Text

Forwards Errata Sheet for Rev 17 to Sgs UFSAR IAW 10CFR50.71(e).Errata Is in Response to Errors in 981016 Package Which Were self-identified by Pse&G.With Insertion Instructions
	ML18106A981
Person / Time
Site:	Salem
Issue date:	12/04/1998
From:	Storz L; Public Service Enterprise Group
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18106A982	List: ML18106A981;
References
	LR-N980557, NUDOCS 9812210223
	Download: ML18106A981 (94)
	v • d • e

Public Service Electric and Gas Company Louis F. Storz Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-5700 Senior Vice President - Nuclear Operations DEC4~ __ -B8 LR-N9B0557 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

UPDATED FINAL SAFETY ANALYSIS REPORT REVISION 17, ERRATA SHEET SALEM GENERATING STATION UNITS I AND 2 DOCKET NOS. 50-272 and 50-311 Public Service Electric and Gas Company (PSE&G) hereby submits an errata sheet for Revision No.17 to the Salem Generating Station Updated Final Safety Analysis Report (UFSAR) in accordance with the requirements of 10CFR50.71(e).

On Qctober 16, 1998, PSE&G issued the Salem UFSAR Revision No.17 update*

package. This errata submittal is in response to errors in that update package which were self-identified by PSE&G. These include the omission of SAR figure changes as well as 11Jinor typographical errors. The errors have been captured under our corrective action program with this submittal serving as a remedial corrective action.

Should there be any questions with regard to this submittal, please do not hesitate to contact us.

Sincerely,

~~1~

( 9812210223 981204 - - :

PDR ADOCK 05000272:

K PDR,___;

l:l.\. Printed on

~ Recycled Paper

DEC 4 1998 Document Control Desk LR-N980557 Attachments:

-Affidavit

- Summary of Errata Sheet Changes

- Salem UFSAR Errata Sheet Package C Document Control Desk - Original & ten copies Mr. Hubert J. Miller, Administrator - Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Patrick Milano, Licensing Project Manager - Salem U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. S. Morris - Salem (X24)

USNRC Senior Resident Inspector Mr. K. Tosch, Manager, IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625

Document Control Desk LR-N980557

/rbd BC (w/o attachments):

CNO & President - Nuclear Business Unit (N09)

Senior Vice President- Nuclear Engineering (N19)

Director- QA/NT/EP (120)

Director- Licensing/Regulation and Fuels (N21)

General Manager - Salem Operations (SOS)

Manager - Nuclear Review Board (N38)

Manager - Joint Owners/Ext Aff Interface (N28)

Licensing Operational Manager - Salem (N21)

J. Keenan, Esq. (N21)

(w/attachments except Salem UFSAR Errata Sheet Package):

Records Management Microfilm File Nos. 1.2.1, 2.1.3

. 95-4933

1 - - ...

REF: LR-N980557 STATE OF NEW JERSEY )

)SS COUNTY OF SALEM )

L. F. Storz, being duly sworn, states that he is Senior Vice President - Nuclear Operations of Public Service Electric and Gas Company, that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this certification; and that in accordance with 10CFR50.71 (e)(2), the information contained in the attached letter and Updated Final Safety Analysis Report accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or p,repared pursuant to Commission requirement, and contains an identification of changes made under the provisions of 10CFR50.59 but not previously submitted to the Commission.

Subscribed and Sworn to before me this 2"d day of December, 1998

,

DELORIS D. HADDEN

- * ~ Notary Public of New Jersey

,. My Commission Expires

- o~-is-2000

LR-N980557 ATTACHMENT 1 Summary of Errata Sheet Changes

ATTACHMENT 1 SSAR 97-082 Figure 1.2-1 Adds depiction of new Unit No. 2 Service Water Accumulator Tanks DCP 2EC-3590 pkg. 7 to the general site plan.

SSAR 97-143 Figure 3.8-35 Adds hinge pin option - Equipment Hatch Bolting Configuration DCP lEA-1304 DCP 2EA-1136 SSAR 98-014 Page 13.1-8a Deletes responsibility for Informa-tion Systems (IS) from Director -

Nuclear Business Support. IS now reports to Corporate.

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TABLE OF CONTENTS (Cont)

Section 4.4.4.2 4.4.4.3 Title Initial Power and Plant Operation Component and Fuel Inspections Page 4.4-55 4.4-56 4.4.4.4 Augmented Startup Test Program 4.4-56 4.4.5 References for Section 4.4 4.4-56 4.5 RELOAD ANALYSIS 4.5-1 4.5.1 References for Section 4.5 4.5-2

SGS-UFSAR 4-vii Revision 17 October 16, 1998

LIST OF TABLES 4.1-1 4.1-2 Reactor Design Comparison Table Analytic Techniques Incore Design 4.1-3 Design Loading Conditions for Reactor Core Components 4.2-1 Maximum Deflections Allowed for Reactor Internal Support Structures 4.3-1 Reactor Core Description I 4.3-2 Nuclear Design Parameters 4.3-3 Reactivity Requirements for Rod Cluster Control Assemblies 4.3-4 Axial Stability Index-PWR Core With a 12-Foot Height 4.3-5 4.3-6 4.3-7 Typical Neutron Flux Levels at Full Power Comparison of Measured and Calculated Doppler Defects Benchmark Critical Experiments SGS-UFSAR 4-viii Revision 17 October 16, 1998

the entire- core barrel and thermal shield),. the. upper- core. support structure and the in-core 1nstrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between the fuel assemblies and control rod drive mechanisms, direct coolant flow past the fuel elements and to the pressure vessel head, provide gamma and neutron shielding, and provide guides for the in-core instrumentation.

The nuclear design analyses and evaluation establish physical locations for control rods and burnable absorbers, and physical parameters such as fuel enrichments and boron concentration in the coolant such that_ the reactor core has inherent characteristics which, together with corrective actions of the Reactor control, Protection and Emergency Cooling Systems provide adequate reactivity control even if the highest reactivity worth rod' cluster control assembly is stuck in the fully withdrawn position. The design also provides for inherent stability against diametral and azimuthal power oscillat~ons.

