ML17335A149
| ML17335A149 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/18/1998 |
| From: | Feliciano A INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17335A147 | List: |
| References | |
| ENSM970128AF, ENSM970128AF-R02, ENSM970128AF-R2, NUDOCS 9808100093 | |
| Download: ML17335A149 (169) | |
Text
ATTACHMENT 1 TO AEP:NRC:1260G7 COMMITMENT ASSOCIATED WITH RESPONSE TO REQUEST FOR INFORMATION RELATED TO 2.206 PETITION
Attachment 1 to AEP:NRC:1260G7 Page 1
The following is our specific commitment associated with this response to the request for additional information (RAI) regarding the 2.206 petition.
No other statements should be considered to be regulatory commitments.
1.
Identified UFSAR discrepancies that meet the condition report threshold, including those of the twenty-one systems covered under the restart plan system readiness
- reviews, will be dispositioned in accordance with the restart plan.
These UFSAR discrepancies will be dispositioned by correcting the non-conformance, performing a
evaluation, performing an operability evaluation in accordance with generic letter 91-18, revision 1,
or requesting a license amendment.
ATTACHMENT 2 TO ABP:NRC:1260G7
RESPONSE
TO REQUEST FOR INFORMATION RELATED TO THB 2.206 PETITION IDENTIFIED ON ENCLOSURE 1 OF THB JUNE 8,
- 1998, LETTER
gi-i>g'p>
Ews' NUCLEAR ENGINEERING DEPARTMENT Calcu1ation Cover Sheet Cook Nuc1ear P1ant AF.,g 4v
@C SHEET I OF~
CALCULATION No. r=wsw
) o i Z INDIANA MICHIGAN POWER COMPANY SAFETY RELATED SYSTEM Wc c w YES ~
NO UNIT No.
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K CALCULATED BY; TITLE Ecc: z Pu No, ~nc ~,
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ATE FILE LOCATION VERIFIED BY:
APPROVED BY:
CULATION DESCRIPTION: ~ ~c c. ~c ~~~~
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~7 r 0<<'-4
~ 79/(oM)
METHOD OF VERIFICATION:
ALTERNATE CALCULATION
+2 ~ H 't 7/9'f (C'e 97-A233
-db'42 jf Lo u/q7 can-ask, REVIEW NO.
REASON FOR CHANGE REVISION Calculated B
Date Verified B
Date Approved By Date ro S. ~ y- ~ hg
<<I=
c-ca~
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Z'~~ ~ t
) t)&r 9808100093 9808i7 PDR ADOCK 050003i5 '
PDR i
E N-'@5-,'c.:$i'p i DONALD C.
COOK NUCLEAR PLANT VERIFICATION CHECKLIST CALCULATIONS Calculation Number Rev. +
Si ture of rifier 1.0 Mere the inputs/iietse:,"eocuccee correctly selected, and documen'teed into th'e calculation' Basis:
Date incorporated
2.0 Basis
Are assumptions necessary to perform the calculation adequately described and reasonable?
Yes N/A a~
s 44 0 '6' Basis:
Are the applicable
- codes, standards and "regulatory requirements identified and requirements for design met?
Yes N/A ~
O EldPCt
~ ~ ~/ur (c~ la g.p Basis:
Was an appropriate design method used?
Lh&
'c Ei 4M r
Yes N/A
5.0 Basis
Is the output reasonable compared to input?
Yes + N/A u
C,
6.0 Basis
Are the results numerically correct?
Ct V a
Q, jd6td U
Yes
'N/A
'I g
~
I I
. (7 0
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page 2.
OF 66 DATE 12/9/97 BY A.Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coo!in S stem SUBJECT Emer enc Core Coolinc System Pumos NPSH Available Table ofContents Design Inputs.
References..
Purpose Methodology.
Assumptions.
Figure Results Analysis..
Recirculation Sump Configuration.
Recirculation Sump Head Loss..
RHR Pump Operating Point.
Pipe Segment Head Loss....................~.......................................
RHR Pump's Suction Pressure South SI Pump NPSH Available................................
West Centrifugal Charging Pump NPSH Available.
East Centrifugal Charging Pump NPSH Available..
RHR Pump NPSH Available Containment Spray Pump NPSH Available.......
Conclusions.
Pump Curves..
Graphs..
HFLC5 Input/Output.
Pipe Segment Data Sheets.
Attachment to Calculation ENSM971028AF 3
..4 5
5
.7
.8
.9
.9 10 11
.11
.12-~3 23-'74 75
'75 26 26
.27 28 28 29-32 33-35 36-46
..47-66 Attachment pgs 1-9
I t
I 0
J
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Paging~OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coolin S stem SUBJECT Emer encv Core Coolina S stem Pumos NPSH Available Design Inputs 1
Required Net Positive Suction Head (NPSHR) for the safety injection, charging, residual heat removal, and containment spray pumps is obtained from the pump curves (Ref 3), attached.
2 The flow path for this calculation is in accordance with OHP4023.ES-1.3 "Transfer To Cold Leg Recirculation."
(Ref 2) 3 - Piping configuration (length, diameter, fittings, e1evations, etc.)
from the Recirculation sump through the residual heat removal (RHR) pump through the RHR supoly to the safety inject'on and centrifugal charging pumps suction are obtained from the isometric or physical drawings (Ref 4),
as shown on attached "Pipe F iction Data Sheets".
4 RHR heat exchanger pressure drop of 15 psi at 2960 cpm from the heat exchanger' specification data sheet.
5 Safety Injection pumps flow balanced at 700 gpm pe" Technical Specification 4.5.2 h.
The flow balance is performed to meet flow conditions in accordance with **12 EHP 4030 STP.208SI "Ul and U2 ECCS Flow Balance Safety Injection System" step 4.11 page 8 of 44.
6 Centrifugal charging pumps are flow balanced at 550 gpm per Technical Specification 4.5.2 h.
The flow balance is performed to meet flow conditions in accordance with **12 EHP 4030 STP.208BI "Ul and U2 EMERGENCY CORE COOLING SYSTEM FLOW BALANCE BORON INJECTION SYSTEM" step 5 '.20 page 48 of 58.
7 Containment spray pump flow 3200 gpm (2000 gpm uppe and 1200 gpm lower spray flows) per DB-12-CTSr pg.
47 section 4.1.1
~ 1 and pg.
57 section 4.1.7.1.
I 8
Fluid vapor pressure of 9.34 psia at 190'f, temperature of recirculation sump fluid during the recirculation phase from UFSAR table 6.1-1 pg.
6.1-12 for U2 which bounds Ul temperature of 160'f.
9 Recirculation sump level 602' 10",
DB-12-CTS, pg.
34 section 3.9.3.3 10- Recirculation Sump Screen dimensions from calculation ENSM971128TWF (Ref 12) approved 12/8/97 for current configuration.
The current configuration is being revamped to conform to the design and installation performe'd by RFC-2361 in 1979. Calculation ENSM971210TWF (Ref 12) approved 12/11/97 determine a screen open area for the revamped (RFC-2361) installation.
11-50% design basis blockage based on ALDEN Labs modeling (Ref 13).
l t
~
6'It/IERICAN ELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~OP 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 182 System Emer en Care Coolin S stem SUBJECT Emeraenc Core Coolina S stem Pum=s NPSH Available Re ferences 1-NRC Information Notice 96-55 "Inadequate Net Positive Suction Head of Emergency Core Cooling and Containmen" Heat Removal Pumps Under Design Basis Accident Conditions" 2-OHP4023.ES-1.3 Transfer to Cold Leg Recirculation 3-Safety Injection (SI)
Pump Pacif' Pump curve 39890A Residual Heat Removal (RHR)
Pump - Inge soll-Rand Pump curve N-315 Centrifugal Charging (CC)
Pump Pacif' Pump curve 34617-L Containment Spray CTS)
Pump-Byron-Jackson curve T-32852-7 4-Isometric/Physical Drawings
'-2-5338-7 2-SI-9 Rev 19 2-5415-15 2-SI-7 sh 1 Rev 2, sh 2 Rev 1
RH-14 sh 1 Rev 2, sh2 Rev 1 2-RH-18 Rev 3 2-RH-22 Rev 10 2-SI-10 sh 1 Rev 11, Sh 2 Rev 4
2-SI-44 Rev 6
2-CS-79 sh 1 Rev 9, Sh 2 Rev 3 2'-RH-23 Rev 12 2-CS-80 Rev 6
2-CS-81 Rev 5
5-Friction Losses in pipe fittings from Cameron Hydraulic Data book 18'" Ed pg.
3-111 through 3-117 6-Pipe flow velocity and friction losses from Cameron Hydraulic Data book 18'" Ed pg.
3-12 through 3-33 7-Related Calculations:
NESM961021AF approved 12/2/96 HXP840301JN approved 12/14/85 8-U2 FSAR Appendix Q question 212.29-4 amendment 78 10/77 attachmen-
"A" NPSH calculation 9-Flow Diagrams 2-5143-39 2-5142-37 2-5129-34 10-Hydraulic friction loss calculation, program revision 5 1988 (HFLC5) will be used to determine the frictional losses through the flow path.
HFLCS is an in-house developed
- program, which was approved for use on Feb.
28, 1988.
This program was validated and approved in accordance with the requirements of GP 2.6 Software Quality Assurance Standard 'n use in 1988.
11-Crane Technical Paper No.
- 410, "Flow of Fluids Through Valves,
- Fittings, And Pipe" 12'" printing 1972 12-Calculation ENSM971128TWF titled "Flow Area of recirculation Sump Screen",
approved 12/8/97 and ENSM971210TWF approved 12/11/97.
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page
~
OF
. 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nudear Plant Unit 1&2 System Eme en Core coolin S stem SUBJECT Emeraenc Core Coolina S stem Pu'-..
s NPSH Available 13-ALDEN recirculation sump studies dated 9/78 14-I.E. Idelchik, Handbook Of hydraulic Resistance, 2nd edition 1986
~Pur oee NRC Information Notice 96-55 addresses inadequate NPSH of emergency core cooling and containment heat remova'umps under design basis accident conditions.
The information..otice addresses this condit'n under the ECCS Recirculation mode of ope"ation.
This calculation will determine the NPS:: available to the SI and CC pumps during the ECCS Recirculation mode when one RHR pump is used
=o supply their flow requirements.
This calculation will also check the CTS and RHR pumps NPSH available.
Nese parameters were originally determined in response to U2 FSAR Appendix Q
Question 212.29-4.
However, currently the RHR system is aligned with the RHR crosstie valves closed due to potential deadheading concerns.
This calculation will check the CTS and RHR =umps NPSH 'available unde the flow conditions, used to determine the SI and CC pumps NPSH availabl Revision 2 will determine the pressure drop across the recirculatic.".
sump screen and if it impacts the RHR p ap' available NPSH.
The flow path used in this calculation '
shown on figure 1.
Method In order to obtain the frictional losses, associated with the flow
- path, the isometric and physical drawincs were used to determine th piping configuration.
Figure 1 shows t.".e flow path and branching flows to the CTS pump and RHR cold leg injection.
The data obtained from the drawings was compiled on the a"tached pipe friction data sheets The totals shown on the data sheets are used as input to HFLC5.
HFLC5 calculates the segments frictiona'osses and is based on the Darcy-Weisbach formula obtained from Cameron page 3-110:
Ht = f L v D 2g where:
Hq frictional loss, ft f
friction factor, dimensionless L pipe length, ft D pipe diameter, ft v pipe velocity, ft/sec g
gravitationa
- constant, 32.2 ft/sec
- ~
C I
r<
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AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page Cn OF 66 DATE 12/9/97 BY A. Feliciano CK.
~
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core coolin Svstem SUB'JECT Emeraenc Core Coolina Svstem Pumps NPSH Available The HFLC5 results are shown on the attached output sheets.
This information is then used to determine the operating point of the RHR pump.
The operating point occurs at the intersection of the pum" performance (head-capacity) curve with the frictional loss (system-head) curve.
The resultant flow is then used to determine the =low distribution and resultant pipe friction.
The resultant pipe friction was determined from the following fowiula obtainec from Cameron page 3-110:
He where:
Ht K
v g
K v 2g head loss, ft resistance coefficient velocity, ft/sec gravitational constant 32.2 ft/sec-Based on the preceding the NPSH available to the SE and CC pumps can be determined.
The NPSH available is determined.from the following formula obtained from Cameron pg.
1-10:
= ha hvpa
+ hsr h.s where:
NPSH net positive suction head, ft abs h,
absolute pressure, ft hp, fluid vapor pressure, ft h
static elevation difference, ft hr, pipe friction losses, ft However, before the SI and CC pumps NPSH available can be determined, it is f'st necessary to determine the RHR pump' suction pressu"e.
The suction pressure can be determined from Bernoulli' equation obtained from Crane Technical paper 410 pg. 1-5.
Bernoulli's equation is written as:
Zl
+ 144P1
+
(vl)
= Z2
+
144P2
+
(v2)
+
h Px 2g p2 2g where:
Zl, Z2 Pl, P2 vl, v2 px I
p2 g
hr elevation, ft
- pressure, psig velocity, ft/sec density of fluid, lbs/ft k
gravitational constant 32.2 ft/sec2 head loss, ft
C I
I
RQERICAN ELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page 7
OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Emer en Core Coolin S stem SUBJECT Emeraencv Core Coo'ina Svstem Pumos NPSH Available Assum tions 1
Containment pressure a" 14.7 psia.
To be conservative this calculation will not include the containment design pressure of 12 psig per FSAR chapter 5 section 5.2.2.2 Design Load Criteria pg. 5.2-16.
2 Single active failure criteria is failure of one RHR pump.
No pump degradation is assumed since bounding highest flow results from this approach.
a) 2 CTS pumps operating,
- however, only one is supplied from the same source that supplies the operable RHR pump since each CTS pump and RHR tra's are supplied individually.
b) 2 CC pumps total low of 840 gpm or 420 gpm each.
This flow condition is obtained from the intersection of the two parallel pump head-capacity curve with the system-head curve at approximately 840 gpm or 1.5 times the 550 gpm flow requiremen" (see attached curve 34617-L).
Note:
1.5 factor is obtained from 840/550.
c) 2 SI pumps total low of 920 gpm or 460 gpm each.
This factor is obtained from the intersection of the two parallel pump head-capacity curve with the system-head curve at approximately 920 gpm or 1.314 times the 700 gpm requirement (see attached curve 39890A).
Note:
1.314 factor is obtained from 920/700.
For conservatism the 1.5 factor determined for the charging pumps will be used since it yields a higher flow requirement of 1050 apm (700 x 1.5) or 525 gpm eacn.
3-Head loss through the Recirculation sump and sump's mesh screen 's less than 1 ft as determined in Amendment 78 Appendix Q.
For purposes of this calculation the pressure drop will be determined based on the open area and 50% blockage for the maximum potential flow.
The flow for 2b and 2c is based on parallel pump operation.
One of the first steps is to draw the system-head curve.
The system-head curve consists of the sum of the static head, pipe-friction head, and head losses in valves and fittings. The parallel head-capacity curve is drawn by adding the capacities at the same heads.
The head-capacity curves of the single and parallel pumps are plotted on the same drawing and their intersections with the system-head curve represent the operating points.
The system-head curve plotted on the pump performance curves was determined from the known Technical" Specification requirements.
The system-head curves were developed by multiplying the resistance factor ft/gpm times the square of the flow.
Minimum and maximum resistance factors are given in the Technical Specification for the CC pumps
l f
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~OP 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 18,2 System Emer en Core Coolin S stem SUBJECT Emer encv Core Coolino S stem Pumos NPSH Available (Design Input 6).
The calculation used the minimum factor to obtain the maximum CC pump flow.
For the SI pumps, the Technical Specification stipulates a flow requirement (Design Input 5) rather than a resistance factor.
The SI resistance factor ft/gpm-(1440/700-
)
is obtained at the 700 gpm head-capacity point.
The resistance factor is comprised of the sum of the static head, pipe-friction head, and head losses in valves and fittings. It is acceptable to use this
- factor, based on the Technical Specification requirements, since it represents the head-capacity operating point of the pump in the system.
The intersection of the system-head and head-capacity curves provides the total flow delivered to the system.
The resultant total flow is generally less than 1.5 times the single pump design flow supplying the same flow path, as stated in assumption 2 above.
ILtgh got'P t 6P9 Qgo Di GPGPSQ 8
Qi) 5 A
Figure 1
ECCS Recirculation Flow Path One RHR Pump Out Of Service
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~OP 66 DATE 12/9/97 BY A. Feliciano CK.
PlANT Cook Nuclear Plant Unit 182 System Erne en Core Coolin S stem SUBJECT Emeraenc Core Coolina S stem Pumos NPSH Available Results The results of this analysis indicate that the S1 pump' NPSH available is 132 ft abs for the south pump and 122 ft abs for the north pump at the assumed single pump flow of 525 gpm.
The NPSH required at this flow is 13 ft abs.
That is, the NPSH available exceeds the NPSH required by 109 to 119 ft abs.
The CC pump' NPSH available is 48 ft abs for the east and 49 ft abs for the west, pump at the assumed single pump flow of 420 gpm.
