ML18019A367
| ML18019A367 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/03/1985 |
| From: | Cutter A CAROLINA POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM GL-85-12, NLS-85-302, NUDOCS 8509110261 | |
| Download: ML18019A367 (27) | |
Text
Pf ff REGULA ORY INFORMATION DISTRIBUTI SYSTEM (BIDS)
ACCEQSIOQ N8R: 6509110261 DOC ~ DATE: 85/09/03 NOTARIZED:
NO FACIL;50-400 Shear on Harris Nuclear. Power Planti Unit ii Carolina AUTHiNAME AUTHOR AFF ILIAT'ION CUTTERgA,BE Carolina Power L Light Co.
RECIP ~ NAME RECIPIENT AFFILIATION DENTONiHRe Office of Nuclear Reactor Regulationi Director DOCKET' 05000400
SUBJECT:
Forwards response to Generic Ltr 85-i2 re TMI Action Item II.K.3+Siincluding plant-specific info necessary to document acceptable reactor coolant pump criteria,Info considered adequate to resolve SER Confirmatory Item 33+
DISTRIBUTION CODE:
8001D COPIES RECEIVED:LTR ~
ENCL SIZEi TITLE> Licensing Submittal:
PSAR/FSAR Amdts 8 Related Correspondence NOTES:
RECIPIENT ID CODE/NAME NRR/DL/ADL NRR'B3 LA INTERNAL; ACRS 41 ELD/HDS1 IE/DEPER/EPB 36 NRR ROEiM ~ L NRR/DE/CEB 11 NRR/DE/EQB 13, NRR/DE/MEB 18 NRR/DE/SAB 24 NRR/DHFS/HFEB40.
NRR/DHFS/PSRB NRR/DSI/AEB 26 NRR/DS I/CP 8 10.
NRR/DSI/ICSB 16 NRR/DS I/PSB 19 NRR/DS I/RS8 23 RGN2 EXTERNAL: 24X DMB/DSS (AMDTS)
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Carolina Power & Light Company SF.P 03 19SS SERIAL: NLS-85-302 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. l DOCKET NO.50-000 RESPONSE TO TMI ACTION ITEM II.K.3.5, GENERIC LETTER 85-12
Dear Mr. Denton:
Carolina Power R Light Company (CPRL) hereby submits information in response to TMI Action Item II.K.3.5. The enclosure provides the plant specific information necessary to document an acceptable reactor coolant pump trip criteria. This information is provided as a item-by-item response to Section IV of the SER attached to Generic Letter 85-l2.
The attached information is considered adequate to resolve SER Confirmatory Item 33.
If you have any questions on this subject, please contact Mr. 3. Eads at (919) 362-2985.
Yo very ly, ABC/3DK/rtj (l8283DK)
Attachment cc:
Mr. B. C. Buckley (NRC)
Mr. G. F. Maxwell (NRC-SHNPP)
Dr. 3. Nelson Grace (NRC-RII)
Mr. Travis Payne (KUDZU)
Mr. Daniel F. Read (CHANGE/ELP)
Wake County Public Library A. B. Cutter - Vice resident Nuclear Engineering R Licensing Mr. Wells Eddleman Mr. 3ohn D. Runkle Dr. Richard D. Wilson Mr. G. O. Bright (ASLB)
Dr. 3. H. Carpenter (ASLB)
Mr. 3. L. Kelley (ASLB)
I. 8509ii026i 850903 PDR, ADOCK 05000400 E.
PDR 411 Fayettevilte Street
~ P. O. Box 1551
~ Raleigh, N. C. 2'r602
ATTACHMENTTO NLS-85-302 Item:
A.
Determination of RCP Trip Criteria-l.
Identify the instrumentation to be used to determine the RCP trip setpoint, including the degree of redundancy of each parameter signal needed for the criterion chosen.
~Res nse:
The RCP trip criterion chosen for SHNPP is RCS pressure, the instrumentation used is RCS Wide-Range Pressure, PT-000, and Redundant Instrument PT-OOI.
(1828JDK/rtj) '.
2.
