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 4 of 21 1 Initials 101.4 DISABLE the operable and standby "HIGH HEAD SAFETY INJECTION PUMPS" by racking out and tagging the respective pump breakers with an Equipment Clearance Order. ECO #* "IA(2A) EI A(E2A)/5"* "IB(2B) E1B(E2B)/5" S "1 C(2C) E1 C(E2C)/5" ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.101.5 CLOSE HHSI cold leg injection valves. {CPOO1}* "LOOP A Tc INJ MOV-0006A"* "LOOP B Tc INJ MOV-0006B"* "LOOP C Tc INJ MOV-0006C" NOTE Plant cooldown may continue while performing Steps 10 1.6 through 10 1. 11.101.6 PLACE LHSI pump hand switches in the PULL TO LOCK position.

{CP0OO }* "LHSI PUMP 1A(2A)"* "LHSI PUMP IB(2B)"* "LHSI PUMP IC(2C)" 101.7 PLACE all three containment spray pumps hand switches in the PULL TO LOCK position.

{CP002} (Ref 2.18, Ref 2.41)* "CSS PUMP 1A(2A)"* "CSS PUMP IB(2B)"* "CSS PUMP IC(2C)" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 144 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 5 of 21 Initials 101.8 IF the control rods are NOT fully inserted, THEN manually INSERT all control rods by selecting the appropriate bank on the "ROD BANK SEL" switch.NOTE" Performance of Step 101.9 will generate a feedwater isolation signal due to low Tavg.* Rod Control Urgent Alarm and associated control room annunciator "ROD CONT URGENT ALARM" (5M03-B5) will be received when the reactor trip breakers are opened 101.9 IF the reactor trip breakers are closed, THEN OPEN reactor trip breakers by placing the "REACTOR TRIP" switch in the TRIP position.

{CP005}NOTE The "MASTER RELAY DEFEAT SWITCH" (MRDS) must be pulled out to move it from RX TRIP to DEFEAT/CVI AVAIL.101.10 WHEN the reactor trip breakers are open, THEN PLACE the "MASTER RELAY DEFEAT SWITCH" in each SSPS Logic Cabinet in either the DEFEAT/CVI AVAIL or DEFEAT ALL position as directed by the Shift Manager/Unit Supervisor.

101.10.1 "PROTECTION SYSTEM LOGIC TRAIN R, LOGIC CABINET, LOGIC TEST PANEL" (SSPS) (ZRR001)101.10.2 "PROTECTION SYSTEM LOGIC TRAIN S, LOGIC CABINET, LOGIC TEST PANEL" (SSPS) (ZRRO08)This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 IRev. 71 1Page 145 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 6 of 21 Initials NOTE Steps 101.22, 101.25, 101.27 and 101.28 have been identified as candidates for early start.Early starts of these steps have been evaluated as having NO safety impact for current plant conditions.

The Shift Manager/Unit Supervisor may authorize early start of procedure steps to enhance plant performance.

CAUTION Use of a pressurizer steam bubble during periods of low pressure, low temperature operation is preferred.

This steam bubble will dampen the plant's response to potential transient generating inputs, providing easier pressure control with slower response rates. (UFSAR 5.2.2.11.3) 101.11 IF directed by Operations Management to take the PZR Water Solid at this point in the cooldown, THEN PERFORM Addendum 9 "Plant Cooldown with the PZR Water Solid" while continuing with this procedure.

101.12 IF the RCS cooldown is being performed due to an unisolable RCS leak, THEN PERFORM the following (Ref 2.59): 101.12.1 CONTINUE RCS cooldown to between 190'F and 1957F.101.12.2 GO TO Step 101.22 AND N/A Steps 101.13 through 101.21.2.101.13 IF RCS drain down is to be performed, THEN NOTIFY I&C to place Core Exit T/C Alarm System in service.101.14 CONTINUE RCS cooldown to 179°F-181°F.

NOTE I&C PM to place Core Exit T/C Alarm System in-service is required prior to RCS Tave less than 160°F.101.15 IF RCS drain down will be performed, THEN NOTIFY I&C to place Core Exit T/C Alarm System in service (PMs Unit 1-95003780, Unit 2-9500378 1).WAN #101.16 NOTIFY Chemistry to sample the VCT for H2 concentration.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 146 of216 I Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 7 of 21I P ~dd~ndurn10 -I Initials CAUTION The stepping or tripping of CRDMs during periods when coolant crud levels are high should be kept to a minimum by applying the following:

Reactor Trip Breakers should be open prior to Hydrogen Peroxide addition to minimize the attracting of crud particles in the CRDMs coils. (Ref 2.63 & 2.96)Reactor Trip Breakers should remain open after a Hydrogen Peroxide addition until coolant turbidity levels have returmed to 1.1 NTU or less. (Ref 2.63 & 2.96)Control Rods should NOT be moved until coolant turbidity levels have returned to 1.1 NTU or less. (Ref 2.63 & 2.96)IF coolant turbidity levels are greater than 1.1 NTU AND Control Rod movement is to be performed, THEN PERFORM Addendum 17, Moving Control Rods with High Reactor Coolant Turbidity. (Ref. 2.96)IF a RCS boundary opening is planned, THEN RCS H2 Concentration should be less than 4 cc/kg prior to opening the RCS.IF Hydrogen Peroxide addition is required, THEN dilution water amount used in the chemical addition SHALL be accounted for to maintain required shutdown margin.NOTE IF OPSP03-CV-0014, CVCS Equipment Verification, has been perfomled AND Hydrogen Peroxide addition is desired, THEN the ECO tag on "1 (2)-CV-0201 A, CHEMICAL MIXING TANK REACTOR MAKEUP WATER ISOLATION VALVE" may be released for the time necessary to add Hydrogen Peroxide.101.17 IF opening of the RCS is planned AND a Hydrogen Peroxide addition is required, THEN PERFORM the following:

101.17.1 ENSURE RCS boron concentration is greater than RCS Shutdown Margin Concentration (Cb) requirements of Step 5.7.6 by sufficient amount to account for the dilution water used to perform Hydrogen Peroxide addition.

REFER TO Step 5.18.3.RCS Cb/Time PZR Cb/Time This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 147 of'216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 8 of 21 Initials 101.17.2 IF Rapid Refueling is planned, THEN ENSURE RCS boron concentration is greater than SHUTDOWN MARGIN LIMIT CURVE Plant Curve Book Figure 5.5, Mode 5 Cb, ARO, 68°F to 200'F, K <= 0.95 by sufficient amount to account for the dilution water used to perform Hydrogen Peroxide addition.

REFER TO Step 5.18.5. (Ref 2.35)RCS Cb/Time PZR Cb/Time 101.17.3 NOTIFY Chemistry to add Hydrogen Peroxide to the RCS per OPOP02-CV-0001, Makeup to Reactor Coolant System, and OPCP03-ZC-0005, Chemical Addition to the Reactor Coolant System.101.18 NOTIFY Chemistry to sample the RCS for fission products and H2 concentration.

101.19 OBTAIN concurrence from Health Physics and Chemistry to continue the RCS cooldown below 180'F.101.20 WHEN concurrence from Health Physics and Chemistry obtained, THEN COMMENCE RCS Cooldown to 150'F, MAINTAIN RCS pressure between 325 psig and 400 psig.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Pag 148of216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 9 of 21 Initials 101.21 IF Core Exit T/C Alarm System was placed in service, THEN PERFORM the following:

101.21.1 WHEN RCS temperature is less than 1607F, THEN VERIFY "RC MID LOOP CORE EXIT TEMP HI" Lamp box I M02 E-1 alarm clear. {CP001 _)101.21.2 WHEN RCS temperature is less than 1507F, THEN CONTINUE with Steps 101.22 through 101.25.NOTE Plant cooldown may continue while performing Step 101.22.101.22 ENSURE an inspection of the RCS for boron deposits and other evidence of primary system leakage is performed per OPGP03-ZE-0033, RCS Pressure Boundary Inspection For Boric Acid Leaks, as determined necessary by SED.(Ref 2.46)CAUTION In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.101.23 ENSURE idle RHR trains have been placed in service per OPOP02-RH-0001, Residual Heat Removal System Operation, to equalize boron concentrations with the RCS.* RHR Train A* RHR Train B* RHR Train C This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 149 of 216 Plant Cooldown E Addendumr 10 MODE 5 Cooidown with MSIVs OPEN PagQe1 f21 Initials NOTE RHR Train A or B should remain inl operation to supply low pressure letdown.101.24 SECURE selected RHR trains per OPOP02-RH-0001, Residual Heat Removal System Operation, as directed by the Shift Manager/Unit Supervisor.

• RHR Train A* RHR Train B* RI-R Train C 101.25 MAINTAIN RCS pressure between 325 and 400 psig.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 150 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MS1Vs OPEN Page 11 of 21I Initials NOTE" Step 101.26 is required by Calculation NC-07090, Evaluation of Boron Dilution Flow Paths in Mode 5b (Mode 5 with reactor coolant loops NOT filled) and Mode 6, to prevent low boron water fr'om being introduced into the RCS. (Refs. 2.108, 2.109 & 3.13)" Step 101.26 will perform one of the following to comply with Calculation NC-07090: " The CVCS Loop A or C charging line already isolated will remain isolated by ECO throughout Mode 5b (Mode 5 with reactor coolant loops NOT filled) and Mode 6.OR" Swap CVCS Loop A and C charging lines so that both flowpaths will be borated with shutdown boron concentration water prior to establishing 0PSP03-CV-0014, CVCS Equipment Verification.

