provides justification for allowing continued short term plant … · 2012. 12. 4. · this letter...

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ACCESSION NBR:9608150120 DOC.DATE: 96/08/08 NOTARIZED: NO FACIL:50-305 Kewaunee Nuclear Power Plant, Wisconsin Public Servic AUTH.NAME AUTHOR AFFILIATION

MARCHI,M.L. Wisconsin Public Service Corp. RECIP.NAME RECIPIENT AFFILIATION

Document Control Branch (Document Control Desk)

DOCKET # 05000305

SUBJECT: Provides justification for allowing continued short term plant operation w/pin hole leak in MFW chemical injection line.

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11 ENCL . 10

NRC- 96-076

WISCONSIN PUBLIC SERVICE CORPORATION

August 8, 1996 10 CFR 50.55a

U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Ladies/Gentlemen:

Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Justification for Continued Operation

This letter provides justification for allowing continued short term plant operation with a pin hole

leak in the main feed water chemical injection line. The leak is located in the chemical injection

soc-o-let (S-O-L) to the 1A main feedwater line weldment, refer to Figure #1 of Attachment 2.

The basis for this justification is provided in Attachment 1.

The leak was initially detected at approximately 0000 hours on August 6, 1996, when the plant was at 98 percent power. Since the line was insulated with asbestos materials, the exact location

of the leak could not be immediately identified.

The following morning, insulation removal was started and the leak was determined to be

located in the chemical injection to lA main feedwater line weldment. Once the leak was

identified the chemical injection line was verified to be secured and was then danger tagged.

Action was also taken to identify the applicable code requirement and to characterize and size

the flaw. Examination techniques included:

Radiography Ultrasonic Test Visual Examination Magnetic Particulate Examination Thermography

As a result of these exams, WPS has initially determined the root cause of the steam leak is

undercutting/lack of fusion and porosity in the soc-o-let weldment. Our assessment of the flaw determined that continued operation of the plant does not significantly impact the health and

safety of the public or place personnel at the plant at increased risk. This assessment is based on the following:

N:\home\mronski\rr2-1.wp

Kewaunee Nuclear Power Plant * North 490, Hwy.42 * Kewaunee, WI 54216-9510 * (414) 388-2560 9608150120 960808 PDR ADOCK 05000305 p PDR

* 0 Document Control Desk August 8, 1996 Page 2

1. There is sufficient weld material to ensure the continued structural integrity of the soc-o

let. A conservative review of our pipe stress analysis determined that 30 percent weld loss could occur without catastrophic chemical injection line failure.

2. Based on the NDE characterization of the flaw, the flaw is not expected to grow prior to the start of our scheduled refueling outage on September 20, 1996. At this time the

weld will be examined for root cause determination and repaired in accordance with the

Section XI requirements.

3. The area has been roped off to prevent unnecessary personnel access.

4. A failure of the chemical injection line is bounded by current accident analysis and, therefore, for the much larger main feedwater line break. This will not place the plant in an unanalyzed condition.

5. A temporary engineered restraint of the soc-o-let will be installed on the line prior to

August 10, 1996, subject to NRC review. The restraint has been engineered to ensure

structural stability of the chemical injection line and limit leakage if a complete failure

of the weld were to occur. Attachment 1 provides a description of the restraint

preliminary design.

Additionally, WPSC commits to the following actions:

1. If a plant shutdown or trip occurs prior to our scheduled outage in September, a code qualified repair will be performed.

2. The weld will be checked daily by quality assurance and once per shift by operations to

identify increased leakage. If leakage increases, an assessment will be made to determine the appropriate action.

3. When the permanent repair is made, a root cause analysis will be performed and corrective actions will be taken to preclude recurrence of the event.

The Plant Operations Review Committee has reviewed the justification for continued operation

and will review the final design of the temporary engineered restraint.

The bases for the justification for continued operation and the proposed temporary modification

were discussed with the Nuclear Regulatory Commission staff on August 7 and 8.

If you need additional information, please contact my staff.