Identify the instrumentation uncertainties for both normal and adverse containment conditions.
Describe the basis for the selection of the adverse containment parameters.
Address, as appropriate, local conditions such as fluid jets or pipe whip which might influence the instrumentation reliability.
~Res onse:
Calculation of instrument uncertainties for normal and adverse containment conditions is documented in the SHNPP Setpoint Study.
The portion applicable to RCP trip criterion (RCS hVide-Range Pressure) is included as Attachment A.2. The SHNPP Setpoint Study basis is documented in Reference 6 to that study, Letter TMI-OG-132, dated December 27, 1979, "3ustification of Instrument Setpoints Used in Emergency Operating Instruction Guidelines."
(1828JDK/rtj
)
3.
In addressing the selection of the criterion, consideration to uncertainties associated with the WOG-supplied analyses'alues must be provided.
These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant-specific features not representative of the generic data group.
If a licensee determines that the WOG alternative criteria are marginal for preventing unneeded RCP trip, it is recommended that a more discriminating plant-specific procedure be developed.
For example, use of the NRC-required inadequate core-cooling instrumentation may be useful to indicate the need for RCP trip. Licensees should take credit for all equipment (instrumentation) available to the operators for which the licensee has sufficient confidence that it willbe operable during the expected conditions.
~Res ense:
Calculation of RCP trip criterion for SHNPP is based on WOG Emergency Response Guidelines and is documented in the SHNPP Setpoint Study.
The portion applicable to RCP trip criterion is included as Attachment A.3-1. The results of the WOG limiting parameters when compared against SHNPP calculated values is included as Attachment A.3-2. Finally, a summary of the basis for selection of a specific SHNPP RCP trip criterion is included as Attachment A.3-3.
(1828JDK/rtj
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Attachment A.2 Page 1 af 2
3.1 Instrument
PT-440 "RCS Wide Range Pressure (PI-4'40,)
'ange:
0 3000 PSIG Inside ContainmentP Yes Smallest Graduation:
50 PSIG Transmitter:
ITT'arton Model 763 PZ-440 Instrument Uncertainties 3.1.1 Normal Transmitter Reference Accuracy (includes linearity< hysteresis, and repeatability)
[Reference 2.1]
+0.5% of span = 0.005 x 3000 psi
++5.0 psi 3.1.2 Accuracy of pressure gauge used in calibration
[Reference 2.9]
+0.5% of span = 0.005 x 3000 psi = ++5 psi 3.1.3 Allowed Calibration Tolerance
[Reference 2.9]
+0.21% of span
= 0.0021 x 3000 psi = +6.3 psi 3.1.4 Ambient Tanperature Effect on Transmitter
[Reference 2.3]
+0.5% of span/(fram 40oF to 130oF)
= (.005)
(3000 psi) = +Q5 psi 3.1.5 Maxinum Transmitter Drift [Reference 2.3]
++.0% of span = 0.01 x 3000 psi = +~0 psi 3.1.6 Normal Indicator Accuracy
[Reference 2.4]
++% of span = 0.02 x 3000 psi = +60 psi 3.1.7 Indicator Reading Error
++/4 of smallest graduation
= 0.25 x 50 psi = ++2.5 psi 3.1.8 Normal Instrument Accuracy
[Reference 2.4]
+0.5% of span
= 0.005 x 3000 psi = +Q5 psi Dunno 1'9
Attachment A.3-1 Page 1 of 7
6.0 RCP Trip Criteria 6.1 Calculation For RCP.Trip Based on RCS Pressure Notes:
1.
Decay heat two minutes following reactor trip is 3.5%
2.
SHNPP 2775 t%T/100% steam flag = 12.2 x 106 ltd/hr 3.
Three loop RCP heat input 10 le 4.
S/G safety set pressure (lowest) ; 1170 PSIG, capacity
= 881, 980 1hz/br.
5.
3.3 psid across S/G flow limiter at 100% steam flow.
Calculations:
1.
Decay heat level = 3.5%
2.
