ML20004E834

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Forwards Info Re Seismic re-evaluation Program,In Response to NRC .Analysis & Mods of Anchorage of Emergency Power,Distribution Sys & Instrumentation Will Be Completed by End of Next Refueling Outage in Early 1982
ML20004E834
Person / Time
Site: La Crosse File:Dairyland Power Cooperative icon.png
Issue date: 06/12/1981
From: Linder F
DAIRYLAND POWER COOPERATIVE
To: Crutchfield D
Office of Nuclear Reactor Regulation
References
TASK-03-06, TASK-3-6, TASK-RR LAC-7590, NUDOCS 8106150355
Download: ML20004E834 (73)
v • d • e


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l o DA/RYl.AND hhk COOPERAWhovst7 2615 EAST AV SOUTH LA CROSSE WISCONS:N 64601 e

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In reply, please

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refer to LAC-7590 84 hy LOCKET NO. 50-409 g

0) p Director of Nuclear Reactdts k?

ion ATTN:

Mr. Dennis M. Crutchfield Operating Reactors Branch #5 Division of Operating Reactors U. S.

Nuclear Regulatory Commission Washington, D. C.

20555

SUBJECT:

DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)

PROVISIONAL OPERATING LICENSE NO. DPR-45 SEP TOPIC III-6, SEIS"IC DESIGN CONSIDERATIONS LA CROSSE

REFERENCES:

(1)

NRC Letter, Crutchfield to Linder, Dated April 24, 1981.

(2)

DPC Letter LAC-7484, Linder to Crutchfield, Dated April 23, 1981.

(3)

DPC Letter LAC-7181, Linder to Crutchfield, Dated October 14, 1980.

Gentlemen:

Your letter, Reference (1), requested DPC to provide a 30-day response to expand on the La Crosse BWR's seismic re-evaluation program.

DPC perconnel attended a meeting on this subject in Bethesda, Maryland, on May 19, 1981; at that time W.

Russell, Chief, SEP Program Branch, granted an extension on the response to June 15, 1981.

DPC has provided its seismic re-evaluation methodology for structural and piping system analysis in Attachment 1.

Reference (2) previously discussed the methodology and status of LACBWR electrical equipment anchorage.

During the May-June 1981 maintenance outage, additional work was accomplished.

has been updated to reflect the recent progress on the electrical equipment.

Reference (3) summarized DPC's historical approach to the seismic evaluation _of the La Crosse site since 1973..To date, DPC has undertaken substantial ef forts to evaluate the integrity of the plant and systems essential to safety and safe shutdown. is DPC's proposed schedule to complete the analytical portion of the seismic re-evaluation program.

8106150 1 p

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.q Mr. Dennis M.

Crutchfield LAC-7590 Operating Reactors Branch #5 June 12, 1981 Due to recent discussions with the NRC staff, a resolution to the site specific liquefaction question and an apparent assignment of a

.12 g ground level acceleration, the scope of the seismic te-evaluation program has increased from the previously submitted plan.

However, DPC will attempt to complete the increased analytical work in the same time frame identified in Reference (3) i.e. October 1982.

DPC is continuing the analysis progran and has commenced adding piping restraints and electrical equipment anchorage where the analysis has indicated the need for additional support. projects a tentative schedule for completing the modifications as identified in the analysis.

DPC will continue the modification program while the remaining analysis are in progress to move the seismic program forward.

In this manner, advanced planning can optimize the installation of modifications.

Without the analysis of all systems and structures completed, it is difficult to accurately estimate the total scope of time, materials, and plant conditions required to fully implement the necessary nodifications.

The estimated completion date of January 1, 1985, will be revised at the conclusion of the analytical evaluation portion of the seismic tack.

This overall approach will permit a large and complex task to be completed in an orderly, economical, and radiologically acceptable manner.

Operation of LACBWR is justified in the interim.

The analysis and modification of the anchorage of emergency power, distribution systems, and instrumentation is underway and will be conpleted by the end of our next refueling outage in early 1982.

DPC has completed 22 of 31 of the High Pressure Core Spray and associated System restraint modifications and will complete the remaining work by the next refueling c utage.

It is anticipated that 6 of the 9 remaining restraints will be added by September 1, 1981.

With this system seismically hardened, there will be a high assurance of system operability if the postulated seismic event occurs.

As a result of the NRC seismic evaluation of LACBWR, a new Emergency Service Water Supply System has been added to the LACBWR facility.

This system will provide an independent and reliable backup cooling water source to the reactor with its supply being obtained from the Mississippi River.

Although, the NRC has assigned a return period of 1000 to 10,000 years for a peak acceleration of.12 g, Technical Reference 10 in Reference (3) determines the site specific return frequency to be more likely between 10,000 and 100,000 years.

Based on the original LACBWR design, the continuing efforts to improve supports and anchorage and probability of occurrence, DPC feels continued operation, is j ustified.

f

+f Mr. Dennis M.

Crutchfield LAC-7590 Operating Reactors Branch #5 June 12, 1981 If there are any questions concerning this topic, please contact us.

Very truly yours, DAIRYLAND POWER COOPERATIVE kAadv$

Frank Linder, General Manager PL:RMB: abs ATTACHMENTS CC:

J. G. Keppler, Reg. Dir., NRC-DRO III Resident Inspectors i

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The original study was performed by Gulf United Nuclear Fuels Co. and was issued in 1974 (Reference 1).

(This company is no longer in business).

Analysis methods and criteria for this study are discussed in Section (A) below.

A second, on-going study, is currently being performed by Nuclear Energy Services, Inc.

Analysis methods and criteria for this evaluation is presented in Section (B).

A)

Original Study (1974) by Gulf United Structural Information a)

Description of Component Reactor Containment Building b)

Modeling Techniques (See Figure I-1)

Structural Damping - 7 percent SSE, 3 percent OBE Mathematical Model - stick beam Mass Distribution - lumped mass (36 masses)

Model Degree of Freedom D c)

Seismic Analysis Method Dynamic Method - timd history Selection of Significant Modes - first three modes analyzed Relative Dicplacement - N/A Modal Combin=. tion - N/A 3 Components Input - only one horizontal acceleration applied Floor Spectra Generation - for RCS subsystems Peak Broadening - N/A Load Combination - seismic only d)

Analysis Criteria Codes - ACI (1971) Concrete AISC (1970) Steel e)

Computer Codes Code - SIM (no further information was presented)

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II.. Genoa Unit 3 Stack The Genoa Unit 3 Stack was originally analyzed for seismic loading by Gulf United, and the results were reported in Reference (1).

Section (A) below describes analysis.

methods and criteria used for this study to the extent'that they can be determined.

(This company is no longer in business and detailed back-up information for the analysis is not available).

This structure was re-analyzed by Nuclear Energy Services in 1980 and the results of this evaluation were documented in Reference (2).

Details of the analytical methods and criteria used in the recent NES study are included below in Section (B).

A)

Original Study (Reference 1) by Gulf United a)

Description of Component Genoa III Stack b)

Modeling Technicues Structural Damping - 7 percent Mathematical Model - stick beam (See Figure II-1)

Mass Distribution - lumped mass (29 masses)

Model Degree of Freedom D c)

Seismic Analysis Method Dynamic Method - time history Selection of Significant Modes - first 10 mode up to 24.92 Hz Relative displacements - N/A Modal Combinations - N/A Three Component Input - one horizontal acceleration applied Floor Spectra Generation - N/A Peak Broadening - N/A Load Combination - N/A d)

Analysis Criteria Codes - not available e)

Computer Codes Code - SIM (no further information was presented)

Error in Assumption According to plant drawings, there are no_ piles under the Genoa III stack as assumed in the Gulf United Study.

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This evaluation was made for the SSE event only (See Fi ~ ure II-2 for structure description).

Structural Damping 7 percent for SSE event from NRC Reg. Guide 1.61.

Seismic Input The site specific spectra developed for LACBWR was used for this analysis (Figure II-3).

This was scaled from 5 percent damping to 7 percent damping by use of co-efficients presented in NRC Reg. Guide 1.60.

A three-dimensional model was used with two horizontal and one vertical input.

The vertical input responses were found by using the procedure of NRC Reg. Guide 1.60 (for frequencies from 0.25 Hz to 3.5 Hz the ratio varies from 2/3 to 1.

For frequencies greater than 33 Hz, maximum ground acceleration is used.

The combining of modal responses and spatial components were completed as per NRC Reg. Guide 1.92.

The responses of medes up to a frequency of,35 Hz were used in the analysis.

Soil Structure Soil-structure interaction effects were' examined by the use of an inertial interaction analysis.

Kinematic interactions effecrs were neglected.

The dynamic forces applied to the structure are simply the product of the design ground motion and the mass of the structure.

Since the loading is applied only on the structure, the soil is replaced by springs ( impedances-) and prescribed design motions are applied directly at the support for the springs.

For spring constants (see Figure 11-4).

To account for any variation in the soil, a range of soil modulus were used in the analysis.

The soil modulus was taken as 1000 KSF and 3000 KSF.