The NPSH required at this flow is 17 ft abs.
That is, the NPSH available exceeds the NPSH required by 31 to 32 ft abs.
At the RHR pump flow of 4600 gpm, determined from the graphical
- analysis, the NPSH available was determined to be 29 ft abs.
The NPSH required at this flow is 20 ft abs.
That is, the NPSH available exceeds the NPSH required by 9 ft abs.
A similar check of the CTS pump's NPSH available determined that at 3200 gpm the NPSH available is 31 ft abs.
The NPSH required at 3200 gpm is 9 ft abs.
That is, the NPSH available exceeds the NPSH required by 22 ft abs.
The determination of the pressure drop across screens indicates that the assumed revision 0
acceptable.
That is, a pressure drop of less and the revision 1
NPSH available results are pressure drop across the screen.
the recirculation sump pressure drop was than 1 ft was determ'ned not impacted by the The calculated recirculation sump head loss is based on the empirical results obtained during the ALDEN sump testing.
The ALDEN test results obtained a pressure drop of.77 ft for 50% blockage.
This compares well with the analytical value of.82 ft determined for 50'lockage.
Based on the results of this calculation adequate NPSH is available to assure that the ECCS pumps are capable of performing their safety function under the Recirculation mode of operation.
A~nal sis Determine the recirculation sump's increased pressure drop due to the addition of a second grating at the maximum expected flow.
The maximum expected flow is based on 2
RHR pumps operating at 4600 gpm each and two CTS pumps operating at 3200 gpm each.
This results in a total flow of 15, 600 gpm through the recirculation sump.
Per the Alden report (Ref 13) the sump test configuration consisted of a single coarse grating and a single fine mesh screen.
The existincr sump configuration is somewhat different in that a second coarse grating is installed after the fine mesh screen.
The Alden report details that the sump head loss (hr,) can be obtained from the loss
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~WOF 66 DATE 12/9/97 BY A. Feiiciano CK.
PLANT Cook Nuclear Plant Unit 18,2 System Emer en Core Coolino S stern SUBJECT Emeraencv Core Coolina Svstem Pum=s NPSH Available coefficient Ct,.
That is, the Alden report pg 26 section "d" indicates C as consisting of the total losses including the screen,
- grating, and entrance los es at the outlet pipes (see below figure).
The report also indicates that the loss of head across the gratinc and screen was evaluated a..d found to be about lle5 times the approach velocity head just upstre~~
of the grating.
This information will be used to determine the existing sump' pressure drop.
6 ea ao 6-lrrr r I
ggg -6
/
r Recirculation Sump Configuration The loss coefficients were determined for various tests
- schemes, an' are represented by the equation:
r C.= h> /(v /2g),
where Ce ht,v loss coefficient, dimensions sump head loss, ft fluid velocity, ft/sec gravitational constant, 32.2 ft/sec2 From the report table 10, the highest head loss occurs in test number 3 at a loss coefficient and fluid velocity of
~ 26 and 13.79 ft/sec, respectively.
Therefore, ht, is determined as follows:
Ct.
= ht. /(v /2g)
.26 = ht. /(13.79 /64.4) the sump's head loss.
solving for ht, yields
.77 ft as The approach velocity is determined using the height of fluid at the
I l
S
>>tp
, A
/JP I
6 AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Peg. ~OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Emer en Core Coolin S stern SUBJECT Emer enc Core Coolina Svstem Pum=s NPSH Available sump entrance elevation of 602 ft 10 in. minus the curb height elevation of 599 ft 4 in.
This results 'n a fluid height of 3.5 ft or 42 in.
From reference 12, the grating consists of eight 25.125 in.
long sections and one 20 in. long section.
The submerged area is determined as follows:
Submerged
- area, A, =
(42 x 25. 125 x 8)
+
(20 x 42)
=
9282 in-or 64.46 ft.
The approach velocity is determined by d'viding the flow by the submerged area as represented by velocity =
Q/A =
15600<5>m x
i
.54 ft/se" 64.46ft (7.48gal/ -'
60sec/min)
Therefore, the head across the g ating and screen is determined frc.=
the Alden relation as follows:
(.54 /64.4)
=.052 ho,
=
11.5 x approach velocity head
=
(11.5)
This represents the head loss across the test I
configuration's grating and screen.
Since the existing sump configuration includes a
secon"'oarse grating it is conservative to augment the test configuration' head loss of.77 ft by the calculated grating and screen head loss.
This results in a sump head loss of.822 ft for the existing configuration.
Based on the above, it will not be necessary to evaluate the impact on the RHR pump' available NPSH since the recirculation sump sc een head loss was determined to be less-than 1 which is consistent with assumption 3.
Determine RHR um o eratin oint In order to determine the RHR pump operating point it is necessary
=o determine the total system-head.
The total system-head is comprisec of pipe segments 1-2 through 25-26.
The individual segment resistance is obtained from the HFLCS outputs.
It is necessary tha-flow values not shown on the output be determined.
The flow values can be determine by the following relationship:
H2
=
Hy (Q2)
(Ql) where:
Hz HL Q2 Ql unknown head loss at known flow, ft head loss at known flow, ft known flow at unknown head loss, gpm known flow at known head loss, gpm
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~g'F 66 DATE 12/9/97 BY A. Feliciano CK.
PIANT Cook Nuclear Plant Unit 182 System Eme en Core Coolin S stem SUBJECT Emeraencv Core Coolina System Pumos NPSH Availa'"le For example, segment 1-2 HFLC5 output shows this as 3 "" at 8650 g"m.
Using this information a head loss can be determined at another
= ow, say at 9000 or 7000 gpm.
Hgppp
=
3 (9000)
'8650) 3.25 ft H1ooo
=
3 (7000)
=
1.96 ft (8650)
This method is used to generate the various pipe segment'ystem-head curves shown and labeled as sheet 1 through 3.
Xt should be note that the addition of these curves is based on the pip'ng arrange,.e.".t.'hat is, pipes in parallel are added at the same head values whi'e pipes in series are added at the same flow value.
The following describes the process of adding the various segments and obtaining the total system-head curve shown as curve c sheet 3
~
The process begins by starting at the east CCP or the last flow distribution point (see fig 1).
The summation of the curves star=s on sheet 1 and ends on sheet 3.
Sheet 1 Curves curve 1
curve 2
curve 3
curve 4
curve 5
curve 6
curve 7
segments 22-25
+
25-26 are added in series segments 22-23
+
23-24 are added in series curve 1
+ curve 2 in parallel segments 16-19
+ 19-20
+ 20-21
+ 21-22 in series curve 3
+
4 in series segments 16-17
+ 17-18 in series curve 5
+
6 in parallel Sheet 2
curve 1
curve 2
curve 3
curve 4
curve 5
segments 11-14
+ 14-15
+ 15-16 in series curve 7 from sheet 1
curve 1
+
2 in series
- segments 11-12
+ 12-13 in series curve 3
+
4 in parallel Sheet 3
curve A segments 1-2
+ 2-3
+ 3-4
+ 4-5
+ 6-7
+ 7-8
+ 8-9
+ 9-10
+
10-11 in series curve B
curve 5 from sheet 2
curve C
curve A +
B in series represents'otal system-head curve D
RHR pump performance curve Note: at the intersection of curves C and D the RHR operating flow of 4600 gpm is obtained.
This flow is then used to determine the SX and
l
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~OF 66 DATE 1219197 BY A.Feliciano CK.
PLANT Cook Nuclear Plant Unit 182 System Emer en Core Coolin S stem SUBJECT Emeraenc Core Coolinc S stem Pumos NPSH Available charging pumps NPSH available.
Based on an RHR flow of 4600 gpm the total required flow from the Recirculation sump could be determined.
The total flow 's comprised of one CTS pump at 3200 gpm
+ one RHR pump at 4600 gpm o" 7800 gpm total.
Of the total, 3200 gpm flows to containment sp" y, 1050 gpm to SI, 840 gpm to CC, and the remainder (2710 gpm) to RHR cold leg injection.
The cold leg injection flow is simulated as caving the system at node 10.
The following provides the flow dist"ibution based on the preceding and figure 1:
Segment 1-2I 2-3 3-4 through 9-10-11 11-12, 12-13 11-14 through 16-17, 17-18 16-19 through 22-23, 23-24 22-25, 25-26 10 15-16 21-22 Flow 7, 800 4, 600 1, 890 525 1, 365 525 840 420 420 gpm gpm gpm gpm gpm gpm gpm gpm gpm Determine the head loss fo" each segment:
Determine the head loss per the above listed flows for the respective segments.
The head loss for each segment can be determined from the relationship shown below and detailed on page 5:
Hf
= K v-2g Segment 1-2 configuration obtained from pipe friction calculation data sheet flow 7800 gpm, diameter 18", pipe length
- 26. 66 ft,
- 1. cate valve, 1
reducer
- 18x24, 1 entrance Note: the velocity and head loss/100ft can be obtained f om Cameron pg.
3-12 through 3-33 as follows:
flow 8000 7800 7000 velocity 11.5 v
10.0 h/100 ft 1.94 h
1.49 By interpolation the velocity (v) and head loss/100'h) can be determined as follows:
8000-.7800
=
11.5-v 8000-7000 11.5-10.0 1.94-h 1 '4-1.49
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING a
Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~tOP 66 DATE 12/9/97 BY A. Feliciano CK.
PlANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coo!in S stem SUBJECT Emeraencv Core Cool'nc S stem Pumos NPSH Available v =
11.5
1.5(200/1000)
11.2ft/sec h
1.94
.45(200/1000)
= 1.85 ft/100 From Cameron pg.
3-111 through 3-118 K values can be obtained for the various fittings as follows:
18" gate valve K=.1, reducer K =.5(1-(dl/d2) ), increaser K = (1-(d1/d2)
)
pipe entrance K =
1
- h. =
L (h/100')
=
26 66 (1 85/100)
=.49 ft heace
=
K (V
) /2g
=.1 (11.2") /64
~ 4
=.19 ft hogpe en' 1 (11.2 )/64.4
= 1.95 ft K-eu =.5(1-(18/24)
)
=.219 ft hoeu
=.219 (11.2")/64.4
=.43 ft
~
~
2
~
h, p =
1 ft (see pg.
)
hga
= hr,
+
hgare
+ hotpe eno
+
h=egt
+
hauop 4 06 ft Segment 2-3 configuration obtained from pipe friction calculation data sheet flow 7800 gpm, diameter 18",
pipe length 26'5 ft, 2 90'ong radius elbows, 1 tee branch 90'R elbow K =.19, tee branch K =.72 hL =
L (h/100')
=
26.15 (1.85/100)
.48 hegbo
K (V ) /2g = 2(.19)
(11.2
) /64.4
=.74 ft htgranch e 72
( 1 1"e 2
) /64
~ 4 1
e 4 ft h2 a ht,
+
hegtgow
+
htgranoh
= 2 62 ft Segment 3-4 configuration obtained from pipe friction calculation data sheet flow 4, 600 gpm, diameter 14",
pipe length 42.93 ft, 6 90'ong radius elbows, 1 tee run, 1 reducer
- 14x18, 1 gate valve 90'R elbow K =.21, tee run K =.26, gate valve K =.1, reducer K =.198
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page
/ ~
OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coolin S stem SUBJECT Emer enc Core Coolina Svstem Pu.-..os NPSH Available Flow 5000 4600 4500 h/100 2.79 h
2.27 v
- 11. 86
~
v
- 10. 67 h/100' 2.37/100'
10.9 ft/sec ht,
L (h/100')
=
42.93 (2.37/100)
= 1.02 ft hegbow
=
K (v
) /2g = 6(
~ 21)
(10.9
) /64.4
= 2.32 ft hgun 26 (10
~ 9
) /64 4
i 48 ft hgeee
=
K (v
) /2g = el (10o9
) /64 o4 184 ft ha 4
= ht,
+
helbow
+
h un
+
hgeee
=
4 0 ft Segment 4-5 configuration obtained from pape sheet flow 4,600 gpm, diameter 14",
pipe length friction calculation da-a
- 3. 33 ft h4 s =
L (h/100')
=
3.33 (2.37/100)
=.079 ft Segment 6-7 configuration obtained from pipe friction calculation cata sheet flow 4,600 gpm, diameter 8",
pipe lencth 59.85 ft, 8 90'ong radius elbows, 1 gate valve,1 check valve 90'R elbow K =.22, gate valve K =.1',
check valve K = 1.4 Flow 5000 4600 4500 h/100 v
35.6 32.1 h
v 28.9 28.9 h/100' 30.24/100'
29.54 ft/sec ht,
L (h/100')
~
59.85 (30.24/100)
= 18.1 ft hegbow
= K (v ) /2g
=
8 (
~ 22)
(29. 54
) /64. 4
= 23. 85 ft hoheok~
1 4
(29 54
) /64 e 4 18 97 ft hg,ee
=
K (v-) /2g =.11 (29.54
) /64.4
=
1.49 ft II hs 7 = ht,
+
hezbow
+
honeok
+
hgege
= 62, 4 1 ft
'a AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page t>
OF 88 DATE 12/9I9?
BY A.Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coolin S stem SUBJECT Emeraencv Core Coolina Svstem Pur.".os NPSH Available Segment 7-8 configuration obtained from pipe friction calculation da=a sheet flow 4, 600 gpm, diameter 8",
pipe length 10 ft, 1 90'ong radi~ s
- elbow, 1 gate valve,1 tee branch 90'R elbow K =.22, gate valve K =.11, tee branch K =.84 ht. =
L (h/100')
=
10 (30.24/100)
3.02 ft hetbo.
K (V
) /2g = (.22)
(29.54
) /64.4
= 2.98 ft h~ee 84 (29
~ 54
) /64 o 4 1 1 38 ft hga-e
=
K (v
) /2g =.11 (29.54
) /64,4
=
1,49 ft h1 g = ht,
+
helbo~
+
h<<e
+
haa-e
= 18.87 ft Segment 8-9 configuration obtained from pipe friction ca cu1ation data sheet flow 4, 600 gpm, diameter 14",
pipe length 0 ft, 1 90'ong rad'
- elbow, 1 red 8x14,1 inc.
- 8x14, hx delta P = 15 psi 8 2960 g"m 15 psi x 2.386 (Cameron pg.
4-4 at 190'f)= 35.79't v = 10.91 ft/sec 90'R elbow K =
~ 21, red K =.337, inc K =.454 heybo~
= K (V
) /2g = (.21)
(10.9
) /64.4
.39 ft heart
~ 337 (10
~ 9
) /64. 4 =
. 622 ft hto = K (v ) /2g
=.454 (10
~ 9
) /64.4
=
.838 ft hh>, = 35.79 (4600/2960)
= 86.44 ft ha-9
=
heybow
+
hred
+ htno
+
hhz = 88
~ 29 ft Segment 9-10 configuration obtained from pipe friction calculation data sheet flow 4,600 gpm, diameter 8",
pipe length 3 ft h9 to =
L (h/100
)
=
3 (30 24/100)
=
~ 91 ft
'AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~7OF 66 DATE 12/9/97 BY A.Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coolin Svstem SUBJECT Emeraenc Core Coolina Svstem Pumos NPSH Available Segment 10-11 configuration obtained from pipe friction calculat'n data sheet flow 1890
- gpm, diameter 8",
pipe length 68.695 ft, 8 90'ong radius elbows, 1 gate valve,4 45'lbows, 1 tee branch 90'R elbow K =.22, gate valve K =.11, 45'lbow K =.22, tee branch K =.84 Flow h/100 2000 5.91 1890 h
1800 4.81 h/100' 5.31/100' 12.8 V
11.5 v = 12.09 ft/sec ht, =
L (h/100')
=
68 '95 (5.31/100)
=
3'5 ft hgo=
K (v
) /2g
=
8 (.22)
(12.09
) /64.4
3.99 ft h4s
.22 (12.09
) /64.4
=.49 ft hggg e
K (v ) /2g 11 (12 09
) /64 4
25 ft hbrarcb
= K (V ) /2g e84 (12e09 )/64 4
1e91 ft hzo 6 ~
= ht.
+
hgo
+
h4s
+ hg,-,
+ hqn,b = 10.3 ft Segment 11-12 configuration obtained from pipe friction calculat'on data sheet flow 525 gpm, diameter 6",
pipe length 5.156 ft, 1 90'ong "ad'us
- elbow, 1 red 6x8,1 tee branch 90'R elbow K =.24, red K =.219, tee branch K =.9 Flow 550 525 500 h/100 1.97 h
1
~ 64 V
- 6. 11 V
5.55 h/100' 1.81/100'
5.83 ft/sec ht,
L (h/100')
=
5.156 (1.87/100)
=.093 ft K(V)/2g(
~ 24)(5
~ 83) /64e413 ft hbn,w = K (v )/2g =.9 (5.83 )/64.4
=.48 ft hggagt K (V ) /2g e 219 (5
~ 83
) /64
~ 4
~ 1 16 ft
Epl I
'AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page IF OF 66 DATE 12/9/97 BY A. Feliciano CK.