Steam flow rate Q (heat input/S/G)
=
(NSSS Power f%/number of loops)
(Decay heat fraction) +
(RCP heat input/nurrber of loops)
Q = [(2775 MH/3) (.035) + (10 HH/3) ] (3,412 x 106 BTU/hr/MH Q = 1.22 x 108 BTU/hr M (Steam flow rate/loop) =
Q latent heat of vaporization (at 1185 PSIA)
M = 1.22 x 108 BTU/hr/615 BTU/ibm = 1.98 x 105 lb'/hr 3.
Verification that only lowest S/G safety is adequate.
100 (1.98 x 105 le/hr/8.81 x 105 l~r = 22.5% which is less than 60% capacity Therefore only using lowest safety valve provides acceptable capacity.
4.
Three percent of liftpressure (1170 PSIG)
(.03)
= 35 psi
)
Attachment A.3-1 Page 2 of 7
5.
Differential pressure between S/G tube region and S/G safety valves.
Assume maximum pressure drop from S/G tube region to S/G safeties steam flow limiter at entrance to steam pipe (3.3 psid at 100% steam flaw)
Make conservative assumption that 30 psi pressure drop exists between S/G tube region and S/G safeties.
The head loss (hL) in a piping system is calculated as follows hL = kV2 2gc where k is a constant dependent on piping configuration gc is a physical constant Therefore hL is proportional to (velocity of stean) 2 hL (at reduced flow) = hr, (rated flow) [Reduced stgam flow velocity/
ted steam flow velocity]W
- Reduced flow (%) = (1.98 x 105 Um/hr/4.1 x 106 ltdr) 100 approximately 5%
- Cross sectional area of steam pipe (32" diameter)
.A =~/ =
Qg (l6 in./l2 in/ft)2 = 5.6 ft2
- For 5% flow, assume pressure is 1185 psia, vg =.3686 ft3/ibm
- For 100% flow assume pressure is 964 psia, vg =.4657 ft3/lcm m = PAV = AV/v ; V = mv/A Velocity at 5% = mv/A = (1.98 x 105 ~r)
(.3686 t3 1.3 x 104 ft/hr 5.6 ft.
Velocity at 100% = mv = (4.1 x 106 ibm/hr) (.4 3
)
A 5.6 ft.
= 3.4 x 105 ft/hr Therefore head loss at reduced flow hL (reduced flow) = hL (rated flow) [Velocity r uced flow/Velocity rate flow]~
hL (reduced)
= 30 psid
(
4
)2 =.04 psid 3.4 x 10 ft/hr
~refore the stated upper bound value of 1 psid is acceptable.
Attachment A.3-1 Page 3 of 7
6.
Delta temperature across S/G tubes Step 1 - Calculate T~ Value TCOLD TSEC (Delta T(full power) x Power Fraction) where TS~ = TSAT (1200 PSIA) ~ 569.4oF Delta T(full power)
= 618oF 556oF 62oF Power Fraction = (2775 HH) (.035) +.10M'/2785 l% =.0385 T~ = 569.4oF + (62oF x.0385)
= 571.8oF Step 2 Calculate THOI Value H(Z = T~ + (Delta T(full range) x Power Fraction)
THOI = 571.8oF + (62oF x.0385)
= 574.2oF Step 3 Calculate ZOOID 574.
I 571.8oF 569.4oF
= 3.5oF For S/G set at 1185 Psia gives a Tsat of 565 'oF 565.7oF + ZAN = 565.7oF + 3.5oF = 569.2oF PSAT (569.2oF)
= 1217.5 PSIA Therefore pressure correction error = 1217.5 PSIA 1185 PSIA
= 32.5 PSIA 7.
Delta pressure between RCS wide range instrument and top of S/G U-tubes Dynamic Delta P = Primary pressure drop across S/G/2 + RCS hot leg pressure drop Delta P = 39.8 psi/2 + 1.4 psi = 21.3 psi Static Vertical distance from top of KS hot leg (RCS wide range pressure) to top of U-tubes (From SHNPP FSAR fig. 5.1.3-1 make conservative assumption that bottom of S/G is at same elevation as h of hot leg)
(Hot leg outer diameter is approximately 34 inches)
Attachment A.3-1 Page 4 of 7
~
~
Therefore vertical distance
~ 34 inches/2
+ 386 inches estimated from S/G Tech. Nanual for RKPP K 16-S120-3005
~
402 inches
~
approximately 33.6 PZ.