The actual modulus is expected to be about 1500 to 1600 KSF from existing soil data.

Model Three-dimensional lumped mass model (35 masses) as shown in Figure II-5.

No eccentricities were used in the analysis.

q w.

w-iw--

+,

g

_ ~ _

- 500 FEET TALL-i

'l 17.42' y

I I

)-

i STEEL. STACK LINER i

A 7" REINFORCED CONCRETE

\\

%u_/ /

t 4

I 1

f 4

e

-24" ISINFORCED CONCRETE i

/

y 1

j CONCRETE PEDESTAL w7 I

lOCTAGONALMATFOUNDAT 1 ".!

"*'# # M f-

?l '.i t*

,,1... 7.ll >;.
  • t .,.'*..M:i :c:

4'MI M L,r,.., L.

n r. :.., * -

. s o. -

i 26.25'

-38. 98'

-c I

75.0' w

FIGURE II-2 SCHEMATIC SKETCH OF GENOA 3 STACK-

'~ ~ ~'-

(NES-1980 STUDY, REF. 2) wr

.*7 v5

=-

M e m pa y.----

g piiyr ge s r-99 g,sug--ra

-,y m p m e gr W.-.-

p w e -.y er--W yw-Wy vy up y v-est we*v t--e-twwem rt y-y yye.---wwg=-w-&q-y g g m= sy-r m*-7Piy+----rwvm-sem waa-

LACCi$C ggg _

i i

E p

N

~

e,,,

/

A

/

4

!V

,('g g j

~

4 xi xi\\

y'%,

l i

q y

-[.

l f,

\\

,~.

V I

/!

/

0

/ Nj Ni

,rNg<a u

e, 9

x "a

7 A

NA

/

e, C.C I C. I L.0 tC "

l PERICC sic.

5. ca. :i r.g c

F.IGURE II-3 P00R[ RENAL

t Table 30-13. Spring Constants for Rigid Circular Footing Resting on Elastic Half-Space Mccon Spring Constant Reference Vertcal A,, 46'.

Tirnoshenko and Goodict (1951) 1-t Horcental

,,,32(I - y)Cr.

Bycroft (1956) 1 - Br Roc 14g 8Gr!

Borowicka (1943) k,= 3(I - t)

Torse.n A e== 8 g,s Reissner and Sagoci (1944)

~ 2(1 + y) 2.0 i

!-2r -4 f.9

++++e+U H

i.e 1-hvas N 4

1.7 l

(ol 1.6 4 Gr.

i k se g

y, I

l.5 h2r.-d 8.4 wt rp-g,3

\\.

s. 2'-

(b) 1.1 l

Figure 1017; Efrect of ep:5 of I~o e

i i

i i

i O

O.4 0.0

8. 2 f.6 2.0 2.4 2.8 3.2 3.6 embedment on the spring con-

/

j stant for ver.ecally loaded circular i

Ocpth f o Rodius Rotso, H/r.

footings (from Kold lon,1969).

f t

l l

FIGURE II-4 SOIL SPRING CONSTANTS j

P00R ORGMA l

v

'3 1

A.

sciisi.o' gog vogts y;

z x

4 80

.,. -0)

L

{ f^ DYNA.MIC DEGREES OF FREEDOM (Tip) g, i

4 ?o' 3

T 4

8 e 9t

,___I 3 M'

?1 u

12 o I3 o n-o

/I--il I:

  • i i

ai s4 i

-2 4',

-4

$~

1

o a

)

2/ f i

/aS'

{2j m

N 2

~T i

24 ?

)

25 t j')O' k?

27 4 b

z= i

/

ki @

/

j 3msuT e' l

rat.641.c b

lh, te i

/..]

v>~~<

vswv&C<rd y

' ),@ 0

\\

,s.o' _

v. e
,y.

ST AC.K MODEL FIGURE II-5 MATHEMATICAL MODEL OF GENOA 3 STACK P00R ORGINAL y

e 4

Foundation The 75' octagonal mat foundation was analyzed by using the resulting loads from the stack model.

It was analyzed using a finite element model.

(see Figure II-6).

Computer Analysis 4

The Dynamic Modeling of the Genoa III Stack was completed using MRI/STARDYNE 3 (c).

The following programs were used:

STAR, Dynre 4, Lanczos and Post.

The static Modeling of the 7 5' octagonal mat was comp'.cted using ANSYS computer program. of Swanson Analysis Systems, Inc.

Loads The following loadings were considered:

Nl X2+X2+X2 Live Load + Dead Load ;

2 3

(seismic)

-e

.e e

f w

i l

i X

L W

id I

$W

[\\

/\\

\\

/

I

\\

\\

=

I

\\

t-

'x ",\\ a\\ d e l e /M'M'N eN s'Vs\\gy

/s/ "4

\\

{s t

/,5

~

s

/N a

/,-

\\

i 5/ 9o O

s\\ * \\ N '

/

E d

!~

_~

$ Y:; *$

)

f s

l

  • \\

'/*m us

/

s c

e T

j k

e l

4 4

g g

g i

t G

N

/

s

[~r EMI-f f> ' /I/

~

t o

5 I\\

g m

3 \\ \\To NN b S,,/ s 0

e 5

N

/ f e E \\[Y, i' yi 5

E i

i

'/ = / ;

E \\XY 5

5 E

g E

.iO

~

e I

E E

E i

i \\'

\\ S E /

.e \\

a v

5/

s

/

L^

<l

\\., R

~

f

'/ e/ 2 c \\ a \\"

1 FIGURE II-6 GENOA 3 MAT FOUNDATION FINITE ELEMENT MODEL P00R ORIGINAL

III.

LACBWR Stack The LACBWR stack was originally analyzed by Gulf United (Reference 1).

For historical purposes, available information describing analysis methods and criteria is summarized in Section (A) below.

Section (B) addresses current-analyces.

A.

Original (Gulf United) Study a)

Description of Components LACBWR b)

Modeling Techniques Structural Damping - 7 percent Mathematical Model - stick beam Mass Distribution - lumped mass Model Degree of Freedom D c)

Seismic Analysis Method (See Figure III-1)

Dynamic Method - time history Selection of Significant Modes - first 10 mode -

24.92 Relative Displacements - N/A Modal Combinations - N/A Three Component Input - One Horizontal Acceleration Applied Floor Spectra Generation - N/A Peak Broadening - N/A Load Combination - N/A d)

Analysis criteria Codes - not available e)

Computer Codes Code - SIM (nc further information was presented)

B.

Current LACBWR Stack Studies An evaluation of the LACBWR stack will be made in accordance with the enclosed schedule.

The overall analysis method will be similar to that applied to the Genoa III stack.

The foundation will be treated differently, however, since the LACBWR stack has a pile foundation, unlike the Genoa III stack.

As a result, foundation analysis for the LACBWR stack will be similar to that applied to the reactor containment building discussed previously.

Q

-- 1000 o 23 4.5* R e 6" Wall

+> 22 o 21

20

-- 900 o 19 o ig i

o 17 o 16

-.- S00 o 15 Reir. forced l

Cor crete 5

/ Chimney 3

u g U

o 12

-- 700 o 11

" 'O Stack i

1 11* R p m 15" Wall l

5'

g. a a c.rzg o9 i

l '5,'/d

  • oB l fI j o7 Piles

. 5 h Pile Foundation

-- 600 lll Cluster "5

i l l o4 o3 8

o2 e

aI 1

o 20' R 4

- - 540 (a) Structure (b) Stick Model Configuration FIGURE III-l MODEL FOR DYNAMIC ANALYSIS OF LACBWR STACK (FROM 1974 GULF UNITE TU Y)

/

W lW

-. _j IV.

Piping Seismic & Stress Analyses

-1.___

A)

Some in-containment piping systems have 41 ready been analyzed by Nuclear Energy Services, InC.

.Results have been reported in References 3 through 7 A summary of results is presented in Table 2.

A discussioh' of analysis methods, analysis criteria, and assumptions for the completed analyses is presented below.

The piping systems that were analyzed were the main steam, feedwater, recirculation, and the HPCS discharge and suction.

It was realized early in the analysis that lateral support would be required, therefore, all analysis included the addition of supports where judged to be necessary.

Since the analysis of these five piping systems took place in 1974 to 1975, all seismic input was taken from the 1974 Gulf United Report (Ref. 1).

Analyses were completed prior to Reg. Guide 1.92.

The seismic loading on the Piping System was taken to be the absolute combina-tion of one vertical and one horizontal event.

The remain-ing horizontal and the vertical were then evaluated.

Stresses on the pipe were expected to be within the appro-priate allowable for each load condition.

The loading conditions included:

1.

Dead Weight & Other Sustained Mechanical Loads 2.

Internal Pressure 3.

Thermal Loading 4.

Seismic Loading (a) Seismic (OBE =

SSE)

(b) Anchor Movement The horizcntal SSE acceleration spectrum used was that corresponding to the appropriate elevation in Ref. 1 (see inditidual report).

For the OBE event, SSE was used.