'LANT Cook Nuclear Plant Unit 182 System Eme en Core Coolin Svstem SUBJECT Emeraenc Core Coolina Svstem Pumos NPSH Available h1 1-12 hL
+
hegbcw
+
hb<<ncb
+ h ee =
. 82 ft Segment 12-13 configuration obtained from pipe friction calcula='on data sheet flow 525 gpm, diameter 4",pipe length 1.167 ft,1 red 4x6 K =.278 Flow 550 525 500 h/100 15.8 13.1 v
13.9 v
12.6 h/100' 14.4/100'
13.25 ft/sec h
L (h/1'00')
=
1.167 (14.45/100)
.169 ft h<<d
K (v ) /2g
=
. 278 (13. 25-) /64. 4 =
. 758 ft h)P gz = hL
+h<<d
=
~ 93 ft Segment 11-14 configuration obtained from pipe friction calcula="on data sheet flow 1365 gpm, diameter 6",
pipe length 13.885 ft, 1 gate va ve, 1
red 6x8,1 tee branch gate valve K =.12, red K =.219, tee branch K =.9 Flow 1400 1365 1300 h/100 12 h
10.4 v
15.5 v
13.2 h/100' 11.44/100'
14.69 ft/sec hL
L (h/100')
=
13.885 (11.44/100)
= 1.59 ft K (v ) /2g
~ 12 (14 69
) /64+ 4 402 ft hb<<ncb
= K (v ) /2g =.9 (14.69
) /64.4
= 3.02 ft h<<d
= K (v ) /2g =.219, (14. 69
) /64.4
=.73 ft, hing 14
= hL
+
htee
+ hb<<nch
+ h,,d = 5. 74 ft Segment 14-15 configuration obtained from. pipe friction calculation data sheet flow 1365 gpm, diameter 8",pipe length 1.86 ft,1 90'R Elbow K =.22
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page ~IOF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Cooiin S stem SUBJECT Emeraenc Core Coolinc S stem Pumos NPSH Available Flow 1400 1365 1300 h/100 2.96 h
2.56 v
8.98 v
8.34 h/100' 2.82/100'
8.76 ft/sec ht,
L (h/100')
=
1.86 (2.82/100)
.052 ft helbow
K (v
) /2g
=
~ 22 (8
76
) /64
~ 4
=
262 ft h14-ls
= ht,
+
helbow 314 ft Segment 15-16 configuration obtained from pipe friction calculation data sheet flow 1365 gpm, diameter 6",
pipe length 18.313 ft, 1 gate val'e, 1
red 6x8r 1 tee run, 1 90'B. e~bow gate valve K =.12, red K =.219, tee run K =.3, 90bow K =.24 I
~~
~
~
~
~
I
~
~
I
~~
~
~
~
I 0
1 I
~
e I
\\
h/100' 11.44/100'
= 14.69 ft/sec
- h. =
L (h/100')
= 18.313 (11.44/100)
= 2.09 ft hgeze K (v ) /2g e 12 (14 e 69
) /64
~ 4 e 402 ft h.= K (v ) /2g
=.3 (14.69
) /64.4
=
1 ft h<<d = K (v )/2g =.219 (14.69 )/64.4
.73 ft helbow
K (v ) /2g =
~ 24 (14
~ 69
) /64
~ 4 =
. 8 ft his-16 hL
+
htee
+ hb<<n=b
+
hosea
+
helbow
=
5
~ 02 ft Segment 16-17 configuration obtained from pipe friction calculation data sheet flow 525 gpm, diameter 6",
pipe length 4.542 ft, 1 tee run K =.3 h/100' 1.92/100'
5.83 ft/sec ht,
L (h/100')
=
4 '42 (1.92/100)
.087 ft h,= K (v )/2g =.3 (5.83 )/64.4
=.158 ft a
h16 lg = ht. + h,,
=.245 ft Segment 17-18 configuration obtained from pipe friction calculation data sheet
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page 2 ~
OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coolin S stem SUBJECT Emeraencv Core Coolina S stem Pumos NPSH Available flow 525
- gpm, diameter 4",pipe length 1.167 ft,1 red 4xo K =.278 h/100' 14.4/100'
13.25 ft/sec ht.
L (h/100')
=
1.167 (14.45/100)
=.169 ft h<<d = K (v ) /2g 278 (13 25
) /64 4
758 ft h17-ie
= ht,
+h<<d =.93 ft Segment 16-19 conficuration obtained from pipe friction calculation data sheet flow 840 gpm, diameter 4",
pipe length 15.163 ft, 1
ed 4x6, 3
90'R
- elbows, 1 tee branch 90'lbow K =.27, red K =.278, tee branch K = 1.02 Flow 840 800 h/100 37 h
32.8 v
21.4 v
20.2 h/100' 36.16/100'
21.16 ft/sec ht.
L (h/100')
=
15.163 (36.16/100)
= 5.48 ft herbe~
= K (V ) /2g
=.27 (21. 16
) /64 '
= 1. 88 ft hb<<nch K (v ) /2g = 1. 02 (21. 16
) /64. 4
= 7. 1 ft K(v)/2g 278(21+16) /64+4193 ft h16-19
= ht.
+
helbow
+ hb<<nch
+
hzed
=
1 6. 39 ft
, Segment 19-20 configuration obtained from pipe friction calculation data sheet flow 840 gpm, diameter 4",
pipe length 5 ft, 1 gate valve, 2 90'R
- elbows, 2 tee branch 90'lbow K =.27, gate valve K =.14, tee branch K = 1.02 h/100' 36.16/100' v = 21 '6 ft/sec
~
~
~
~
~
ht, =
L (h/100'
=
5 (36. 16/100)
=
1
~ 81 ft heave
= K (v ) /2g
=.14 (21
~ 16
) /64.4
=.97 ft
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page
~ ~
OF 66 DATE 12/9/97 BY A.Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coofin S stem SUBJECT amer encv Co"e Coolina S stem Pumos NPSH Available hbrancb K (V ) /2g (2) 1 02 (21 o 16
) /64 4
14 o 18 ft helbow = K (V ) /2g = (2)
.27 (21.16
) /64.4
= 3.75 ft hie 2o = hL
+
helbow
+ hbrancb
+ hgare
= 20
~ 71 ft Segment 20-21 configuration obtained from pipe friction calculation data sheet flow 840
- gpm, diamete 4",
pipe length 46.125 ft, 1 gate valve, 8
90'R elbows 90'lbow K =.27, gate valve K =
~ 14 h/100' 36.16/100'
21.16 ft/sec hL
L (h/100')
=
46.125 (36 '6/100)
= 16'8 ft hgate
=
K (V ) /2g =
14 (21
~ 16
) /64
~ 4 =
. 97 ft helbow K (V ) /2g (8) e27 (21 16
) /64 4
15e02 ft h20-21 hL
+
helbow
+
hgare 32 67 ft
,Segment 21-22 configuration obtained from pipe friction calculation data sheet flow 840
- gpm, diameter 8",
pipe length 17.167 ft, 1 90'R elbow 1 tee branch 90'lbow K =
~ 22, tee branch K =.84 Flow 850 840 800 h/100 1 '4 h
F 01 v
5'5 v
5.13 h/100' 1.11/100'
5.39 ft/sec hL
L (h/100')
=
17 '67 (1
~ 11/100)
=
~ 19 ft helbow = K (V ) /2g
=
~ 22 (5.39
) /64
~ 4 =
~ 099 ft hbrancb
= K (v ) /2g
=.84 (5.39
) /64.4
=.38 ft h21-22 hL
+
helbow
+
hbranch
. 67 ft
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Caiculation Number ENSM970128AF Rev 2 Page 4>
OF 66 DATE 12/9/97 BY A. Feliciano CK.
PlANT Cook Nudear Plant Unit 1&2 System Emer en Core Coolin S stem SUBJECT Emeraencv Core Coolincr S stem Pt~-..os NPSH Available Segment 22-23 configuration obtained from pipe friction calculation data sheet flow 420 gpm, elbow 1 tee branch, diameter 6",
pipe length 10.5 ft, 1 red 6x8, 1 90'R 1 45'lbow 90'lbow K =
~ 24, red K =
~ 219,tee bra..ch K =.9,45'lbow K =
~ 24 Flow 450 420 400 h/100 1.34 h
1
~ 07 v
5 v
4.44 h/100' 1
~ 178/100'
4.66 ft/sec hL
L (h/100')
=
10 '
(1 '78/100)
.124 ft h>p = K (v') /2g
. 24 (4. 66') /64. 4 =
~ 08
~
~
~
hbra<<h
K (v-) /2g
. 9 (4. 66
) /64. 4
=
. 303 ft h<<d =
K (v-) /2g
=.219 (4. 66') /64.4
=.074 ft h4s
= K (v ) /2g
=.24 (4.66
) /64.4
=.081 ft h22-23 hL
+
helbow
+
hbranch
+
hred
+
hgs
=
. 66 ft Segment 23-24 configuration obtained from pipe friction calculat'on data sheet flow 420 gpm, diameter 6",
pipe length 8 '69 ft, 1 90'R elbow 1 tee b'ranch, 1 gate valve 90'lbow K =
~ 24, gate valve K =.12, tee branch K =
~ 9, h/100' 1 '78/100'
4. 66 ft/sec hL
L (h/100'
=
8
~ 969
( 1
~ 178/100)
=
. 106 ft hap
K (v') /2g
~ 24 (4
~ 66
) /64
~ 4 =
~ 08 ft hba ch = K (v )/2g =.9 (4.66 )/64
~ 4 =
~ 3 ft hg
<e K (v ) /2g 12 (4
66
) /64.4
=.P4 ft 3
~
~
~
~
~
23-24 hL
+
helbow
+ hb<<nch
+
hgare
+
e 526 ft
0 l
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page 2 a OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1K2 System Emer en Core Coolin S stem SUBJECT Rmeraenc Core Coolina Svstem Pumps NPSH Available Segment 22-25 configuration obtained from pipe friction calculat'on data sheet flow 420 gpm, diameter 8",
pipe length 15.95 ft, 1 tee branch tee branch K =.84 Flow 450 420 400 h/100
.341 h
.284 v
2.89 v
2.57 h/100'
.31/100'
2.69 ft/sec hr,
L (h/100')
=
15.958
(.31/100)
=.05 ft hb,h = K (v-)/2g =.84 (2.69 )/64.4
=
.Oo ft h22 2<
hg
+
hbranch o 14 ft Segment 25-26 configuration obtained from pipe friction calcula-.ion data sheet flow 420 gpm, diameter 6",
pipe length 15.73 ft, 1 90'R elbe" 1 tee branch, 1 gate valve 90'lbow K =.24, gate valve K =.12, tee branch K =.9, h/100' 1.178/100'
= 4.66 ft/sec I
ht, =
L (h/100'
=
- 15. 73 (1. 178/100)
=
. 185 ft hso
= K (v ) /2g
=.24 (4.66-) /64.4
=.08 ft hbranch
=
K (V
) /2g =.9 (4.66') /64.4
=.3 ft haate K (v ) /2g 12 (4
66
) /64 4
04 ft h2s-2s
= ht.
+
helbow
+
hbranch
+ hgate
+
=
~ 61 Determine the RHR um 's suction ressure The RHR pump' suction pressure is determined by the relationship shown below and detailed on page 5 (Note:
All elevations are obtained from Pipe Friction calculation data sheets):
~,
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page
~+
OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Emer en Core Coolin Svstem SUBJECT Emer enc Core Coolina Svstem Pumos NPSH Available solve equation for P2 P2
= @ ((Z1-Z2)
+ 144(P1)
+ (vl)
(v2) h
)
144 PL 2g 2g Note: the 1 and 2 shown in the equations represent the first pipe segment's circled nodes 1 to 2
from figure 1.
For succeeding p'pe segments the node numbers are changed in the equations.
Pressure at P2 Zl = 602.83 ft Z2 = 589.33 v2
11.2 ft/sec h> 2 = 4.06 ft p~
p2 = 60.32 lbs/ft Pl = 14.7 psia vl = 11.2 ft/sec P2 60 32 ((602 83 589 33)
+ 144(14
- 7) +(11 2)2 144
- 60. 32
- 64. 4
- 11. 2)
-'. 06)
- 64. 4 18.65 psia
~
~
Pressure at P3 Z2
589.33 ft Z3 = 586.43 v3 = 10.9 ft/sec h2 s = 2.62 ft p2
p3 = 60. 32 lbs/ft P2
= 18. 65 psia v2 = 11. 2 ft/'sec P3
=
60.32
((589.33 586.43)
+ 144(18.65)
+(11.2) 144 60.32
- 64. 4 10.9)
-2.62) 64.4 18.76 psia Pressure at P4 Z3 v4 ps 586.43 ft Z4
575.17 10.9 ft/sec hs 4
4 ft p<
60.32 lbs/ft P3
= 18.76 psia v3 = 10.9 ft/sec P4
=
60.32
((586.43 575.17)
+ 144(18.76)
+(10.9) 10.90
4) 144 60.32 64.4 64.4 21.8 psia Pressure at P5
RHR Suction Pressure Z4 = 575.17 ft Z5 = 575.08 ft P4
= 21.8 psia v4 10.9 ft/sec v5 10.9 ft/sec h4 s =.079 ft p4
=
ps = 60.32 lbs/ft
~
~
~
~
~
~
~
~
~
~
~
~
P5
=
60.32
((575.17 575.08)
+ 144(21.8)
+ (10.9) 10.9)
.079) 144 60.32 64.4 64.4 21.8 psia
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page
~
OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coolin S stem SUBJECT Emer encv Core Coo'inc S stem Pumps NPSH Available Determine the ECCS um '
= ha h~a
+ h~r hga where:
NPSH net positive suction head available, ft abs h,
absolute pressure, ft h,p, fluid vapor pressure, ft h
- static elevation difference, ft hf, pipe friction losses, ft NPSH Avaiiable South Safet In'ection Pum hf, = the sum of segments hz
+
h>
a
+
ha g
+
hg ip
+
hip ii
+
+
hia-i3 hi h, = 62.41
+ 18.87
+ 88.29
+.91
+ 10.3
+.82
+..93
= 182.53 ft h,
=
RHR PP suction pressure
+
RHR PP TH
~
~
(21.8
- 2.386)
+ 300 (from pp curve at 4600 gpm}
352.01 ft hypa9 34 *2s38622o29 ft hst
=
575 589
~ 21
=
14
~ 21 ft NPSH available
= h, - h,
+ h,
- hf, 352. 01
- 22. 29
+ (-14
~ 21) 182. 53 132.98 ft abs 8 525 gpm NPSH'equired is 13 ft abs at 525 gpm from curve 39890A Available NPSH exceeds required NPSH by 119.98 ft abs NPSH Available North Safet In'ection Pum hf, = the sum of segments hs;
+
h~
a
+
ha g
+ hg ip
+
hip ii
+
+
hi4-is
+
his-is
+
his-i~
+
hi1-ia hfs- =
62 '1
+ 18 '7
+ 88 '9
+
~ 91
+ 10 '
+
5'4
+
~ 31
+
5'2
+
~
~
~
~
~
~
~
~
~
+
",.93
= 193.03 hii-i4
~ 245
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related MechanicaI Systems Calculation Number ENSM970128AF Rev 2 Page Z +
OF 66 DATE 12/9/97 BY A.Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Eme en Core Coolin S stem SUBJECT Emer encv Core Coolincr S stem Pumos NPSH Available h>
=
RHR PP suction pressure
+
RHR PP TH (21.8
- 2.386)
+ 300 (from pp curve at 4600 gpm) 352.01 ft hypa 9 34
- 2 e 386 22 29 ft hst
=
575 589.21
= -14.21 ft NPSH available
=
hg hyps
+ hst hrs C
352.01 22.29
+ (-14.21) 193.03 122.48 ft abs 8 525 gpm NPSH required is 13 ft abs at 525 gpm from curve 39890A Available NPSH exceeds required NPSH by 109.48 ft abs NPSH Available West Centrifu al Char in Pum hq, = the sum of segments ha
~
+
h7 a
+
ha-9
+ ha-to
+
hto-ii 2
+
hg4 ts
+
h.*)a
+
has
$ 9
+
h19-20
+
hao
+
h" aa
+
h23-z4
+
hti g4 has aa
+
h, = 62.41
+ 18.87
+ 88.29
+.91
+ 10.3
+ 5.74
+.31
+
5.02
+
16.39
+
20.71
+ 32.67
+
.67
+
~ 66
+.526
= 263.48 ft h,
=
RHR PP suction pressure
+
RHR PP TH (21.8
- 2.386)
+ 300 (from pp curve at 4600 gpm) 352.01 ft h pa
=
9'4
- 2 '86
22 '9 ft hst
575 592.5
= -17.5 ft NPSH available
=
ha hvpa
+ hst hfs 352.01 22.29
+ (-17.5) 263.48
= 48.74 ft abs 8 420 gpm NPSH required is 17 ft abs at 420 gpm from curve 34617-L Available NPSH exceeds required NPSH by 31.74 ft abs
AMERICANELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page
>7 OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nuclear Plant Unit 1&2 System Emer en Core Coolin S stem
=J SUBJECT Erre enc Core Coolina S stem Pum=s NPSH Available NPSH Available East Centrifu al Char in Pum hf, = the sum of segments
+
h14
'=
+
hts ts
+
h2s-2e hs-7
+
h7-s
+
hs 9
+ h9 1p
+
h1p 11
+
h11 14 h16-19
+
h19 2p
+
h20"21
+
h21 22
+
h22-25
+
hrs
= 62.41
18.87
+ 88.29
+.91
+ 10.3
+ 5.74
+.31
+
5.02
+
16.39
+
20.71
+ 32.67
+
.67
+.14
+
~ 61
= 263.04 ft h,
=
RHR PP suction pressure
+
RHR PP TH (21.8
- 2.386)
+ 300 (from pp curve at 4600 gpm) 352.01 ft hyp4L 9 34
- 2o386 22e29 ft
- hsr,
=
575 592.5
= -1/.5 ft NPSH available
= h, hp,
+ h-hr, 352.01 - 22.29
+ (-17.5) 263.04
= 49. 18 ft abs 8 420, gpm NPSH required is 17 ft abs at 420 gpm from curve 34617-L Available NPSH exceeds required NPSH by 32.18 ft abs NPSH Available Residual Heat Removal Pum hq,
= the sum of segments h1 2
+
h2 3
+
h3 4 +
h4 s
hrs = 4.06
+
2 '2
+
4
+
~ 079 = 10.76 ft h,
=
atmospheric pressure 14.7
- 2.386 35.075 ft hp,
= 9.34 " 2.386
= 22.29 ft
~
~
~
hst
=
602.83 575
= 27.83 ft NPSH available
= h,
,h~,
+ h h<,
35.075 22.29
+ 27.83 10.76
= 29.86 ft abs 8 4600 gpm
1 JV I
V
I AQERICAN ELECTRIC POWER NUCLEAR GENERATION GROUP NUCLEAR ENGINEERING Safety Related Mechanical Systems Calculation Number ENSM970128AF Rev 2 Page Z 5 OF 66 DATE 12/9/97 BY A. Feliciano CK.