Therefore static Delta P
~
(33.6 ft)
(.491 psi/ft)
(.8 density correction)
~ 13.2 psi
'herefore total Delta P ~ Dynamic + static
= 21.3 psi + 13.2 psi +
34.5 psi Therefore, RCP trip pressure is 1.
3% accumulated pressure,.-.
2.
Steam line differential pressure 3.
Primary to secondary Delta T effect 4.
Dynxtu.c and static primary Delta P 35 32.5 103 psi From section 3.0 the following uncertainties are obtained:
KS wide range pressure errors (normal CV) +80.6 psi (adverse CV) + 309.9 psi RCS trip pressure is (Normal CV)
(Adverse CV) 1170 psig + 103 psi + 80.6 psi = 1354 psig (rounded to 1360 PSIG for use in EOP's) 1170 psig + 103 psi = 309.9 psi = 1583 psig (rounded to 1600 PSIG for use in EQP's)
Attachment A.2 Page 2 of 2 3.1.9 Naximan Instrument Drift [Reference 2.6]
+0 of span 0.01 x 3000 psi ~ 0 psi Therefore, Maxim'ormal Instrument Error is:
[4(15)2+ (6 3)2+ (60)2+
(12 5)2+ 2(30)2~1/2 80.6 psi Adverse CV effects on PT-440 Ref. 2.3 gives value of transmitter accuracy within ++0% after Loss of Coolant Accident.
To calculate adverse CV effects on RCS wide range pressure, the "normal" CV reference transmitter error (0.5%)
and the anbient tarp. effects on transmitter (0.5%) will be replaced by +~0% [(Q.l)
(3000 psi)) = +F00 PSI for adverse.
CV effects on transmitter.
Therefore, maxUMm total adverse CV errors on RCS wide range pressure is:
[(3QQ)
+ 2(15)
+ (6 3)
+ (60)
+ (12.5)
+ 2(30) ]+
= 309 9 psj Paae 13
Attachment A.3-1 Page 5 of 7 6.2 Calculation For RCP Trip Based on RCS Subcooling Minimum pressure for SGTR and non-LOCA events is specified in the WOG ERG Executive Volume.
Use 1300 PSIA to calculate subcooling errors.
T(sat)
(1300 PSIA) i 577.60 F
From section 3 the following uncertainties are obtained:
Hmzz~l hdwarm.M RCS loop RID (non-bypass) errors
......17.3oF 18.2oF Wide range KS pressure instrument error 80.6 psi 309.9 psi For normal QT, add temperature and pressure error. in non~nservative direction (i.e. add temperature errors, subtract pressure errors)
T(sat)
(1300PSIA 80.6 psi) = Tsat (1219.4 PSIA) = 569.2 P
Therefore subcooling error due to KS pressure error = 577..6 569.2 = 8.0oF Subcooling error due to RID instrument error = 17.3oF Therefore subcooling error total
= 8.0oF
+ 17.3oF
= 25.3oF rounded to 25oF normal CV For adverse CV, add temp.
and pressure errors in non-conservative direction Tsat (1300 PSIA 309.9 psi) = Tsat (990.1 PSIA) = 543.5oF Subcooling error due to RCS pressure error = 577.6oF 543.5oF = 34.1oF Subcooling error due to RID inst. error = 18.2oF Total subcooling error = 34.1oF + 3.8.2oF = 52.3oF rounded to 55oF adverse CV
~
~
Attachment A.3-1 Page 6 of 7
6.3 Calculations For BCP Trip Based Ch BC'.~Secondary Pressure Differential 1.
- S/G pressure variable 2.
Delta P between steam pressure measurement and S/G shell 1 psi (see calculation for KS pressu're) 3.
Delta P across the S/G tubes a.