The thermal-hydraulic design analyses and evaluation establish coolant flow parameters which assure that adequate heat transfer is assured between the fuel cladding and the reactor coolant. The thermal design takes. into account local variations in fuel rod dimensions, power generation, flow distribution, and mixing. The mixing vanes incorporated in the fuel assembly spacer grid design induces additional flow mixing between the various flow channels within a fuel assembly as well as between adjacent assemblies

Instrumentation is provided in and out of ~he core to monitor the nuclear, thermal-hydraulic, and mechanical performance of* the reactor and to provide-inputs to automatic control functions.

The reactor core design together with corrective actions of the Reactor Control, Protection and Emergency Cooling Systems can meet the reactor performance and safety criteria specified.in Section 4.2 *

S.GS;..UFSAR 4.1-3 Revision 1 7-0ctober.-16, 1998

To illustrate the effects of: the.--change in fuel design, Table 4.-1-1' presents-principal nuclear, thermal-hydraulic, and mechanical design parameters for the Salem 17 x 17 STD, Vantage SH, and Vantage+ fuel assemblies.

The effects of fuel densification were evaluated(l).

The analytical techniques employed in the core design are tabulated in Table 4 .1-

2. The loading conditions considered in general for the core internals and components are tabulated in Table 4.1-3. Specific or limiting loads considered for design purposes of the various components are listed as follows: fuel assemblies in Section* 4. 2 .1.1. 2; reactor internals in Section 4. 2. 2. 3 and Table 5.1-10; neutron absorb.er rods, burnable absorber rods, neutron source rods, and thimble plug assemblies (if used) in Section 4.2.3.1.3; control rod drive mechanisms in Section 4.2.3.1.4.

4.1.1 Reference for Section 4.1

1. Davidson, S.L. (Ed.), et al., "Vantage SH Fuel Assembly Reference Core Report," WCAP-10444-P-A and Appendix A, September 1985; Addendum 2-A, March 1986; Addendum 2-A, April 1988.

2. Davidson, S.L., Nuhfer, D.L. (Eds.), "Vantage+ Fuel Assembly Refernece Core Report," WCAP-12610-:P-A, April 1995.

3. Hellman, J. M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Application," WCAP-8218-P-A (Proprietary) and WCAP-8219-A (Nonproprietary), March 1975.

SGS-UFSAR 4.1-4 Revision. 17-0ctober 16, 1998

00000000000 0~ ~ I 1.973 FUEL ASS'Y AND CONTROL ROD 0 () 0 PITCH. ,

c o~ /

i ,,

0 ~ 0

/

GUIDE 00 THIMBLE 8.~66 TYP

.040 I

TYP 00 I

0 0 I I

0 0 i 0 0 INSTRUMENTAT! ON SHEATH 00 FUEL ASSE~BU WITHOUT ROD I CLUSTER CONTROL 1_ CONTROL CLUSTER ELEMENT FUEL RODS 261+ REQ'D 00 = 0.3'N CLAD THICKNESS= 0.0225 CLAD ~TERIAL - ZIRC-~

REVISION 6 FEBRUARY 15, 1987 Fuel Assembly Cross Section - 17 x 17 PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.2-1

159.97'

~---...-- 3 .47 152.160 I 8.,26

153 60 133 10 112 ' ' '2 00 71 . , 30 3' 17Xl7 VANTAGE 5-H FUEL ASSEMBLY PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION

Updated FSAR 17 X 17 VANTAGE 5-H FUEL ASSEMBLY Revision 11, July 22, 1991 Sheet 1of1 Fig. 4.2-2A

END PLUG SPRING 6.5" (TYP)

U02 PELLETS 151.6" (REF) FUEL-CLAD GAP 144.0" (TYP)

ZI RCA LOY CLAD SPECIFIC DIMENSIONS DEPEND ON DESIGN VARIABLES SUCH AS PRE-PRESSURIZATION. POWER HISTORY. AND DISCHARGE BURNUP REVISION 6 FEBRUARY 15, 1987 Fuel Rod Schematic PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.2-3

DIM 17Xl7 V5-H A 152. 160 B 7.319 c 144.00 DIA D .329 DIA E .374 DIMENSIONS ARE IN INCHES BOTTOM END PLUG SHOWS INTERNAL GRIP TYPE FOR V5-H FUEL RODS. -

DIA E - - - - -

TOP DIA D - BOTTOM

, - .!' i I J , r) - ..

~ I I I\

I~PLENUM I

DIM B DIM C FUEL STACK DIM A FUEL ROD LENGTH 17Xl7 VANTAGE 5-H FUEL ROD PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION

Updated FSAR 17X17 VANTAGE 5-H FUEL ROD SCHEMATIC Revision 11, July 22, 1991 Sheet 1 of 1 Fig. 4.2-3A

MID GRID EXPANSION JOINT DESIGN STA I HLESS STEEL.SLEEVE GRID STRAP

EXPAHSIOH LOBE EXPANSION DIAMETER REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Plan View SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.2-5

BUTT WELD ALL AROUND TOP NOZZLE ADAPTER PLATE STAINLESS STEEL SLEEVE EXPANSION LOBE

TOP .GRID REVISION 6 FEBRUARY 15, 1987 Top Grid to Nozzle Attachment

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.2-7

')

w

...::::c w

0 u :z:

a::: 0

..... N VJ

z:

w c...