101.26 PERFORM the following to ensure CVCS Loop A and C isolation lines are not a source of RCS dilution in Mode 5b and Mode 6 (Refs. 2.108, 2.109 & 3.13): 101.26.1 IF the NOT In-Service CVCS CHARGING TO RCS LOOP COLD LEG MOV OPERATOR, (the closed valve) CV-MOV-0006 or CV-MOV-0003 will remain isolated throughout Mode 5b and Mode 6, THEN PERFORM the following for the NOT In-Service valve, OTHERWISE GO TO Step 101.26.2: 101.26.1.1 De-Energize the MOV and hang ECO on supply breaker.ECO #101.26.1.2 CAUTION tag Control Room handswitch to identify"Potential dilution source must remain isolated in Mode 5b and Mode 6".ECO #101.26.1.3 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.101.26.1.4 GO TO Step 101.27.This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 151 of216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MS1Vs OPEN Page 12 of 21 Initials 101.26.2 ENSURE one of the following (Ref 2.68): 101.26.2.1 The differential pressure between the charging header and the RCS is less than 2500 psid OR 101.26.2.2 SECURE the CVCS charging pump per OPOP02-CV-0004, Chemical and Volume Control System Subsystem, OR 10 1.26.2.3 CLOSE CVCS Charging Line "OCIV MOV-0025" 101.26.3 IF "LOOP A ISOL MOV-0003" is in service, THEN PERFORM the following:

101.26.3.1 OPEN "LOOP C ISOL MOV-0006" 101.26.3.2 CLOSE "LOOP A ISOL MOV-0003" 101.26.3.3 GO TO Step 101.26.5 101.26.4 IF "LOOP C ISOL MOV-0006" is in service, THEN PERFORM the following:

101.26.4.1 OPEN "LOOP A ISOL MOV-0003" 101.26.4.2 CLOSE "LOOP C ISOL MOV-0006" 101.26.5 RESTORE CVCS charging pumps AND CVCS Charging Line"OCIV MOV-0025" as directed by Unit Supervisor.

101.26.6 ENSURE flow established through CVCS Loop A or C isolation line that was placed in service for at least two minutes at greater than 50 gpm flowrate prior to establishing OPSP03-CV-0014, CVCS Equipment Verification.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 152 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 13 of 21 Initials 101.27 WHEN Rapid Refueling rod holdout operations are desired, THEN PERFORM the following:

CAUTION* During rod holdout operations, RCS temperature should be stable when moving control rods to prevent control rod thermal binding and two positive reactivity changes during rod motion.* Reactor trip breakers must be open during a RCS cooldown when any control rods are on the bottom.OPSP03-CV-0014, CVCS Equipment Verification MAY isolate the letdown flowpath (Example:

closing the Dernin outlets before opening the Demin Bypass), ENSURE the letdown flowpath restrictions are planned for PRIOR TO performing OPSP03-CV-00 14, CVCS Equipment Verification.

101.27.1 ENSURE RCS is borated to SHUTDOWN MARGIN LIMIT CURVE Plant Curve Book Figure 5.5, Mode 5 Cb, ARO, 68°F to 200'F, K <= 0.95. (Ref 2.35)RCS Cb/Time PZR Cb/Time 101.27.2 PERFORM OPSP03-CV-0014, CVCS Equipment Verification, to isolate dilution paths while rods are in the Rapid Refueling position.An Equipment Clearance Order SHALL be hung to ensure dilution paths remain isolated.* ECO #ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.101.27.3 IF Control Rods are to be locked out for GREATER THAN 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with Reactor Vessel Head on the flange and fuel is in the vessel, THEN REFER TO OPOP03-ZG-0012, Operation with Rods in the Rapid Refueling Position.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 1Page153of2l6 Plant Cooldown AddenduIn 10 I MODE 5 Cooldown with MSIVs OPEN Page 14 of 21 Initials 101.27.4 IF coolant turbidity levels are greater than 1.1 NTU AND Control Rod holdout is to be perfonned, THEN PERFORM Addendum 17, Moving Control Rods with High Reactor Coolant Turbidity. (Ref.2.96)101.27.5 PERFORM the appropriate sections of OPMP07-DM-0003, Rapid Refueling Rod Holdout Operation.

101.27.6 WHEN the Rapid Refueling rod holdout operations are complete, THEN ENSURE the Reactor Trip Breakers are open.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 7 Page 154 of 216]Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 15 of 21 Initials NOTE Valve exercising during cold shutdown SHALL commence within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of achieving cold shutdown and continue until all testing is complete or the plant is ready to return to operation at power. For extended outages, testing need not be commenced in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided all valves required to be tested during cold shutdown will be tested before plant startup.However, it is not the intent to keep the plant in cold shutdown to complete cold shutdown testing.Cold Shutdown Testing reference:

IST Plan Rev. 13, Section 5.4.2, Valve Exercise Test OM-2004 CODE, ISTC-3521(g) 101.28 INITIATE Cold Shutdown testing within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of Mode 5 entry. Cold Shutdown testing consists of performance of the following surveillances:

• OPSP03-CV-001 1, Chemical and Volume Control System Valve Operability Test (Cold Shutdown)* OPSPO3-FW-0002, Feedwater System Valve Operability Test (Cold Shutdown)0 OPSP03-HC-0004, Reactor Containment Building Normal Purge System Valve Operability Test (Cold Shutdown)* OPSP03-MS-0002, Main Steam System Cold Shutdown Valve Operability Test* OPSP03-RC-0010, Pressurizer Power Operated Relief Valve Operability Test* 0PSP03-RH-0007, Residual Heat Removal System Valve Operability Test (Cold Shutdown)* OPSP03-RH-00 10, RHR System Valve Leak Tests* 0PSP03-SI-0028, SIS Accumulator Valves Operability Test (Cold Shutdown)This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 155 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 16 of 21 Initials 101.29 NOTIFY the Plant Operations Surveillance Coordinator to identify the specific procedures required to be performed prior to restart.10 1.30 IF RCS Cooldown is being performed as the result of an unisolable RCS leak, THEN GO TO Section 8.0, RCS Depressurization.

101.31 CONTINUE with plant cooldown.CAUTION LHSI cold leg injection valves SHALL remain open on operating RHR trains.101.32 WHEN RCS temperature is less than 150'F, THEN CLOSE LHSI cold leg injection valves on all idle RHR trains. {CP00I }* "LOOP A Tc INJ MOV-0031A"* "LOOP B Tc INJ MOV-003 1B"* "LOOP C Tc INJ MOV-0031C" 101.33 IF necessary to control ambient temperature in the RCB to between 65'F and 70°F, THEN ADJUST the number of RCFCs in operation per OPOP02-HC-0001, Containment HVAC.101.34 WHEN RCS temperature is less than 150'F or WHEN desired, THEN CLOSE MS1Vs. {CP006}* SG IA(2A) "MSIV FSV-7414" 0 SG IB(2B) "MSIV FSV-7424" 0 SG I C(2C) "MSIV FSV-7434" 0 SG 1D(2D) "MSIV FSV-7444" I This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 156 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 17 of 21 Initials NOTE Secondary plant shutdown SHALL be completed through Step 10.15 prior to placing the SGs in wet layup per Step 101.37.101.35 ENSURE secondary plant shutdown performed as required per Section 10.0.101.36 IF establishing a SG Nitrogen blanket, THEN PERFORM the following:

NOTE A set of hoses is normally stored in the OUTAGE BOX. IF performing this in a Forced Outage, hoses may need to be fabricated.