The vertical response spectrum for the SSE loading was taken as 2/3 of the horizontal SSE ground response spectrum assuming no amplification of vertical response in the structure.

For the OBE earthquake, the vertical piping response spectrum is taken as b of the SSE vertical response spectram.

Damping values used were 1 percent for the OBE and 2 percent for the SSE.

The spectra are presented in the appendix of each report.

i n

Loadings were combined in order to fulfill Class I or i

Class II ASME Criteria (EQ 8-9-10-11).

Seismically induced anchor movements were estimated by calculating low frequency displacements from the containment building response spectra at the different anchor point elevations.

Support flexibility was accounted for by the use of springs as non-rigid supports (frequency below 33 Hz).

Eccentric masses were introduced in the systems to account for the weight of such items as valve controllers, operators, or hand wheels.

Each piping stress report contains tables showing eccentric masses placed on the system.

All snubbers will be designed to have a frequency of greater than 33 Hz.

For locations where rod hangers exist and the rod was found by analysis to have a compression force, the support was designed to be capable of resisting that compressive force without buckling.

Where supports were modified, no changes or reruns of the piping analyses were made to determine the effect of change i.' stiffness of the cupport en the pipe strerses.

3)

Current Piping Syster Stress Analysis Ef forts NES is currently analyzing the 14" ven: line from the shutdown condenser, the Manual Depressurizati. system piping and the piping for the' dedicated safe sautdown system (High Pressure Service Water te Auxiliary Core Spray System).

The 1974 Gulf United Report is still being used as the source of seismic input for the ongoing piping system analyses, since NES has not yet developed new floor spectra fror its three-dimensional model.

The seismic l

input frcm the Gulf United Report (1974) is being used in three dimensions (two hori: ental and 1 vertical).

Thc2e inputs are being used in accordance with NRC Reg. Guide 1.92.

Since the spectra of the Gulf United Report is more severe than the site specific spectra,

/

NES feels that the results will remain conservative.

A representative comparison will be made to ensure this fact.

All pipe supports (HPCS suction and discharge) are being designed using the results of existing analyses.

The reaction forces due to the two seismic loading conditions are being added by use of SRSS technicues.

This means that the reactions of the absolute additions cf the vertical plus one horizontal is squared and added i

to the square of absolute addition of the vertical l

plus the remaining horizontal.

The square root'of that l

addition is then used to design the support.

i 1

III.

STRUCTURAL INTEGRITY OF MECIIANICAL AND ELECTRICAL COMPONENTS,

~2 PIPING AND SUPPORTS COMPONENT MODELING ANALYTICAL ANALYSIS COMPUhER DESCRIPTION TECllNIQUES PROCEDURE CRITERIA CODES Piping The following techniques were A finite element Dynamic modal The computt utilized for each piping system dynamic analysis responses were code used 5 analyzed.

with X, Y,

& Z combined by the PIPSD, devt spectra applied SRSS method, ed by URS/

Eccentric masses simultaneously stress limits John Blume Mass distribution was used to are from ASME Associates Support flexibility analyze each B&PV, Section contained C piping system.

III for classes CDC cybernt Response spectrums corres-Structural damping listed.

systems.

ponding to the following values of 1% and elevations were used:

27, were used for OBE and SSE respec-tively.

Main Stear-664.5' OBE-SSE Class - 2 Feedwater 664.5' OBE-SSE Class - 2 FIPCS Suction 745'-700' OBE-SSE Class - 1 FIPCS Discharge 695' OBE-SSE Class - 1 Recirculation not OBE only Class - 2 given

V.

TURBINE BUILDING As for the other structures, the Turbine Building had been analyzed to some extent by Gulf Unitad in 1974 (Reference 1).

Section (A) below describes analysis methods and criteria used for this study to the extent that they can be determined.

(This company is no longer in business and detailed back-up information for the analysis is not available.)

A)

Original Study (Reference 1) by Gulf United a)

Description of Component Turbine building b)

Modeling Techniques (see Figures V-1 thru V-3)

Structural damping - 7 percent Mathematical model - stick beam Mass distribution - lumped mass Model degree of freedom - single DOF c)

Seismic Analysis Methods Dynamic method - response spectrum Selection of significant modes - not available Relative displacement - N/A Moda; combination - not available Three. component input - one horiLontal accelera-tion applied Floor spectra generation - N/A Peak broadening - not available Load combination - seismic only d)

Analysis Criteria Code - ACI e)

Computer Codes Codes - not available B)

Additional Work to be Performed NES has not yet analyzed the Turbine Building.

NES is currently developing an approach to this analysis.

It is felt that the methods used in previous (Gulf United) analyses are over-simplified and must be up-graded.

Details of the proposed analysis methodology, criteria and assumptions will be made available to the NRC as soon as they become fully developed.

i 1

700 220K o (floof) t

~k = 1730 K/11 G8G-7%

(Vlater Tank) o G43 K 423 K o

i j

+

1

~k = 1730 K/11

{

GGil n

D V

~k = 1730 K/11 I

I

~k = 1730 K/11 CA O mm mm

+

27'-9 22'-3 4

20*-0 e

F C

D A

"E

..k*=

- 577 K/11 3

1 Column Line 1 k'

W2=

= 32G

  • total W

f = 27 - 2.9 cps co FIGURE V-1 DYllAMIC t10 DEL OF COLUMN LINE 1 ib

~

![!!!!

!!ilib

.i 700 o 220 K (Roof) 220K

~k - 1730 K/11 C8 G-71/16 Q<>

D

~k' G84

~k = 1730 K/11 GG8 mm mm, 24*

2G' 20*

1 1

. k' = 1730

,-r 2

F D

D A

4

- 83 k/11 k

Column Line 10 w2.

- 1528

  • to tal i=

= 6 2 cps i

t l

1 PIGURP, V-2 DYliAMIC MODEI, Ol' COLUMN LINE 10 co I

U1 e

t b

i

=E

'5_h5 Elev. 686 o 1000 K

-k = 164.3 k/11 "R/C Columns Elev. 640 wssxxN.

wsxxx mm Elevation of Turbine Support

2. 1=63.5 W

M 2' X 2' 2' X 2' x

m.,

A h

Cru f = 27

  • l 27 *PS 25' 3' X 2.25' 3' X 2.25' T

M T

4_.%

_?_*,'.

c 35*

l Plan View of Turbine Support t

FIGURE V-3 DYNAMIC MODEL OF TURBINE SUPPORT o

a~.?.

. i.i.i.i.i.i.i.i.d

. 6

i VI.

SPENT FUEL RACKS & FUEL POOL New, high density spent fuel racks have recently been installed-and licensed.

The following references should be consulted for design and analysis details:

1.

Structural Analysis Design Report for the Lacrosse Boiling Water Reactor High Density Spent Fuel Storage Racks, prepared by Nuclear Energy Services, Inc.,

Danbury, CT (NES Report No. 81A0546, 1978).

2.

Structural Analysis Report for the Lacrosse Boiling Water Reactor Spent Fuel Pool Structure, prepared by Nuclear Energy Services, Inc., Danbury, CT (NES Report No. 81A0095, 1978).

+

VII.

"B" DIESEL GENERATOR BUILDING Ref. " Description and Design Criteris -for Diesel Generator Building and Secondary Emergency 0:4-Site Electrical Power System LACBWR - Genoa Station - t' nit 2" dated July 31, 1974 (Docket 50409-174).

This report set forth the criteria to be used for design of the "B" Diesel Generator Building.

NES is attempting to learn from Sargent & Lundy Engineers the scope and results of any seismic analysis performed during the building design phase.

When this information becomes available, it will be reviewed to determine if the SEP requirements have been met.

r l

l l

r.

l

I REFER 2NCES 1.

" Seismic Evaluation of the Lacrosse Boiling Water Reactor" January 11, 1974 (Docket 50409-172) Gulf United Services.

2L.

" Seismic & Structural Analysis of the Genoa 3 Stack Using the NRC Site-Specific Ground Response Spectra, November 20, 1980, Nuclear Energy Services, Inc., Danbury, CT (report 81A004 0).

3.

" Seismic and Stress Analysis of LACBVR Feedwater Piping System", June 18, 1975, Nuclear Energy Services, Inc.,

Danbury, CT (report 81A0087).

4.

" Seismic and Stress Analysis of LACBWR Main Steam Piping System", August 1, 1975, Nuclear Energy Services, Inc.,

Danbury, CT (report 81A0088).

5.

" Seismic and Stress Analysis of LACBWR Recirculation Piping System", November 17, 1975, Nuclear Energy Serviccs, Inc.,

Danbury, CT (report 81A0089).

6.

" Seismic and Stress Analysis of the LACBWR High Pressure Core Spray Suction Line Piping System", Nucle-Energy Services, Inc., Danbury, CT (report 81A0090).

7.

" Seismic and Stress Analysis of the LACBWR High Pressure Core Spray Discharge Line Piping System", May 10, 1977, Nuclear Energy Services, Inc., Danbury, CT (report 81A0091).

l l

5 4

KEY TO ATTACHMENT 2 1'

. EXAMPLES OF TYPE '.UPPORT

. 1.