>C Cl)

I

)

I I

\

)

I 0

a::

Cl REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Elevation View, Grid to Thimble Attachment SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.2-9

~ZIRCALOY THIMBLE

/BOTTOM GRID SPOT WELD

~SS INSERT

THIMBLE/END PLUG INERT-GAS FUSION WELD INSERT/END PLUG WELD ZIRCALOY THIMBLE END PLUG

~SS THIMBLE SCREW WITH INTEGRAL LOCKING CAP REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Guide Thimble to Bottom Nozzle Joint SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.2-11

vaiues given in *-reference 28. The above, along with- thee.design ..- basis*.in* r of Section 4.3.1.3, Control* Power Distribution, satisfies GDC-10

Discussion Fuel burnup is a measure of fuel depletion which represents the integrated energy output of the fuel (MWD/MTU) and is a convenient means for quantifying fuel exposure criteria.

The core design lifetime or design discharge burnup is achieved by installing sufficient initial excess reactivity in each fuel region and by following a fuel replacement program (such as that described in Section 4.3.2) that meets all safety-related criteria in each cycle of operation.

Initial excess reactivi~y installed in the fuel, although not a design basis, must be sufficient to maintain core criticality at full power operating conditions throughout cycle life with equilibrium xenon, samarium, and other fission products present. The end-of-design cycle life is defined to occur when the chemical shim concentration is essentially zero with control rods present to the degree necessary for operational requirements. In terms of chemical shim boron concentration this represents approximately 10 ppm with no control rod insertion *

A limitation on initial installed excess reactivity is not required other than as is quantified in terms of other design bases such as core negative reactivity feedback and shutdown margin discussed below.

4.3.1.2 Negative Reactivity Feedbacks (Reactivity Coefficient)

The fuel temperature coefficient will be negative and the moderator temperature coefficient of reactivity will be non-positive for power operating conditions, thereby providing

SGS.,..UFSAR 4.3-3 Rev~sion 1'7-

October 16, 1998.-

negative reactivity feedback characteristics. The design basis meets- GDC-11.

Discussion

When compensation for a rapid increase in reactivity is considered, there are two major effects. These are the resonance absorption effects (Doppler) associated with changing fuel temperature and the spectrum effect resulting from changing moderator density. These basic physics characteristics are often identified by reactivity coefficients. The use of slightly enriched uranium ensures that the Doppler coefficient of reactivity is negative. This coefficient provides the most rapid reactivity compensation. The core is also designed to have an overall negative moderator temperature coefficient of reactivity so that average coolant temperature or void content provides another, slower compensatory effect.

Nominal power operation is permitted only in a range of overall non-positive moderator temperature coefficient. The non-positive moderator temperature coefficient can be achieved through use of fixed burnable absorber, integral fuel burnable absorber (IFBA) and/or_ control rods by limiting the reactivity held down by soluable boron.

Burnable absorber content (quantity and distribution) is not stated as a design basis other than as it relates to accomplishment of a non-positive moderator temperature coefficient at power operating conditions discussed above.

4.3.1.3 Basis Control of Power Distribution The nuclear design basis is that, with at least a 95 percent confidence level:

1. The fuel will not be operated at greater than 13 kW/ft under normal operating conditions including an allowance SGS-,.UFSAR 4.3-4 Revision 17 October 16-, 1998

9* 0 8 -4

-I-

E OI

.i::

7 -8

~

E Cl en 1.1.J c...

6 -12 en Li.J 0 c...

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a::

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c

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u...

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~

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(,,) ::::t I

> :=l 0

0 2 -28 (/')

z:

a:: 0 c... Pu2~ I (,,)

I

~ -32 0 -- ----- -36 0 8 12 16 20 24 28 32 36 40 BURr~uP. ( GWD/ MTU)

REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Production and Consumption of Higher Isotopes SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-2

0 0

c co 0

0 0

w <.D

z:

0

z:

w a:::

I

"' w =>

I 80 I- 0 L.L.. I-

=> 0.. L.L.. :E 0

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w

~

a:>

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Cl

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=>

Cl

/ ~

a::: :z: :z: ....

I a::: <D 0 :::>

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.... 8 r-a:::

w

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=>

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=>

a:> - I N

E Cl 0 =>

0.. 0

~

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..... I 0 0

=>

L.L.. )I: I 0 0

0...

=:i

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- :z:

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z:: a::i 0

UJ w

0

z:

I 0 0

c:o

<!:)

<a::

UJ I 0 0

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(..)

I w

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.<(

0 0

I :z:

a:::

=>

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0

t" I ::z::

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0 0

0 I N 0

0 0 0 80 0 0

<.D 0

N 800 0

t" N

lHdd) H011VM1H3JH0J HOMOS 1VJl11MJ REVISION 6 FEBRUARY 15, 1987 Boron Concentration vs First Cycle Burnup PUBLIC SERVICE ELECTRIC AND GAS COMPANY With and Without Burnable Poison Rods (Typical)

SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-3

cgi 181 cgi cgi 181 181 181 181 181 181 181 181 181 181 r81 181 D 181 D 181 181 181 181 181 181 181 r81 181 181 f8l f8l f8l 181 r81 D 181 D 181 181 cgi f8l f8l f8l 181 r81 181 181 181 24 BP'S 20 BP'S t CORE CENTE R D l8l D 181 D 181 181 D 181 181 181 D 181 D 181 181 D D D 181 D 181 181 D D D D D 181 D 181 D 181 D D D D D

D 181 D 181 D CORE CENTER 12 BP'S Do D DD 6 BP'S

\

181 181 181 Do 181 D D D D t?3l D D D D D D D D Do D oD 5 BP'S REVISION 6 FEBRUARY 15, 1987 Burnable Poison Rod Arrangement Within an PUBLIC SERVICE ELECTRIC AND GAS COMPANY Assembly - Unit 1 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-4A

~

rBI 181 0

~

181 0 181~

~

181 181 0

0

~

181 0

l8l

~

181 l&1 181 181 D 181 181 0 181 0 181 0 l&1 ~ 0 181 0 181 181 181 ~ 181 ~

~ ~ D r8I 181 20 BP Is 16 BP'S CORE CENTER

~ ~ 181 D

~

~ D 181 D D 181 ~ D D

~ 0 181 0 t8l ~ D D D D D ~ O ~ D DD D DD 12 BP'S CORE CENTER 9 BP'S t

l8J t8I 181 l8l 181

~ 0 t8l D ~

~ 0 D 181 D D D D D Do D 0 D 10 BP'S REVISION 6 FEBRUARY 15, 1987 Burnable Poison Rod Arrangement Within an PUBLIC SERVICE ELECTRIC AND GAS COMPANY Assembly - Unit 2 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-48