101.36.1 INSTALL bulk Nitrogen supply hoses to each SG connection:

• "1 (2)-MS-0510 N2 BLANKET ISOL"{57 ft IVC Loop A}* "1(2)-MS-0508 N2 BLANKET ISOL"{57 ft IVC Loop BI* "1(2)-MS-0506 N2 BLANKET ISOL"{57 ft IVC Loop C1* "1(2)-MS-0504 N2 BLANKET ISOL"{57 ft IVC Loop D}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 j Rev. 71 Page 157 of216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 18 of 21 Initials 101.36.2 ENSURE OPEN the following nitrogen valves:* 1 (2)-NL-0080 "LOW PRESSURE NITROGEN TO STEAM GENERATOR BLANKETING ISOLATION VALVE" (YDB Elv 29' between Unit 1(2) TGB and IVC on the South Side, West End)* 1(2)-NL-0186 "LOW PRESSURE NITROGEN TO S/G 1D(2D) BLANKETING ISOLATION VALVE" {IVC Rm 404 Elv 51', behind FSV-7444 between MS Line and Wall S I(2)-NL-0188 "LOW PRESSURE NITROGEN TO S/G IC(2C) BLANKETING ISOLATION VALVE" {IVC Rm 401 Elv 51', behind FSV-7434 between MS Line and Wall)* 1 (2)-NL-0 190 "LOW PRESSURE NITROGEN TO S/G I B(2B) BLANKETING ISOLATION VALVE" {IVC Rrn 402 Elv 51', behind FSV-7424 between MS Line and Wall)* 1 (2)-NL-0 192 "LOW PRESSURE NITROGEN TO S/G I A(2A) BLANKETING ISOLATION VALVE" {IVC Rm 403 Elv 51', behind FSV-7414 between MS Line and Wall)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 158 of 216 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 19 of 2 Initials 101.36.3 ENSURE SG PORVs CLOSED. {CP006}* SG I A(2A) "PORV PV-74 11 0 SG IB(2B) "PORV PV-7421" 0 SG 1 C(2C) "PORV PV-7431" 0 SG I D(2D) "PORV PV-7441" 101.36.4 OPEN "N2 BLANKET ISOL" valves. {IVC 57 ft}* SG 1 A(2A) "1 (2)-MS-0510 N2 BLANKET ISOL" SG 1A(2A) "1 (2)-MS-0509 N2 BLANKET ISOL" SG 1B(2B) "1(2)-MS-0508 N2 BLANKET ISOL" SG IB(2B) "I (2)-MS-0507 N2 BLANKET ISOL" SG IC(2C) "1(2)-MS-0506 N2 BLANKET ISOL" SG 1C(2C) "1(2)-MS-0505 N2 BLANKET ISOL" SG ID(2D) "1(2)-MS-0504 N2 BLANKET ISOL" SG ID(2D) "l(2)-MS-0503 N2 BLANKET ISOL" This procedure.

when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Pg19o 1 Plant Cooldown Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 20 of 21 Initials 101.36.5 USING Control Room pressure instrumentation for reference, ADJUST each regulator to maintain a positive pressure less than 2.0 psig on the Steam Generators.

• 1(2)-MS-PCV-8393 "LOW PRESSURE NITROGEN TO S/G 1(2)D BLANKETING PRESSURE CONTROL VALVE" {57 ft IVC Loop D}* 1(2)-MS-PCV-8394 "LOW PRESSURE NITROGEN TO S/G ](2)C BLANKETING PRESSURE CONTROL VALVE" 157 ft IVC Loop C}1(2)-MS-PCV-8395 "LOW PRESSURE NITROGEN TO S/G 1(2)B BLANKETING PRESSURE CONTROL VALVE" {57 ft IVC Loop B} _S "1(2)-MS-PCV-8396 "LOW PRESSURE NITROGEN TO S/G 1(2)A BLANKETING PRESSURE CONTROL VALVE{57 ft IVC Loop A}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 160 of216 Plant Cooldown I Addendum 10 MODE 5 Cooldown with MSIVs OPEN Page 21 of 21 Initials CAUTION In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.101.37 IF placing SGs in wet layup, THEN PERFORM the following:

10 1.37.1 VERIFY secondary plant shutdown completed through Step 10.15.NOTE Step 101.37.2 may generate a feedwater isolation signal depending on position of MRDS switch due to high water levels.101.37.2 PLACE SGs in wet lay up by feeding each SG per OPOP02-AF-0001, Auxiliary Feedwater, until Wide Range level indicates between 98%and 100%.* SG 1A(2A)* SG IB(2B)* SG 1 C(2C)* SG 1D(2D)101.37.3 SECURE AFW operation per OPOP02-AF-0001, Auxiliary Feedwater.

101.37.4 PLACE SGs in recirculation per OPOP02-SB-0002, Steam Generator Wet Layup Recirc.101.37.5 ESTABLISH SG wet lay up chemistry as required per OPCPO1-ZA-0038, Plant Chemistry Specifications.

101.38 ENSURE AFW operation is secured per OPOP02-AF-0001, Auxiliary Feedwater.

101.39 GO TO this procedure Section 8.0.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 161 of216 Plant Cooldown Addendum 11 I RCS/PZR Pressure Operations Guideline Page 1 of 5 1.0 Purpose and Scope 1.1 This Addendum provides Operations Guidelines RCS/PZR Pressure Control Operations during primary plant heatup/cooldown and Solid Plant Operations below NOP/NOT.2.0 References 2.1 LOT 201.14.HO.01, PRESSURIZER PRESSURE AND LEVELCONTROL SYSTEM 3.0 Prerequisites 3.1 NONE 4.0 Notes and Precautions 4.1 IF a PZR PORV OPENS with the RCS solid, THEN do NOT block the PORV closed until the source of the pressure transient is understood.

BLOCKING a COMS PORV during a pressure transient MAY result in lifting the mechanical relief devices in the RHR system, which is undesirable.

4.2 WHEN performing testing on an Instrument Channel OR performing an Electrical Bus outages that provides power to an Instrument Channel, THEN ENSURE that the affected channel is De-Selected from Actuation, Control, Blocking or Backup Functions to the system controlling RCS/PZR Pressure & Level.4.3 SSPS Actuation Cabinets "A" and "B" and Logic Cabinets "R" and "S" must be energized for both COMS to be OPERABLE.

Logic Cabinets "R" and Actuation Cabinets "A" required for PCV 0655A "Pressurizer PORV Train A". Logic Cabinets "S" and Actuation Cabinets "B" required for PCV 0656A "Pressurizer PORV Train B". {LER 2-05-003, SSPS ECO to support FWIV energize to actuate MOD in 2REI 0 caused COMS to be inoperable during solid plant operations in violation of TS 3.4.9.3.(CR 05-5960, CR 05-3 071)}.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 162 of 216 Plant Cooldown Addendum 11 RCS/PZR Pressure Operations Guideline Page 2 of 5 5.0 Pressurizer Master Pressure Controller Operations Guideline 5.1 WHEN Operating Below Normal Operation Pressure, THEN ENSURE the following Controllers are in MANUAL:* Pressurizer Pressure Controller RC-PK-0655A (PZR Press Master Controller)" Pressurizer Spray Valve Controller "PRZR SPR PCV-06551B"* Pressurizer Spray Valve Controller "PRZR SPR PCV-0655C" 5.2 RCS/PZR Pressure Control during Heatup/Cooldown Operations SHOULD be performed by varying the Controller Output the following: " Pressurizer Pressure Controller RC-PK-0655A (PZR Press Master Controller) to Control the output of the PZR CONTROL HTRS* Pressurizer Spray Valve Controller "PRZR SPR PCV-06551B" and "PRZR SPR PCV-0655C" to Control RCS/PZR Pressure" Backup Heaters may be cycled as necessary to aid in PZR turnover flow 5.3 The following is the Master Controller outputs in Manual Operation:

Function Controller Output VDC Controller Output % Signal Direction PCV-0655A OPENS 8.75 87.5 INC (4)PCV-0655A CLOSES 7.50 75.0 DEC (+)PZR PRESS DEV HI 7.19 72.0 INC (4)ALARM SPRAY FULL OPEN 7.19 72.0 INC (4)PRES PRESS DEV HI 6.56 65.5 DEC (+)ALARM RESETS SPRAY FULL 4.06 40.5 DEC (+)CLOSED CONT HTRS 0% PWR 3.44 34.5 INC (4)CONT HTRS 50% 2.50 25.0 N/A PWR CONTROL HTRS 1.56 15.5 DEC (+)100% PWR PZR PRESS DEV LO 0.94 9.5 DEC (+)ALARM & BU HTRS ON This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 163 of216 Plant Coold own Addendum 11 RCS/PZR Pressure Operations Guideline Page 3 of 5 6.0 Pressurizer Spray Valve Controller Operations Guideline 6.1 Pressurizer Spray Valve Controllers "PRZR SPR PCV-0655B" and "PRZR SPR PCV-0655C" SHOULD remain in Manual and throttled while the associated Spray Valve's RCP is operating.

This allows the Pressurizer Spray Valve Controllers operating in Manual to control RCS/PZR pressure.6.2 WHEN RCP 1A(2A) is NOT running, THEN ENSURE Pressurizer Spray Valve Controller "PRZR SPR PCV-0655C" in "MAN" and CLOSED.6.3 WHEN RCP ID(2D) is NOT running, THEN ENSURE Pressurizer Spray Valve Controller "PRZR SPR PCV-0655B" in "MAN" and CLOSED.6.4 IF using Auxiliary Spray, THEN control Pressurizer Spray Valves per the Auxiliary Spray Operations Guideline.

7.0 Auxiliary

Spray Operations Guideline 7.1 Auxiliary Spray SHALL NOT be initiated with a temperature differential greater than 621-F. (TRM 3.4.9.2.c) 7.2 The temperature differential between the pressurizer liquid and the reactor coolant SHALL NOT exceed 320'F to minimize the effects of surge line thermal stratification.

7.3 WHEN using Auxiliary Spray, THEN THROTTLE with the Normal Pressurizer Spray Valves to establish the desired flow rate into the Pressurizer.