Bolted to Equipment 2.

Bolted to Concrete Wall 3.

Bolted to Ccncrete Slab 4.

Bolted to Block Wall 5.

Welded to Embedded Channel 9

REFERENCE OR STATUS A.

To Be Evaluated B.

Evaluation in Progress C.

Attachinent No. 2, LAC-7484 D.

To Be Modified E.

Modification Commenced F.

Intentionally Blank

- 0.

NES-Task-063 H.

Test Data On File at Site I'

I t

m-~

e..,,

-..--c..~..

,m-

,-,.,..--mm

-m%,-,,,,

-.w..

-.w-,.-

-,.-.~,, - - --

b

SUMMARY

OF INVESTIGATION OF ANCHORACE AND SUPPOR) 0F SAFETY RELATED ELECTRICAL EQUIPMENT AND l' EMS THAT MAY DAMAGE THIS EQUIPMENT ITEM 5 THAT CDU!.D PDTEMIIALLY INTERNALLY ATTMHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. Of SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT WPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCE 1/1/80

& I.D.

SUPPORT EVAL.'D NAME 8 I.D.

SUPPOET EVAL.*D CONCLUSION Turbine Bldg.

Contatraent Turbine Bldg.

'l A

MSIV Drive Isolation Grade Floor Motor 647' Binchboard Control Turbine Bldg.

3,5 huclear Instr.

1 C

B

  • E" Room Control Room Indicators Log 668' Co*unt Ch. 1 43-38-804 Period Ch. I 1

C 42-38-805 l

Log Count Ch. 2 1

C 42-38-806 f,

Period Ch. 2 1

C 42-38-807 j

Log N Ch. 3 1

C 42-38-808 Period Ch. 3 1

C 42-38-809 Log N Ch. 4 1

C

(

42-38-810 Period Ch. 4 1

C 42-38-811 i

I E

i h

.L

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF i,

SAFETY REL ATED ELECTRICAL EQUIPMINI AND IitMS THAT MAY DAMAGE THIS EQUIPMENT ITEMS THAT COULD POTENTIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. OF l

SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MDDIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLOG. & ELEV. SUPPORT SINCE 1/1/80

& I.D.

SUPPORT EVAL.'D NAME A I.D.

SUPPORT EVAL.'D CONCLUSION Benchboard Nuclear Instr.

1 C

"E" (Cont 'd)

Indicators Power Level 1

C Ch. 5 42-38-812 Power Level 1

C Ch. 6 42-38-813 Power Level 1

C Ch. 7 42-38-814 Power Level 1

C Ch. 8 42-38-815 i

(

i 6

I I

.[

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTRICAL EQUIPPENT AND ITEMS THAT MAY DAMAGE THIS EQUIPPENT ITEMS THAT COULD PDTENTIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH TH15 EQUIPMENT

!.0. OF SYSTEM IN WAS ANCHORAGE WAS

_WAS DOCLPMENT EQUIPPENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE DE SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. 8 ELEV. SUPPORT $INCE 1/1/80 8

.D.

SUPPORT EVAL.'D NAME 8 I.D.

SUPPORT EVAL.'D CONCLUS!0F Panel G Control Turbine Bldg.

3,5 Nuclear Instr.

1 C

R Room Control Room Ch. 7 668' 42-38-507 Nuclear Instr.

1 C

Ch. 8 42-38-508 Scaler Source 1

C Range Monitor Ch. I 8 Ch. 2 42-38 509 Radiation 1

C Monitoring Pecorder 45-43-801 Containment None C

Radiat f ori Montitors 73-43-501 73-43-502

^

73-43-503 l

)

i

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF t

SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT ITEMS THAT COULD POTENTIALLY I

INTEltNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. OF SYSTEM IN WAS ANCHORAGE WAS WA5 DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLE0 BLDG. 8 ELEV. SUPPORT SINCE 1/1/80 8 I.D.

SUPPORT EVAL.'D NAME & I.D.

SUPPORT EVAL.'D CONCLUSION Panel F Control Turbine Bldg.

Nuclear Instr.

1 C

B R.om Control Room AGS Channels 668' (Rear Panel F)

Contairrnent 1

C Vessel Level Power Supply 37-42-401 ATWS Relay 1

C 50-42-602AT ATWS Relay 1

C 50-42-604 AT ATWS Relay 1

C 50-42-601AT ATWS Relay 1

C 50-42 712AT ATWS Relay 1

C 50-42-605AT

, L

f t

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT ITEM 5 THAT CDULD PDTENTIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. OF i

SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT j-EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. 8 ELEV. SUPPORT SINCE 1/1/80 8 f.D.

SUPPORT EVAL.'D NAME 8 I.D.

SUPPORT EVAL.'D CONCLU$l0N I

Panel E Control Turbine Bldg.

3,5 Nuclear Instr.

1 C

B Roan Control Room Dual Pen Rec.

668' Power Range Ch. 7 8 8 42-38-803 Reactor Water 1

C Level #2 Power Supply 50-42-401 Reactor Water 1

C Level #2 Remote Amp 1tfier 50-42-303 i

AT..i Relay 1

C 50-42-503 AT Reactor Water 1

C Level Pemote Anipli fier 50-42-302 Power to Flow 1

C Square Root Converter 2A 50-37-503 e k

n t

I l'.

1 i

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT ITEMS THAT COULD POTENTIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMFNT I.D. OF

. SYSTEM IN WAS ANCHORAGE WA5 WA5 DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG & ELEV. SUPPORT SINCE 1/1/80 4 1.0.

SUPPORT EVAL.'D NAME A 1.D.

SUPPORT EVAL.'D CONCLUSION-Power to Flow 1

C Panel E Square Root (Cont'd)

Converter 2B 50-37-504 Power to Flow 1

C Square Root Converter IA 50-37-501 Power to Flow 1

C Square Root Converter 19 50-31-502 Nuclear Instr.

1 C

Dual Pen Rec.

Ch. 3 & 4 42-38-801 Nuclear Instr.

1 C

Dual Pen feet.

Wide Range Ch. 5 & 6 42-38-802 Reactor Flow 1

C Recorder 50-37-801 n

h I

i

?

0

$UMMARY OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEM 5 THAT MAY DAMAGE THIS EQUIPMENT I TEM 5 THAT CDULD POTENTI ALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THl5 EQUIPMENT I.D. OF SYSTEM IN WAS ANCHORAGE WAS WAS

. DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT.

TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLOG. & ELEV. SUPPORT $1NCE 1/1/80

& I.D.

SUPPORT EVAL.'D NAME & !.D.

SitPPORT EVAL.'D CONCLUSION Panel E Reactor Water 1

C i

(Cont'd)

Level #3 Power Supply 50-42-403 Reactor Water 1

C Level #3 Level Indicator 50-42-811 Reactor Water 1

C Level #1 P/S 47-85-406 (Dist. Panel)

Reactor Water 1

C Level #2 indicator 50-42-802 Reactor Water 1

C Level #1 Rec.

50-42-801 l

i

~

7-m

!t i

SUPNARY OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY REL ATED ELECTRICAL EQUIPEENT AND ITFMS THAT MAY DAMAGE THIS EQUIPMENT ITEMS THAT CDULD POTENTI ALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT

!.D. OF SYSTEM IN WAS ANCHORAGE WAS WR5 DOCUMENT EQUIPMENT WHICH LDCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME

!.D.

INSTALLED BLDG. 8 ELEV. SUPPORT SINCE 1/1/80 A I.D.

SUPPOR T EVAL.'D NAME & I.D.

SUPPORT EVAL.'D CONCLUSION '

Safety Panel Panel D Reactor Turbine Bldg.

3.5 Rx Water Level 1

C 8

Protection Control Room

  1. 1 Safety 668' Drawer 50-42-501 Rx Water Level 1

C

  1. 2 Safety Drawer 50-42 502 Rx Pressure #1 1

C Safety Drawer 63-35-501 Rx Pressure #2 1

C Safety Drawer 53-35-302 P/F #1 Safety 1

C Drawer 50-37-505 P/F #2 Safety 1

C Drawer 50-37-506 a

I l '

f it

i c

l i

i I

l

-[

l

SUMMARY

OF IN'.'ESTIGATION OI ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTRICAL ECUIPMENT ANFs ITFM5 THAT M E D VV.E-THIS EQUIPMENT

~

ITEMS THAT COULD POTENTI ALLY INTERNALLY ATT ACHED COMPONENTS INTEPACT WITH THIS EQUIPMfNT I.D. Of SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPMENT WHICH LOCAil0N TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPOPTING i

EQUIPMENT NAME I.D.

INSTALLED BLDG. A ELEV. SUPPORT SINCE 1/1/80 A 1.D.

SUPPORT EVAL.'D NAME 8 f.D.

SUPPORT EVAL.'D ' CONCLUSION l

l Panel D Control Turbine Bldg.