R p H Ml K J HG FE D C B A 180° 6 6 6 2 23 5 20 24 15 .20 5 3 45 5 24 24 20 24 24 5

~

24 20 24 24 20 24 5 20 20 12 24 24 24 12 20 20 6 24 24 24 24 24 24 7 6 24 24 24 24 24 24 24 6 8 6 20 24 24 24 24 20 6 90° 270° 9 6 24 24 24 24 24 24 24 6 10 24 24 24 24 24 24 11 20 20 12 24 24 24 12 20 20 12 24 20 24 24 20 24 13 45 5 24 24 20 24 24 5 .

I~ 5 20 24 23 20 15 5 15 6 6 6 oo NUMBER INDlCATES NUMBER OF BURNABLE POISON RODS S INDICATES SOURCE ROD REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Burnable Poison Loading Pattern - Unit 1 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-5A

R p N M L K J H G F E D c B A 180° 10 10 10 2 9 12 20 19 12 9 IS 3 20 16 16 16 20 9 9 4S

~ 20 20 16 16 20 20 5 12 20 16 16 16 20 12 6 10 16 16 20 20 16 16 10 7 20 16 20 20 20 16 20 8 10 16 16 20 20 16 16 10 900 9 20 16 20 20 20 16 20 10 10 16 16 20 20 16 16 10 II 12 20 16 16 16 20 12 12 20 20 16 16 20 20 13 4S 20 9 20 16 16 16 9 19 I~ 9 12 20 IS 12 9 15 10 10 10 oo NUMBER INDICATES NUMBER OF BURNABLE POISON RODS S INDICATES SOURCE ROD REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Burnable Poison Loading Pattern* Unit 2 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-SB

G..

x 1H I

G F E D C B A G.. --1.072 -->- - ..... _.,_,_ - __ ,_ --- - - - - G..

8 I

9 0 65

.~ GDDDDDD I

. . _.~_ . . BBDDDDD 10 1 2 I

11 0 89

~ *~ BEJBDDDD I

12 1 1 39 GBBBDD 1_.~_,o_o_ BB EJI0.9461 GD I

13 __

14L-1--~+-2_0 GB 888 I

1 I

I s...._o_.~_1_0_ o.8&611 o.93911 o.6821 CA Leu LA TED F~H ~ 1.38 I

KEY:

VALUE REPRESENTS ASSEMBLY RELATIVE POWER REVISION 6 FEBRUARY 15, 1987 Normalized Power Density Distribution Near BOL, PUBLIC SERVICE ELECTRIC AND GAS COMPANY Unrodded Core, HFP, No Xenon*

SALEM NUCLEAR GENERATING STATION Unit 1 Updated F~AR FIG. 4.3-6A

q_

1H G F E D C B A I

ct. - . . . 1.051 -.. . . . -- -.. -- -.. . -.. . .... . . - -ct 8

I 9

0 2

-~ BDDDDDD 10 17 EJBDDDDD

.-----t--.1.1, I

11 1 1

~ * ~ EJBBDDDD I

12

___ * ~ GBBBDD 12 13 .__1_.1:1-.27__,J GB BI 0.9381 EJ D I

I 14 -~-35-' 1.0431810.999188 15 _o_.~_2s_BBB CALCULATEDF~H=1.36 I KEY.

Ci. VALUE REPRESENTS ASSEMBLY RELATIVE POWER REVISION 6 FEBRUARY 15, 1987 Normalized Power Density Distribution Near BOL, PUBLIC SERVICE ELECTRIC AND GAS COMPANY Unrodded Core, HFP, No Xenon - Unit 2 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-68

Cf_

1 H G F E D C B A x1 Cf_ -- 1.117 - -- - - - .. - .. - .... - --- - - Cf_

8 I I

9

---+----'

~ BDDDDDD 10 0 10

_ * ~ BBDDDDD 11 0 I

11 11 7 12

. ._ -+-~*~BBBBDD

13 I

1 10 0

...___._:_ _ BBBBBD BI I

14 __1_.1_ :09__ 11.0591 0.980 11 0.90611 0.5841 Io.8461 BI o.6681 I

15 __ o_.~+-52__

N CALCULATED F ~H == 1.36 I

KEY:

VALUE REPRESENTS ASSEMBLY RELATIVE POWER REVISION 6 FEBRUARY 15, 1987 Normalized Power Density Distribution Near BOL, PUBLIC SERVICE ELECTRIC AND GAS COMPANY Unrodded Core, HFP, Equilibrium Xenon - Unit 1 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-7A

q_

1H G F E D c B A I

- ... - - _*,_ --- - -ct_

I I

9 10 0

~ BDDDDDD I

10

--1.1~& EJBDDDDD I

11

~ * : EJBBDDDD 1 1 61

12 .2-:22__. BBBB DD B EJ EJ I IEJ D I

13 _1_.,_:26_ 0.920

.~-29__. EJ I B EJ B I

14 0.9951 1s ...__o_.~_1a__.