This minimizes thermal cycles on the Auxiliary Spray nozzle and prevents unwanted short cycle of the CVCS flow into the RCS Loops.7.4 THROTTLING with CHG FLOW CONT VLV FCV-0205 will impact Auxiliary Spray flow. CHG FLOW CONT VLV FCV-0205 should be used to maintain the desired PZR Level and Normal Pressurizer Spray Valves should be used to establish the desired flow rate into the Pressurizer.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 164 of 216 Plant Cooldown Addendum 11 I RCS/PZR Pressure Operations Guideline Page 4 of 5 8.0 Solid Plant Operations NOTE Whenever the plant is water solid AND the reactor coolant pressure is being maintained by low pressure letdown, it is recommended to keep all three letdown orifices open. (Ref 2.65.1 and 2.95)8.1 PRESS CONT PCV-0135 may NOT respond fast enough to limit pressure transients, IMMEDIATE MANUAL Operation/Intervention to control pressure is AUTHORIZED.

This includes manual operation of PRESS CONT PCV-0135 and CHG FLOW CONT VLV FCV-0205 and SECURING RCPs to protect the RCP seals.8.2 If all RCPs have stopped for more than 5 minutes and the RCS temperature is greater than the charging and seal injection water temperature, DO NOT start the first RCP until a steam bubble is formed in the pressurizer. (UFSAR 5.2.2.11.3) 8.3 If all RCPs are stopped and the reactor coolant is being cooled down by the residual heat removal heat-exchangers, a non-uniform temperature distribution may occur in the RCS.DO NOT attempt a start of a RCP unless a steam bubble exists in the pressurizer or an acceptable temperature profile can be demonstrated. (UFSAR 5.2.2.11.3) 8.4 When the reactor coolant pressure is being maintained by tile low pressure letdown control valve during water solid operation, changes to the flow rate through the RHR loop by throttling of valves or starting and stopping the RHR pumps may cause a rise in reactor coolant pressure.

For example, stopping the RHR pumps may cause a rise in the reactor coolant pressure of between 100 and 150 psig.8.5 It is recommended where possible that the RHR train that is used for low pressure letdown NOT be used for RCS cooling. This will minimize pressure transients while adjusting RCS heatup/cooldown rates.8.6 During plant cooldown, all SGs will normally be connected to the steam header to assure a uniform cooldown of the RCS. (UFSAR 5.2.2.11.3) 8.7 In Solid Plant Operations, pressure/temperature transients can be extremely fast. ENSURE all START/STOPPING of equipment thermally or hydraulically coupled to the RCS is evaluated for expected effect on RCS pressure.8.8 DO NOT Isolate all RHR Trains inlet lines from the reactor coolant loop unless there is a steam bubble in the pressurizer or the charging pumps are stopped.This procedure, when completed, SHALL be retained.

lI I C OPOP03-ZG-0007 Rev. 71 Page 165of216 Plant Cooldown I Addendum 11 RCS/PZR Pressure Operations Guideline Page 5 of 5 8.9 Use of a pressurizer steam bubble during periods of low pressure, low temperature operation is preferred.

This steam bubble will dampen the plant's response to potential transient generating inputs, providing easier pressure control with slower response rates.(UFSAR 5.2.2.11.3)

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 166 of 216 Plant Cooldown Addendum 12 Solid Plant Operations Entry Checklist Page 1 of 9 UNIT 1 (Circle Unit Performing Checklist)

UNIT 2 1.0 PROCEDURE 1.1 Shift Manager or designee performs a thorough review of the outage schedule to ensure safety of the plant during Solid Plant Operations.

This review is performed by comparing the outage schedule to the guidelines on Addendum 12. Shift Manager or designee ensures compensatory measures or other remedial actions address Solid Plant risk concerns.1.1.1 All applicable issues will be analyzed to determine if they meet the Solid Plant Operations guidelines identified in this procedure.

1.1.2 IF the guidelines are NOT met, THEN: 1.1.2.1 Operations SHALL document those issues in the Remarks Section of this Addendum.

Operations identifies any HIGHER RISK EVOLUTIONS which require additional reviews.1.1.2.2 Corrective actions and/or contingencies SHALL be developed and implemented to minimize risk (i.e., schedule changes, deletion from outage, compensatory measures, training, etc.).1.1.3 All corrective actions should be evaluated for the following:

1.1.3.1 Effectiveness to ensure actions taken, especially schedule revisions, have NOT affected shutdown issues elsewhere.

1.1.3.2 To ensure safety of the plant and plant personnel have been maintained.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 167 of 216 Plant Cooldown Addendum 12 Solid Plant Operations Entry Checklist Page 2 of 9 UNIT 1 (Circle Unit Performing Checklist)

UNIT 2 NOTE A "NO" answer does NOT necessarily represent an unacceptable risk situation, but identifies an area that may require further evaluation or other compensatory measures.IF the answer to any of the questions below is "NO", THEN determine the actions required to minimize risk. These actions may include new or additional outage activities, schedule changes, procedure changes or temporary modifications.

Comments to any "NO" responses that were recorded during the read through of this Addendum SHALL be addressed in the REMARKS.YES/NO PORVs & COMS: A. Is Cold Overpressure Mitigating System (COMS) operable as required?

{SSPS Actuation Cabinets "A" and "B" and Logic Cabinets "R" and "S" must be energized for both COMS to be OPERABLE.

Logic Cabinets "R" and Actuation Cabinets "A" required for PCV 0655A "Pressurizer PORV Train A". Logic Cabinets "S" and Actuation Cabinets "B" required for PCV 0656A "Pressurizer PORV Train B". LER 2-05-003, SSPS ECO to support FWIV energize to actuate MOD in 2RE10 caused COMS to be inoperable during solid plant operations in violation of TS 3.4.9.3. (CR 05-5960, CR 05-3071)}B. IF one or both pressurizer PORVs are inoperable, THEN are tile required RHR pumps AND discharge relief valves operable and in service? (REFER TO OPOP03-ZG-0001, Addendum 6, RCS Water Solid and COMS PORV(s) Administration Controls)C. Are the required electrical distribution systems operable to support COMS?D. Are the required electrical distribution systems operable to support PZR level and PZR Pressure instrumentation?

E. Is the current RCP configuration capable of supplying PZR Spray flow and in the recommended configuration for PZR Water Solid conditions?

This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 168 of 216 Plant Cooldown Addendum 12 Solid Plant Operations Entry Checklist Page 3 of 9 I UNIT 1 (Circle Unit Performing Checklist)

UNIT 2 YES/NO High Pressure Makeup Sources Inoperable:

F. Are high pressure injection systems rendered inoperable during Solid Operation conditions?

G. Are all but one Centrifugal Charging Pump rendered inoperable during Solid Operation conditions?

H. Is the PDP rendered inoperable during Solid Operation conditions?

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 169 of216 Plant Cooldown I Addendum 12 Solid Plant Operations Entry Checklist Page 4 of 9 NOTE A "NO" answer does NOT necessarily represent an unacceptable risk situation, but identifies an area that may require further evaluation or other compensatory measures.IF the answer to any of the questions below is "NO", THEN determine the actions required to minimize risk. These actions may include new or additional outage activities, schedule changes, procedure changes or temporary modifications.

Comments to any "NO" responses that were recorded during the read through of this Addendum SHALL be addressed in the REMARKS.YES/NO Pressure Control Devices: I. Are PRESS CONT PCV-0135 and CHG FLOW CONT VLV FCV-0205 functioning properly?J. Is work on valves or other components that could lead to an RCS Inventory/Pressure transient been minimized or avoided while in Solid Plant Operations?

Electrical Bus Status: K. Is work on Electrical Bus outages or other components that could lead to an RCS Inventory/Pressure transient being minimized or avoided while in Solid Plant Operations?

SSPS Actuation Cabinets "A" and "B" and Logic Cabinets "R" and "S" must be energized for both COMS to be OPERABLE.

Logic Cabinets "R" and Actuation Cabinets "A" required for PCV 0655A "Pressurizer PORV Train A". Logic Cabinets "S" and Actuation Cabinets "B" required for PCV 0656A "Pressurizer PORV Train B". {LER 2-05-003 , SSPS ECO to support FWIV energize to actuate MOD in 2RE10 caused COMS to be inoperable during solid plant operations in violation of TS 3.4.9.3.(CR 05-5960, CR 05-3071)}.

L. Are switchyard activities coordinated with Outage Management to minimize or avoid impact on operable power supplies?M. Is the use of non-standard electrical line-ups minimized or avoided to meet electrical requirements?

OTHER: N. Evaluation of this Plant Condition using the ORAM program indicates all Safety Function colors are either green or yellow?This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 170 of 216 Plant Cooldown Addendum 12 Solid Plant Operations Entry Checklist Page 5 of 9 NOTE A "NO" answer does NOT necessarily represent an unacceptable risk situation, but identifies an area that may require further evaluation or other compensatory measures.IF the answer to any of the questions below is "NO", THEN determine the actions required to minimize risk. These actions may include new or additional outage activities, schedule changes, procedure changes or temporary modifications.