Document Control Desk August 8, 1996 Page 3

Sincerely,

M. L. Marchi Manager - Nuclear Business Unit

RPP/dlf

Attach.

cc - US NRC - Region III US NRC Senior Resident Inspector Mr. Lanny Smith, PSCW


ATTACHMENT 1

To

Letter from M. L. Marchi (WPSC)

to

Document Control Desk (NRC)

Dated

August 8, 1996

Justification for Continued Operation


Document Control Desk August 8, 1996 Attachment 1, Page 1

INTRODUCTION

This attachment provides the basis for our Justification for Continued Operation (JCO). The attachment is divided into the following sections:

1. Components Affected

2. Section XI Requirements

3. Justification for Continued Operation a. NDE Results, b. Analytical Evaluation/Root Cause, c. Consequences of Failure, d. Alternative Considerations,

4. Preliminary Design of the Temporary Restraint

Section 1 - COMPONENTS AFFECTED

A single line diagram of the feedwater line and chemical injection connections is provided in Figure #3 of Attachment 2.

The 3/4 inch chemical injection line is seldom used for secondary chemistry control; therefore, continued plant operation will not be hampered by maintaining the system isolated. Furthermore, there are no Technical Specifications on its operability and it is not assumed to

be required by the USAR.

The feedwater system has three safety significant functions, they are:

1. Provide feedwater isolation during a secondary line break inside containment,

2. Containment Isolation, and

3. Not to rupture and result in a high energy line break outside of containment.

Items 1 and 2 are discussed in the following paragraphs, while item 3 is discussed in section 3.c.

Feedwater isolation for the A header is provided by:

1. A trip of the main feedwater pumps in conjunction with closure of motor operated valve FW-12A, or



2. Closure of air operated valves FW-7A and FW-10A.

Although not covered by Technical Specifications, the electrical components of the valves have been qualified for the harsh environment that would occur if the feedwater line were to rupture. Further, the feedwater pump motor breaker is not located in the area of the leaking weld. Therefore this function of the system is still operable.

Containment Isolation is discussed in section 5 of the USAR. The feedwater lines are classified as class 4 penetration. Class 4 penetrations are described as:

"Normally operating incoming and outgoing lines which penetrate the Reactor Containment Vessel and are connected to closed systems inside the Reactor Containment Vessel, and which have a low probability of being ruptured by an assumed accident are provided with at least one remotely-operated valve outside of the Reactor Containment Vessel.

Steam Generator secondary side isolation valves receive special treatment because their function is not containment isolation for a loss-of-coolant accident, but only containment isolation for a main steam line rupture within containment"

As discussed previously the operability of feedwater isolation function is not affected by this failure and therefore the design basis of the system is maintained.

In conclusion, there are no Technical Specifications which apply to the affected components and all USAR assumptions continue to be valid.

Section 2 - SECTION XI REQUIREMENTS

Section XI requires that the feedwater system piping may be inspected doing periodic system pressure tests. This component was most recently inspected in 1993.

Article IWA-5250 "Corrective action" specifies that, "The source of leakage detected during the conduct of a system pressure test shall be located and evaluated by the owner for corrective measures as follows: .... (3) Repairs or replacements of components shall be performed in accordance with IWA-4000 or IWA-7000 respectively." These articles require the repairs be performed in accordance with the owner's design specification and the original construction code at the component or system. Not withstanding that the leakage was not detected during a system pressure test, WPSC has elected to follow NRC guidance (GL 91-18) and apply this requirement to the leakage. In accordance with IWB-3142.4, "Acceptance by Analytical Evaluation," WPSC has performed an analytical evaluation to demonstrate the component's acceptability for short term continued operation. The calculation is included in Section 3 of this Attachment. For longer term operation, WPSC proposes to install a temporary restraint of the class 2 portion of



the chemical addition piping that attaches to the main feedwater system as described in Section

4 of this attachment. WPSC will submit the proposed final restraint design as a separate

transmittal.

Section 3- JUSTIFICATION FOR CONTINUED OPERATION

al NDE RESULTS

The branch connection configuration and location of the leakage area is shown on Figures 4 and

5 of Attachment 2. The branch connection consists of a carbon steel (ASTM A-105 Grade 2), 3/4" extra heavy soc-o-let, welded to a carbon steel (ASTM A-106 Grade B seamless) 16" diameter schedule 140 pipe. The original fabrication of the P-1 to P-1 fillet weld was manually made by the SMAW process with E-6010 filler material used for the root pass and E-7018 used for the remainder. The fabrication process required a liquid penetrant examination of the final

weld surface.

To assess the structural condition of the branch connection, several nondestructive techniques

were used. These techniques included a visual inspection of the leak, a straight beam ultrasonic inspection, radiography, magnetic particle examination, and thermography.