I I

I BI o.7841 o.5631 CALCULATED F~H" 1.35 q_ - KEY:

VALUE REPRESENTS ASSEMBLY RELATIVE POWER REVISION 6 FEBRUARY 15, 1987 Normalized Power Density Distribution Near BOL, PUBLIC SERVICE ELECTRIC AND GAS COMPANY Unrodded Core, HFP, Equilibrium Xenon - Unit 2 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-78

-1.04 1.03 1.04 1.03 1.06 1.09 1.03 1.07 1.14 l><

1.04 1.09 1. 15 1. 18 1.17 1.04 1.11 x 1.18 1.03 1.09 1. 15 1.15 1.19 x

1. 16 1.19 1.17 1.03 1.09 1. 15 1.15 1. 16 1.19 1. 17 1.17 1.04 1.11 x 1.17 1. 18 x 1.20 1.20 x

1.03 1.09 1. 15 1.15 1.03 1.09 1 .15 1.15 1.04 1. 11 x 1 .18 1.16 1. 19 1.17 1.17 1.20 1.17 1 .16 1. 19 1.17 1.17 1.20 1.17 1.17

1. 19 x 1.19 1 .19 x 1. 19 1.19 x

1.04 1.09 1. 15 1.18 1 .17 1.19 1. 16 1.16 1.18 1.16 1.16 1. 19 1.17 1.03 1.07 1.14 l>< 1.18 1.18 1. 15 1.15 1 .17 1.15 1. 15 1. 18 1.18 x

1.03 1.06 1.09 1.14 1.03 1.04 1.06 1.07

1. 15 x

1.09 1.11 1 .15 1. 15 x

1.09 1.09 1 .11

1. 15 1.15 x

1.09 1.09 1 .11

1. 15 1.13 1.09 1.09 1.07 1.06 1.04 1.04 1.03 1.03 1.03 1.04 1.04 1.03 1.03 1.04 1.03 1.03 1.04 1.03 1.03 1.03 1.03 1.031 REVISION 8 FEBRUARY 15. 1987 Rodwise Power Distribution in a Typical Assembly (Assembly G-9) Near BOL, HFP, Equilibrium Xenon, PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Unrodded Core - Unit 1 Updated FSAR FIG. 4.3-11A

-I . 02 I. 02 I. 03 I. 02 1.05 1.08 I. 02 1.06 I. 12 [X 1.03 1.08 I. Ill I. 17 I. 16 I. 03 I. 10 [X 1. 17 I. 18 [X I. 03 1.08 I. Ill I. Ill I. 16 I. 18 I. 17 1.03 1.08 I. Ill I. 14 I. 16 I. 19 I. 17 I. 17 1.03 I. 10 [X I. 16 I. 18 [X I_ 19 I. 20 x

1.03 1.09 1.03 1.09 1.04 I. 11 I. Ill x

I. Ill I. 15 I. 15 I. 18 I. 16 I. 16 I. 18 x

I. 19 I. 19 I. 17 I. 17

19 I. 17 I. 17 I. 19 1.20 I. 20 x

I. i8 I. 18 I. 20 I. 18 I. 20 I. 04 1.09 I. 15 I. 18 I. 17 I. 19 I. 17 I. 17 I. 19 I. 17 I. 17 I. 20 I. 18 I .Oll 1.08 I. Ill I .Oll 1.06 I. 10

[X I. Ill I. 18 I~ 16 I. 18 I. 16

[X '* 16 I. 16 I. 16 I. 18

[X I. 16 I. 16

'* 16 I. 16 I. 19

[X I. 19 I. 17 I. 15 I. 11 I .Oll 1.05 1.07 1.08 I. 10 I. 12 I. 10 I. 10 I. I 2 I. 10 I. 11 I. I 2 I. 11 1.09 1.08 I. 07 I .Oll 1.0ll 1.0ll 1.04 1.05 I. 05 1.05 1.05 I .05 1.05 1.05 1.06 1.06 1.06 1.06 1.06 1.061 REVISION 6 FEBRUARY 15, 1987 Rodwise Power Distribution in a Typical Assembly PUBLIC SERVICE ELECTRIC AND GAS COMPANY (Assembly G-9) Near BOL, HFP, Equilibrium Xenon, SALEM NUCLEAR GENERATING STATION Unrodded Core - Unit 2 Updated FSAR FIG. 4.3-118

FREQUENCY OF AXIAL FREQUENCY SIZE FREQUENCY GAPS PER DISTRIBUTION OF DISTRIBUTION OF ROD-F g GAPS-F.J GAPS-Fk SPIKE DUE TO SINGLE

~ "

i--~~~-'-~~~--r

~ MTG FROM DENSIFICATION MODEL (MAX. GAP SIZE GAPS-Sg ~ DRAW COMPUTER / AT GIVEN HEIGHT)

CODE MODE I "'~M_o_o_E_2~~~~~

'\- EXPECTED VALUES

' . SEEN BY INCORE DETECTOR

PROBABILITIES OF EXCEEDING GIVEN SPIKE SIZE FOR EACH AXIAL LOCATION

.ROD CENSUS 1 r CONVOLUTION POWER SPIKE FACTOR-S(Z}

REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Flow Chart for Determining Spike Model SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-17

12.0 10.0

- - - - 15 x 15 17 x 17 a..

~

Cl 8.0 UJ

-I Cl

z:

Cl) 0 t-UJ

J 6.0 Q

UJ

ii.:

a..

Cl) t-z: --- - -- - ---

UJ

(..)

Cl: ~.o UJ a..

2.0 0

0 1.0 2.0 3.0 ~.o GAP SIZE (INCHES)

REVISION 6 FEBRUARY 15, 1987 Predicted Power Spike Due to Single Nonflattened PUBLIC SERVICE ELECTRIC AND GAS COMPANY Gap in the Adjacent Fuel SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-18

\ ,...

0

\

\ II

\ I-

\ Cl) z:

..... 0

\ 0 N

\ a:

I-

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c

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c.o <

iI 0 N

0 N

0 0 8 REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Power Spike Factor as a Function of Axial Position SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-19

2.4 2.3 2.2 Q) 2.1

~

0 Q. 2.0 2.32@0' 2.32@ 6'

-u.0 N

1.9 2.18@ 10.8' x<( 1.50@ 12' 1.8

?:

1.7 1.6 1.5 1.4 0 1 2 3 4 5 6 7 8 9 10 11 12 CORE HEIGHT (ft)

REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Maximum F 0 - Power vs Axial Height During Normal Operation SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-20

20 0

oO 0

8 0

15 I-s: . 10

~

5 0

-50 -25 0 25

~I REVISION 6 FEBRUARY 15, 1987 Peak Power Density During Control PUBLIC SERVICE ELECTRIC AND GAS COMPANY Rod Malfunction Overpower Transients SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-21 I_

20 15 0

10 0

0 0

CD 5

-50 -25 0 25

.6 I REVISION 6 FEBRUARY 15, 1987 Peak Linear Power During Boration/

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Dilution Overpower Transients SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-22

o. 759

o. 792

4. 31!.