PLANT Cook Nucjear Plant Unit 1&2 System Eme en Core Coolin S stem SUBJECT Emeraenc Core Coolina Svstem Pum=s NPSH Available NPSH required is 20 ft abs at 4600 gpm f om curve N-315 Available NPSH exceeds required NPSH by 9.86 ft abs NPSH Available Conta'nment S ra Pum ht, = the sum of segments ht 2
+
h (from app Q
amendment 78) hrs
= 4.06
+ 4.45
= 8.51 ft h,
=
atmospheric pressure 14.7
- 2.386 35.075 ft hyp~
9 34
- 2e386 22e29 f't
~
~
~
~
~
hsc
=
602.83 575.29
= 27.54 ft NPSH available
= h h p
+
h3<
hf, 35.075 22.29
+ 27.54 8.51 31.82 ft abs 8 3200 gpm NPSH required is 9 ft abs at 3200 gpm from curve T-32913-1 Available NPSH exceeds required NPSH by 22.82 ft abs Conclusions The results of this calculation demonstrate that sufficient NPSH is available to assure that the ECCS pumps perform their safety function when aligned to the Recirculation sump.
That is, sufficient NPSH is available with one RHR pump supplying two SI and two CC pumps without taking credit for the containment design prcssure'he attachment to this calculation was performed to evaluate the impact of an RHR pump degraded by 10% from the baseline head-capacity curve.
The results detailed in the attachment indicate that an RHR pump degraded by 10% from the baseline head-capacity curve still provides sufficient NPSH in excess of the required NPSH for the SI and CC pumps.
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JAQ 21 '97 12:48 FROM PLANT-ENG PAGE.881 rhx'out FRICTION CALO - INPUT FILE IS-rhrrecir
'pa
\\
THE XNPUT DATA FOR THE HFLCS SYS.
RES.
CALC.
CONSXSTS OF THE FOLLOWING DATA:
T -. TEMPERTURE DEG F E - PIPE ABSOLUTE ROUGHNESS (FT.)
N FXRST PIPE SEGMENT NUMBER Nl LAST PIPE SEGMENT NUMBER QDES DESIGN FLOW THRU PIPE SEGMENT (GPM)
QMIN NXNIMUM FLOW THRU PXPE SEGMENT (GPM)
QMAZ MAXIMUMFLOW THRU PIPE SEGMENT (GPM)
QDELT FLOW XNCREMENT THRU PIPE SEGMENT (GPM)
D PIPE SEGMENT INTERNAL DIA. (XN.)
L PIPE SEGMENT LENGTH (FT. )
p()st gx b~d ~t~smittg >~0~
p<poooo s
K PIPE SEGMENT K FACTORS L/D PIPE SEGMENT L/D FACTORS
'OLLOWXNG IS YOUR INPUT DATA T
E N
N1 190-00
.00015 1
4 Oopt.
Vt:
Fax o Fax O QDES 7700 00 7700.00 0-00
.00 QMXN QNAX 7700.00 7700.00 7700.00 7700.00 4500.00 4500.00 4500.00 4500.00 QDELT D
. 00 17. 124
. 00 16. 876
.00 13.124
-00 13.124 L
26-66 26.15 42.93 3'3 K
.97
.00
.20
.00 L/D 10.00 100.00 160-00
.00 FOLLOWING XS HFLC5 RESULTS WATER TEMP.(F)
DENSITY(LBM/CUFT)
ABS VISCOSITY(LBM/SEC/FT)
PIPE ABS ROUGHNESS(FT) 190.00 60.32
.217609E-03
.150000K-03 PXPE SEG NO FLOW-GPM, 7700.0 1
PIPE DZA(ZD-ZN) ~
- 17. 124 VEL(FPS)
LHD(FT)
KHD(FT)
LDHD(FT) 10.73
-42 1.73
.22 TOT HD(FT) 2.38 PIPE SEG NO 2
PIPE DIA(ZD-IN) =
16.876 FLOW-GPM VEL(FPS)
LHD(FT)
KHD(FT)
LDHD(FT) 7700-0 11.04
.44
.00 2.38 TOT HD(FT) 2-83 2- '3 PIPE SEG NO FLOW-GPM 4500.0 PIPE SEG NO FLOW-GPM 4500.0 3
PIPE DZA(ID-IN) =
13.124 VEL(FPS)
LHD(FT)
KHD(FT)
LDHD(FT) 10.67
.92
.35 3-74 4
PXPE DZA(ID-XN) ~
13.124 VEL(FPS)
LHD(FT)
KHD(FT)
LDHD(FT) 10.67
.07
.00
.00 TOT HD(FT) 5-01 TOT HD(FT)
.07 OLDS NUMBER FRICTION PIPE SEG DES.
FLOW 1
7700.0 2
7700.0 3
4500.0 FACTOR TABLE RE.NO.
F-FACTOR 4243402. 0
. 0126 4305761.0
.0126 3235750.0
.0132 HEAD LOSS 2.38 2.83 5.01
~~g ~q 101'C npy5 ~W~~
~, JAN.21
'97 12: 48 FROH PL'ANT-ENG PAGE.882 4500.0 3235750.0
. 0132
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FRICTION CALC - INPUT FILE IS-c:icalcirsnpsh THE INPUT DATA FOR THE HFLC5 SYS.
RES.
CALC.
CONSISTS OF THE FOLLOWING DATA:
T TEMPERTURE DEG F E
PIPE ABSOLUTE ROUGHNESS (FT.)
N - FIRST PIPE SEGMENT NUMBER N1 LAST PIPE SEGMENT NUMBER QDES - DESIGN FLOW THRU PIPE SEGMENT (GPM)
QMIN - MINIMUM FLOW THRU PIPE SEGMENT (GPM)
QMAX -
MAXIMUM FLOW THRU PIPE SEGMENT (GPM)
QDELT -
FLOW INCREMENT THRU PIPE SEGMENT (GPM)
D - PIPE SEGMENT INTERNAL DIA. (IN.)
L PIPE SEGMENT LENGTH (FT.)
K - PIPE SEGMENT K FACTORS L/D PIPE SEGMENT L/D FACTORS QC t-J(
FOLLOWING T
E 190.00 QDES 5050.00 5050.00 050.00 IS YOUR
.00015 QMIN 1050.00 1050.00 1050.00 INPUT DATA N
N1 1
3 QMAX QDELT D
5050.00 1000.00 7.981 5050.00 1000.00 7.981 5050.00 1000.00 13.124 L
59.85 10.00
.00 K
L/D
.00 308.00
.00 "93 F 00
.79 20.00 OWING IS HFLC5 RESULTS WATER TEMP.(F) 190.00 DENSITY(LBM/CUFT)
- 60. 32 ABS VISCOSITY(LBM/SEC/FT)
=
.217609E-03 PIPE ABS ROUGHNESS(FT)
.150000E-03 PIPE SEG NO FLOW-GPM 1050.0 2050.0 3050.0 4050.0 5050 '
PIPE SEG NO FLOW-GPM 1050.0 2050.0 3050.0 4050.0 5050.0 E
SEG NO W-GPM 050.0 2050.0 3050.0 4050.0 5050".0 1
VEL(FPS) 6.73 13.15 19.56 25.97 32.39 2
VEL(FPS) 6.73 13.15 19.56 25.97 32.39 3
VEL(FPS) 2.49 4.86 7.23 9.61 11.98 PIPE DIA(ID-IN)
LHD(FT)
KHD (FT) 94
.00
.3.52
.00 7.72
.00 13.55
.00 21.01
.00 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT)
.16
.00
.59
.00, 1.29
.00 2.26
.00 3.51
.00 PIPE DIA(ID-IN)
LHD(FT)
KHD (FT)
.00
.08
.00
.29
.00
.64
.00 1.13
.00 1.76
- 7. 981 LDHD(FT) 3.23
- 12. 03 26.41 46.37 71.91 7.981 LDHD(FT)
.97 3 '3 7.98 14.00 21.71 13.124
.LDHD(FT)
.03
.10
.22
.38
.59 TOT HD (FT)
- 4. 17
- 15. 55
- 34. 13
- 59. 92 92.92 TOT HD (FT) 1.13 4.22 9.27 16.27 25.22 TOT HD (FT)
.10
.39
.86 1.51 2.35 c-7
)
OLDS NUMBER FRICTION FACTOR TABLE SEG DES.
FLOW RE.NO.
F-FACTOR 1
5050.0 5971215.0
.0143 2
5050.0 5971215.0
.0143 3
5050.0 3631231.0
.0132 HEAD LOSS 92.92 25.22 2.35 THE INPUT DATA FOR THE HFLC5 SYS.
RES.
CALC.
CONSISTS OF THE FOLLOWING DATA:
T - TEMPERTURE DEG F E - PIPE ABSOLUTE ROUGHNESS (FT.)
N - FIRST PIPE SEGMENT NUMBER N1 - LAST PIPE SEGMENT NUMBER QDES - DESIGN FLOW THRU PIPE SEGMENT (GPM)
QMIN - MINIMUM FLOW THRU PIPE SEGMENT (GPM)
QMAX - MAXIMUM FLOW THRU PIPE SEGMENT (GPM)
QDELT FLOW INCREMENT THRU PIPE SEGMENT (GPM)
D - PIPE SEGMENT INTERNAL DIA. (IN.)
L PIPE SEGMENT LENGTH (FT.)
K - PIPE SEGMENT K FACTORS L/D - PIPE SEGMENT L/D FACTORS FOLLOWING IS YOUR INPUT DATA T
E N
N1 1
.00
.00015 4
20
~~p f'70l 7d i
~)@g ~ ~l p
S 5050.00 2500.00 700.00 700.00 1800.00 1800.00 1800.00 700.00 700.00 1100.00 1100.00 1100.00 1100.00 550.00 550.00 550.00 550.00 QMZN 1050.00 1000.00 100.00 100.00 1000 F 00 1000.00 1000.00 100.00 100.00 1000.00 1000.00 1000.00 1000.00 150.00 150.00 150.00 150.00 QbQD 5050.00 2500.00 800.00 800.00 2600.00 2600.00 2600.00 800.00 800.00 1200.00 1200.00 1200.00 1200.00 550 F 00 550.00 550.00 550.00 QDELT 1000.00 500.00 100.00 100.00 800.00 800.00 800.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100 F 00 100.00 D
7.981 7.981 6.357 4.260 6.357 8.329 6.357 6.357 4.260 4.260 4.260 4.260 8.329 6.357 6.357'-
8.329 6.357 L
3.00
- 68. 69
- 5. 16 1
~ 17 13.89 1.76 18.81.
4.54 1.17 15.16 5.00 46.13 17.17 10.50 8.97 15.96 15.73 "
K
.00
.00
.22
.28
.22
.19
.22
.00
.28
.28 "
.00
.00
.00
.22
.00 "
.00
.22 L/D
.00 297.00 80.00
.00 73.00 20.00 53.00 20.00
.00 120.00
,193.00 173.00 80.00 96.00
-93.00
- 60.00 93.00
~
~
C I
/7V-2 6 FOLLOWING IS HFLC5 RESULTS WATER TEMP.(F) 130. 00 DENSITY(LBM/CUFT)
- 61. 54 ABS VISCOSITY(LBM/SEC/FT)
=
.342668E-03 E ABS ROUGHNESS(FT)
=
.150000E-03
'I PIPE SEG NO 4
PIPE DIA(ID-IN) =
- 7. 981 FLOW-GPM VEL(FPS)'HD (FT)
KHD (FT)
LDHD(FT)
TOT HD (FT) 1050.0
- 6. 73 '":..05
.00
.00
.05 2050.0 13.15
.18
.00
.00
.18
~
~
)
~
~
~
p 3050.0 4050.0 050.0
- 19. 56 25.97 32.39
.39
.68 1.06
.00
.00
.00
.00
.00
.00
.39
.68 1.06 SEG NO FLOW-GPM 1000.0 1500.0 2000.0 2500.0 PIPE SEG NO FLOW-GPM 100.0 200.0 300.0 400.0 500.0 600.0 700.0 800.0 PIPE SEG NO FLOW-GPM 100 '
200.0 300.0 400.0 500.0 00.0 700.0 800.0 5
VEL(FPS) 6.41 9.62 12.83 16.03 6
VEL(FPS) 1.01 2.02 3.03 4.04 5.05 6.07 7.08, 8.09 7
VEL(FPS) 2.25 4.50 6.75 9.00 11.25 13.51 15.76 18.01 PIPE DIA(ID-IN)
LHD (FT)
KHD (FT) 1.01 F 00 2.22
.00 3.89
.00 6.04
.00 PIPE DIA(ID-IN)
LHD (FT)
KHD (FT)
.00
.00
.01
.01
.02
.03
.04
.06
.06
~ 09
.09
.13
.12
.17
.16
.22 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT)
.00
.02
.02
.09
.04
.20
.07
.35
.11
.55
.16
.79
.21 1.07
.28 1'0 7.981 LDHD(FT)
- 2. 89 6.37 11.20 17.37 6 '57 LDHD(FT)
.02
.09
.19
.34
.52
.73
.99 1.28 4.260 LDHD(FT)
.00
.00
.00 F 00
.00 F 00
.00
.00 TOT HD (FT) 3.90 8.59 15.09 23.41
/Q
//
p/-
< Z.
TOT HD (FT)
.03
.11
.24
.42
.66
.94 1.29 1'8
/0- <3 TOT HD (FT)
.03
.12
.25
.43
.67
.95 1.28 1.66 PIPE SEG NO FLOW-GPM 1000.0 1800.0 2600.0 8
VEL(FPS) 10.11 18.20 26.28 PIPE DIA(ID-XN)
LHD (FT)
KHD (FT)
.65
.35 2.06 1.13 4.26 2.35 6 '57 LDHD(FT) 1.81 5.74 11.87 TOT HD (FT) 2.81 8.93 18.48 II-Ig r PIPE SEG NO FLOW-GPM 1000.0 1800.0 2600.0 PIPE SEG NO FLOW-GPM 1000.0 1800.0 2600.0 9
VEL(FPS)
- 5. 89 10.60 15.31 10 VEL(FPS) 10 '1 18.20 26.28 PIPE DIA(ID-IN)
'LHD (FT)
KHD (FT)
.02
.10
.07
.33
.13
.70 PIPE DIA(ID-IN)
LHD(FT)
KHD (FT)
.86
.35 2.72 1
~ 13.