Visual Inspection

A visual inspection revealed that the area of leakage is located in the weld adjacent to the toe

of the weld cap. This area is indicative of cluster porosity with leakage occurring at one or two

pores of less than 1/16" in diameter. Seepage from the indication appears to stream horizontally out from toe of the weld. With the insulation removed, the volume of leakage is not sufficient to condense into visible water droplets beneath the piping.

Ultrasonic Inspection

A straight beam ultrasonic examination of the pipe wall surrounding the branch connection and

adjacent areas of the feedwater piping was performed to identify if any areas of significant erosion or corrosion from inservice operation exist. The location of the measurement and the

resultant wall thickness are shown in Figure 6 of Attachment 2. Based on a nominal wall thickness for schedule 140 piping of 1.4538", the results clearly indicate no evidence of erosion or corrosion of the feedwater piping.



Radiography

Radiographs were taken to assess the structural condition of the pressure boundary materials. The orientation of the radiographic source and film is shown in Figure 7 of Attachment 2. Due to the geometry of the weld joint, penetrameters are not feasible for these exposures. Performance of an examination of 100% of the adjacent feedwater piping was not possible due to the operation of the system and geometric constraints. These radiographs do however show:

1. No evidence of corrosion or erosion in the soc-o-let. 2. No evidence of corrosion or erosion in the feedwater piping. 3. No evidence of circumferential indications in the feedwater piping 4. No evidence of defects affecting the structural integrity of the weld joint.

Magnetic Particle Inspection

A magnetic particle inspection was performed on the feedwater piping and soc-o-let weldment. This inspection found no linear indications or other areas of degradation.

Thermography

An infrared thermographic inspection was performed to depict the formal profile of the leakage area. Figures 8 and 9 of Attachment 2 provide a visual depiction of the small area of steam emitted. The thermography indicates no additional areas of leakage.

b) ANALYTICAL EVALUATION/ROOT CAUSE OF FAILURE

Operation of the chemical injection piping has been reviewed to determine potential loads that likely contributed to the degradation of the 3/4" on 16" x/hvy S-O-L weldment. The configuration of the piping is such that it shares common piping supports with another chemical injection line and is connected to piping that provides auxiliary feedwater during startup and shutdown operations. Through most of the life of the plant this chemical injection line has been isolated, however it was used for phosphate injection during the first operating cycle. Under this situation, the phosphate injection temperature would have been at ambient while the feedwater would have been around 450 *F. The resultant delta temperature of approximately 380*F would have caused high stresses in the area of concern. The other chemical injection line is used to inject boric acid into the secondary side of the steam generators and operates continuously. KNPP has been using the other chemical injection line for boric acid injection over the last five cycles. At times the chemical injection piping with the leak is known to have been subjected to pipe vibration; however, the exact cause is not fully documented. It could be possible that vibration of the chemical injection piping occurs during boric acid injection or during auxiliary feedwater operation. The total time of auxiliary feedwater operation since 1979 was recently estimated at 2981 hours. Degradation of the 3/4" on 16" x/hvy soc-o-let weldment


Document Control Desk August 8, 1996

Attachment 1, Page 5

is believed to have been caused by the combination of the resulting bending stresses associated

with pipe vibration and thermal stresses caused by the large delta temperature. Upon discovery of the leak, KNPP took appropriate steps to minimize these loads by isolating both of the

chemical injection lines since they share some common hangers. Future vibration of the

chemical injection line should now only occur during shutdown when the auxiliary feedwater

system is started. Historically, no vibrational problems have been encountered at this time. This

leaves only a pressure stress which is essentially constant from approximately 700 psi to 1200

psi (100% to 0% power).

With knowledge of how the piping system was operated, an upper bound crack growth rate is

established. The most severe load is judged to be the thermal stresses. To be conservative, through wall cracking is postulated to have occurred in approximately six months of thermal

loading during phosphate injection. This assumption of time is far less than the actual loading time associated with boric acid injection, phosphate injection, and auxiliary feedwater operation.

The original weld thickness is postulated to be 0.75 inches which was taken from measurements

on recently exposed radiographs and is consistent with design requirements. The resultant crack

growth rate is 0.00017 in./hr. (0.75 in./4320 hrs.).