I. 2 17 I. 224 0.6~

o. 774 1. 255 0.800 I. 249

3. 4C'. -0.51:

I. 229 I. 229 I. 225 I. 2 18

-0. 3" -0. 9" I. 109 I. 077 I. 107 1.09 2

-o.r I. 4".

1. 202 I. 223 I. 170 I. 256

- 2. 7" 2.7 ..

0.523 I. 217 I. 22.1 I. 217

o. 548 I. 203 I. 233 I. 210

4. 6". I. I'l'. I .O*" -0 .6 ..

I. 229 I. 189

-3.3".

I. 229 I. 220

-o.r I. 2 17 ...-- CALCULATE D I. 21 I

-0. 5 ..

....-- MEASURED

......__ DIFF£RENC E

-F PEAKING FACTORS z

I. 5

~H 1.357 FNQ 2.07 LOCATED AT M-8 SOUTH Comparison Between Calculated and Measured PUBLIC SERVICE ELECTRIC AND GAS COMPANY Relative Fuel Assembly Power Distribution SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.J..23 REVISION 6 FEBRUARY 15, 1987

I

8 Q.

0 t-0 UJ (I) 0 z: m 0z: """0 -

z ............

UJ

>< t- "'

N 0

z 0::

Q UJ ""' 0

....J ::& z: 00 0

cc 0~ cc""

0 r--

~

. UJ

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c:.:> UJ

"" 0:: 0 0:: 0 UJ UJ <...> f- t-

> :z 0

c N

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<O UJ % UJ 0:: 0 II ....J 0 0:: 0 II

c

<...> u... "" "" 0

<...> "" UJ c:z::

0 0 u I.Cl u..

0 I-

z:

UJ 0 u

1" c:z::

UJ Q.

0 N

0 0 ao <D :I" N 0 0 0 CJ

~3M0d 3AllV13~ lVIXV REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Comparison of Calculated and Measured Axial Shape SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-24

X REACTOR I

\l REACTOR 2 0 REACTOR i+

REACTOR 5 0 e REACTOR 3 FQ 0

0 3.0 0

2.5 o*

0 x V'

o

0 x

xx

~ ~

0 V'@°'

x

\7 x V'

x x

2.0

O

i' 0

oO

x

0 ~ 8

- - 8"'~ f ~ *~

111xO x 0

~ <%).. *0:

670~~

0 co. 0

~

o~

I. 5 ij5 -ijQ 30 -25 15 -10 -5 0 5 I0 15 20 25 30 INCORE AXIAL OFFSET (%)

REVISION 6 FEBRUARY 15, 1987

PUBLIC SERVICE ELECTRIC AND GAS COMPANY Measured Values of F 0 for Full Power Rod Configuration SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.J.25

20 u.

0...._

E CJ 0.

I-

a: 10 2000 PPM

<O

i..

<O I-zw u 0 u.

u.

w 0

u w

er:

> -10 I-

<(

a:

w Q..

a:

w I-er: -20 0

I-

<(

er:

w Cl 0

a:

-30 0 100 200 300 400 500 600 MODERATOR TEMPERATURE (°F)

REVISION 6 FEBRU.ARY 15, 1987 Moderator Temperature Coefficient - BOL, PUBLIC SERVICE ELECTRIC AND GAS COMPANY Cycle 1, No Rods - Unit 1 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-29A

20 0

LL E

(,,)

Q.

~ 10

-I-

"°Cl.

ro 1-'

zw 0 u

LL LL w

0 u -10 w

a:

I-

<(

a:

w -20 Q..

~

w I-a:

0 I- -30

<(

a:

w c

0

~

-40 0 100 200 300 400 500 600 MODERATOR TEMPERATURE (°F)

REVISION 6 FEBRUARY 15, 1987 Moderator Temperature Coefficient, BOL, PUBLIC SERVICE ELECTRIC AND GAS COMPANY Cycle 1, No Rods - Unit 2 SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-298

50 Cl..

w I- ijQ

(/)

E

(.)

Cl..

z::

I- 30 0::

i Cl 0

0::

....I 20 I-z:

w 0::

U.J u..

LL-10 Cl 0

0 50 100 150 200 250 STEPS WITHDRAWN REVISION 6 FEBRUARY 15, 1987 Accidental Simultaneous Withdrawal of 2 Control PUBLIC SERVICE ELECTRIC AND GAS COMPANY Banks EOL, HZP, Banks D & B Moving in the Same SALEM NUCLEAR GENERATING STATION Plane Updated FSAR FIG. 4.3-36

160 150 lijO 130 120 110 100 z: 90 z:

0 80

-I-en THIMBLE DASHPOT 0

Cl.. 70 60 50 ijO 30 20

. 10 0

0 .2 . ij .6 .8 1.0 1.2 I .ij 1.6 1.8 2.0 2.2 2.ij 2.6 TIME OFT.RAVEL (SECONDS) REVISION&

FEBRUARY 15, 1987 Design Trip Curve PUBLIC SERVICE ELECTRIC AND GAS COMPANY

SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-37

1.00 0.5 0

100 50 0 ROD POSITION (PERCENT INSERTED)

REVISION 6 FEBRUARY 15, 1987 Normalized Rod Worth vs Percent Insertion PUBLIC SERVICE ELECTRIC AND GAS COMPANY All Rods But One SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-38

N r-co

~

0 U"I N

Li'>

co

~

0 z:

~

a<

~

0

~

0 0

~

-°"