5.62 2.35 8.329 LDHD(FT)
.16
.51 1.06 6.357 LDHD(FT) 1.31 4.17 8.62 TOT HD (FT)
.29 91
/f-IJ 1.89 TOT HD (FT) 2.52 8.01 I
'6.59 PIPE SEG NO FLOW-GPM 100.0 200.0 300.0 400.0 500 '
600.0 700.0 11 VEL(FPS) 1.01 2.02 3.03 4.04 5.05 6 '7 7.08 PIPE DIA(ID-IN)
LHD (FT)
KHD (FT)
.00
.00,
.00
.00
.02
.00
'04
.'00
.06
.00
.08
.00
.11
.00 6.357 LDHD(FT)
.00
.02
.05
.08
.13
.18
.25 TOT HD (FT)
.00
.03
.07
.12
.18
.26
.35 ibz- (7
)7 yes P8 p~~~6b p'c
is, I ~
800.0 8.09
.14
.00
.32
.46 SEG NO W-GPM 00.0 200.0 300.0 400.0 500.0 600.0 700.0 800.0 PIPE SEG NO FLOW-GPM 1000.0 1100.0 1200 '
PIPE SEG NO FLOW-GPM 1000.0 1100.0 1200.0 PXPE SEG NO FLOW-GPM 1000.0 00.0 PIPE SEG NO FLOW-GPM 1000.0 1100.0 1200.0 PIPE SEG NO FLOW-GPM 150.0 250.0 350.0 450.0 550.0 PIPE SEG NO FLOW-GPM 150 '
250.0 350.0 450.0 550.0 E
SEG NO W-GPM 50.0 250.0 350.0 450.0 550.0 12 VEL(FPS) 2.25 4.50 6.75 9.00 11.25 13.51 15.76 18.01 13 VEL (FPS) 22.51 24.76 27.01 14 VEL(FPS) 22.51 24.76 27.01
. 15 VEL (FPS) 22.51
- 24. 76 27.01 16 VEL(FPS) 5.89 6.48 7.07 17 VEL(FPS) 1.52 2.53 3.54 4.55 5.56 18 VEL(FPS) 1.52 2.53 3.54 4.55 5.56 19 VEL(FPS)
.88 1.47 2.06 2.65 3.24 PIPE DIA(ID-IN)
LHD (FT)
KHD (FT)
.00
.02
.02
.09
.04
.20
.07
.35
.11
.55
.16
.79
.21 1.07
.28 1.40 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT) 5.56 2 '9 6.72 2.65 7.98 3.15 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT) 1.83
.00 2.22
.00
- 2. 63
.00 PIPE DIA(ID-IN)
LHD(FT)
KHD (FT) 16.93
.00 20.44
.00 24.28
.00 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT)
.20
.00
.24
.00
.29
.00 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT)
.01
.00
.03
.02
.06
.04
.10
~ 07
.15
.11 PIPE DIA(ID-IN)
LHD (FT)
KHD(FT)
.01
.00
.03
.00
.06
.00
.09
.00
.13
.00 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT)
.00
.00
.01
.00
.03
.00
.04
- 00
.06
.00 4.260 LDHD(FT)
.00
.00
." 00
.00
.00
.00
.00
.00 4.260 LDHD(FT)
- 15. 63 18.88 22.42 4.260 LDHD(FT)
- 25. 14 30.36 36.06 4.260 LDHD(FT) 22.54
- 27. 21 32.33 8.329 LDHD(FT)
.66
.79
.93 6.357 LDHD(FT)
.06
.16
.31
.51
.74 6 '57 LDHD(FT)
.06
.16
.30
.49
.72 8.329 LDHD(FT)
.01
.04
.07
.11
.16 TOT HD (FT)
.03
.11
.24
.42
.66
.94 1.29 1.68 TOT HD (FT) 23.39
- 28. 24 33.56 TOT HD (FT)
- 26. 98 32.57 38.70 TOT HD ( FT) 39.46
- 47. 65 56.61 TOT HD (FT)
.86 1.03 1.22 TOT HD (FT)
.08
.22
.42
.68 1.00 TOT HD (FT)
.07
.19
.36
.58
.85 TOT HD (FT)
.02
.05
.09
.15
.22 I
~~$ 7 d7 7W
~g vF~ g
~
g
~ I l
I
PIPE SEG NO OW-GPM 50.0 50.0 350.0 450.0 550.0 20 VEL(FPS) 1.52 2 '3 3.54 4.55 5.56 PIPE DIA(ID-IN)
LHD(FT)
KHD(FT)
.02
.00
.05
.02
.10
.04
.16
.07
.23
.11 6.357 LDHD(FT)
.06
.16
.30
.49
.72 TOT HD (FT)
.09
.23 44
.72 1.06 REYNOLDS PIPE SEG 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 NUMBER FRICTION DES.
FLOW 5050.0 2500.0 700.0'00.0 1800.0 1800.0 1800.0 700.0 700.0 1100.0 1100.0 1100.0 1100.0 550.0 550.0 550.0 550.0 FACTOR TABLE RE.NO.
F-FACTOR 3868241.0
.0144 1914971.0
.0146 673170.9
.0159 1004542.0
.0168 1731011.0
.0153 1321171.0
.0147 1731011.0
.0153 673170 '
.0159 1004542 '
.0168 1578565.'0
.0165 1578565.0
.0165 1578565.0
.0165 807382.4
.0151 528919.9
.0161 528919.9
.0161 403691.2
.0159 528919.9
.0161 HEAD LOSS 1.06 23.41 1.28 1.29 8.93
.91 8.01
.35 1.29 28.24 32.57 47.65 1.03 1.00
.85
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Attachment to Calculation ENSM971028AF Purpose Calculation ENSM970128AF determined the available ECCS pumps NPSH. This attachment willreview the impact ofan RHR pump de>>raded by 10% from the baseline head-capacity curve. This attachment is performed under the same basis/assumptions that the calculation is performed.
I
==
Conclusion:==
An RHR pump degraded by 10% from the baseline head-capacity curve willstill assure adequate NPSH is available to the SI and CC pumps.
The NPSH is tabulated as follows:
Pump Flow (gpm)
NPSHA'ft abs.)
NPSHR (ft abs.)
CCS>'I $I el~7 $I "W" CC cc+71 CC RHR CTS 500 500 400 400 4400 3200 124 114 47 48 30 12 12 11 11 18 9
Method:
Plotted on the attached RHR head-capacity curve (N-315) is the system-head curve for the calculation.
The intersection is the operating poim for the conditions stipulated (4600 gpm) in the calculation.
A head-capacity curve for an RHR pump degraded by 10% from the baseline is determined by taking 10% ofthe head value (370 ft) at 3000 gpm (design flow). This value is then subtracted from each ofthe head values at each ofthe low points to generate the degraded head-capacity curve.
The intersection ofthe degraded head-capacity curve with the system-head curve represents the new operating point for the RHR pump. The head and flow is 275 ft 4400 gpm. A percent flow reduction can be obtained as follows:
% Flowa,~,<= [(4600-4400)/4600]
- 100 = 4,35%
This % flowreduction willbe applied to the SI and CC pump flows assumed in the calculation..
4 ~
( (
~
~
~
~
Attachment to Calculation ENSM971028AF Calculation:
CC pump flow based on% reduction 420 gpm - 420*.0435 = 401.73 gpm use 400 gpm SI pump flow based on% reduction 525 gpm - 525*.0435 = 502.16 gpm use 500 gpm Based on these flows the corresponding pipe segment flows are indicated below:
Segment 1-2,2-3 3-4 through 9-10 10-11 11-12,12-13 11-14 through 15-16 16-17,17-18 16-19 through 21-22 22-23,23-24 22-25,25-26 Flow 7,600 4,400 1,800 500 1,300 500 800 400
'400 Determine the RHR pump's suction pressure:
Pressure at P2 8000-7600
=
11.5-v 8000-7000 11.5-10.0 v2= 11.5 - 1.5(400/1000) = 10.9 fUsec h).g = 4.06(7600/7800)
= 3.85 ft P2
= 60.32 ((602.83 - 689.33) + ~144 14.2
+ ~10.9
- ~10.9
- 3.86}
144 60.32 64.4
'64.4 18.74 psia Pressure at P3 4500-4400
=
10.67-v 4500-4000 10.67-9.49 v3 = 10.67 - 1.18(100/500) = 10.43 ft/sec
~ +@7 2 a<)
Attachment to Calculation ENSM971028AF hp.s = 2.62(7600/7800)
= 2.49 ft P3 = 60.32 {(589.33-586.43)+~144 18.74
+ ~10.9 '
~10.43 '
2.49}
144 60.32 64.4 64.4 18.91 psia Pressure at P4 h, =4(4400/4600) =3.666 P4= 60.32 {(586.43-575.17)+~144 18.91
+ (10.43' 10.43'
- 3.66}
144 60.32 64.4 64.4 22.09 psia Pressure at P5 - RHR Suction Pressure l~g =.079(4400/4600)
=.072 ft P5= 60.32((575.)7-575.08)+~14422.09
+ ~10.43
- ~10.43
-.072}
144 60.32 64.4 64.4 22.1 psia NPSH Available South Safe In ection Pum ha = the sum ofsegments h().g+ he~ +
h))-g + hg.)o + h)o>> + h>>.n
+ h)p.>>
Q7 62.4 1 (4400/4600) = 57. 1 ft h~.)) = 18.87(4400/4600)
= 17.26 I h{).g = 88.29(4400/4600)
= 80.78 ft hg.)p =.91(4400/4600)
=.83 ft h>o-) t = 10 3(1800/1890)'
9 34 <
h>>.)p =.82(500/525) =.74 ft h)p.tg =.93(500/525) =.84 ft ha = 166.89 ft
+t e'
Attachment to Calculation ENSM971028AF h, = RHR PP suction pressure+ RHR PP TH
= (22.1
- 2.386) + 275 (from degraded pp curve at 4400 gpm) 327.73 ft h~ = 9.34 *2.386 = 22.29 ft hg= 575-589.21
=-14.21 ft NPSH available = h, - h~ + h-hr,
= 327.73 - 22.29 + (-14.21) - 166.89
= 124.34 ft abs 500 gpm NPSH required is 12 ft abs at 500 gpm from curve 39890A Available NPSH exceeds required NPSH by 112.34 ft abs NPSH Available North Safetv In ection Pum hr,= thesumofsegments 4.~+ h~~ + h).p +hp.)p + h)p.)) + h)).)~ + h)~)q +
h)5-)p + h)g)p + h)7-)3
+p = 62.41(4400/4600)
= 57.1 ft hv.)) = 18.87(4400/4600)
= 17.26 ft h)) p = 88.29(4400/4600)
= 80.78 ft h) )p = 91(4400/4600)
=
83 ft h)p.)) = 10.3(1800/1890)
= 9.34 ft h)).)4 = 5.74(1300/1365)
= 5.21 ft h)4.)$ =.314(1300/1365)
=.28 6 h)5.)g = 5.02(1300/1365)
= 4.55 ft h) p.)p =.245(500/525)
=.22 ft h)v.))) =.93(500/525)
=.84 ft
Attachment to Calculation ENSM971028AF hr, = 176.41 ft
= (22.1
- 2.386) + 275 (from degraded pp curve at 4400 gpm) 327.73 ft h~ = 9.34
- 2.386 = 22.29 ft h= 575 - 589.21
= -14.21 ft NPSH available = h, - h~ + h, - hr,
= 327.73 - 22.29 + (-14.21) - 176.41
= 114.82 ft absI500 gpm NPSH required is 12 ft abs at 500 gpm Gism curve 39890A Available NPSH exceeds required NPSH by 102.82 ft abs I~
NPSH Available West Centrifu al CharLin Pum ha = the sum ofsegments 4.7+
h7~ + h~p +hp.ip + hio.u + hu.ig + hi4.ig +
his.i6 + hip.ip+ hip.~p + hagi
+ h i.~ + hpz.g + h23.~
- 4. = 62.41(4400/4600)
= 57.1 ft h7.ii = 18.87(4400/4600)
= 17.26 ft hg.9 = 88.29(4400/4600)
= 80.78 ft h9-io = 91(4400/4600)
=.83 ft hip.ii = 10.3(1800/1890)
= 9.34 ft hii.i4= 5.74(1300/1365)
= 5.21 ft hi4.is =.314(1300/1365)
=.28 ft his-i6 = 5.02(1300/1365)
= 4.55 ft hi6.ip = 16.39(800/840)
= 14.87 ft hip.~o = 20.71(800/840)
= 18.78 ft S~~
5 gpfy7 0
Attachment to Calculation ENSM971028AF hgp.g) = 32.67(800/840)
= 29.63 ft hei.~ =.67(800/840)
=.61 ft h~~ =.66(400/420)
=.59 ft hg).~ =.526(400/420)
=.48 ft ha = 240.31 ft h, = RHR PP suction pressure+ RHR PP TH
= (22.1
- 2.386) + 275 (from degraded pp curve at 4400 gpm) 327.73 ft h~ = 9.34*2.386 = 22.29 ft h= 575-592.5
=-17.5 ft NPSH available = h, - h,~ + h-hr,
= 327.73 - 22.29 + (-17.5) - 240.31
= 47.63 ft abs Qa 400 gpm NPSH required is 11 ft abs at 400 gpm from curve 34617-L Available NPSH exceeds required NPSH by 36.63 ft abs NPSH Available East Centrifu al. Chargi~nP~um hr, = the sum ofsegments Q~+ h>< + h>> + h~ip +
h~o.>~ + hi>.lc + h:-'!5 +
h>s.)g + h)I5)9+ h/9.20 + hagi
+ heal.ri + hn.~ + h~..g
+q = 62.41(4400/4600)
= 57.1 ft h~.~ = 18.87(4400/4600)
= 17.26 ft h>> = 88.29(4400/4600)
= 80.78 ft h9.1p =.91(4400/4600)
=.83 f1 hio.n = 10.3(1800/1890)
= 9.34 ft
Attachment to Calculation ENSM971028AF hu.ip = 5.74(1300/1365)
= 5.21 ft hi~>s = 314(1300/1365)
= 28 ft h>5.ig = 5.02(1300/1365)
= 4.55 ft h)gip = 16.39(800/840)
= 14.87 ft h)y.~ = 20.71(800/840)
= 18.78 ft hm.z> = 32.67(800/840)
= 29.63 ft hz).zz =.67(800/840)
=.61 ft hzz-z5 =.14(400/420)
=.13 ft hzg.zg =.61(400/420)
='.55 ft h(, = 239.92 ft h, = RHR PP suction pressure+ RHR PP TH
= (22.1
- 2.386) + 275 (from degraded pp curve at 4400 gpm) 327.73 ft h~ =9.34*2.386 =22.29 ft hs,= 575-592.5
=-17.5 ft NPSH available = h, - h~ + h-hr,
= 327.73 - 22.29 + (-17.5) - 239.92
= 48.02 ft abs 400 gpm NPSH required is 11 ft abs at 400 gpm from curve 34617-L Available NPSH exceeds required NPSH by 37.02 ft abs NPSH Available Residual Heat Removal Pum hr, = the sum ofsegments h~.z+
hz.z + h3~+~
Attachment to Calculation ENSM971028AF h).2 = 4.06(7600/7800)
= 3.85 ft h2.3 = 2.62(7600/7800)
= 2.49 ft h3% = 4(4400/4600)
= 3.66 ft h4.$ =.079(4400/4600)
=.072 ft hp, = 10.07 ft h, = atmospheric pressure
= 14.7*2.386 35.075 ft h~ = 9.34
- 2.386 = 22.29 ft h>> = 602.83 - 575 = 27.83 ft NPSH available = h, - h,~ + h>> - ha
= 35.075 - 22.29 ~ 27.83 - 10.07
= 30.55 ft abs @4400 gpm NPSH required is 18 ft abs at 4400 gpm from curve N-315 Available NPSH exceeds required NPSH by 12.55 ft abs NPSH Available Containment S rav Pum hr, = the sum ofsegments hi.2+ h (from app Q, amendment 78) ha = 4.06(7600/7800)
+ 4.45 = 8.3 ft h, = atmospheric pressure
= 14.7*2.386 35.075 ft
\\
h~ =9.34*2.386 = 22.29 ft h>> = 602.83 - 575.29 = 27.54 ft
Attachment to Calculation ENSM971028AF NPSH available = h, - h,~ + h-hr,
= 35.075 - 22.29 + 27.54 - 8.3
= 32.03 ftabs@3200 gpm NPSH required is 9 ft abs at 3200 gpm from curve T-32913-1 Available NPSH exceeds required NPSH by 23.03 A abs
J 4'a~~
Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 491071395 ENCLOSURE TO AEP:NRC:1260G7
Attachment 2 to AEP:NRC:1260G7 Page 1
The following presents the questions contained in your June 8,
- 1998, request for additional information (RAX), with our response following.
ReceR eel "The October 9,
1997 petition from the Union of Concerned Scientists (UCS) raised concerns that the Nuclear Regulatory Commission (NRC) design inspection in August and September of 1997 identified significant operability issues in systems that have recently been evaluated and approved by the D.C. Cook design basis documentation reconstitution program.
Following the inspection, the NRC issued a Confirmatory Action Letter (CAL) on September 19, 1997.
The CAL references letters that you have docketed and that describe the long-and short-term action plans to be used at D.C. Cook to find and correct engineering problems in other safety-related systems.
Please provide specific details of the programs that will be used to identify significant deficiencies in safety-related systems before restart of either D.C.
Cook Unit 1 or Unit 2.
Your response should include the following detailsc a ~
b.
C ~
d 0 B ~
systems to be reviewed and the logic for selection of the systBIRsg review methodology, including milestones, system deficiencies, corrective actions, and whether each system is in full conformance with the licensing and design basis as described in the Updated Pinal Safety Analysis Report (UPSAR).