Comments to any "NO" responses that were recorded during the read through of this Addendum SHALL be addressed in the REMARKS.YES/NO OPERATIONS:

0. Has the Operations Crew conducting Solid Plant Operations been provided with recent (within last 90 days) training on the following:

a) Performing Solid Plant Operations?

b) Theory of Solid Plant Operations and Operations margins?c) Procedures that may affect Solid Plant Operations?

i.e.* RCP OffNormal* Loss of RHR* COMS Actuation d) Planned Outage Equipment status during the Solid Plant Operations?

e) Lessons learned from current Solid Plant Operations events?f) Over-riding automatic functions of COMS PORVs?This procedure, when completed, SHALL be retained.

oPOP03-ZG-0007 Rev. 71 Page 171 of 216 Plant Cooldown Addendum 12 Solid Plant Operations Entry Checklist Page 6 of 9 NOTE A "NO" answer does NOT necessarily represent an unacceptable risk situation, but identifies an area that may require further evaluation or other compensatory measures.IF the answer to any of the questions below is "NO", THEN determine the actions required to minimize risk. These actions may include new or additional outage activities, schedule changes, procedure changes or temporary modifications.

Comments to any "NO" responses that were recorded during the read through of this Addendum SHALL be addressed in the REMARKS.YES/NO P. Critical System requirements associated with Solid Plant Operations:

• Reactor Coolant System -intact?* Reactor Coolant Pumps -current functional RCP configuration capable of supplying PZR Spray flow and in the recommended configuration for PZR Water Solid conditions?

• RHR -A&B operable w/ low pressure letdown in accordance with TS -NO maintenance work in progress.* Low Pressure Letdown -NO maintenance work in progress.* CVCS* Letdown -NO maintenance work in progress, NO system lineup changes, i.e.filter/demin changes* Charging -NO maintenance work in progress, NO system lineup changes* Seal Injection

-NO maintenance work in progress, NO system lineup changes* CCW- operable in accordance with TS, NO system lineup changes* ECW -operable in accordance with TS, NO system lineup changes* SSPS -Out of service, actuation trains in test* SI -removed from service, all activities that could introduce high pressure to the RCS suspended* SG's- NO heat input activities

• 7300 -NO maintenance work in progress, instruments/controls associated with PZR indications all operable* QDPS -all operable* ICS -operable w/trends of RCS pressure, RCS temperature, and RCS flow* IA -Two compressors available, normal configuration

-NO maintenance work in progress.* Normalization Status -PZR Level and PZR Press, NO Changes during Solid Plant Operations.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 172 of216 Plant Cooldown Addendum 12 Solid Plant Operations Entry Checklist Page 7 of 9 NOTE A "NO" answer does NOT necessarily represent an unacceptable risk situation, but identifies an area that may require further evaluation or other compensatory measures.IF the answer to any of the questions below is "NO", THEN determine the actions required to minimize risk. These actions may include new or additional outage activities, schedule changes, procedure changes or temporary modifications.

Comments to any "NO" responses that were recorded during the read through of this Addendum SHALL be addressed in the REMARKS.YES/NO Q. Any Electrical Load that impacts the following critical systems* 13.8 KV Aux/Stdby buses* 4.16 KV class 1 E -all channels operable* 120 V Vital -DP1201,1202,1203,1204 operable* 120 V Non-vital

-DP003,004,005,006 operable* 125 VDC class 1E -all channels operable 0 SSPS Actuation Cabinets 'A" and "B" and Logic Cabinets "R" and "S" must be energized for both COMS to be OPERABLE.

Logic Cabinets "R" and Actuation Cabinets "A" required for PCV 0655A "Pressurizer PORV Train A". Logic Cabinets "S" and Actuation Cabinets "B" required for PCV 0656A "Pressurizer PORV Train B".(CR 05-5960, CR 05-3071)}.

R. The following Annunciators are functional:

Lampbox 4M08* C-3, LETDN HX OUTL TEMP HI* C-4, LETDN FIX OUTL PRESS HI" C-5, PLANT COMPUTER SYSTEM ALARM* C-7, PRZR PRES DEV HI* D-3, LETDN HX TEMP HI DEMIN DVRT* D-4, LETDN HX OUTL FLOW HI/LO" D-5, PRZR PORV BLK VLV NOT OPEN* E-2, VCT LEVEL HI/LO* E-3, VCT PRESS HI/LO* E-5, PRZR PORV OPEN COMMAND" F-1, CHG PMP HEADER PRESSURE LO* F-2, VCT LEVEL LO-LO* F-3, CHG FLOW HI/LO This procedure, when completed, SHALL be retained.

II OPOP03-ZG-0007 Rev. 71 Page 173 of 216 Plant Cooldown I Addendum 12 Solid Plant Operations Entry Checklist Page 8 of 9 NOTE A "NO" answer does NOT necessarily represent an unacceptable risk situation, but identifies an area that may require further evaluation or other compensatory measures.IF the answer to any of the questions below is "NO", THEN determine the actions required to minimize risk. These actions may include new or additional outage activities, schedule changes, procedure changes or temporary modifications.

Comments to any "NO" responses that were recorded during the read through of this Addendum SHALL be addressed in the REMARKS.YES/NO Lampbox 4M07* D-l,PRTPRESSHI" E-lPRTTEMPHI" F-1, PRT LEVEL HI/LO Lampbox 5M02* B-5, RC TEMP LO ARM COMS* B-6, RCS COLD OVERPRESS ALERT-TRN A* B-7, RCS COLD OVERPRESS ALERT-TRN B* B-8, ERF SYSTEM ALARM Lampbox 1M02* C-4, RHR/LHSI PUMP I A(2A) DISCH PRESS HI* C-6, RHR/LHSI PUMP 1B(2B) DISCH PRESS HI" C-8, RHR/LHSI PUMP 1C(2C) DISCH PRESS HI* E-3, RHR PUMP 1A(2A) CURRENT LO* E-4, RHR PUMP 1A(2A) TRIP* E-5, RHR PUMP IB(2B) CURRENT LO* E-6, RHR PUMP I B(2B) TRIP* E-7, RHR PUMP 1C(2C) CURRENT LO" E-8, RHR PUMP I C(2C) TRIP* F-3, RHR TRN A ISOL VLV N/CLOSED* F-5, RHR TRN B ISOL VLV N/CLOSED* F-7, RHR TRN C ISOL VLV N/CLOSED This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 174 of 216 Plant Cooldown Addendum 12 Solid Plant Operations Entry Checklist Page 9 of 9 REMARKS: This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 175 of 216 Plant Cooldown Addendum 13 Manual Blowdown of Main Steam lines upstream of MSIVs Page 1 of 6 I I Initials NOTE 0 This Addendum MAY be performed as necessary to remove moisture from the Main Steam Lines.0 IF Condenser is NOT available, THEN REFER TO Step 4.0.1.0 IF performing Manual Blowdown of Main Steam Lines AND the Condenser is available, THEN PERFORM the following on the selected Main Steam Lines: NOTE The "MAIN STEAM OUTLET-DRAIN" valves SHALL be closed within one hour of opening the corresponding MSIV, OR Prior To reaching 20% power, UNLESS under direct control of an Operator performing "Manual Blowdown".

1.1 UNLOCK

and OPEN the following "MAIN STEAM OUTLET DRAIN" valves to remove moisture from the MS Lines: {IVC 44 ft}1 (2)-MS-0543 "S/G 1A(2A) MAIN STEAM OUTLET DRAIN VALVE" 1(2)-MS-0544 "S/G 1B(2B) MAIN STEAM OUTLET DRAIN VALVE" 1(2)-MS-0545 "S/G 1 C(2C) MAIN STEAM OUTLET DRAIN VALVE" 1l(2)-MS-0546 "S/G ID(2D) MAIN STEAM OUTLET DRAIN VALVE" This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 176 of216 Plant Cooldown I Addendum 13 I Manual Blowdown of Main Steam lines upstream of MSIVs Page 2 of 6 Initials 2.0 MONITOR the following "MAIN STEAM OUTLET DRIP LEG LEVEL SWITCH" Plant Computer points for indications of successful blowdown the Main Steam Lines:* LD7900, S/G IA(2A) MS LN DRN FROM MS-2001* LD7901, S/G IB(2B) MS LN DRN FROM MS-2002* LD7902, S/G IC(2C) MS LN DRN FROM MS-2003* LD7903, S/G ID(2D) MS LN DRN FROM MS-2004 3.0 WHEN it is NO longer desired to blowdown the Main Steam Lines from Step 1.0, THEN PERFORM the following on the selected Main Steam Lines: 3.1 CLOSE and LOCK the following "MAIN STEAM OUTLET DRAIN" valves as required: 1(2)-MS-0543 "S/G 1A(2A) MAIN STEAM OUTLET DRAIN VALVE" Perform Ind. Verify 1*(2)-MS-0544 "S/G 1B(2B) MAIN STEAM OUTLET DRAIN VALVE" Perform Ind. Verify 1(2)-MS-0545 "S/G 1 C(2C) MAIN STEAM OUTLET DRAIN VALVE" Perform Ind. Verify 1*(2)-MS-0546 "S/G ID(2D) MAIN STEAM OUTLET DRAIN VALVE" Perform Ind. Verify This procedure, when completed, SHALL be retained.