Visual examination conclusively revealed that the flaw is a small cluster of porosity with one or

more small round holes less than 1/16 inch in diameter located in the 3/4" on 16" x/hvy soc-o

let weldment. Radiographic examination has shown that degradation is minor in nature; no evidence of any circumferentially oriented extent of cracking was noted on the film. The film

confirms that degradation is caused by porosity. Radiography did not reveal any evidence of

lack of structural integrity; no degradation of the feedwater piping or 3/4" on 16" x/hvy soc-o-let was observed. All degradation was observed to be in the 3/4" on 16" x/hvy soc-o-let (S-O-L) weldment. Magnetic particle examination confirmed that the external surfaces of the feedwater

piping and 3/4" on 16" x/hvy S-O-L weldment are free of linear indications. Straight beam

ultrasonic examination of the 16 inch diameter feedwater piping produced readings consistent

with original design values verifying that the degradation has not extended into the feedwater

piping. This information suggests that area of degradation is small with respect to the total weld

volume and weld length. Degradation was probably initially caused by a small amount of

undercut/lack of fusion in the shielded metal arc 6010 root weld and porosity/inclusions in the

weld which were aggravated by the high thermal and vibrational loads.

The postulated failure mechanism for this type of situation is believed to be leakage similar to

what is currently being observed. Stains on the pipe insulation indicate that very small amounts

of leakage have probably been present for some time. The volume of leakage is miniscule. No

dripping or water can be seen on the pipe or in the area. The small amount of leaking steam

evaporates readily once it comes in contact with the external surface of the feedwater piping. No evidence of steam cutting can be observed in the feedwater piping. Since the high thermal

and vibration loads have been minimized, essentially no driving force exists to propagate a crack

in the near future. In fact, this situation would remain stable for years.



Under this situation, at approximately 400'F and system pressure without fatigue, the long term failure mechanism would be limit load. Experience has shown that the critical flaw size under this situation is approximately equal to the O.D. pipe diameter. The O.D. of the 3/4" on 16" x/hvy S-O-L is approximately 2.5 inches. As stated above, radiography confirms that the extent of degradation is small with respect to the weld length. A thru-wall crack having a postulated length of 0.5 inches would grow to 2.5 inches assuming a crack growth rate of 0.00017 inches/hour under operation with high thermal stresses and vibrational loads after approximately 1.3 years of loading.

During the next month and a half, the 3/4" on 16" x/hvy S-O-L weldment will only see system pressure while the loads of concern have been eliminated except for minor pipe vibrations which may be encountered during operation of the auxiliary feedwater system which is connected to the chemical injection piping. Limiting plant operation to six weeks ensures a factor/margin of safety of 11.6 assuming an upper bound crack growth rate. Since the loads of concern have been eliminated, the real margin of safety is much greater than 11.6.

c) CONSEQUENCES OF FAILURE

As previously discussed, the catastrophic, sudden failure of the chemical injection line is not expected, however, an evaluation of the potential consequences of the complete failure of the chemical injection line was performed. This evaluation included the effect on other plant equipment, the effects on reactor safety, and effects on the USAR safety analysis.

1) Effect on plant equipment USAR Chapter 10 analysis of feedwater pipe breaks conservatively bounds the potential failure of a 3/4 inch branch connection to the main feedwater piping. Circumferential breaks in branch runs greater than 1" were considered with the break area equivalent to the pipe diameter.

The feedline break analysis reviewed all needed equipment to handle the transient and its location relative to the feedwater piping location. The needed equipment provides for reactor trip, capability to maintain hot shutdown after the break, and ultimately achieve cold shutdown. All needed equipment was shown to be protected from adverse environment, or qualified to operate in conditions which meet or exceed those from the break. Feedline breaks in the compartment of interest were analyzed. The consequences of the pipe break on plant equipment are therefore shown not to present any unanalyzed challenges to plant safety.

2) Effects on reactor safety The failure of a 3/4" branchline form the main feedwater header will not result in a significant challenge to reactor safety. The failure will result in an increased demand on feedwater for the faulted train. The operators have multiple indications which will



provide a prompt indication of the problem. Any significant steam leakage will be detected by the steam exclusion system and annunciated in the control room. Mismatch in feedwater flow between the generators will alarm the steam generator tilt annunciator. Should the feedwater system be unable to maintain an adequate supply to steam generator and simultaneously feed the break, additional alarms on steam generator level and steam flow > feed flow will provide an alert to the operators. The operators will take actions to maintain adequate feed flow (reduced power) or trip the unit if deemed necessary. Upon a reactor trip, main feedwater will be terminated and the auxiliary feedwater system will supply feedwater to the steam generators. The plant will respond as it would for any normal reactor trip since main feedwater is no longer required.