Res onse 1
The AEP Nuclear Generation Group (AEPNG) has expanded the scope of our actions to identify and correct discrepancies in safety-related systems that were identified through the Nuclear Regulatory Commission (NRC) Architect Engineer (A/E) inspection and other internal inspections beyond that described in previous submittals.
This expanded response is embodied in the Cook Nuclear Plant Restart
- Plan, which was formally initiated on March 7, 1998.
The plan was discussed with NRC personnel at the SALP board meeting on April 3,
- 1998, and again at the pre-decisional enforcement conference on May 20,
- 1998, and was docketed
- under AEP:NRC:1303.
The restart plan is similar to those recently used at several other plants.
The specific details of the programs that will be used to identify discrepancies in safety-related systems before restart of either Cook Nuclear Plant Unit 1 or Unit 2 are defined below in our combined response to requests 1.a and 1.b.
a ~
b.
Systems to be reviewed and the logic for selection of the systems Review methodology, including milestones The Cook Nuclear Plant Restart Plan and other ongoing efforts are currently underway to provide reasonable assurance that significant discrepancies in the systems evaluated have been identified and are properly dispositioned prior to restart.
These actions include:
Attachment 2 to AEP:NRC:1260G7 Page 2
Restart plan system readiness reviews Review of non-risk significant maintenance rule systems Review of non-maintenance rule systems Containment spray safety system functional inspection (SSFI)
Vertical slice inspections of containment and containment systems Additional SSFI-type inspection UFSAR revalidation project These actions are discussed below in more detail.
Pl n v
w
~
The maintenance
- rule, which provided a
pre-existing classification for systems into risk significant categories, was used as the basis for selecting systems for these comprehensive reviews.
Various probabilistic risk assessment results (e.g.,
core damage frequency, risk reduction
- worth, risk achievement
- worth, and Fussel-Vessely values) were re-,examined to provide additional assurance that the maintenance rule system classification did not exclude important systems.
The selected systems encompass risk significant maintenance rule systems at Cook Nuclear Plant, as well as systems classified as important non-risk significant standby maintenance rule systems, as follows:
~
120 volt AC/CRID Inverters
~
Air Recirculation/Hydrogen Skimmer
~
~
250 volt DC Station Batteries
~
Component Cooling Water
~
Containment
~
~
Control Air
~
~
ECCS Charging/CVCS High Head Injection
~ -
~
ECCS Safety Injection
~
Electrical Safety Busses (4000 volt/600 volt)
~
~
Essential Service Water
~
Ice Condenser
~
~
Non-essential Service Water
~
Plant Air Compressors
~
Reactor Coolant System/RCS Pressure Relief
~
Reactor Protection System/Solid-state Protection/ESFAS
Attachment 2 to ABP:NRC:1260G7 Page 3
~
These reviews are led by system engineers, with input from operations and maintenance personnel.
~
Materiel condition and design basis conformance are reviewed to determine whether there exists reasonable assurance that the
- systems, following resolution of discrepancies identified during the
- reviews, will be capable of start-up and operation within their design bases.
~
The materiel condition reviews include:
- 1. system walkdowns by an interdisciplinary team;
- 2. review of outstanding condition reports;
- 3. review of corrective and preventive maintenance backlog for the affected system;
- 4. review of maintenance rule system performance; and
- 5. review of operability determinations in effect.
~
The design basis conformance reviews include:
- 1. review of UFSAR and technical specification design requirements;
- 2. review of surveillance tests for the affected system;
- 3. review of pre-operational testing;
- 4. evaluation of design modifications
- approved, but not implemented;
- 5. review of design modifications in service;
- 6. review of temporary modifications currently in service; and
- 7. review of industry operating experience.
~
Composite results of the system readiness reviews will be examined to determine if horizontal expansion into programmatic areas is warranted.
~
Qualifications of system engineers performing the system reviews were specifically evaluated by oral examination before a
panel of industry
'and Cook Nuclear Plant engineering peers and managers.
~
Initia1. review of all twenty-one systems is complete.
Final presentations of system readiness to the system engineering 'eview board (SBRB) are in progress.
Presentations to the restart oversight committee (ROC) are xn progress.
vi w N
-Ri i ni i n
M 'n n n R
~
The remaining non-risk significant maintenance rule systems will be reviewed under the plant engineering functional area review.
~
The reviews of non-risk significant maintenance rule systems by system engineers will include:
- 1. review of outstanding condition reports;
Attachment 2 to ABP:NRC:1260G7 Page 4
2
~ review of corrective and preventive maintenance backlog for the affected system;
- 3. review of maintenance rule system performance;
- 4. review of operability determinations currently in effect;
- 5. overview of design changes in service; and
- 6. review of temporary modifications in service.
~
Identified discrepancies that meet the restart criteria will be addressed in accordance with the Cook Nuclear Plant restart plan.
~
Additionally, non-risk significant maintenance rule systems that are reviewed will be evaluated to determine whether significant materiel condition problems or significant design basis non-conformances exist to warrant additional reviews.
f
~
The reviews of non-risk significant maintenance rule systems will be initiated during July 1998 and will be completed prior to restart.
R vi w N
in n
ul
~
Condition reports and.
maintenance backlogs for plant systems not covered under the mai.ntenance'ule will be used as indicators to determine ifi further functional reviews of individual systems are warranted.
Generally, these systems are required for plant operation and are monitored in service.
F n in n
i n
~
Based on issues identified during the A/B inspection, we determined that containment spray will be evaluated in more detail prior to restart.
~
An independent contractor was used to conduct an SSFI-type inspection of containment spray.
Issues identified during the inspection are currently being addressed.
n f n
'nm n
~
Based on. lessons learned during and following the A/E inspection, it was determined that the containment and accident response systems that it houses will be evaluated-in more detail prior to restart.
~
An independent contractor was used to conduct a vertical slice inspection of the containment and the containment systems.
Issues identified during these inspections are currently being addressed.
Attachment 2 to AEP:NRC:1260G7 Page 5
A 'nl FI-I
~
A performance plan for an SSFI of one additional risk-significant system is currently under development and is scheduled to begin in August 1998.
F vli
~
This ongoing project involves a line-by-line review and revalidation of design bases as described in the UFSAR.
Identified UFSAR discrepancies that meet the condition report threshold, including those of the twenty-one systems covered under the restart plan system readiness reviews, will be dispositioned in accordance with the restart plan.
These UFSAR discrepancies will be dispositioned by correcting the non-conformance, performing a
10 CFR 50.59 evaluation, performing an operability evaluation in accordance with generic letter 91-18, revision 1, or requesting a license amendment.
~
The UFSAR reviews are performed by an independent team of consultants under the direction of AEPNG.
~
UFSAR reviews for twenty-one systems covered by the restart plan system readiness reviews will be completed
, prior to restart.
c.
System deficiencies As of July 27,
- 1998, approximately 3366 discrepancies have been identified in the system readiness reviews and vertical slice inspections.
Of this
- number, approximately 69%
are materiel condition issues and 15% are design basis issues.
About 494 of these have been classified as restart items.
Open items generated during the system readiness reviews are classified according to System Engineer Review Board (SERB) criteria.
The SERB criteria contains twenty-five categories related to materiel condition and design basis.
The SERB criteria uses attachment C of the restart plan to establish the threshold for restart items.
Each open item is categorized to the SERB criteria and is cross-referenced to the restart plan screening criteria.
The application of the SERB criteria provides a
systematic, uniform method to classify items identified during the system readiness reviews.
d.
Corrective actions Corrective actions will be taken prior to restart for items
- meeting, the restart criteria.
Other discrepancies will be addressed through normal corrective action and work control systems.
e ~
Whether each system is in full conformance with the licensing and design basis as described in the Updated Final Safety Analysis Report (UPSAR).
As described
- above, the various system review efforts are intended to identify discrepancies in safety-related
- systems, including non-conformances with the design basis as described in
32 Attachment 2 to ABP:NRC:1260G7 Page 6
the UFSAR.
As discussed
- above, identified UFSAR discrepancies that meet the condition report threshold, including those of the twenty-one.
systems covered under the restart plan system readiness reviews, will be dispositioned in accordance with the restart plan.
These UFSAR discrepancies will be dispositioned by correcting the non-conformance, performing a
10 CFR 50.59 evaluation, performing an operability evaluation in accordance with generic letter 91-18, revision 1,
or requesting a license amendment.
R~ee t 2
"If a system will not be in conformance with its licensing and design bases, please provide the details of the deficiency, and a justification for the system's operability."
Res onse 2
The system review efforts currently undexway are intended to identify discrepancies in safety-related
- systems, including non-conformances with the design basis as described in the UFSAR.
Identified UFSAR discrepancies that meet the condition report threshold, including those of the twenty-one systems covered under the restart plan system readiness
- reviews, will be dispositioned in accordance with the restart plan.
These UFSAR.
discrepancies will be dispositioned by correcting the non-conformance, performing a
10 CFR 50.59 evaluation, performing an:
operability evaluation in accordance with generic letter 91-18, revision 1, or requesting a license amendment.
~Re eet 3
Describe the programmatic changes that will be implemented at D. C.
Cook before restart and that in the long term will provide reasonable assurance that safety-related systems as described in the VFSAR will perform their intended safety function.
Re~ggn<~
~B~~un The December 2,
- 1997, response to the NRC Confirmatory Action Letter transmitted, as attachment 4,
our short-term assessment program results.
This assessment was performed., to determine the extent of. the previously identified CAL issues.
Subsequent to this. submittal; the NRC requested additional information on the programmatic i'mplications of the issues raised in the A/E inspection.
In response: to this request and to support resolution of issues associated:,
with the
- CAL, AEPNG initiated a
comprehensive assessment of the A/E inspection findings and their potential broader implications, and consolidated this information from a programmatic perspective.
An integrated multi-discipline team, the A/B Inspection Programmatic Issues Team (ABPIT), reporting to senior management, was formed in January 1998 to carry out this comprehensive assessment.
This assessment examined the program areas of design
- control, 10 CFR 50.59, corrective
- action, and relevant parts of other programs related to design control (developing and maintaining procedures, generic NRC operating experience (OE) information review, and quality assurance related
/
'S
Attachment 2 to ABP:NRC:1260G7 Page 7
to A/B inspection issues).
Nine programs were evaluated by the ABPIT to evaluate the nature and extent of programmatic issues affecting design and configuration control.
Separate from the ABPIT initiative, an additional evaluation was performed on the surveillance program.
The AEPIT recommendations, including those for the surveillance
- program, are being dispositioned in accordance with the Cook Nuclear Plant restart plan.
Pr ammatic Chan es The programmatic findings and our
- complete, restart listed to reflect changes developed from the A/E Inspection subsequent evaluations are summarized below as or post restart actions.
Completed actions are the extent of changes made to date.
Design control is the
- process, used by ABPNG to engineer and document changes to design basis information or physical features of plant structures,
- systems, and components.
'The design control process is intended to ensure that regulatory requirements are met and good engineering practices are followed when changing technical,
- quality, or functional requirements, or performance characteristics of the plant.
D h n This program encompasses the processes and procedures used by AEPNG to engineer and documeht changes to the design of the plant.
The scope of this program includes engineering,
- design, installation, and testing of design changes.
Based on the results of,the assessment, AEPNG has taken or will take the following steps to address specific A/E inspection issues and areas requiring program improvements.
~
Selected system descriptions, design standards, and design guidelines were revised to incorporate design changes or corrective actions related to A/B inspection
'issues'
, Completed selected design changes to address A/E inspection issues such as the modification of the control air system.
~
Developed a
new procedure to enhance program controls for installation of insulation inside containment.
~
Developed' new procedure to improve the design change determination process to provide added assurance that the design change process will not be bypassed.
~
Revised procedures to establish design review teams for design
- changes, clarified the use of technical direction, and addressed the practices for abandoned plant equipment.
Attachment 2 to AEP:NRC:1260G7 Page 8
~.
Complete specific design changes in accordance with the restart plan such as upgrading ice condenser door shock absorbers.
~
Revise design change procedure to strengthen ties to the design basis and licensing basis update processes.
~
Conduct familiarization training on the use of specific design standards, procedures governing abandoned equipment, and the revised design change procedures.
P R
~
Complete implementation of the Engineering Improvement Program (EIP).
~
Conduct self-assessments to monitor the effectiveness of procedural and process enhancements for design changes.
P i n f h
i Li n in B
Various processes are utilized to document and evaluate the plant design and licensing bases at Cook Nuclear Plant.
AEPNG has initiated the following actions to address specific A/E inspection issues and areas requiring program improvements.
m l n
, ~
Procedures and familiarization training have been implemented to clarify the definitions of
'change,'licensing basis,'nd
'design basis' Established a
design basis reconstitution project to integrate and improve the effectiveness of the UPSAR revalidation
- project, the design basis document reconstitution
- project, and the normal operating procedure upgrade project.
R A
~
Revise procedures to improve work processes for maintaining design and licensing basis documents.
~
Complete UFSAR revalidation activities for the twenty-one systems covered by the restart plan.
R A
n
~
Complete the design basis reconstitution project.
~
Change the calculation index database to improve access and retrievability, of design basis information.
Attachment 2 to AEP:NRC:1260G7 Page 9
i n See response to request 4c for a detailed discussion of our ongoing plan to address issues associated with the accuracy and quality of engineering calculations at Cook Nuclear Plant.
In r n
n r i This program captures the process used to provide assurance that instrument uncertainty is appropriately addressed in our calculations and to account for instrument uncertainties in our procedures.
AEPNG is taking several steps to address and resolve instrument uncertainty issues, as described.below:
A i n
~
Additional guidance has been developed and incorporated in the plant-specific methodology manual.
~
Level instruments similar to the refueling water storage tank (RWST) level instruments were reviewed to determine whether problems similar to those encountered with the RWST level instruments exist elsewhere in the plant
~
~
Procedural improvements to control the use of uncertainties in'rocedures,
- analyses, and tests.
~
Actions have been taken to modify the design of the RWST level instrumentation to address the flow induced error effects identified in the A/E inspection.
~
Operator procedures, used shiftly and daily to verify technical specification compliance, were revised to address instrument uncertainties.
~
Engineering standards associated with the design of level measurement systems have been revised.
Enhanced training for affected personnel and interfacing departments.
This training will focus on critical parameters, process measurement uncertainties, and
'nstrument uncertainty calculations.
Required changes resulting from the programmatic review of calculations will be incorporated into the'instrument and control (IaC) information system procedures, in accordance with the restart plan.
n
~
New calculations are being generated to address instrument uncertainties that were not referenced in existing calculations.
Attachment 2 to AEP:NRC:1260G7 Page 10
~
Conduct an assessment
'of the instrument uncertainty program effectiveness.
~
Inputs for instrument uncertainties will be incorporated into the normal operating procedures upgrade program.
~
Emergency operating procedures will be reviewed to identify and validate footnote values.
1 FR Im mn The 10 CFR 50.59 program defines the process by which proposed changes to the plant or procedures, as described in the
- UFSAR, are reviewed to determine if they can be implemented without prior NRC approval.
The process used to perform 10 CFR 50.59 screening and 10 CFR 50.59 evaluations was evaluated.
A detailed discussion of the old 10 CFR 50.59 process is provided in our response to request 4b.
Additionally, a review was performed to evaluate the controls used to provide assurance that the screening and evaluation processes are not bypassed.
The following steps have been or will be taken to address specific A/E inspection issues and areas requiring program improvements.
m l A i
~
Procedures and familiarization training have been implemented to clarify the definitions of
'change,'licensing basis,'nd
'design basis'.
~
An industry expert was retained to review the program and procedures, recommend appropriate improvements, and provide training on the new 10 CFR 50.59 procedures.
~
Training was conducted, by an industry expert, on new 10 CFR 50.59 procedures.
~
Process implemented
, to communicate management expectations regarding change via the 10 CFR 50.59 process to appropriate personnel.
A n
~
Revisea procedures to address potential 10 CFR 50.59 bypass, mechanisms identified by our internal assessment.
This program encompasses the process used to identify and address conditions'dverse to quality.
The review of this program focused on our capability to take timely corrective and preventive action when non-conformances are identified and to determine whether the program supports maintaining the plant design bases and licensing bases.
The following actions have been or will be taken to resolve corrective action program issues:
P
Attachment 2 to ABP:NRC:1260G7 Page 11
~.
Established dedicated corrective action group to own the corrective action
- process, and
- monitor, motivate, and mentor line management implementation of the corrective action program.
~
Ownership of the program has been defined and communicated within the organization.
~
Enhanced procedural guidance to establish daily review of condition reports through a
management review board to improve classification of observed conditions.
~
Reduced the number of significance levels for condition reports to optimize root cause analysis efforts.
~
Procedures revised to improve effectiveness, timeliness, and to clarify when 10 CFR 50.59 screenings are required.
~
Bffectiveness measures have been developed to monitor program performance.
R A i n
~
Reduce and maintain the backlog of overdue corrective action items within established standards.
~
Clarify line management responsibility and accountability in the implementation of the corrective action program.