O POP03-ZG-0007 Rev. 71 [Page 177 of 216 Plant Cooldown PAddendum 13 IManual Blowdown of Main Steam lines Upstream of MSIVs Pag 3of 6 Initials 4.0 IF performing Manual Blowdown of Main Steam Lines AND the Condenser is NOT available, THEN PERFORM the following on the selected Main Steam Lines: NOTE The "MAIN STEAM OUTLET DRAIN" valves SHALL be closed within one hour of opening the corresponding MSIV, OR prior to reaching 20% power.4.1 ENSURE the following "MAIN STEAM OUTLET DRAIN DRIP LEG SECOND DRAIN ISOL VALVE" are CLOSED on the selected Main Steam Lines:* 1(2)-MS-0464 "S/G 1A(2A) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 303)* 1(2)-MS-0466 "S/G 113(2B) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 302)0 1(2)-MS-0468 "S/G IC(2C) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 301)* 1(2)-MS-0470 "S/G 1D(2D) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 304)4.2 REMOVE the pipe cap down steam of "MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" on the selected Main Steam Lines: Pipe Cap at 1(2)-MS-0464 "S/G 1A(2A) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC., EL 44 ft, Rm 303)Pipe Cap at 1(2)-MS-0466 "S/G IB(2B) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 302)Pipe Cap at 1(2)-MS-0468 "S/G 1C(2C) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 301)Pipe Cap at 1(2)-MS-0470 "S/G ID(2D) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 304)This procedure, when completed, SHALL be retained.

[OPOP03-ZG-0007 Rev. 71 Page 178 of 216 Plant Cooldown Addendum 13 Manual Blowdown of Main Steam lines upstream of MSIVs Page 4 of 6 Initials 4.3 UNLOCK and OPEN 2 a turn the following "MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE" valves on the selected Main Steam Lines:{IVC 44 ft)0 1(2)-MS-0463 "S/G IA(2A) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)"* 1(2)-MS-0465 "S/G IB(2B) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)"* 1(2)-MS-0467 "S/G I C(2C) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)"* 1(2)-MS-0469 "S/G ID(2D) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)" CAUTION The following Step will vent HOT STEAM and HOT WATER from the piping down stream of the "MAIN STEAM OUTLET DRAIN DRIP LEG SECOND DRAIN ISOL VALVE"." IF required for personnel Safety, THEN connect a suitable hose to the threaded pipe end and direct the HOT STEAM and HOT WATER to a safe area." BURN protection (Heavy Gloves) SHALL be used when handling HOT Hose components.

4.4 "VERY SLOWLY" THROTTLE OPEN the following "MAIN STEAM OUTLET DRAIN DRIP LEG SECOND DRAIN ISOL VALVE" on the selected Main Steam Lines, WHEN adequate flow is obtained, THEN STOP Opening and WAIT for DRY Steam to pass:* 1(2)-MS-0464 "S/G 1A(2A) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN 1SOL VALVE" (IVC, EL 44 ft, Rm 303)* 1(2)-MS-0466 "S/G 1B(2B) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 302)* 1(2)-MS-0468 "S/G 1C(2C) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 301)* 1(2)-MS-0470 "S/G 1D(2D) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 304)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 I Page 179 of216 Plant Cooldown Addendum 13 Manual Blowdown of Main Steam Lines Page 5 of 6 Initials 4.5 WHEN DRY Steam is passing, THEN CLOSE the "MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" on the selected Main Steam Lines: 1(2)-MS-0464 "S/G 1A(2A) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 303)1(2)-MS-0466 "S/G IB(2B) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 302)* 1(2)-MS-0468 "S/G IC(2C) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 301)* 1(2)-MS-0470 "S/G ID(2D) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 304)4.6 CLOSE and LOCK the "MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE" valves on the selected Main Steam Lines: {IVC 44 ft}* 1(2)-MS-0463 "S/G 1A(2A) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)" Perform Ind. Verif 1(2)-MS-0465 "S/G 1B(2B) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)" Perform Ind. Verif 1(2)-MS-0467 "S/G 1C(2C) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)" Perform Ind. Verif* 1(2)-MS-0469 "S/G ID(2D) MAIN STEAM OUTLET DRIP LEG FIRST DRAIN ISOL VALVE (ORC)'Perform Ind. Verif This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 180 of216 Plant Cooldown Addendum 13 Manual Blowdown of Main Steam Lines Page 6 of 6.Initials 4.7 INSTALL the pipe cap down steam of "MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" on the selected Main Steam Lines:* Pipe Cap at 1(2)-MS-0464 "S/G IA(2A) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 303)0 Pipe Cap at 1(2)-MS-0466 "S/G 1B(2B) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 302)* Pipe Cap at 1(2)-MS-0468 "S/G 1C(2C) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rm 301)* Pipe Cap at 1(2)-MS-0470 "S/G ID(2D) MAIN STEAM OUTLET DRIP LEG SECOND DRAIN ISOL VALVE" (IVC, EL 44 ft, Rmn 304)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 181 of 216 I Plant Cooldown Addendum 14 MOV-0016A, B & C Emergency Operations Guideline Page 1 of I 1.0 Purpose and Scope 1.1 This Addendum provides instructions for the Emergency Operation of SI-MOV-0016A, B & C during Shutdown conditions that require aligning LHSI pumps to the Emergency Sumps.2.0 References 2.1 CR 03-5747 3.0 Prerequisites 3.1 NONE 4.0 Notes and Precautions 4.1 The MAB Watch is the normally assigned "dedicated" Operator to release the ECO and align the Breakers for the SI-MOV-0016A, B & C valves. Alternate Operations personnel may be assigned the responsibility of the "dedicated" Operator to release the ECO and align the Breakers for the SI-MOV-0016A, B & C valves, as determined by the Shift Manager/Unit Supervisor.

5.0 Procedure

5.1 WHEN instructed by procedures to Hang an ECO on MOV-0016A, B & C and associated power breakers AND the LHSI train is required to be functional for injection/recirculation purposes, THEN ESTABLISH a "dedicated" Operator watch position to release the ECO and align the Breakers for the SI-MOV-0016A., B & C valves, as determined by the Shift Manager/Unit Supervisor.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 182 of 216 Plant Cooldown Addendum 15 Rx Head Venting and RCS level Instruments Disagreements Page 1 of 6 Evaluation Guideline NOTE This Addendum is similar to OPOP03-ZG-0009.

Addendum 11, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

Either Addendum can be used.1.0 Purpose and Scope 1.1 This Addendum provides instructions for the methodology used to vent the Reactor Head and evaluate RCS Instruments Level Disagreements.

2.0 Prerequisites

2.1 This Addendum may be performed in Mode 5, 6 or Core Off Loaded to the Spent Fuel Pool when the RCS is DEPRESSURIZED.

3.0 Notes

and Precautions 3.1 CREE 97-14388-1 states that RC-0070 is at the 47 ft. 47/8 in. elevation.

3.2 Draining

the RCS has the potential to draw a vacuum in the Reactor Vessel Head.3.3 A vacuum in the Reactor Vessel Head will cause the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" indication to be in error.3.4 Draining should be done at a slow controlled rate with ALL differences in level indication resolved prior to continued draining.3.5 Raising RCS level above the top of the RX Head (47 ft. 4 in) will re-establish the water plug in the RX Head vent to atmosphere.

Periodic venting of the RX Head to remove gasses that come out of solution WILL be required to maintain the RX Head full. The RX Head vent to atmosphere requires opening 1 (2)-RC-0509 to remove the water plug (Loop Seal) during each venting cycle.3.6 Due to ALARA concerns from Dose when venting the Head, it is recommended that RCS level NOT be maintained above the top of the RX Head for long periods of time when the RCS is de-pressurized This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 183 of216 Plant Cooldown Addendum 15 Rx Head Venting and RCS level Instruments Disagreements Page 2 of 6 Evaluation Guideline 3.7 WHEN RCS instruments level disagreements are understood OR corrected, THEN the Shift Manager authorization is required before continuing to drain down.3.8 The RCP diffuser effectively separates the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" (connected on the intermediate leg) from the Reactor Vessel when reducing water level at a rate of greater than 20 gpm. The top of the RCP diffuser is at + 10 inches from centerline.

The RCP contains very limited flow paths below the top of the diffuser.

It is recommended that the RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" be considered as inaccurate below +9 inches.3.9 WHEN Borated water drips/spills onto Carbon steel components, THEN ENSURE it is immediately cleaned up or a CR written to document the condition.

3.10 WHEN Reactor Vessel Head voiding is indicated by Pressurizer level rising with constant or rising VCT level and RVWL Sensors 1, HJTC Train "A" or Train "C" Computer Points IITE2004 and IITE3004 respectively temperature rising, THEN Addendum 20, Venting Reactor Vessel Head Using Head Vent Throttle Valve(s), provides the steps for venting the Reactor Vessel Head to PRT.3.11 IF PRT is NOT lined up to GWPS, THEN a "PRT PRESS HI" alarm (4M07-D1) may be received.

Use 0POP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation, to vent PRT as necessary.

This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 184 of 216 Plant Cooldown Rx Head Venting and RCS level Instruments Disagreements Page 3 of 6 Addendum 15 Evaluation Guideline Initials 4.0 Venting the Rx Head with RCS level greater than 47 ft. 4 in. with Head Vent rig installed NOTE WHEN the RCS level is greater than 47 ft. 4 in. the Reactor Head/Pressurizer Equalization Line has a loop seal that prevents gases in the head from escaping directly to the Pressurizer.