For lowest S/G safety valve set pressure value is 32.5 psi (see calculation for KS pressure) b.
For no-load pressure
..(assume 1100 PSIA)
Step 1 Tsec Tsat (1100 PSIA)
~ 556.4oP Delta T full power = 62oF Power fraction =.0385 T~ = 556.4oP + (62 P x.0385)
= 558.8oF Step 2
THOI = 558.8oP + (62oF x.038)
= 561.2oF Step 3
561.
8 8o LKZD = ln f 61.2o
= 3.5oF 558.8oF 556.4oF]
Psat (556.4oF + 3.5oF)
= Psat (559.9oF)
= 1131.8 PSIA 1131.8 PSIA 1100 PSIA = 31.8 PSIA Therefore Delta P due to ?RID 32.5 (limiting case) 4.
Delta P between wide range RCS pressure and top of S/G tubes
= 34.5 psi (see calculations for RCS pressure) 1.
S/G pressure 2.
Steam line differential pressure 3.
Primary to secondary delta pressure 4.
Dynanic and static primary delta p (total)
Variable 1 psi 32.5 psi 68 psi
Attachment A.3-1 Page 7 of 7 5.
From section 3.0 the following uncertainties are obtained:
RCS wide range pressure errors S/G pressure errors 80.6 psi 34.9 psi 309.9 psi 34.9 psi For normal cv total Total instrument error = f80.62 + 34.92]1/2 ~ 87.8 psi rounded to 88 psi For adverse CV t%
Total instrument error i [309.92 + 34 ~92]l/2 311.9 psi rounded to 312 psi Therefore (normal Vfj RCS to S/G pressure
= 68 psi + 88 psi 156 psi (Mverse CV)
RCS to S/G pressure
= 68 psi + 312 psi = 380 psi
Attachment A.3-2 Page 1 of 1
REF:
1, Executive Volume, Generic Issues; Section TABLE 1 (Cont) dealing with RCP Trip/Restart N
LIMITING RESULTS OF SGTR AND NON-LOCA ANALYSIS PLANTS MINIMUM RCS MINIMUM RCS
~PRE MRE PEI K
LI G
MINIHUM RCS/SECONDARY DI FFERENTIAL PRESSURE PSI Indian Point 2
1175 31 293 Indian Poin.
3 1196 32 315 Virgil Summer Shearon Harris 1 and 2
lfqlccrs fi"~ A'ac/nrmf i%.3"l:
Farl ey 1 and 2
North Anna 1
and 2
Surry 1 and 2
Beaver Valley 1
1421 I peon CIaoc J 1219 51 zs'ss 37 549 (5( L3b'c 350 Beaver Valley 2 1132 30 278 Robinson 2
Turkey Poirt 3 and 4
1232 31 309 Prairie Island 1 and 2
1348 39 389 Kewaunee 1238 38 361 Ginna Point Beach 1 and 2
1166 29 305 Connecticut Yankee San Onofre Yankee Rowe Results were not obtained for these plants RCP TRIP 0069V:1 22 HP/LP-Rev
~
1
Attachment A.3-3 Page 1 of 1
A3.
Specific SHNPP Parameter Used for RCP Pump Trip Criteria During SGTR Events As mentioned previously, RCP pump trip criteria for SHNPP was calculated using the generic ERG Revision 1 guidelines as a basis for calculating plant specific RCP trip criteria.
For SHNPP, all three RCP trip parameters, namely:
1.
RCS Pressure 2.
RCS Subcooling 3.
RCS to Steam= Generator Differential Pressure were within the cutoff limits allowed by the generic analysis.
(Operations Engineer has the calculations.)
For SHNPP
The decision to chose RCS Pressure for SHNPP EOPs was based on the following items:
RCS to Steam Generator Differential Pressure This parameter requires the operator to MCB indications and then perform mental places an. additional requirement on the stress situation and is, therefore, not continuously read two subtraction.
This operator during a high desirable.
RCS Subcooling This parameter was not selected since there are non-LOCA situations where.the operator could lose RCS subcooling and not need to trip RCPs thereby losing the preferred means of core heat removal (i.e., forced circulation).