VJ 0

z::

- 0

c 00 (")

<D ~

I I LJ.J

i;:

"' N

(") I-

- 0~ ~ ~ ~

Q::

L.o..I a<

~1 0

~

0

~

z: .:::. "

3:

i:

N MN 0

0 o.n ' 0,._

N co

~

N 0

N

~I- N N

-I

-I REVISION 6 FEBRUARY 15, 1987 Axial Offset vs Time - PWR Core With a PUBLIC SERVICE ELECTRIC AND GAS COMPANY 12-Ft Height and 121 Assemblies SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-39

10 8

6 2

N

STABILITY INDEX: -0.076 (HR.- I) 1-

z:

ct 0

ci:::

0

5 0-

- 2

N42 N44

0 0

I-

_J I-

z:

ct ci:::

~ -12

-10 8

0 N43 4

3 2

RCCE-11 0-

-14 *

-16 0 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 HOURS AFTER WITHDRAWAL OF RCC E-1 I REVISION 6 FEBRUARY 15, 1987 XV Xenon Test Thermocouple Response PUBLIC SERVICE ELECTRIC AND GAS COMPANY Quadrant Tilt Difference vs Time SALEM NUCLEAR GENERATING STATION

Updated FSAR FIG. 4.3-40

(wJd) 1J3j3Q H3ldd00 l~MD31NI g 0 g 8 g 8 ~ 0 8 g N

N 8

N 00<0

t'N 0

Cl 00<0 :::t' 0

0 N 0 I I I . I I I I I I I I 0

0 0

00 0

a:::

L.U

<O 3:

0 a..

z: ::::>

~

'.2 0 l.L'I

""':z:

en 0

..., ~

....:::c 0-:z: ~ z:

L.J.J u

0

'-' en < ~ 0

t' a:::

L.U a..

i ....

z:

....:::c >- 0

z:

~

cc :z: -

_, 0

'-' :::c 0

z: :;; ""'

..... '-' 0

z:

0 0

cc ..., '-'

0 >'.

0 <l 0 N

0 0

0 0 0 0 0 0 0 ci ci ci ci ci

<O I

i.n I

t' I ""I N

I

-0 I REVISION 6

( d ~/WJd) 1N31 JI jj]QJ M3ldd00 FEBRUARY 15, 1987 Calculated and Measured Doppler Defect and PUBLIC SERVICE ELECTRIC AND GAS COMPANY Coefficients at BOL, Two Loop Plant, 121 Assemblies, 12-Ft Core SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-41 L

1400 1200 1000 E

800 a.

a co u

600

~00 CALCULATED /

200 0

0 2000 400Q 6000 8000 10000 12000 I 4000 16000 I 8000 BURNUP. ~ff/D /MTU REVISION 6 FEBRUARY 15, 1987 Comparison of Calculated and Measured Boron PUBLIC SERVICE ELECTRIC AND GAS COMPANY Concentration for 2-Loop Plant, 121 Assemblies SALEM NUCLEAR GENERATING STATION 12-Ft Core Updated FSAR FIG. 4.3-42

8 0

0 N

Cl) 8 C")

<.O 0

0 LCl

=

f--

0 :;:;::::

0 ........

<.O Cl C")

3 0..

z:

0:::

0 ::::::>

0

,..._ co N

0 0

<.O

....er 0

=>

C(

E: 8 O')

0 8 8 8,..._ 0 0 8 0 0

("') <.O LCl =

REVISION 6 FEBRUARY 15, 1987 Comparison of Calculated and Measured CB 2-Loop PUBLIC SERVICE ELECTRIC AND GAS COMPANY With 121 Assemblies, 12-Ft Core SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-43

0 0

0 C"')

0 0

0 N

0 0

0 80 0

0 8en s~

o~

cc -Cl 8~

0

,...._ a..

~

0 :z::

8 §3

<.o CD 0

0 0

LO 0

0 0

~

0 0

0

("")

80 N

8 0

0 8 8en 0 0

CD 0

0

~

8 l.l')

8 N

8 0 REVISION 6 FEBRUARY 15, 1987 Comparison of Calculated and Measured c 8 3-Loop PUBLIC SERVICE ELECTRIC AND GAS COMPANY Plant, 157 Assemblies, 12-Ft Core SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 4.3-44

Status TABLE 6.2-10 (Continued)

Inside Containment Outside Containment Auto I sol.

Time Figure service Class N. ~ l Valve(s2 ~ Pwr-Signal Valve(s2 ~ Pwr-Sig~al ~ Fluid Te~p.

6.2-24 Reactor Coolant B Open Open Closed 1CC190 # Auto- A p 1CC131 # Auto- c p ~10 Liquid Co~d Pump Thermal trip trip Barrier Cooling Closed Closed Closed 1CC208 # Non- N/A N/A Water Discharge return 6.2-25 Gas Analyzer B Int. Int. Closed 1WL96 # Auto- c T 1WL97 # Auto* B T ~10 Gas Cold from RCDT trip trip 6.2-25 N2 Supply B Open Open Closed 1WL98 # Auto- c T 1WL108 # Auto- B T ~10 Gas Cold to RCDT trip trip J

6.2-25 Reactor Coolant B Open Open Closed 1WL98 # Auto- c T 1WL99 # Auto- B T ~10 Gas Cold Drain Tank Vent trip trip 6.2-26 Reactor Coolant B Int. Int. Closed 1WL12 # Auto- c T 1WL 13 # Auto- B T ~10 Liquid Hot Drain Tank Pump trip trip Discharge B N/A N/A N/A 1WL476 Relief N/A N/A 6.2-27 Accumulator N2 B Closed Closed Closed 1NT34 # Non- N/A N/A 1NT32 # Auto- D T ~10 Gas Cold*,

Supply return trip 6.2-28 Safety Injection B Closed Closed

Closed 1SJ123 # Auto- A T 1SJ53 # Auto- D T <10 Liquid Cold Test line trip 1SJ60 # trip B T ~10 Liquid Cold 6.2-29 RHR Outlets to D Open Open Open 11SJ49 (2)Rem. A N/A Note Liquid Co~~

Safety Injection 12SJ49 Manual B 14 System & Closed System 6.2-30 Safety Injection D Open Open Open 11-14SJ144 Non- N/A N/A 1SJ135 Rem. B N/A Note Liquid Cold

" Pumps Outlet to return Manual 14 Cold Legs and Closed Closed Closed 1SJ158 Rem. B N/A Test Line Manual

For !Jni t 2 Only 3 of 12 SGS-UFSAR Revision 17 I* *: October 16, 1998

Status TABLE 6.2-10 (Continued)

Inside Containment Outside Containment Auto lsol.