4.1 Vent the Rx Head by PERFORMING the following:

4.1.1 ENSURE

the Reactor Head vent manifold installed at "1(2)-RC-0 132 RX VESSEL HEAD ATMOSPHERIC VENT VALVE".4.1.2 ENSURE suitable drain hose is connected to "1(2)-RC-0509 RX VESSEL HEAD" "VENTING MANIFOLD DRAIN"{RCB On RV Head}.4.1.3 ENSURE the hose is routed to an appropriate drain and the hose is properly restrained.

4.1.4 ENSURE

the hose is restrained in a manner to prevent gross contamination the LISA.4.1.5 ENSURE NO Kinks, Blockages OR Loop Seals exist which could block venting.4.1.6 CLOSE "1 (2)-RC-0508 RX VESSEL HEAD VENTING"MANIFOLD PRZR EQUAL LINE ISOL" {RCB On RV Head}4.1.7 OPEN "1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN" {RCB On RV Head}4.2 WHEN Rx Head venting is complete, THEN PERFORM the following:

4.2.1 CLOSE

"1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN" {RCB On RV Head)4.2.2 OPEN "1(2)-RC-0508 RX VESSEL HEAD VENTING"MANIFOLD PRZR EQUAL LINE ISOL" {RCB On RV Head}This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 185 of216 Plant Cooldown Addendum 15 Rx Head Venting and RCS level Instruments Disagreements Page 4 of 6 Evaluation Guideline Initials 5.0 Venting the Rx Head with RCS level less than 47 ft. 4 in. with Head Vent rig installed 5.1 ENSURE suitable drain hose is connected to "1(2)-RC-0509 RX VESSEL HEAD' ..VENTING MANIFOLD DRAIN"{RCB On RV Head}.5.2 ENSURE the hose is routed to an appropriate drain and the hose is properly restrained.

5.3 ENSURE

NO Kinks, Blockages OR Loop Seals exist which could block venting.NOTE WHEN the RCS level is less than 47 ft. 4 in. THEN "I1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN" may remain open to provide a continuous Reactor Head vent.5.4 OPEN "1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN" {RCB On RV Head}6.0 Venting the Rx Head with RCS level less than 47 ft. 4 in. and Head Vent rig NOT Installed CAUTION*IF conditions exist that may cause RCS liquid to vent from the RX VESSEL HEAD ATMOSPHERIC VENT and Head Vent rig is NOT Installed, THEN ENSURE adequate hose/catch devices are in place to PREVENT borated water from contacting the Rx Head.* 1(2)-RC-0070 RX VESSEL HEAD VENT ISOL located in high dose area, Due to ALARA concerns after valve has been initially checked open, the Shift Manager/Unit Supervisor may direct that alternate methods are used for position verification such as previous alignment, etc.6.1 ENSURE OPEN "1(2)-RC-0070 RX VESSEL HEAD VENT ISOL"{RCB On RV Head)6.2 ENSURE OPEN "I(2)-RC-0 132 RX VESSEL HEAD ATMOSPHERIC VENT"{RCB On RV Head)6.3 ENSURE the Blind Flange removed and FME cover in place.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 186 of 216 Plant Cooldown Rx Head Venting and RCS level Instruments Disagreements Page 5 of 6 Addendum 15 Evaluation Guideline Initials NOTE Normally Head Vents are used In Mode 5 with the RCS pressurized (Refer to Step 3.36).However Step 7.0 may be required when RCS is NOT pressurized and injecting Nitrogen into SG U-tubes.7.0 Venting the Rx Head with RCS level areater than 47 ft. 4 in. with Head Vent rig NOT installed 7.1 OPEN Head Vent Isolation Valves: " ISOLHV-3657A and ISOL HV-3658A OR* ISOL HV 3657B and ISOL HV-3658B 7.2 OPEN Head Vent Throttle Valve(s):* HCV-0601* HCV-0602 7.3 WHEN the Reactor Vessel Head void is vented (as indicated by RVWL Sensor 1 temperature lowering and pressurizer level lowering with an associated PRT pressure rise), THEN PERFORM the following:

7.3.1 ENSURE

CLOSED the Head Vent Throttle Valve: " HCV-0601" HCV-0602 7.3.2 ENSURE CLOSED the Head Vent Isolation Valves:* HV-3657A* HV-3657B" HV-3658A* HV-3658B This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 187 of216 Plant Cooldown Rx Head Venting and RCS level Instruments Disagreements Page 6 of 6 Addendum 15 Evaluation Guideline Initials 8.0 RCS level Instruments Disagreements Evaluation Guideline 8.1 IF a difference of greater than the specified band in indication exists between RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" and RVWL OR RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" and Pressurizer Cold Cal, THEN: 8.1.1 ENSURE the Rx Head and the Pressurizer are equalized in pressure.8.1.2 ENSURE the Rx Head and the Pressurizer equalizing line in service.8.1.3 PERFORM a walkdown of the sightglass.

8.1.4 IF RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" is in service, THEN ENSURE "OPOP07-RC-0001, LINEUP 1, "RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)" in Service Lineup" PERFORMED daily and DOCUMENTED in a temporary log.8.1.5 CORRECT any discrepancies noted in the sightglass walkdown.8.1.6 VENT the reactor vessel head as required.8.1.7 OBSERVE CAUTION when determining which level indication is to be used. Inadequate head venting due to a rapid draining rate can cause RVWL to remain higher than actual loop level.8.1.8 RVWL OR the Pressurizer Cold Cal level channel instrumentation SHALL be used as the correct level indication until proven inaccurate.

8.1.9 MONITOR

the applicable instruments

{ RVWL, Pressurizer Cold Cal, Narrow Range Hot Leg Levels and RC-LG-3662 "RCS LEVEL SIGHTGLASS (SLINKY)"}

for level convergence.

8.1.10 PROCEED with RCS drain down with Shift Manager's permission.

This procedure, when completed, SHALL be retained.

0POP03-ZG-0007 Rev. 71 Page 188 of 216 Plant Cooldown I Addendum 16 Throttling "1(2)-CV-0198 RMW ISOL" for Technical Specification 3.4.1.3 I CAUTION* This Addendum SHALL ONLY be used for setting the throttle position of "1 (2)-CV-0 198 RMW ISOL" for Technical Specification 3.4.1.3.This Addendum bypasses normal automatic termination functions and alarms designed to alert the operator of any flow deviations which might cause an over dilution/boration." An operator SHALL be stationed at the controls with NO concurrent duties while using this Addendum to monitor for, and secure over dilution of the RCS.1.0 PERFORM the following to set the throttle position of "1(2)-CV-0 198 RMW ISOL" for Technical Specification 3.4.1.3: 1.1 DISPATCH an Operator locally to "1(2)-CV-0198 RMW ISOL"{51 ft MAB Rm 335)1.2 THROTTLE "1(2)-CV-0198 RMW ISOL"{51 ft MAB Rm 3351 approx 4 3/turns open (z32% open) to limit flow to the blending tee to less than or equal to 110 gpm.CAUTION MONITOR total RMW added to the VCT to ENSURE the RCS is NOT diluted below current Shutdown Margin Boron Concentration requirements.

1.3 ENSURE

NO Borations or Dilutions are in progress.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 189 of 216 Plant Cooldown I Addendum 16 Throttling "1(2)-CV-0198 RMW ISOL" for Technical Specification 3.4.1.3 1.4 ENSURE the following valves are CLOSED:* "BA FLOW CONT VLV FCV-01 10A"* "RMW FLOW CONT VLV FCV-0l 1 A"* Makeup Stop to Charging Pumps "TO VCT OUTL FCV-0101B" valve.* Makeup Stop to VCT "FILL FCV-01 1 B" valve.1.5 ENSURE ONLY one (1) of the Reactor Makeup Water Pumps is running." "RMW PUMP IA(2A)" OR* "RMW PUMP IB(2B)" 1.6 ENSURE the standby RMW pump in "PULL TO LOCK". {CP004}"RMW PUMP 1A(2A)""RMW PUMP 1B(2B)" 1.7 CHECK the value indicated on the flow totalizer "TOTAL / GAL FQI-01 111B".CAUTION From full open to full closed, the stroke times of the makeup valves are approximately 1 -2 seconds. These stroke times should be anticipated while making up to the RCS to minimize any makeup overshoot.

For very small amounts of makeup, the valves may need to be placed in the CLOSE position before the valve reaches full open.1.8 OPEN Makeup Stop to VCT "FILL FCV-01 1 IB" valve.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 190 of 216 I Plant Cooldown Addendum 16 Throttling "1(2)-CV-0198 RMW ISOL" for Technical Specification 3.4.1.3 S NOTE It is permissible to cycle (open/close) "RMW FLOW CONT VLV FCV-01 1 A" while performing adjustments on "1(2)-CV-0198 RMW ISOL" to minimize the dilution of the RCS.AC 1.9 OPEN "RMW FLOW CONT VLV FCV-01 1 A", WHILE coordinating with the Operator at 1(2)-CV-0198

{51 ft MAB Rm 335} to THROTTLE"1(2)-CV-0198 RMW ISOL" as required, to limit maximum RMW flow to the VCT between 100 gpm to 110 gpm, as read on Inst F-I ll on recorder "BA BLEND FLOW" FR-01 10 (CP004) (CR 98-14576).