3,5 Alternate Core 1

C R

Roon Control Roon Spray Low Flow 668' Alann 38-37-601

[

ACS Flow 1

C 1ndicator-36-37-801 ACS AC Valve 1

C Motor Control Switch ACS DC Valve 1

C Fot or Cor:f rol Switch l

C ont airnc n*

1 C

4 Vettel tr:;rN!

I

' rot t tre Irmitti.Lo, 37-35-315 F t d '*!% 'I' 1

C C:

I 7(5501 lF. fee'r.nl

!'r m sare Indicator U-E-310 9

r.

t a

.s t

(-

f

$:I' SUMMARf 0F INVESTIGATION OF ANCHORAGE AND SUPPORT OF I

SAFETY REL ATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT ITEMS THAT COULD POTENTIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT

!.D. OF SYSTEM IN WAS ANCHORAGE WAS WAS DOCIPENT EQUIPENT WHICH

' LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME 1.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCE 1/1/80

& I.D.

SUPPORT EVAL.'D NAME & l.D.

SUPP0fti EVAL.'D CONCLUSION Pcnel D Cont airunent 1

C (Cont'd)

Vessel Internal Pressure Power Supply 37-35-401 Containment 1

C Vessel Liquid Level Selector Switch 37-31-701 Containment 1

C Vessel Licuid Level

  • ndicator 37-43-801 Reactur Plant 74-92-002 Essential Turbine Bldg.

3 D

Br eders, Pet *y C

SK-5101-003 Dettery Power Electrical Peters

-7 CNrger Equipm nt 4o09 El. 654' Reactor Plant 74-91-0U1 Essential Turbine Bldg. Sitting No Emergency 1

A SK-5101-063 Battecies Power Electrical On t a,ap

-6 E quip.nent Battery R oo, E 1. 6S4 ' Racir

.I e'

f

SUMMARY

OF INVESTICATION OF ANCHORAGE AND SU? PORT OF I

SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEMS THAT HTY DAWAGE THis EQt!!PMENT 1

T ITEMS THAT CDULD POTENTI ALLV INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. OF F

SYSTEM IN WAS ANCHORAGE WAS WAS DOCUME NT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TTPE OF SUPPORT SUPPORTING EQUIPMENT NAME 1.D.

INSTALLED BLOG. & ELEV. SUPE 'RT SINCE 1/1/80

& I.D.

SUPPORT EVAL.'D NAME & I.D.

SUPPORT EVAL.*D CONCLUSION Reactor Plant Essential Turbine Bldg.

3 E

SK-5101-063

-6 Battery Rack Power Electrical l

Equipment Room E1. 654' l

I 4

i

3 t,

i I

i

SUMMARY

OF INVESTIGATION Of ANCHORACE AND St!PPORT OF l

SAFETY RELATED ELECTRICAL EQUIPMENT AND ITIMS THAT MAY DAMAGE THIS EQUIPMENT I

ITEMS THAT COULD POTENTIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT

!.D. OF SYSTEM IN WAS ANCHORAGE KAS WAS DOCUMENT' EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF-SUPPORT SUPPORTI NG EQUIPMENT NAME I.D.

INSTALLED BLOG. & ELEV. SUPPORT SINCE 1/1/80 A l.0.

SUPPORT EVAL.'D NAME 8 f.D.

SilPPORT EVAL.'D CONCL USION Ginerator Essential Turbine Bldg.

3 0

Input A Output I

C SK-5101-06J Plant Battery Power Electrical Breakers

-7 Charger Equipment Room 654' G:nerator Essential Turbine Bldg. Sitting No SK-5101-063 Plant Power Electrical On

-2 Betteries Equipment Battery Room 654' Racks Generotor Essential Turbine Bldg.

None E

SK-5101-063 Plant Power Electrical

-2 Batterj Equipment Racks Roon 654'

)

I e (

I

.{

".ttMMARY OF INVESI!GATION OF ANCH0RACF AND SUPP0PT OF g

SAFETY REL ATED ELECTRI( Al EQUIP"ENT AND ITEMS THAT MAY PAMAGF THIS EQUIPMENT ITEMS THAT COULD POTEMIIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT 1.0 Of SYSTEM IN WAS A'4CHORAGE WAS WAS DOCU"ENT EQUIPMENT WHICH LOCATION TYPE OF M00!FIED EQUIPMENT NAME TYPE JF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME 1.D.

INSTALLED BLOG. 8 ELEV. SUPPORT SINCE 1/1/80

& !.D.

SUPPORT EVAL.'D NAME A I.D.

SUPPORT EVAL.'D CONCLUSION l

Recctor Relay Reactor Turbine Bldg.

3 0

Relays C

SK-5101-063 Cabinet Protection Electrical

-1 Equipment Room 654' Turoine Bldg.

Essential Turbine Bldg.

3 D

Rx Plant 1

C Ladder to SK-5101-063 MCC 1A Power Electrical Battery BRD

-1 Equipmer'.

Charger Bkr.

Room 654' 480-120 Volt C

Breaker G.P. Main C

Steam Shutoff Valve nreaker 64-11-U03 Control Rom, 1

C rnery ary L lyiit i ng.E,k e.

O ACS AC M.D.

I C

^

Valve Breaker 30-30-001 Turbine Bldg.

1 C

170- Volt Dist. Panel 7_ _ -__ -

' 3=

5 _

k

.t..j i

1 t

i-

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY REL ATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS EQUIP

  • TENT l

L ITEM 5 THAT CDULD PDTENHALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. OF SYSTEM IN WAS ANCHORAGE WAS WA5 DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPPENT NAME TYPE OF SUPPORT.

TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. 8 ELEY. SUPPORT $1NCE 1/1/80

& I.D.

SUPPORT EVAL.'D NAME A I.D.

SUPPORT -EVAL.'D CONCLUSION Turbine Bldg.

Turbine Ridg.

I C

MCC 1A 120-Volt Reg (Cort'd)

Dist. Panel Turbine Bldg.

Essential Turbine Cidg.

1 R

120-Volt Bus Power Electrical Aux. Dist.

Equipment Panel Rom 654' s

i i

I

-i i

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT P' SAFETY ret ATE D Elf CIRICAL EQUIPMFNT AND ITEMS THAT MAY DAMAGF IRl! EQUIPPENT ITEMS THAT COULD POTENTIALLY INTERNALLY ATTACHED CDMPONENTS INTERACT WITH THIS EQUfPMENT I.D.- 0F SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF.

MODIFIED EQUIPMENT MAME TYPE DF SUPPORT TYPE OF SUPPOPT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLOG. A ELEV. SUPPORT SINCE 1/1/80 A I.D.

SUPPORT EV'.t.'D NAME & I.D.

SUPPORT EVAL.'D CONCLUSION t

IB Non-Essential Electrical 1

0 Tie to 18 1

C SK-5101-063 I nt errupt ible Power Equipment Static Inverter

-1 Bus Breakers Room El. 654' Breaker Panel Main Control 1

C BB0 "E" Main Control 1

C BD. D Breaker W/D 41-503/04 Radiation 1

C Monitor Panel G3 Breaker Safety Syster.

1 C

Panel D2 (41-503761)

(41-5039ng)

Creater Nuclear instr.

1 C

Panel G1 (41-503306)

Breaker

.l

i I

i I.

?

I

SUMMARY

OF INVESTIGATION OF ANCHORAGE AMI SUPPORT OF SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEM 5 THAT MAY DAMAGE THIS EQUIPMENT ITEMS THAT COULD POTENTIALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. OF -

SYSTEM IN WAS ANCHORAGE WAS WA5 DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCF 1/1/80 A I.D.

SUPPORT EVAL.'D NAME A I.D.

SUPPORT EYAL.'D COMCLUSION Rx Plant 78-87-001 Essential Turbine Bldg.

I D

Main Control 1

C SK-5101-063 125-Volt DC Power Electrical BBD "D" Breaker

-1 Bus Equipment 41-503628 Room E1. 654' 41-503701 480-Volt Ess.

1 C

Switchgear Feed Breaker 41-503666 ACS DC Valve 1

C Breaker 41-503828 4

41-503775 Station Under 1

C Voltage Pelay; Breaker 41-503634 inverter IA 1

C Non-Interruptible Bus Breaker 41-503677 MSIV & Bypass 1

C e

l i

I I

e

SUMMARY

OF INVESTIGATION OF ANCHORAGF AND SUPPORT OF j

SAFETY REL ATED ELECTRICAL EQUIPMENT AND IIEMS THAT MAY DAMAGF THIS EQUIPMENT l

ITEMS THAT COULD POTENTIALLV

~

INTERNALLY ATTACHED COMPONENTS INTERACT WITH 1HIS EQUIPMFNT

!.D. OF SYST5.M IN WAS ANCHORAGt WAS WAS DOCUMENT EQu!PMENT WHICH LOCATION TYFE OF MODIFIED IQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME

!.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCE 1/1/80 A I.D.

SUPPORT EVAL.'D NAME A !.D.

SUPPORT EVAL.'D CONCLUSION Reactor Plant Reactor Plant 1

C I

125-Volt DC Hattery to 5

Bus (Cont'd)

Reactor Plant Bus t-Reactor Plant I

C Char ger to Reactor Plant Battery IC Static Essent t al T1rbine Bldg.