~
Change the process to align the level of root cause analysis and corrective and preventive actions to be commensurate with event or condition significance.
R A
~
Implement improved condition reporting software to enhance condition trending and event analysis.
~
Update. corrective action procedures to address process and enhance reporting capability for tracking and trending.
~
Conduct additional training on root cause
- analysis, apparent cause analysis, and error reduction technology.
~
Participate in an industry project sponsored by the Electric Power Research Institute (BPRI) plant support engineering subcommittee to develop guidance to optimize engineering activities in support of corrective action programs.
~
Conduct assessments of program effectiveness.
r
Attachment 2 to AEP:NRC:1260G7 Page 12 0th x R lat d Pz am Ar a D v l in n
i inin Pr This program includes the processes utilized to incorporate design bases and licensing bases information into procedures, and to maintain the procedures current.
The following actions have been or will be taken to resolve procedure-related issues.
R r
A i n
~
Update specific AEPNG corporate directive describing the current organization.
~
Conduct additional self assessments to evaluate consistency of AEPNG procedural controls.
~
Complete A/E inspection condition report actions identified as restart items related to updating specific operating procedures.
~
The senior management review team determined that a
complete document control and records management functional area assessment will be performed before restart.
P R
r A i n
~
Complete normal operations procedure upgrade
- project, which was instituted in October 1997, to address quality and human performance related aspects of the procedures.
~
AEPNG corporate and plant procedure processes will be integrated.
n i R
'n E
in E
V W
This program is the process used by AEPNG to review generic NRC correspondence related to industry OE to identify potential impacts. on the design and operation of Cook Nuclear Plant.
The following actions have been or wi.ll be taken to address generic NRC OE related process issues:
R A i
~
- Procedure revisions to consolidate, the review process for NRC and OE information.
Evaluate the need for further sampling of past NRC communications for appropriate disposition.
~
Conduct familiarization training on procedure revisions that consolidate the review process.
Attachment 2 to ABP:NRC:1260G7 Page 13 l'
n A
R n I A review was also conducted of various aspects of the QA program.
AEPNG has initiated the following actions to address identified quality assurance issues:
m l A i ns
~
~
A self assessment, joint utility management audit (JUMA),
and root cause analysis were completed to identify issues related to the planning and implementation of QA oversight initiatives.
~
Performance assurance has prioritized significant programmatic issues associated with the QA program and escalated them to senior management for action and accountability (this is a continuing process).
~
Senior line management has assigned ownership for resolution of these identified programmatic issues.
A i n
~
Audit plans are being revised to specifically require performance assurance to challenge design
- inputs, such as assumptions, when calculations are assessed.
~
The method for directing performance assurance resources is being changed to enhance the oversight of the design and condition of systems.
~
Revise performance assurance system surveillance instructions to include passive components.
~
Conduct training on the changes to audit plans and surveillance instructions.
r A i
~
Develop additional procedural guidance to provide direction for follow-up on previously identified adverse conditions.
~
Follow, up assessments will be conducted to determine whether restart actions have effectively addressed identified QA issues.
The surveillance program was added to the list of programs to be evaluated as part of the restart plan.
The following actions have been or will be taken to address surveillance program
- issues, such as those identified with the ice condensers
~
~
A team was formed to perform a root cause analysis of issues related to the surveillance program.
Attachment 2 to AEP:NRC:1260G7 Page 14
~
Operations department superintendent was designated as the owner of the surveillance program.
R r
A ions
~
~
A group will be
- formed, responsible for
- managing, developing, scheduling, and tracking the completion of surveillances for the plant.
~
Training and qualification of personnel performing surveillance testing activities is being evaluated to determine the extent of additional training required.
~
Procedures are being revised to enhance consistency between the different work groups' Assessments are ongoing to evaluate conformance with regulatory requirements and surveillance program acceptance criteria.
P R
r A i n
~
New scheduling tools are being evaluated to improve the efficiency of the surveillance scheduling process.
r r m
Utilizing a
multi-disciplined
- approach, a
comprehensive evaluation of
- programs, procedures, condition
- reports, and
,related processes was completed.
As a
result of these evaluations, corrective actions have been identified related to programmatic issues and will be addressed in accordance with the Cook Nuclear Plant restart plan.
- Further, provisions for measuring and monitoring future programmatic effectiveness, as described in the preceding
- sections, have been or will be developed.
ReceR >~~4 By letter dated January 12,
- 1998, the UCS submitted an addendum to the original 2.206 petition.
The January 12, 1998 letter raised six new concerns.
Please respond in full to the following five concerns from the January 12, 1998, letters a ~
Concern 1
as it pertains to D.
C.
Cook Plant.
- Also, include the detailed action plan for 'the ice melt, ice condenser, inspection, and repair plan.
Re on o 4 The first concern pertained to the Cook Nuclear Plant ice condenser containment and stated, "The NRC Inspector General's office was informed last summer about alleged problems in the configuration and testing of the ice condenser at Watts Bar.
Problems with the bay doors and components of the ice baskets were specifically identified.
The allegations also suggested that many of the problems were generic and therefore affected the other ice condenser
- plants, including D.
C.
Cook.
Finally, it was alleged that the problems were known, but not properly reported by the D.C.
Cook licensee, the McGuire licensee, and even Westinghouse:"
Attachment 2 to AEP:NRC:1260G7 Page 15 The problems discussed regarding the lower inlet doors involved uplift of the ice condenser floor slab at Sequoyah and McGuire nuclear plants due to water intrusion and subsequent
- freezing, which resulted in binding of the ice condenser lower inlet doors.
The problems with the ice basket coupling screws involved the discovery of coupling screw heads and complete screws in the bottom of the Watts Bar ice condenser following thaw of the ice condenser in 1995.
The damaged and intact screws at Watts Bar were attributed to improper torquing during initial installation, and possibly due to thermal cycling.
Cook Nuclear Plant personnel had been made aware of the problems with floor heaving and lower inlet door binding through the sharing of ice condenser operating experience among ice condenser plants.
When this operating experience became available, Cook Nuclear Plant personnel made tours of each ice condenser at the first opportunity and no evidence of floor uplift was identified.
Since that
- time, Cook Nuclear Plant has not experienced any uplift of the ice condenser floor slab and has not experienced binding of lower inlet doors due to floor uplift.
The performance at Cook Nuclear Plant is attributed to an operating practice to perform aggressive floor defrosts to ensure thorough drying of the floor following evolutions where water may have come in contact with the floor.
Awareness of the experience at other plants has resulted in a
heightened sensitivity to the potential to damage the floor due to water intrusion and re-freezing.
For example, following the recent thaw of the unit 1 ice condenser, where water clearly came in contact with the floor, extensive measures are being taken that are intended to ensure the floor is sufficiently dry before the ice condenser is cooled below freezing temperatures.
Review of the ice basket coupling screw issue at Watts Bar indicated that the root cause was attributed to screws being over-torqued during initial installation, and also possibly due to thermal cycling of the screws.
It was the recollection of Cook Nuclear Plant personnel that ice basket coupling screws or screw heads had been found in the ice condenser or ice melt system in past
- years, though not in the same numbers as Watts Bar.
These screws were attributed to known damage to ice basket top rims, and known separated ice baskets.
Ice basket top rim damage and separated ice basket segments have occurred in. the. past during ice basket weighing surveillances.
Ice baskets are weighed by lifting the basket from the top rim.
If an ice basket is frozen in place and a high degree of lifting force is applied, it is possible to distort, the top rim of the ice basket, or to separate two adjacent ice basket segments.
If the distortion or separation is significant, sheet metal screws, which attach the top rim to the ice basket
- cylinder, or which attach adjacent ice basket
- segments, can be sheared.
This condition was known to exist on a number of baskets in both units at Cook Nuclear Plant, and remedial actions were taken, such as replacing fasteners, and restraining ice baskets having separated segments to prevent basket ejection.
The screws and screw head' observed in the bottom of the ice condenser and in the ice melt, system vacuum filter were attributed to these types of basket damage.
Subsequently, during early
- 1998, ice basket inspections were conducted on both units to further investigate the ice basket coupling screw issue.
A number of missing screws were
Attachment 2 to AEP:NRC:1260G7 Page 16 identified, and documented under I ER 315/98-005.
Inspection and repair of ice baskets is
- ongoing, along with metallurgical analysis of failed and intact screws.
In March of 1998 a decision was made to completely thaw both units'ce condensers to allow a
thorough inspection and comprehensive repair and restoration activities. In parallel with inspection and repair activities, a review of the ice condenser surveillance and maintenance
- programs, procedures and practices is being undertaken.
This review is intended to ensure that these activities are adequate to provide reasonable assurance of ice condenser operability.
Upon completion of inspection and repair activities, the ice condenser will be reloaded with ice, and ice condenser surveillances will be performed prior to plant startup.
Unit 1
has been selected as the lead unit for ice condenser refurbishment activities and will have first priority
. for resources.
Activity on unit 2 will proceed following unit 1 and will be worked as resources permit.
Ice condenser refurbishment activities will be completed prior to entry into mode 4 '(hot shutdown),
when the ice condenser is required to be operable.
The following paragraphs summarize the key facets of the ice c'ondenser refurbishment project.
I NDEN E W
n
'n n
Pr i n Prior to beginning the thaw of each ice condenser, each unit's containment will be prepared to handle the water from the ice
- thaw, which is estimated to be approximately 350, 000 gallons per unit.
Containment preparations include primarily: removing the lower inlet door shock absorbers; inspecting and sealing the ice condenser floor slab; installing a
temporary ice melt water collection and transfer system; and protecting lower inlet doors from melt-water.
Prior to the initiation of the ice condenser
- thaw, a floor defrost will be initiated to remove ice from and to dry the floor.
The floor will then be inspected to ensure floor seals, which prevent water from entering the ice condenser floor slab, are in good condition. Floor seals will be repaired as necessary prior to the ice condenser thaw.
Actions are being taken to ensure sufficient drying of the floor of the ice condenser prior to again cooling the ice condenser below freezing.
The normal ice condenser drains consist of a series of twenty-
- one, twelve inch drains spaced around the ice condenser floor.
These drains lead to flapper valves that drain to the lower containment.
To facilitate collection of melt
- water, each ice condenser drain will be fitted with a
screen to collect any debris and an inflatable seal plug to prevent drainage of ice melt water to the lower containment.
A series of temporary sump pumps and piping will be installed on several of the seal plugs to transfer melted ice from the ice condenser, through a
containment penetration, to temporary storage tanks in the plant yard.
Melt water will then be
- pumped, using a
second set of temporary pumps and piping, from the temporary storage tanks to the chemical and volume control system (CVCS) monitor tanks for eventual discharge, via the circulating water system in
Attachment 2 to AEP:NRC:1260G7 Page 17 accordance with applicable permits.
The total melt-water removed from ice condenser will be
- measured, by monitoring tank level changes.,
to provide feedback on the quantity of ice in the ice condenser in the as-found condition.
The ice melt-water collection and transfer system will be installed via temporary modifications.
In order to expedite the melt process, a
heat addition system was designed and installed.
N PE I
R P I AND RE BI HMEN B
k Each ice condenser contains
- 1944, forty-eight feet tall, ice baskets that are approximately one foot in diameter and contain borated ice.
During pre-melt ice basket inspections, several conditions were identified including damaged
- baskets, missing or damaged ice basket coupling screws and undocumented ice basket hardware configurations.
These conditions were documented in LERs 315/98-008 and 315/98-032.
Following melt-out of the ice bed, a
combination of internal and external video inspections and visual inspections, including some basket removal, will be performed on the ice baskets to identify damage and to determine whether the configuration of the basket and associated hardware is in accordance with design.
Bottom rims of i.ce baskets will be removed to facilitate inspection and repair of ice basket hold down bar welds.
A definition of detrimental ice basket damage is being developed.
The threshold of detrimental damage will be accepted via the design change process.
Damaged ice baskets outside the definition of "detrimental damage", will be repaired or replaced.
Any identified missing or damaged coupling screws will be replaced.
The hardware configuration of each basket will be documented, and the configuration will be restored to an approved design configuration.
Ice baskets will meet applicable foreign material exclusion requirements prior to refill.
Lwr In D r Each ice condenser is divided into twenty-four bays, each of which contain two lower inlet doors.
The lower inlet doors separate the ice condenser from the lower containment and are designed to open under differential
- pressure, which would be experienced during a postulated
- accident, to admit blowdown into the ice condenser.
The lower inlet doors will be protected from water during the melt-out process and then inspected in place.
Hardware such as door skins, hinges and seals will be examined for signs of distress and addressed as required.
Any repairs that involve restoring the doors to other than the currently approved., design configuration will be authorized via a
design change..
Attachment 2 to AEP:NRC:1260G7 Page 18 w
Inl D
k r
The ice condenser lower inlet doors have companion shock absorbers, each of which currently consists of a
foam wedge enclosed in a fiberglass reinforced polyethylene bag and steel mesh.
The shock absorber is designed to absorb the kinetic energy associated with opening of the lower inlet door during a
postulated
- accident, through crushing of the foam.
'During ice condenser inspections, the shock absorbers were observed to be deteriorated, as evidenced by worn 'areas and tears in the bags and tears in the mesh.
This condition is being documented via LER 315/98-035.
With the exception of the entrance, end wall shock absorbers, the shock absorbers will be removed from the containment to a
lay down
- area, for further disassembly and inspection.
The shock absorber components (bags,
- foam, mesh) will be replaced with a later generation design "air box", which is designed to absorb the kinetic energy of an opening lower inlet door by collapse of the air box.
The new design is considered to be significantly more durable than the original shock absorber design.
This improvement is being effected via a design change.
The end wall shock absorbers will be replaced with new materials of the current design.
rm kD r Each of the twenty-four ice condenser bays contains eight intermediate deck doors that rest on a steel frame just above the ice baskets.
The intermediate deck doors are designed to open due to differential pressure during a postulated accident.
The intermediate deck doors consist of insulating foam within a steel box.
The intermediate deck doors have experienced
- wear, including dents and punctures, during surveillance and maintenance activities.
The intermediate deck doors will be removed from the ice condenser for repair and refurbishment.
In
- general, the doors will either be replaced, restored to original design specifications, or repaired to an alternate design by design change.
Protective covers are being fabricated for these
- doors, to prevent deterioration during future outages.
Each ice condenser bay has a
top deck door that rests on a
structure approximately twelve feet above the intermediate deck doors.
The top deck doors consist of a
framed layer of insulation.
These doors also open following a
postulated accident to provide a path between the ice condenser and the upper containment.
The top deck doors will be inspected in place and hardware such as door fabric and insulation, hinges and seals will be examined for signs of distress and addressed as required.
Any repairs that involve restoring the doors to other than the currently approved design will be authorized via a design change.
Attachment 2 to AEP:NRC:1260G7 Page 19 Sixty air
- handlers, located in the plenum between the intermediate deck and upper deck doors, circulate cool air in the ice condenser.
Outstanding corrective maintenance on air handlers will be reviewed to ensure the air handlers can support the melt-out process as well as future operation.
Following the ice condenser
- thaw, walkdowns will be performed of the air handlers to ensure hardware is in place and functioning, in accordance with design.
I n en r
r re an Mi ellan u
C The ice condenser system,'tructures, and components will be inspected by a multi-disciplined team for integrity and materiel condition.
Discrepant conditions will be documented and dispositioned in accordance with the Cook Nuclear Plant restart plan.
I E NDE E
EL AD Prior to and following the thaw of the ice condensers, debris was identified 'in ice baskets and adjacent flow passages.
These conditions were documented in LER 315/98-017.
Therefore, following inspections, repairs and refurbishment, each ice condenser will be thoroughly inspected to provide assurance that it is free of foreign material prior to reload with fresh ice.
Controls will be implemented to provide assurance that the ice condenser is, and
- remains, free of foreign material during and following the ice condenser reload.
E NDEN VEI E
P The NRC inspection of the Cook Nuclear Plant ice condenser in early 1998 revealed a number of issues related to ice condenser surveillance testing.
Other examples of discrepancies were documented in LERs 315/98-005,-007,-015,-025, and
-026.
As a
- result, the basis for ice condenser surveillances will be reviewed and a surveillance basis document will be develop'ed for each ice condenser surveillance required by the technical specifications.
The surveillance basis document will serve as a
repository for information pertaining to the surveillances, such as basis information, detailed methodology, and assumptions,
- margins, limitations and quality techniques.
Based on the surveillance basis documents, surveillance procedures will be rewritten for the as-left surveillances prior to declaring the ice condenser operable.
b.
Concern 2 as it pertains to the revie~ and assessment of safety evaluations performed under your old 50.59 process.
Provide the details of the review and corrective actions.
The 10 CFR 50.59 program defines the process by which proposed changes to the plant or procedures, as described in the
- UFSAR, are reviewed to determine if they can be implemented without prior NRC approval.
This evaluation requires an understanding of
Attachment 2 to AEP:NRC:1260G7 Page 20 the potential impact of a
change on the design and licensing basis of the facility as described in the UFSAR to determine if an unreviewed safety question (USQ) exists.