RCS Pressure This parameter waC selected since it:
l.
Is a concept that is already familiar to personnel with previous Vestinghouse PWR experience.
2.
Requires the operator to monitor only one indication on MCB (RCS Pressure).
3.
" Is a parameter that the operator normally monitors following a reactor trip or safety injection.
Item:
B.
Potential Reactor Coolant Pump Problems l.
Assure that containment isolation, including inadvertent isolation, willnot cause problems if it occurs for non-LOCA transient and accidents.
a.
Demonstrate that, if water services needed for RCP operations are terminated, they can be restored fast enough once a non-LOCA "situation is confirmed to prevent seal damage or failure.
~Res oese:
b.
Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.
~e a.
RCP seal injection is continuously maintained during accident conditions; thus, any containment isolation signal willnot damage RCP seals.
b.
There are two isolation signals to be discussed which could affect RCPs.
The first is Containment Isolation Phase A which is generated by either of the following:
Safety Injection Actuation Signal (SIAS)
Manual from MCB On a Containment Isolation Phase A signal, no RCP services are interrupted, therefore, continued RCP operation can occur without damage to seals or pump failure.
The second is Containment Isolation Phase B which is generated by either of the following:
Containment pressure above High-3 (10 psig)
Manual from MCB On a Containment Isolation Phase B signal, CCW to RCP motor-oil coolers and thermal barrier heat exchangers isolates.
The RCP seals willbe maintained since seal injection flow from the charging safety injection pumps is not isolated.
However, loss of CCW to the motor-oil coolers requires the RCPs to be tripped due to inability to adequately cool the RCP motor oil. For this reason, the SHNPP EOPs call for tripping of all RCPs whenever any Phase B isolation signai is received.
Also, Abnormal Operating Procedures call for tripping of affected RCPs (l) whenever any bearing or motor-winding temperatures exceed alarm limits, or (2) within 10 minutes if CCW flow to either RCP motor oil cooler is lost.
(1828JDK/rtj
)
2.
Identify the components required to trip the RCPs, including relays, power supplies, and breakers.
Assure that RCP trip, when determined to be necessary, willoccur. If necessary, as'a result of the location of any critical component, include the effects of adverse containment conditions on RCP trip reliability. Describe the basis for the adverse containment parameters selected.
~Res ense:
Components required for RCP trip:
MCB Control Switch 6.9kV Auxiliary Bus Breaker RCP Trip Coil No.
1 125V DC Control Power 102SA
~"-
100SA Bus A, Cub 5 Bus B, Cub 9 Tc 1-102 TC 1-100 DP-1A-SA DP-1A-SA 106SA Bus C, Cub 2 TC1-106 DP-lA-SA Allcomponents are located in the Reactor Auxiliary Building and, thus, are not affected by adverse containment conditions.
(1828JDK/rtj
)
C.
Operator Training and Procedures (RCP Trip)
I.
Describe the operator training program for RCP trip. Include the general philosophy regarding the need to trip pumps versus the desire to keep pumps running.
~Res ense:
Candidates for RO or SRO level licenses receive classroom instruction on RCP trip criteria, including: I) basis for requiring RCP trip under small-break LOCA conditions,
- 2) transient analyses for RCP trip at varying times before and after RCP trip criteria are met, 3) methodology for calculating RCP trip criteria'and selection of SHNPP specific criterion, and 0) RCP trip parameters and bases for nonaccident conditions.
In addition, plant-specific simulator instruction provides an opportunity to practice implementation of RCP trip criterion under both accident and nonaccident conditions.
(1828JDK/rtj )
I e
2.
ao b.
C.
d.
e.
f.
Identify those procedures which include RCP trip-related operations:
/
RCP trip using WOG alternate criteria RCP restart Decay heat removal by natural circulation Primary System void Removal Use of steam generators with and without RCPs operating RCP'trip for other reasons
~Res onse:
SHNPP EOPs use RCP trip and..restart criteria consistent with the WOG Emergency
Response
Guidelines (Items C.2a-e).