2 No B

I nve r* er Power Electrical Equipment loon El. 654'

_ _ _. _ - ~ _ _ _ _. _.

LM e m !!09 Tsw1t f al T arbine ".ldg.

3 0

ifrGrvol t ar e.

1 0

v.5101-063 Re t ay '.,abi ne.

Power L lect rical RclJy5 for

-1 L quipment J8O-f el* E ss.

Roon El. 654'

!A R 14 T

480- Vol t I

r Turbine Ridq.

Q MCC 1A

23 h__

J:=

P !

4 I

t F

b SUMMARf 0F 1NVESTICAT10N OF ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTPICAL EQUIPMENT AND ITEMS THAT PAY DAPAGE TH15 EQUIPMENT

{

l Ilta THAT COULil POIt RII ALLY i

INTERNALLY ATTACHED COMPONFNTS INTERACT WITH THl$ EfyllPMENT l.D. Or i

SYSTEM IN WAS ANCHORAGE M5

. WL5 DOCU"E NT j

EQUIPENT WHICH LOCATION TTPE OF M00lFIED EQUIPENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPOPTING EQUIPMENT NAME I.D.

INSTALLED BLDG. A ELEY. SUPFORT SINCE 1/1/80 A 1.D.

SUPPORT EVAL.*D NAME A 1.0.

SUPPORT EVAL.*D EONrLUSION 6-Gznerator Essential Turbine Bldg.

3 D

Aux. Dist.

1 C

SK-5101-063 Plant 125-Volt Power Electrical Panel Breaker

-1 PC EC Equipment Room El. 654' Control Power 1

C

  1. 1 Breaker Control Power 1

C

  1. 2 Breaker DBD "D" Breaker 1

C 3 Breakers for 1

C C.B. Control Power Relay Panel I

C Gen. Rattery I

r, to Cen. Plant 125-Volt DC P.a:

Gen. Battery 1

C Charger to Gen.

125-Volt DC l'us Gen. Battery 1

C Charger to Gen. Pattery

. m

t I

1 1

i i

t t

. i.

SUMMARY

OF INVESTIGATION OF ANC ERAGE AN3 SUPPORT OF i

SAFETY RELATED Elt.TRICAL EQulPMcNT AND IIIMS THAT MAY DAMAGE THl$ EQUIPMENT ITEM 5 THAT C(NLD POT [NTI ALLY t

INTERNALLY ATTACHED COMPONENTS INTERACT WITH Tril5 EQUIPMENT l.D. OF SYSTEM IN WAS ANCHORAGE WAS WA5 DOCIMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. 8 ELEV. SUPPORT SINCE 1/1/80 8 I.D.

SUPPORT EVAL.'O NAME A I.0.

SUPPORT EVAL.'D CONCLUS!nN Aux. Dist.

Essential Turbine Bldg.

4 1C Static 1

C B

PKnel for Power Electrical Inverter Bkr.

125-Volt DC Equipment Gin. Plant Bus Room E1. G54' Gen. Plant 1

C Batteries Breaker Gen. Plant I

C Battery Charger &

AssoClated Breakers Diesel Bldg.

1 C

Protective Pelays Solatron For 74-al-On! Essential Turbine Bldg.

3 No P

Turbine Bidy.

Power Electrical Reg Bus Equipment Room E1. 654' transtnemer 13-80-001 Essential turnine Bldg.

2 No For Turbine Power Electrical Building Equipment 120-Volt Bus Roan E1. 654' 430-240/12u Volt 1

l h

t 4

i f

i l

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF-

)f SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DA" AGE THIS EQUIPMENT i*

IItro IMRI COULD PUit nII ALLY INTERNA!LY ATTACHED COMPONENTS INTERACT WITH THl5 EQUIPMENT

.D. OF SYSTEM IN WAS ANCHORAGE WA5 WR5

'#.IMENT l

EQUIPENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF, SUPPORT TYPE OF SUPPORT '/.40RTING i

EQUIPMENT NAME I.D.

INSTALLED BLDG. & ELEV. SUPPORT $1NCE 1/1/80

& l.D.

SUPPORT EVAL.'D NAff & I.D.

SUPPORT EVAL.'D Cf VtUSION I

1A 1 KVA Essential Turbine Bldg.

2 No G

Inn rter Power Electrical Equipment Room E1. 654' No IA Inverter Essential Turbine C1dg.

2 G

Input & Output Power Electrical -

Breakers Equipment Room E1. 654' IA Inverter Essential Turbine Bldg.

2 No G

M ter Panel Pcwer Electrical Equipwnt Room El. 654' 120-VAC Non-Essential Turbine Cidg.

2 No

  • G Interruptible Powe-Electrical Bus IA Equipinent Row 1 E1. 654' 120-VAC Nan-Essential Turbine Bldg.

2 No C.G Interruptible Power Electrical Bus IA Fuse E quipment Panel Roon E1. 654' e

m.

r__-_-

Ii i

i

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF l'

SAFETY REL ATED ELECTRICAL EQUitPE NT AND IIt M5 THAT MAY DAMAGE THIS EQUIPMENT ITEM 5 THAT COULD POTENTlaLLy INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUI 4fMT i.D. OF -

SYSTEM IN WAS ANCHORAGE W45 W DOCUMENT EQUIPMENT WHICH IOCATION TYPE OF MG3!FIED EQUIPMEkT NAME TYPE OF SUPPORT TYPE C SHPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCE 1/1/80 A I.D.

SUPPORT EVAL.'D NAME A I.D.

SUPPORT EVAL.'D CLWLUSION IB Emergency 78-83-002 On-Site 18 Diesel 3

Bid. El.

Fuel Oil 3..

R Diesel Emergency 9

Ginerator Power 641 Tank Compressed 1

Air Bottles (8)

Battery 4

Charger 74-91-005 Relay 4

Enclosure 18 Diesel 18-89-001 On-Site IB Diesel 5

Relays A 1

C R

Genu atcr Emergency B1dg. E1.

Controls Control Panel Dower 641' D.C.

Roan IC Energency 78-83-902 On-Site IB Diesel 3

Diesel Gen.

Emergency Generator 8

Starting Power Bldg.

i Batteries S El. 641' Racks

\\

1

~.. -. - - -

1 I

h I

i

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY REL ATED ELECTRICAL EQUIPMENT AND litMS THAT PAY DAMAGE THIS EQUIPMENT ITEM 5 THAT CDULD rustRIIALLY

~

INTERNALLY ATTACHED COMPONENTS INTERACT WITH THl$ EOUIPMENT 1..,. dF STSTEM IN WAS ANCHORAGE WA5 WR$

DOCUMENT EQUIPMNT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING

'g EQUIPME NT NAME I.D.

INSTALLED BLOG. & ELEV. SUPPORT $1NCE 1/1/80 A l.D.

SUPPORT EVAL.'D NAME A l.D.

SUPPORT EVAL.*D CONCLUSION 480-Volt Ess.

On-Site IB Diesel 5

1R D.G. Output I

C R

Bus 13 Emergency Generator Dreaker Power Bldg. El.

(74-79-044) 641' 452 EGB Inter Bus Tie 1

C 452 TOB Bkr.

74-77-043 fmergency Core C

Spray Puy IB 452 ECCB Ekr.

(74-77-013) 4R0-Volt Otesel 1

C r.1dg. 18 reed MCC 462 DP i'reaker 74-//-U45 1

I k'

Il' lL I:

t t

k f

SUMMARY

OF INVESTIGATION OF ANCHORACF AND SUPPORT OF SAFETY REL ATED ElfCTRICAL EQUIPP'ENT AND ITFMS THAT MAY DAMAI.F THIS EQUIPMENT f'

ITEM 5 THAT COULD ruIt 4ITAllY' f

~

INTERNALLY ATTACHED COMPONENTS INTERACT WITH THl5 EQUIPMFNT

!.D. OF SYSTEM IN WAS ANCHORAGE WAS WA5 DOCU*ENT EQUIPMENT WHICH LOCATION TVPE OF MODIFIED EQUIPMENT MAME TVPE OF SUPPORT TYPE OF SUPPORT SUPPOPTING EQUIPMENT NAME I.D.

INSTALLED BLDG. 4 ELEV. SUPPORT SINCE 1/1/80

& l.D.

SUPPORT EV AL. 'D r4AME & I.D.

SUPPORT EVAL.'D CONCLUSION 480-Volt On-Site IB Diesel S

Diesel Bldg.

I C

B Dissel Bldg.

Emergency Generator Battery 18 MCC Power Bldg. El. 641' Charger Bkr.

74-32-061

~

Diesel Engine 1

C

)'

Radiator Een Ilreaker 78-32-004 Diesel Bldg.

1 C

Uninterruptible Supply 74-32-062 Diesel Bldg.