Perform Dual Verif 1.10 WHEN the maximum RMW flow to the VCT is between 100 gpm to 110 gpm, as read on Inst F-I ll on recorder "BA BLEND FLOW" FR-01 10 (CP004), THEN PERFORM the following (CR 98-14576):

1.10.1 ENSURE "RMW FLOW CONT VLV FCV-01 1 A" is closed.1.10.2 LOCK IN PLACE "1(2)-CV-0198 RMW ISOL". {51 ft MAB Rm 3351 Perform Ind. Verif 1.10.3 CLOSE the Makeup Stop valve to VCT "FILL FCV-01 I IB".This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 191 of 216 I7 Plant Cooldown I Addendum 16 Throttling "1 (2)-CV-0198 RMW ISOL" for Technical Specification 3.4.1.3 1.11 ENSURE the following valve handswitches are in the "AUTO" position:* "BA FLOW CONT VLV FCV-01 10A"* "RMW FLOW CONT VLV FCV-01 1 A"* Makeup Stop to Charging Pumps "TO VCT OUTL FCV-01 10B" valve.* Makeup Stop to VCT "FILL FCV-011 I IB" valve.1.12 IF NOT desired to run the RMW pump, THEN STOP the RMW pump by placing the handswitch in "STOP- and return to "AUTO". {CP004}* "RMW PUMP IA(2A)" .. "RMW PUMP IB(2B)" 1.13 PLACE a Caution Tag at the following locations reading "RMW flowrate restricted per Technical Specification 3.4.1.3. Acceptance Criteria: 1(2)-CV-0198 RMW ISOL throttled for less than 110 gpm flow and ONLY one (1) RMW pump in service"{CP004}:

• "1(2)-CV-0198 RMW ISOL". {51 ft MAB Rm 335}* FK-01 11 "RMW TOT M/U FLW CONTR (RMW FLOW CONT FK-0111)" (CP004)The standby RMW pump's Control Room handswitch: "RMW PUMP IA(2A)" "RMW PUMP 1B(2B)" 1.14 RECORD the ECO number: ECO #1.14.1 ENSURE a Shift Manager/Unit Supervisor or designee has accepted (signed on to) the ECO(s) listed above.1.15 SET RMW Flow Control to the desired flow blended flowrate using "RMW FLOW CONT FK-01 11". {CP004}1.16 IF required, THEN adjust RCS boron concentration to maintain Shutdown Margin.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 192 of 216 E ~ddend~rn 17 Plant Cooldown aeIo2 Moving Control Rods with High Reactor Coolant Turbidity Page 1 of 2 Initials CAUTION There is a rise in risk to experiencing a misstepping event during power operations following rod movement at turbidity levels greater than 1.1 NTU.While performing rod control movement activity it is possible to experience rod misstepping.

1.0 NOTIFY

System Engineering that rod movement is to occur with Turbidity GREATER THAN 1.1 NTU.2.0 Record the Reactor Coolant Turbidity Level prior to moving rods.Turbidity

_NTU Time/Date 3.0 REVIEW CREE 08-13703-1.

4.0 Proceed

with rod holdout activity.NOTE Rod rnisstepping is indicated by deviation between a rod(s) DRPI indication and its associated group(s) step counters of greater than or equal to12 steps excluding ROD BOTTOM.5.0 IF any rod misstepping occurs during rod holdout activity, THEN WRITE a Condition Report documenting the misstep and continue with rod holdout activity.6.0 IF any one rod missteps three or more times during rod holdout activity, THEN PERFORM the following:

6.1 TERMINATE

Rod holdout activity.6.2 INSERT all rods to rod bottom per OPMP07-DM-0003, Rapid Refueling Rod Holdout Operation.

6.3 ENSURE

Reactor Coolant cleanup to LESS THAN OR EQUAL TO 1.1 NTU.6.4 PERFORM OPOP07-RS-0001, Control Rod Exercise.6.5 PERFORM OPMP07-DM-0003, Rapid Refueling Rod Holdout Operation.

This procedure, when completed, SHALL be retained.

I OPOP03-ZG-0007 Rev. 71 Page 193 of 216 J Plant Cooldown F Addendum 17 Moving Control Rods with High Reactor Coolant Turbidity Page 2 of 2 ]Initials NOTE Consult with Chemistry, Outage and System Engineering to schedule backend Control Rod Exercise during a period of RCS low Turbidity level.7.0 IF any rods misstepped during rod holdout activity, THEN INITIATE an OPERABILITY ASSESSMENT SYSTEM (OAS) entry to perform rod exercising per OPOP07-RS-000 1, Control Rod Exercise, prior to Reactor Startup.8.0 Send a copy of Addendum 17 to System Engineering.

This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 194 of 216 Plant Cooldown Addendum 18 Indicated Pressurizer Level When Solid vs. Pressurizer Page 1 of 1 Temperature NOTE Graph from CREE 02-1259-4 that can be used to estimate the indicated Pressurizer Cold Calibrated Level and should only be used as an operator aide for the predicted indicated value when the pressurizer indicates full: " % Indicated Level is Pressurizer Cold Calibrated Level* Temperature (F) is Pressurizer Liquid Temperature

%Level Indicated when 101.12% full 104 102 -y7-05x 2 -5.385171 E-03x +I 23585E+02 100 9.997547E-01 98-j 96$1 94-.U -u Series2 92 Poly. (Series2)90 88 86 84 0 100 200 300 400 500 Temperature (F)1. OBTAIN Pressurizer Liquid temperature (TI-0608)2. USE graph to estimate indicated Pressurizer Cold Calibrated Level (LI-0675) when pressurizer indicates full.This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Pace 195 of 216 Plant Cooldown Controlling RCS Inventory at or above Elv 39 ft. 4.9 in. Page I of 10 FAddendLIM 19 1 INITIALS NOTE* This addendum is to be used for controlling RCS/PZR level during the performance of Section 9.0 and allows RCS/PZR level to be adjusted as needed within the band of 20%PZR Cold Calibrated level to 39 ft 4.9 in.* Section 9.0 SHALL BE completed for RCS drain down and re-fill.1.0 Prerequisites

1.1 ENSURE

Seal Injection in service to all coupled RCPs per OPOP02-CV-0004, Chemical and Volume Control System Subsystem.

CAUTION* Do NOT Raise RCS level above 20% pressurizer cold cal level using this Addendum.* IF PZR Cold Calibrated level indicates less than 0% and the RCS level sightglass level indication is at the top of the sightglass, THEN the RCS Fill SHALL be stopped and the level difference between PZR level and RCS level sightglass investigated.

2.0 PERFORM

the following to RAISE RCS Inventory:

2.1 DETERMINE

level to be raised to.2.2 ENSURE a prejob brief has been performed.

2.3 IF RCS level is to be raised above 47 ft. 4 in, THEN CLOSE the following valves:* "1(2)-RC-0507 RX VESSEL HEAD""VENTING MANIFOLD VENT VALVE"{RCB On RV Head}* "1(2)-RC-0509 RX VESSEL HEAD""VENTING MANIFOLD DRAIN"{RCB On RV Head}2.4 ENSURE "CHG FLOW CONT FK-0205" in MANUAL. (CP004)2.5 OPEN "CHG FLOW CONT FK-0205" as necessary to initiate flow to the RCS.(CP004)This procedure, when completed, SHALL be retained.

OPOP03-ZG-0007 Rev. 71 Page 196 of 216 Plant Cooldown Addendum 19 Controlling RCS Inventory at or above Elv 39 ft. 4.9 in. Page 2 of 10 INITIALS 2.6 MAINTAIN letdown flow less than the combined charging and seal injection flow.NOTE PRT level may NOT rise due to low system pressure and head loss restrictions in the vent piping.2.7 MONITOR PRT level.2.7.1 IF PRT level rises, THEN CLOSE the reactor vessel head vent valves to prevent filling the PRT. (CP005)* "HEAD VENT THROT VLV HCV-0601"* "HEAD VENT THROT VLV HCV-0602"* "ISOL HV-3657A"* ~"ISOL HV-3658A"* "ISOL HV-3657B" 0 "ISOL HV-3658B" 2.8 CONTROL "CHG FLOW CONT FK-0205" to ENSURE RCS pressure does NOT exceed 50 psig.2.9 IF RCS pressure rises unexpectedly, THEN REDUCE Charging flow rate as required to maintain pressure control.2.10 WHEN the desired RCS/Pressurizer level is reached, THEN ADJUST the charging, seal injection and letdown flows to MAINTAIN the desired RCS/Pressurizer level.2.11 IF RVWL AND RCS level sightglass, OR RVWL AND pressurizer Cold Calibrated level instrumentation do NOT agree within 6 inches, THEN STOP the fill and REFER TO Addendum 15, Rx Head Venting and RCS level Instruments Disagreements Evaluation Guideline.

2.12 MAINTAIN RCS level per Shift Manager/Unit Supervisor direction.

This procedure, when completed, SHALL be retained.