During the A/E inspection, concerns were raised relative to the adequacy of our 10 CFR 50.59 program and the potential for inadvertently bypassing this program when making changes to plant
- systems, structures, components, or procedures.
The A/E inspection specifically identified instances where 10 CFR 50.59 reviews were required but not performed, and at least one instance where a
USQ determination was
- required, but not performed.
An underlying cause of these discrepancies, as noted by the
- NRC, was our understanding of what constitutes the plant's
.design basis, the role of the
- UFSAR, and how these are affected by 10 CFR 50.59.
Subsequent to the A/E inspection three (3) self-assessments and one independent contractor audit of our 10 CFR 50.59 program were conducted.
These assessments identified areas requiring improvement, including programmatic improvements.
The first self-assessment was conducted in December
- 1997, and reviewed seventy-one 10 CFR 50.59 screenings and USQ determinations performed between January 1996 and September 1997
'everal issues were identified that were administrative or procedural in nature.
Though discrepancies were identified, these issues were determined to have no impact on the technical conclusions of the evaluations.
The second self-assessment examined 10 CFR 50.59 program effectiveness and was performed in January 1998.
The purpose was to determine if the 10 CFR 50.59 program was adequate to support plant restart.
Two statistically significant samples of 10 CFR 50.59 evaluations performed between 1980 and 1995 were examined.
A key element of this assessment was to examine the rigor and accuracy of the 10 CFR 50.59 evaluations and to characterize the acceptability of a particular evaluation's justification or basis in light of lessons learned regarding the design bases and current regulatory guidance.
This self-assessment concluded that administrative issues associated with program, documentation needed improvement, but no programmatic weaknesses existed that would prevent plant restart.'he third self-assessment was conducted in January-February 1998 to evaluate the potential for other programs or processes to inadvertently bypass the 10 CFR 50.59 program when implementing changes'o the'lant or procedures (e.g.,
failure to recognize change)
This review of a statistically'ignificant sample concluded that previous controls had allowed potential changes to be implemented without the benefit of a 10 CFR 50.59 screening.
However, in no case were 10 CFR 50.59 reviews found to result in any operability or USQ issues.
These potential bypass mechanisms were considered to be administrative/procedural in nature and the assessment concluded that there were no broader safety implications.
As a result of these self assessments a
number of programmatic changes have been or will be implemented, including:
~ J
Attachment 2 to AEP:NRC:1260G7 Page 21 A i
~-
Procedures and familiarization training have been implemented to clarify the de'finitions of
'change,'licensing basis,'nd
'design basis'.
~
An industry expert was 'retained to review the program and procedures, recommend appropriate improvements, and provide training on new 10 CFR 50.59 procedures.
~
Process implemented to communicate management expectations regarding change via the 10 CFR 50.59 process to appropriate personnel.
A i n
~
Revise procedures to address potential 10 CFR 50 '9 bypass mechanisms identified by our internal assessment.
In addition to the three self-assessments, an audit was performed by an independent contractor.
The audit involved an examination of the licensee's self-assessments performed in 1997 and 1998 and a critical review of the quality of past 10 CFR 50.59 screenings and evaluations'he quality was based on the application of current standar'ds for acceptability in performance of the 10 CPR 50.59 products'onsistent with the licensee's conclusions, the contractor determined that none of the sampled screenings and evaluations identified unreviewed safety questions (USQ),
improper screening conclusions, or issues involving equipment inoperability.
Discrepancies in some aspects of 10 CFR 50.59 documentation were noted and enhancements to the 10 CPR 50.59 program were recommended.
The recommendations to elevate the program standards for future 10 CFR 50.59 screenings and evaluations have been implemented.
In summary, multiple examinations have been conducted since the end of the A/E inspection to evaluate the effectiveness of the 10 CFR 50.59 program at Cook Nuclear Plant.
In each of the four examinations, it was concluded that there was a
high probability/confidence level that the nature of identified discrepancies has not resulted in unreviewed safety questions or inoperability of equipment.
Notwithstanding, programmatic enhancements have-been made to elevate our standards and improve communication of the increased expectations to personnel performing future 10 CPR 50.59 screens and evaluations.
c ~
Concern.
3 pertains to engineering calculations.
Please provide the details of the review and assessment performed to date of engineering calculations.
The response should include the population and type of calculations
- reviewed, Justification for the population
- selected, findings, corrective actions, and long-term plan to assure accuracy and quality of engineering calculations at D. C. Cook.
w Attachment 2 to AEP:NRC:1260G7 Page 22 Re n
4c
~BkcCk~
The December 2,
- 1997, letter to the NRC (ABP:NRC:1260G3) described our short-term assessment performed in, response to the CAL. Calculations were identified as a contributor to the issues that arose during the A/B inspection.
As a short-term action,
~
peer group reviews were established to analyze and review
~calculations for issues similar to those identified in the A/E inspection and to determine if they lead to equipment or systems being inoperable.
The issues included questions regarding assumptions, calculation errors, and process measurement effects on instrument calculations.
While AEPNG's review revealed both technical and administrative discrepancies, none were identified that resulted in equipment or systems being inoperable.
The short-term assessment included a
review of twenty system functional calculations from a population of 139 calculations.
These calculations were listed in the design basis documents for.
seven risk significant systems (risk significant as identified in our independent plant examination).
Later it was decided to expand the review to the risk significant systems identified in our maintenance rule program and to have the review conducted by an independent consultant.
The primary objective of the expanded review was to conduct a
systematic and procedurally controlled review to document overall
- quality, level of detail, completeness, conformance to current nuclear industry calculation preparation standards and technical accuracy of the reviewed calculations.
In addition, the review evaluated whether any inoperable conditions resulted.
The calculation review process also included overview and acceptance by a
technical overview committee (TOC) consisting of senior engineering personnel from both the consultant and ABPNG.
The expanded program reviewed a
total of eighty-one system functional calculations, including seventeen of the twenty system functional calculations originally reviewed by the AEPNG peer group (three had been superceded),
an'd sixty-four calculations that were randomly selected from AEPNG design basis documents (DBDs) to provide a representative sample of the total population of ABPNG authored system functional calculations.
The sixty-four 'calculations sampled were selected using a
methodology intended to provide an acceptable level of confidence and reliability that the population did not contain a discrepancy resulting in inoperable equipment or systems..
A sample size of sixty-four calculations out of the total population of 239 system functional calculations selected was utilized to establish the confidence and reliability level.
The plant systems in the sample population were:
auxiliary feedwater, component cooling water, chemical and volume control
- system, containment spray, essential service water, residual heat
- removal, 4kV electrical, safety injection, accumulators, reactor protection
- system, ESFAS, emergericy diesel generator
- systems, control air, plant air, offsite power, 120 VAC, 250 VDC, 600,VAC, non-essential service water, RCS pressure relief, and main steam.
~ t 0
Attachment 2 to AEP:NRC:1260G7 Page 23 The review team consisted of twenty-four engineering personnel from an independent consultant with experience in the mechanical, electrical, instrument and controls and civil/structural disciplines.
As a final step, calculation reviews were overviewed by the TOC.
The purpose of the TOC was to provide oversight of the review
- process, ensure consistency and to provide input on issues raised.
Calculation Review Results The calculation discrepancies identified in the seventeen calculations reviewed by both the AEPNG peer groups and consultant were similar.
- However, because the scope and level of documentation for the two reviews were different, the review observations were not identical.
The AEPNG reviews were primarily focused on identifying technical issues that had the potential to affect equipment or system operability rather than discrepancies affecting the administrative quality of calculations.
Also, minor technical discrepancies were not always documented because 'these type discrepancies were often resolved immediately during the peer group reviews.
The consultant's reviewers documented the results of their reviews using detailed checklists while the AEPNG peer group reviewers typically summarized their observations in a brief e-mail format.
The results and conclusion from the sixty-four calculations reviewed in detail by the consultant were similar to those identified above.
The initial review of eighty-one calculations selected for review in the sample is complete.
Sixty-nine calculations have been through the entire review and commeht resolution process including TOC overview and acceptance.
No discrepancies have been identified which resulted in equipment or 'ystem inoperability.
Only one calculation was identified as, having discrepancies that could have a significant impact on results of the calculation.
- However, in this case, it was determined that the calculation discrepancies would not have affected the operation of the system.
Twelve calculations are in various stages between the comment resolution process and TOC acceptance.
Although nine of these
'alculations have been conservatively designated as having the potential for significantly impacting calculation
- results, none are expected to result in design basis limits being exceeded or system-or component inoperability.
The corrective action program is tracking completion of the review process for these twelve calculations in accordance with the restart plan.
Results of Review AEPNG system functional calculations included a large number of calculations spanning the nearly 30 year history of the plant, and included calculations prepared by several engineering disciplines.
As
- expected, the reviews identified discrepancies that were diverse.
- However, there were several types of discrepancies that were
- common, as follows:
Attachment 2 to AEP:NRC:1260G7 Page 24 Unclear or undocumented calculation purpose or objective at times resulted in confusion as to the intent and use of the information developed in the calculation.
~
Some calculations were not well organized or did not contain sufficient detail for the calculation to be easily understood.
Zn these
- cases, the calculation steps could be difficult to follow.
~
Use of undocumented, not referenced, or out of date design input made some of the calculations difficult to review.
For
- example, parameters were used in some of the calculations without providing a
basis for their validity.
Generally, further investigation or evaluation confirmed that the correct parameters had been
- used, although in some instances it required significant levels of effort to establish this fact.
~
Assumptions were used in some calculations without a
clear statement of why the assumption was acceptable, or conservative.
~
Referenced calculations, drawings and other documents in some instances did not include an indication of their revision or
- date, or that the calculation may have been superseded.
~
Unclear statements of acceptance criteria for the calculation did not clearly demonstrate that the calculated results met the acceptance criteria.
~
Zt was not always clear how or where the results of the calculation were to be used.
~
The calculation process was decentralized and fragmented.
Most administrative discrepancies were related to the level of detail or clarity in the calculations and appeared to be related to the lack of prescriptive direction in AEPNG calculation procedures.
Many of the
. calculation reviews required a
significant amount of time and effort on the part of the reviewers and AEPNG personnel to identify and locate the information required to review the calculation and fully understand its
- purpose, design inputs and results.
These types of discrepancies
~
are correctable through the use of detailed calculation preparation.
standards and procedures combined with an increased focus',
by the" calculation preparers and verifiers, on the requirements for comprehensive documentation o'f calculations.
The technical discrepancies identified in the calculation reviews tended to be specific to the individual calculation.
Most technical discrepancies,
- however, were of low significance levels
'nd were resolved during the review by additional
- research, applying reasonable engineering judgement or by performing simple manual calculations to confirm the assumptions or results.
- However, several calculations required additional levels of
- effort, up to and including recalculation, to resolve apparent technical discrepancies.
Attachment 2 to AEP:NRC:1260G7 Page 25 Pro rammatic Chan es We are currently implementing programmatic changes to address the calculation issues identified in these reviews.
The following actions have been or will be performed to address programmatic issues associated with the calculations:
Com leted Actions:
~
Communicated management's commitment to improving the quality of the AEPNG calculations.
~
A practice has been established to subject new or revised calculations to a
peer or consultant review pending implementation of program enhancements.
Restart Actions:
~
Calculation procedure is being revised to address identified process discrepancies.'
Calculation preparers, verifiers, and approvers are being given formal training in the required elements of an.
acceptable AEPNG calculation.
This training emphasizes the necessary calculation characteristics, using specific examples
~
Enhancing the calculation control and indexing process to provide specific information on calculation status and location.
~
Establish a
program to monitor the effectiveness of calculation process improvements.
~
Resolve remaining calculation issues identified as restart items relating to the independent review.
Post Restart Actions:
~
Upgrade the calculation index to provide more detailed information on the interrelationship of calculations to other plant documents.
~
Benchmark external design organizations for calculation development practices and quality improvement.
Calculation Conclusions The. independent.
reviews performed on a
representative sample covering risk'ignificant systems (i.e.,
identified in our maintenance rule) are intended to provide reasonable assurance that the calculations performed in the past by AEP will not lead to inoperable conditions.
As noted previously, calculational activities are on-going.
Issues identified as a result of these activities will be dispositioned in accordance with the Cook Nuclear Plan restart plan.
Attachment 2 to AEP:NRC:1260G7 Page 26 Concern 4.
Please include the NPSH calculations for all safety-related pumps.
Describe the calculational technique and all assumptions'used in the calculations.
Re on to 4d We have reviewed our calculation files to provide assurance that we have acceptable calculations documenting adequate NPSH for the safety-related pumps.
In accordance with your request, we have provided NPSH calculations (listed in Table
- 1) for the safety-related pumps where such calculations are applicable.
Included in the calculations are the techniques and assumptions used in their performance.
Certain safety-related pumps do not utilize NPSH calculations and are therefore not included in this submittal.
These include:
~
Essential service water (ESW) pumps that are wet pit design that is subject to submergence considerations rather than NPSHA.
~
The reactor coolant system (RCS) pumps do not have an NPSH calculation since their safety function is pressure boundary only.
~
None of the pumps associated with the operation of the emergency diesel generators have NPSH calculations as they are typically flooded suction, positive displacement or have only a pressure boundary function.
~
The post accident containment hydrogen monitoring system (PACHMS) pump is a
vacuum pump for pulling containment air into the hydrogen analyzer.
No NPSH calculation is required for the vacuum pump.
Attachment 2 to AEP:NRC:1260G7 Page 27 Table 1
List of NPSH Calculations Attached to this Letter Submitted as Proprietary Information f
~m~n PP-10 PP-26 PP-3 PP-35 PP-4 SI n 'lNm Component Cooling Water Pump Safety Injection Pump Motor Driven Auxiliary Feedwater Pump Residual Heat Removal Pump Turbine Driven Auxiliary Feedwater Pump l l i ENSM970919AF ECCS Recirculation Phase ENSM970128AF, Rev.
2 ECCS Injection Phase NEMP950501JEW, Rev.
0 HXP791121AF, Rev.
1 ECCS Recirculation Phase ENSM970128AF, Rev.,2 ECCS Injection Phase NEMP950501JEW, Rev.
0 RCS HXP900904JEW, Rev.
0 HXP791121AF, Rev.
1 PP-46 Boric Acid Boric Acid Storage NESP032395JJS, Rev.
1 Tank Transfer Pumps PP-50 PP-9 CVCS CTS Centrifugal Charging Pump Containment Spray Pump ECCS Recirculation Phase ENSM970128AF, Rev.
2 ECCS Injection Phase NEMP950501JEW, Rev.
0 VCZ NESM961021AF, Rev.
0 CCP ENSM720719FK, Rev.
1 ECCS Recirculation Phase ENSM970128AF, Rev.
2 ECCS Injection Phase NEMP950501JEW, Rev.
0
P
Attachment 2 to AEP:NRC:1260G7 Page 28 B ~
Concern 5.
Please provide the actions taken to assure the accuracy of the February 6,
- 1997, response to the NRC request for information pursuant to 10CPR 50.54(f) in light of the inspection findings from the design inspection in September, 1997 and the follow-up design inspection in April 1998 R
one 4
The lessons learned from the AE inspection and subsequent inspections have enhanced our understanding of the design and licensing bases and the processes used to maintain
- them, as originally described in our February 6,
- 1997, response.
Following the A/E inspection and subsequent shutdown of both units in September of 1997, the NRC issued a confirmatory action letter (CAL) that led us to evaluate the applicability of the results and discrepancies identified during the inspection to other systems and components throughout the plant.
In addition to the issues identified in the CAL, several new issues axose concerning our containment systems.
In response to these
- issues, we are performing a
comprehensive assessment to provide reasonable assurance of plant system readiness, programmatic readiness, functional area readiness, and containment readiness.
The primary mechanism implementing this assessment is the Cook Nuclear Plant restart plan (previously submitted in AEP:NRC:1303)
The restart plan describes the activities and controls that are intended to ensure the plant is ready for safe start up and power operation.
The details of these readiness reviews have been discussed in detail in previous sections of this letter (attachment 2),
in, response to the specific concerns raised in the 2 '06 petition.
Additionally, as we progress toward restart, many of these issues will be discussed further with the NRC during our ongoing 0350 meetings.
In addition to the readiness assessments and supporting activities described in our restart plan, we have also initiated a revised design basis reconstitution program.
The purpose of this program is to provide assurance that:
~
there is an adequate understanding of, and contro1 over, the plant's design, and licensing basis requirements; and
~
requirements-are being effectively implemented both in plant design'nd. in the procedures that govern plant operation and maintenance; The design basis reconstitution program is an ongoing effort that will continue after startup of the units.
In summary, our actions to date, as described in the preceding sections of attachment 2,
have served to enhance our understanding of our licensing and design bases as discussed in our February 6,
- 1997, response.
In addition, AEPNG is performing a
comprehensive assessment of
- system, functional
- area, and programmatic readiness reviews.
Issues identified will be dispositioned in accordance with the Cook Nuclear Plant restart
C
Attachment 2 to AEP:NRC:1260G7 Page 29 plan; NRC permission for restart will not be requested until the restart plan is complete and reasonable assurance of restart readiness is achieved.
C J
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