1 C

Distribution T rans f orm r 75-32-060 vistribution 1

t' Panel 74-20-004 c

~

- 3' 2

i

SUMMARY

Of I?nt SitGATION OF ANCHORAGE AND StfPPORT Of f

SAFETY Pfl ATFD [LFCTP!f 4ETTJLI2l TNT AND ITEM 5 THAT UU PAMAGF THIS FQUIPMENT j

t I TEMS THAT COULD DOTf NTIALLY l

INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPPFNT I.D. OF s

SYSTEM IN WAS ANCHORAGE WAS bd5 DCCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT (YPE OF SUPPORT SUPPORTING f

EQUIPMENT NAME I.O.

INSTALLED BLOG. A ELEV. SUPPORT SINCE 1/1/80 A l.D.

SUPPORT EVAL.*D NAME A 1.D.

SUPPORT EVAL.*D CONCL USION j

16 Static Essential 18 Diesel 3,5 Trans f onner 1

C 81 Inverter Power Generator Relay 8 Bldg. E l. 641' Controit IB Diesel Essential IB Diesel 3

B Building Power Generator Standby Bldg. El. 641' i!atteries IB Diesel Essential IB Diesel 3.5 Transformer 1

C B

Batiding Power Generator Relays &

hittery Bldg. El. 641' Controls Charger 10'

'.'D; Diesel On-Site IB Diesel 5

IB Static 1

C D

f l !j. '&.

Emergency Generator inverter Bkr.

Ns Power 31dg. El. 681' n - L'-tie 4..

DC Control 1

C Powar E s !..

SwitChgear IB Brea6cr 74-32-063

. mW

'I

)

e t

i

?

SutmARY OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFE TY REL ATED ELECTRICAL EQUIPMENT AND ITEMS THAT MM DAMALT'TIIIS EQUIPMENT ITEMS THAT COULD POTENTI ALLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT l.D. OF SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPMENT WH:r.H LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCE 1/1/80 8 1.D.

SUPPORT EVAL.'D NAME A 1.0.

SUPPORT EVAL.'D CONCLUSION i,i:

D.G. Control 1

C 125 VDC Diesel Panel Feed Bldg. Main Breaker Dist. Bus 74-32-068 (Cont'd)

D.G. Bldg.

1 C

Engine Gen.

T Control Crt.

14-32 064 O

Diesel Bldg.

1 C

] g Main Battery i

Feei 74-32-070 Diesel Bldg.

I C

Main Charger F J. g rent ll 74-32-Of,6 y

"g M

Diesel Bldg.

1 C

i P

Batte ry!

Charger !ic 74-32-065 i

2 l

125 VDC Emergency IB Diesel i

Breaktrs 1

C Otst. Danel Dewer Bldg. E1. 641' s

-o- [

l l

i i

I

.i SUPMARY OF INVE5TIGAT10N OF ANCHORAGE AND SUPPORT OF SAFETY REL ATIF ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT ITEM 5 THAT CDULD PURU[ ALLY

' INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT

!.D. OF SYSTEM IN WAS ANCHORAGE WAS WR$

DOCUM NT EQUIPENT WHICH LOCATION TYPE OF MODIFIED EQUIPffMT NAME TYPE OF SUPPORT TVPE OF SUPPORT SUPPGPTING EQUIDMENT NAME I.D.

INSTALLED BLDG. A ELEV. SUPPORT SINCE 1/1/80 A !.D.

SUPPOPT EVAL.*D NAME A I.D.

SUPPORT E VAL. 'D CONCLUSION 480-Volt Ess.

Essential Turbine Bldg.

3 Yes Feed Breaker 1

C SK-5101-063 Bus IA Power Penetration 452 EGA

-4 Room E1. 640' 78-79-001 IA Emergency 1

C Core Spray Pwnp Breaker 452 ECC A 74-77-012 Turbine Bldg.

I C

MCC 1A Feed B reak er 74-77-011 Inter Bus 1

C T ie Credites-452 THA AC3 valve 36-30-001 Alternate Turbine B1dg.

3 P

Muror (AC)

Core Spray Mezz. Level 654' ACS Valve 38-30-002 Alternate Turbine Bldg.

3 9

Ii Motor (DC)

Core Spray Mezz Level P;

654'

?l

. U

9 i

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT j

I Itro THAT COULD Po rt RII ALL Y INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. OF SYSTEM IN WAS ANCHORAGE WR5 WAS DOCUMENT EQUIPM NT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE DF SUPPORT SUPPORTING EQUIPMENT NAME I.D.

INSTALLED BLDC. 8 ELEV. SUPPORT SINCE 1/1/8D

& I.D.

SUPPORT EVAL.'D NAME 8 f.D.

SUPPORT E VAL.

  • 0 CONCLUSION ACS Valve DC For Alternate Turbine Bldg.

2 M; tor Starter 38-30-002 Core Spray Mezz. Level A

654' IA Emergency 78-83-001 On-Site Turbine Bldg.

3 Yes Output Breaker 1

C Fuel Oil 4

B Facility Diesel Emergency E1. 640' 78-31-701 Day Tank Change G:nerator Power 78-81-3 Battery 1

C Charger IA Diesel 78-83-901 On-Site Turbine Bldg. Strapped R

GIncrator Emergency E1. 640' to Rack Storting Power Bat tery lA Die wt On-S i t e Turbine Bldg.

2 8

Ge n rator Emergency E1. 640' Starting Power Battery Rack C ont ainment Containment Containment lelded No R

Bldg.