3) Effects on USAR Safety Analysis The USAR Chapter 14 safety analysis addresses the loss of normal feedwater in section 14.1.10. The unit is designed to withstand a complete loss of main feedwater from full power conditions under a number of conservative assumptions. The complete loss of normal feedwater does not result in any adverse conditions in the core, because it does not result in water relief from the pressurizer safety or relief valves, or uncovering of the tube sheet of the steam generator being supplied with auxiliary feedwater. The leakage resulting from a 3/4 inch branch line does not alter the conclusions of the USAR bounding analysis.

d) ALTERNATIVE CONSIDERED

Due to the design of the Chemical Injection System, a Section XI code-qualified repair is impractical at power without increasing the potential of challenging plant safety.

fl Plant shutdown

To perform a code repair of the repair of the weldment leak requires that the plant be placed in a cold shutdown condition. This is because the repair would require cooldown and depressurization of the 1A Feedwater line. At this late point in the fuel cycle, substantial boron will be required to achieve the cold shutdown condition with a resultant large use of water to dilute for the return to criticality. The relativity small excess reactivity present at this point in the operating cycle will extend the unit startup. The minimum time to perform the repair has been estimated to be at least 3 days. In addition, the shutdown will result in an unnecessary transient on the facility system and components.

21 Continued Operation as Justified by this Evaluation

This document provides sufficient justification for continued operation until the fall refueling without impacting the health and safety of the public or plant personnel.



Section 4 - PRELIMINARY DESIGN OF THE TEMPORARY RESTRAINT

Reference Figure 10 in Attachment 2.

Carbon steel plate, angles, and U-bolts are used to fabricate a restraint to hold the fitting in place should the weld connecting the fitting to the FW pipe fail. The design completely surrounds the 3/4" Chemical injection pipe with a carbon steel plates which form a collar around

the 3/4 " pipe. The hole for the 3/4" pipe includes a chamfer to include the fillet weld, connecting the 3/4" pipe to the fitting, within the collar. Two (2) carbon steel angles, alternately welded and bolted to the plates or collar halves will extend, perpendicular to the FW

pipe longitudinal axis, beyond the O.D. of the FW pipe. Two (2) U-bolts are connected to the

angles, perpendicular to the pipe longitudinal axis, to attached the restraint to the FW pipe. Any force resulting from the socket is constrained by the restraint which will hold the fitting into the

FW pipe wall.

Design considerations include:

The review of the FW piping analysis of record to reconcile any additional loads placed on the

FW pipe.

Loads on the restraint due to normal operation, seismic, and accident conditions will be

incorporated into the analyses.

Clear angles of vision to permit complete visual inspection of the fitting to FW pipe weld.

The design and installation will be performed in accordance with the Temporary Change

procedures, which includes configuration control, safety analyses, second level reviews and

appropriate work control procedures.


ATTACHMENT 2

To

Letter from M. L. Marchi (WPSC)

to

Document Control Desk (NRC)

Dated

August 8, 1996



LIST OF FIGURES

Figure # Description

1 Cut Away View Pre-Assembled Components

2 Drawing of Pipe, Soc-o-Let Welds

3 Feedwater One Line Diagram

4 Leak Locations

5 Picture of Feedwater Pipe Near Leak

6 Feedwater Line Wall Thickness Measurements

7 Radiography Geometry

8 Infrared Picture of Leak Area

9 Infrared Picture of Leak Area

10 Temporary Restraint Preliminary Design Sketch

The following drawings are not referenced but are provided for your information:

11 Chemical Injection Piping Isometric

12 Chemical Injection Piping Isometric

13 Texas Bending Company Drawing


3/1" Sok-O-Let Connectorne

ick

Note Hol'e is FW line has bed

3/1" hemical Injection Line

nector nto place,

Cut Away View Pre-Assembled Components

FIGURE 1

15" Feed 1,438"'

3/i"

15 Feed

IA38"

ASSEMBLY PROCESS:

1, Sok-O-Let connecto 2. Hole drilled throuc

manufactured hole) 3. Chemical injection I

/ater hick

welded to FW connector (en

id through F W e welded into I

Sok-O-Let Connector

/elds

3/" Cemica In jection Line

pinhole leak

Cut Awayline

place

ViewPost-Assembled Components

FIGURE

MAIN FEEDWATER SYSTEMSTRAICHTENER

W -5-2

FW-31

SAMPLE

AUX FEED

FEED HEATERS

IA CDSR FW PUMP RECIRC

I COSR

FIG URE 3

F112A FWTA

FW PUMP

CD SYS & HD PUMPS

WISONSIN PUBLIC SERVICEcORATION RE.ORG

KEWAUNEE NUCLEAR POWER PLANT ULTRASONIC, LIQUID PENETRANT, MAGNETIC PARTICLE AND VISUAL EXAMINATION LIMITATION TO EXAMINATION RECORD

Feetwa.Th.I Fftow co taa.A. HT9S To SAhOteMO SYSTEM OR COMPONENT:- LLLS lassabe cwTAT

DRAWING NO.: ISTI1- 9I1 56.1 C N I " Saoao o

COMPONENT IDENTIFICATION:\ %U Sc k 14o ?%fe PROCEDURE: ME~lo.155 REVISION: -Rat6

ULTRASONIC: LIQUID PENETRANT:

EXAMINER: YCEVA

MAGNETIC PARTICLE: VISUAL: X

LEVELDATE:

EXAMINER: NA LEVEL

SKETCH TO PROVIDE: APPROXIMATE SIZE, LOCATION, ORIENTATION, TYPE OF LIMITATION AND PERCENTAGE OF REDUCED EXAMINATION COVERAGE.

~'A'

1*1

KEWAUNEE NUCLEAR POWER PLANT REVIEW:

~~1

AUTHORIZED NUCLEAR INSERVICE INSPECTOR REVIEW:

FIGU RE

DATE:&'#If a

DATE:

q

DATE: NA

;oo&4 , ,,

LO LJ

cc

FIGURE 6

VAL7--7

I -~

A

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Description: P-.W. 1,"# PIrE pous4 o0 vwFt-f

Weld Proc. Nos.

Flow Dwgs. Physical Dwgs.

159- 1

Ihim Ot fl~rrantiAn_____________________ Of___ _ D .:a--- ' * t-*-

E-QA Table (1C,3A)i

O-1tem No.

-Weld No.& Symbol

Pneu Hydro PT MT RT UT

7Bill of material

KNPP Weld Map

Dwg. No.

AI

WR No._ ABI No._

8X/b knoq T 5iAJ.- E YPOSuRE

g-7-q, &. loai5

Claar AD se- ,--7r-,

.I

cs

FIGURE

0

7

B aEWAUNEE NUCLEAR POWER PLANT

Infrared Thermography

Thermographic Indication Report No. 96-03

Thermographer Mike Peotter F Report Date 8/7/96 Survey Date 8/7/96 Survey Time 6:19:12 PM

400.0

380.0

320.0

300.0

280.0R

260,0

2400 0

*<229T I

Thermal Image Visible Image

FIGURE 81

TherMonitor 95

A04.TIF

FLGURE TherMonitor 95

98/7/96 5:25:40 PMirgfrarnet~rio9

U-Bolts Loop around FW line

* -16" FW line (not to scale)

Sok-O-Let

nut

U-Bolt

Cc

Angle Iron

Restraint Plate

3/4" Chem. Inj, line

ntact point on weld

Partial Cut away view of Restraint Installed

FIGURE 10

i I1 1110 i, !I: -- -- -- -- --

xvz'-

FIS URF 11

CONT. OF POINT C

C1l22A

% " Chem. Inj. Line

16" Feedwater Line

Cll214

FICURE'

520

VI

p~P AT 43r44

-FOR INFORMATION ONLY

NO FUTUR E REVISIONS WILL BE MADE TO THIS DRAWING.

" FWAUNIFL UNiT I" CHEMICAL INJECTIOW PIPINc AT

Pia sVALV a OCjvb.mb 7.D Pckw Sn o sll s~i o

12

I.

A 12~:a3 fl~~fl5 2D. EVE A~1~c CHG D. NOMEN. PE DAlE 7

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1184"s PAINTING o. OOUSIOP.

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