RCP trip criteria during nonaccident conditions is covered in AOP-018, "Abnormal RCP Operation."
Normal operation of RCPs is covered in OP-I00, "Reactor Coolant System."
See Attachment C.2 for a listing of SHNPP EOPs.
(1828JDK/rtj )
Page I of 2 Attachment C.2 EOP Title Cross Reference CPSL WOG GUIDELINES TITLE EPP-1 F.PP" 2 EPP-3 EPP-4
'PP-5 EPP-6 EPP-7 EPP-8 EPP-9 EPP-10 EPP-11 EPP"12 EPP-13 EPP-14 EPP-15 EPP-16 EPP-17 EPP-18 EPP-19 EPP-20 EPP-21 EPP"22 ECA-O.O FCA-O.l ECA-0.2 ES-0.1 ES-0.2 ES-0.3 ES-0.4 ES-1. 1 ES-1.2 ES-1.3 ES-1.4 ECA-1.1 ECA-1.2 E-2 ECA-2.1 N/A ES-3.1 ES-3.2 ES-3.3 ECA-3.1 ECA-3.2 ECA-3.3 Loss of AC Power to 1A-SA and 1B-SB Busses Loss of Al] AC Power Recovery without SI Required Loss of All AC Power Recovery with SI Required Reactor Trip Response Natural Circulation Cooldown Natural Circulation Cooldown with Steam Void in Vessel (With RVLIS)
Natural Circulation Cooldown with Steam Void in Vessel (Without RVLIS)
SI Termination Post-LOCA Cooldown and Depressurization Transfer to Cold Leg Recirculation Transfer to Hot Leg Recirculation Loss of Emergency Coolant Recirculat'ion LOCA Outside'ontainment Faulted Steam Generator Isolation Uncontrolled Depressurization of All Steam Generators SGTR Isolation Post-SGTR Cooldown Using Backfill Post-SGTR Cooldown Using Blowdown Post-SGTR Cooldown Using Steam Dump SGTR With Loss of Reactor Coolant:
Subcooled Recovery SGTR With Loss of Reactor Coolant:
Saturated Recovery SGTR Without Pressurizer Pressure Control FRP-S.1 FRP-S.2 FRP"C.l FRP-C.2 FRP"C.3 FRP"H.1 FRP-H.2 FRP"H.3 FRP-H.4 FRP"H.5 FRP"P.1 FRP-P.2 FRP-J. 1 FRP-J.2 FRP-J.3 FRP-I.1 FR-S.1 FR-S.2 FR-C.1 FR-C.2 FR-C.3 FR-H.1 FR-H.2 FR-H.3 FR-H.4 FR-H.5 FR-P.1 FR-P.2 FR-Z.1 FR-Z.2 FR-Z.3 FR-I.1
Response
to
Response
to
Response
to
Response
to
Response
to
Response
to
Response
to
Response
to
Response
to
Response
to
Response
to Conditions
Response
to Conditions
Response
to
Response
to
Response
to
Response
to Nuclear Power Generation/ATh'S Loss of Core Shutdown Inadequate Core Cooling Degraded Core Cooling Saturated Core Cooling Loss of Secondary Heat Sink Steam Generator Overpressure Steam Generator High Level Loss of Normal Steam Release Capability Steam Generator Low Level Imminent Pressurized Thermal Shock Anticipated Pressurized Thermal Shock High Containment Pressure Containment Flooding High Containment Radiation Level High Pressurizer Level IBMD-WKLB02-OS2
~
Attachment C.2 EOP Title Cross Reference Page 2 of 2 CP&L MOG GUIDELINES TITIE FRP-I.2 FRP-I.3 FR-I. 2 FR"I.3 Response 'to'Low Pressurizer Level
Response
to Voids In Reactor Vessel PATH-1 PATH-2 E<<,O & E-1 E-3 Reactor Trip or Safety Injection/Loss of Reactor or Secondary Coolant Steam Generator Tube Rupture NOTE:
~ '0 This cross reference lists may change depending on future HOG ERG revisions and plant specific needs.
IBMD-MKLB02-OS2