Wall

o Electrical Penetration Shell Peretration 640' t -

i e

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF SAFETY REL ATED ELE CTRICAL ECUIPM rIT AND IIt P'5 THAT M)LY DA5'1BGF THIS EQUIPMENT ITEM 5 THAT C0tfLD POTENTI ALL'Y

~~~

_ INTERNALLY ATTACHED CO*PONENTS INTERACT WITH THIS EQUlPMENT I.D. TV SYSTEM IN WAS ANCHORAGE WA5 W45 DOCU"F!4T EQUIPMENT WHICH LOCATION TYPE OF MODIFIED ' EQUIPMENT NA.E TYPE OF SUPPORT TYPE OF SUPPORT SUPPORT!Nr.

I EQUIPMENT NAME I.D.

INSTALLED BLDG. A ELEV. SUPPORT $1NCE 1/1/80 8 I.D.

SUPPORT EVAL.'D NAME A I.D.

SUPPORT EVAL.*D CONCLUSION t

Offgas Vent 55-25-014 Containment Turbine Bldg.

No H

Header Isolation Pipe Tunnel Solenoid Valve 635' Contalment 37-35-101 Containment Turbine Bldg.

1 No H

Bldg. Pressure 37-35-102 Building Penetration 1

No H

Scitches 37-35-103 Roan 640' I

No sf ACS Elow 38-37-301 Alternate Turbine Bldg.

2 A

D Transmitter Core Spray Mezz 654'

'q r

C ontainment 37-35-301 Cont alment Turbine R11g.

3 No u

Uessel Vessel Pipe Tunnel Internal vanitoring 629' Pressore T raris. d t' er Cont aima rit 37-35-302 Cont ai rtiier.L Turbine Cidg.

3 No e

Ves wl Vessel Pipe Tunrel onitoring 629' J

Internal v

+

h e>su e 4

i Transai*ter

)

1 i

1 I I

i f

k, I

SUMMARY

OF INVESTIGAT!UN OF ANCHORAGE AND SUPPORT OF SAFETY RELATED ELECTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THIS IQUIPMENT ITEMS THAT COULD POTENTIAlLY INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS IQUIPMENT I.D. OF SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPPENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING f

EQUIPMENT NAME I.D.

INSTALLED BLDG. A ELEV. SUPPORT SINCE 1/1/80 8 f.D.

SUPPORT EVAL.'D NAME & I.D.

SilPPORT IVAL.'D CONCLUSION 4

Contairunent 37-42-301 Contairunent Turbine Bldg.

3 No H

VIssel Vessel Pipe Tunnel Inttrnal Monitoring 629' Ls,vsI h

TraIsmitter Conta t rinent 37-42-302 Contairunent Turbine Bldg.

3 No H

V ssel Vessel Pipe Tunnel Intzrnal Monitoring 629' Levsl Transinitter Cont ainment 73-25-016 Contaiment Containment 1

No H

Ventilation Isolation 643' Solenoid Valve C ontai nment 73-25-017 Contaiment Contaiment 1

No Ventilation Isolation 643' Solenoid Valve Containment 73-25 018 Cor.tairraent Contairrr.ent 1

No H

Ventilation Isolation 643' Solenoid valve 29 -

<c

I-h f

I

SUMMARY

OF INVESTIGAT10N OF ANCHORAGE AND SUPPORT OF SAFETY REL ATED ElfCTRICAL EQUIPMENT AND ITEMS THAT MAY DAMAGE THl$ EQUIPMENT ITEMS THAT COULD POTENTIALLY INTE!!NALLY ATTACHED COMPONFNTS INTERACT WITH THIS EQUIPMfMT I.D. Of SYSTEM IN WAS ANCHORACE WA5 WAS DOCUMENT

[

EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NA!T TYPE OF SUPPORT TYPE M SUPPORT SUPPORTING.

EQu!PMENT NAME I.D.

INSTALLED BLOG. 8 ELEV. SUPPORT SINCE 1/1/80 A 1.D.

SUPPORT EVAL.'D NAE & I.D.

SUPPORT EVAL.'D CONCLUSION Containment 73-25-019 Containment Contalment 1

No H

Vgntilation Isolation 643' Solsnold Valve Centainment 73-c. J03 Containment Containment i

No H

Ventilation Isolation 643' Solenoid Valve C onta inment 73-25-008 Containnent Contalment 1

No H

Vrntilation Isolation 643' Solenoid Valve 9/F 1 A 50-37-301 Reactor Containment 2

No to Tra'isanitter Protection 633' r/F 13 50-37-304 Reactor Contaimi:9t

?

Irans.nt:ter Prct ect ion 633'

\\

P/t 2A 50-3i-302 Heactor Contain4cnt 2

No H

Tr esai t t er P rot ect i on 633' P/F a 50-31-303 React or Containnent 2

No Transoitter Prot ect ion 633' e m o emo

. I

-k

t

$iI:I'

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT )F

'{:

SAFETY REL ATED ELEcrRICAL EQUIPM.NT ANP IliM5 THAT MAY DAPA5F THIS EQUIPMENT ITEM 5 THAT COULD POTENTIALLY

[

SYSTEM IN WAS ANCHORAGE ~ INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPPENT 1.D. Of WAS WAS DOCUMENT EQUIPMENT WHICH LOCATION TfPE OF MODIFIED EQUIPMENT NAMF TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTINf.

EQUIPMENT NAME I.D.

INSTALLED BLDG. 8 ELEV. SUPPORT SINCE 1/1/80 A !.D.

SUPPORT EVAL.'D NAME A 1.D.

SUPPORT E VAL. 'D CONCLUSION MISV Bypass 64-25-002 Containment Containment 1

No H

Solenoid Valve Isolation 633' MSIV Bypass 64-25-003 Contairunent Containment 1

No H

~

Solenoid Valve Isolation 633' Scienoid for 56-25-002 Containment Containment 1

No H

Decay Heat Isolation 633' Isolation Valve 56-25-001 Reactor 55-25-013 Containment Grade Floor 2

No si Offqas isolation Conta t rinent Solenoid 6aj' Reactor 55-22-022 Containment Grade Floor 2

No H

Oftgas isolaticn Containment Solenoid 643' Reactor 50-42-302 Reactor Containment 2

m Water Level Prctection 659'

  1. 1 Transmitter Ni

_m e,..,

E i'

I:

t l'

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF

{-

SAFETY RELATED ELECTRICAL EQUIPPENT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT litm THAT COULD PUt t RII ALLY -

INTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. Of SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME 1.D.

INSTALLED BLOG. 8 ELEV. SUPPORT SINCE 1/1/80 A I.D.

SUPPORT FVAL.'D NAME & l.D.

SUPPORT EVAL.'D CONCLUSION A

Raacter Water 50-42 306 Reactor Containment 2

Laval #3 Protection 659' Transmitter A

RIactor Water 50-42 303 Reactor Contairunent 2

L(vel #2 Prot ection 659' Transmitter A

Radiation 73 43-201 Containment Containment None Monitor Isolation 667' A

Padiation 73-43-202 Containnent Cont ainnent None Monit:r Isolation 667' A

Rediation 73-43-203 Cont alanent Contal m.ent None o g.or Isolation 667' H

PSI. Solenoid 61-22-005 Hydraulic Containment 1

No Valve Grade 643' Accumulator tf e f 1

i I

l i

i

SUMMARY

OF INVESTIGATION OF ANCHORAGE AND SUPPORT OF

[

SAFETY REL ATED ELECTPICAL EQUIPMfNT AND ITEMS THAT MAY DAMAGE THIS EQUIPMENT I

I ITEMS THAT COULD POTENTIALLY l

[NTERNALLY ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT

!.D. Of 2

SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPMENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING EQUIPMENT NAME 1.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCE 1/1/80

& I.D.

SUPPORT EVAL.'D NAME & I.D.

SUPPORT EVAL.'D CONCLUSION HPCS Motor 53-06-001 High Contaiment 3

R l

(IA)

Pressure 667' I

Core Spray HPCS Motor 53-06-002 High Contaiment 3

R (18)

Pressure 667' Core Spray LPCS Inlet to 53-25-005 Low Contalment 2

No H

Reactor Pressure Approx. 672' Solenoid Valse Core Spray MDS Solenoid 62-25-015 Manual Contal mnent 1

No H

Depressuri-Apra ax. /14' 2ation PDS Scienoid 62-25-016 Manual Cont ai nment 1

No H

Depressuri-Approx. 714' 2etion Nuclear Inste.

Reactor Cont ai ment 1,2 A

Detectors, Pr tection Cables, A Junction Boxes Ch. I thru 8 i

y

' ll 1

l

.k

?.

k i

i l

SUMMARY

OF iNVESTIGAT10N OF ANCHORACE AND SUPPORT OF l-SAFETf REL ATED ELECTRICAL EQUIF'MEU AND ITEP'5 THAT PAY DAPAGE THIS EQU!PMENT i

ITEM 5 THAT CDULD PUTENTI ALLY

_ INTERNALLV ATTACHED COMPONENTS INTERACT WITH THIS EQUIPMENT I.D. CF*

SYSTEM IN WAS ANCHORAGE WAS WAS DOCUMENT EQUIPPENT WHICH LOCATION TYPE OF MODIFIED EQUIPMENT NAME ' TYPE OF SUPPORT TYPE OF SUPPORT SUPPORTING b

EQUIPMENT NAME

!.D.

INSTALLED BLDG. & ELEV. SUPPORT SINCE 1/1/80

& 1.D.

SUPPORT EVAL.'D NAME & l.0.

SUPP')RT EVAL.'D CONCLUSION j

f Centainment 73-25-004 Containment Containment 1

No H

Vtntilation Isolation 643' 6

Soltnold Valve Containment 73-25-001 Containment Containment 1

No H

Ventilation Isolation 643' Solenoid Valve

..h

ATTACHMENT 3 ANAIYSl% SCHITAlti

~~~ T9l;l 1982

~

~

~MiE JULY AliG.~5[IL$1D{f{,'IEC{,3,4f. _ FEB. PAR. APR.

MAY JUNE JULY AE.~

l. EMERGENCY POWER EQUIP.

i

  1. 4CHORAGES

-- 3

f. SHUTDOW1 INSTRUMENTATIOi

-I

  1. 4CHORAGES COiTAINMENT BUILDING I

STRUCTURE

). EMERGENCY SERVICE WATER g

i SUPPLY SYSTEM f.MDSPIPING L

l

5. SitJTD0ld CONDENSER.

[___

l

. MSIV l

l 1

h CDITROL ROD DRIVES I

l

. LACB"R STACK l

~~

. IB DIESEL BUILDING I

I l

,1. COiTROL ROOM l

1

. DECAY HEAT SYSTEM.

I-l

3. SPENT FUEL POOL MAKEUP I

l k 7tHtBINE BUILDING l

l

9. FEEDWATEP PIPING COMPLETED I
h. MAIN STEAM PIPING CD4PLETED l

!7. RECIRCULATION PIPING COMPLETED l

B..CS SUCTION PIPING (OMPLETED i

),..CSoiS-GE,,,iNG C.PEmo

. G-3 STACK COMPLETED

ATTACHMENT 4 MODiflCATI(tl Srlitlitif C i

1981 1987

_198J... ___

1984 JULY _.0CI.

. JNJ. ePR. JULY.. OCJ....)AN,. N'R, _ JUQ. OCT. JAM. APR.. JULY OCL

1. EHEPGEriCY POWER Lf}fJlP.

AllCH0PAGES

.-- -.-._)

2. SHUTD0Ti IfGTRUMENTATIU4 i j

NICH0 RAGES q

3, C0'ITAIN'iEfiT C'JILD! fig

  • l

]

STRUCTURE 4, EMER5EilCY SER'lCE WATER L

SUPPLY SYSTEM p

- - - -]-

l l5,MD5 PIP!'lG

[

l f 6. 54JTD0 ire C0rlDEN5tR l ~~ ~' j I

, 7. MSIP B. C0fiTROL R00 DRIVES O **

R **

9, LACBWR STACK *

10. IB DIESEL BUILDING
  • i
11. CONTROL ROOM C**

g e

c12. LECAY HEAT SYSTFt' i

C*

R **

l

13. SPEtiT FUEL POOL MAKEUP C**

14, TURBINE BUILDING *

15. FEEDWATER PIP!flG l

l

]**

1 16, MAIN.5 TEN 1 PIPING l

l

]**

17. RECIRCULATION PIPING R**

t

38. HPCS SUCTION PIPING C(NPLETED 1
19. HPCS DISCHARGE FIPING

~~j 20, G.3 STACK

  • s Esus

.,. ". ; d d.

  • .?

n._,.~.,-,,_..

,

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