Time Figure Service Class* N §. l Valve{s~ ~ Pwr-Signal Valve{s~

I

~ Pwr-Signal (Sec) Fluid Temp.

6.2-,~o Auxiliary Feed- c Open Int. Open Closed 11AF11 Rem. c N/A Note Liquid Cold through water supply Open Int. Open System 12AF11 Manual c 14 6.2-~3 Turbine Driven Open Int. Open 13AF11 c Open Int. Open 14AF11 c 6.2-40 Auxiliary Feed- c Closed Int. Open Closed 11AF21 Rem. B N/A Note Liquid Cold through water supply Closed Int. Open System 12AF21 Manual B 14 6.2-43 Motor Driven Closed Int. Open 13AF21 A Closed Int. Open 14AF21 A 6.2-44 Reactor Cavity B Int. Int. Closed 1\.IL 16 # Auto- c T 1\.IL 17 # Auto- B T ~10 Liquid Cold sump Discharge trip trip to \.laste Disposal B N/A N/A N/A 1\.IL478 Relief N/A N/A 6.2~45 Fire Protection B Closed Closed Closed 1FP148 # Non- N/A N/A 1FP147 # Auto- c T ~10 Liquid Cold Water Supply return trip 6.2-45 Refueling Canal B Closed Closed Closed 1WL190 # Manual N/A N/A 1SF36 # Manual N/A N/A N/A Liquid Cold Supply 6.2-.45 Refueling Canal B Closed Closed Closed 1WL191 # Manual N/A N/A 1SF22 # Manual N/A N/A N/A Liquid Cold Discharge 6.2-~~A Post LOCA B Closed Closed Int. 11VC19 # Rem. A N/A 11VC17 # Rem. A N/A Note Gas Cold

. i Atmosphere Manual Manual 14 Sample B Closed Closed Int. 11VC20 # Rem. A N/A 11VC18 # Rem. A N/A Note Gas Cold Manual Manual 14 B Closed Closed Int. 12VC20 # Rem. c N/A 12VC18 # Rem. c N/A Note Gas Cold Manu<!l Manual 14 B Closed Closed Int. 12VC19 # Rem. c N/A 12VC17 # Rem. c N/A Note . Gas .Cold Manual Manual 14

For Unit 2 Only 8 of 12 SGs-pFsAR Revision 17 October 16, 1998

""C c

en OJ

)> !:

ro men S: m ............

2 ::i:J CD CJl c< CD o

ro - 0) mm

)> m

n rm C> (")

m -c 2 ::i:J RELIEF.VALVE RELIEF VALVE m- * ~ ACTUATION

0 (") * / ACTUATION

)> )>

-c 2

-o 2 C>

G) )>

~ en

)> 0

-c 0

-~

@--e-~

0 ""C 2 )> I 2 SI (SEE SI (SEE I SI (SEE

-< FIG. 7.2 _, l I F I G. 1* 1..-6) I FIG. 1 Z-6)

I I p - p ref c

"'O f t I

a. AVAILABLE FOR AVAILABLE FOR 3 MODE QI CONTROL CONTROL I SIGNAL It>

a. RELIEF "Tl FUNCTION FUNCTION VALVE I~

en

)>

i:J

...It>

""C H1. P. R.l.

SWITCH ING DUR I HG TESTS Hi. P. R.T.

SWITCHING DURING TESTS Hi. P. R.T.

No. 2.

AC TUA TI OH 9

~VALVE RELIEF QI

J ~ No. I

a. c:

""C N

ACTUATION 0 It>

R.T~

It> ""C 0 ...

g. ~ P. R.T. Lo. P. R.T.

~SPRAY

J c: Lo. P. R. T. Lo. Lo. P.

en~

~ (") I VALVES

.... 0 I It> :::J 3 !::t

~

. !2.

IO CH III 10 CH IV TO CH II TO CH I "Tl O.T. ATs.P. O.T.Ms.P. O.T.Ms.P. O.T.Ms.P. VARIABLE

.o*

c:

It>

HEATERS

13.l.L2~1.5 Director - Nuclear Business Support ____ -:-

Nuclear Business Support is responsible for providing support services to the Nuclear Business Unit. Included within this support are direct services to departments within the Nuclear Business Unit and services to the corporation and external stakeholders on behalf of the Nuclear Business Unit. Responsibilities include: project management, purchasing and materials management, integrated site planning, maintenance and outage planning, external affairs including co-owner activities, industry and community affairs, supporting rate counsel and legal affairs, and internal and external communication; financial services including capital, operating and maintenance, and co-owner's budgets; strategic planning, financial planning and cost analysis.

13.1.1.2.1.6 Nuclear Human Resources Manager The Nuclear Human Resources Manager (NHRMGR) directs and controls various human resources program and administrative services functions necessary to support the Nuclear Business Unit *

SGS-UFSAR 13.1-Ba Revision 17~

October 16, 1998

THIS PAGE INTENTIONALLY BLANK 13.1-8b Revision_ 9*

July*-22~" 1989-

operations. Responsibilities- of the Operators __ assigned- to--r.adwaste include the following:

1. Completing checkoff lists, logs, and other shift data associated with radwaste operations to provide continuous surveillance of the equipment assigned

2. Manipulating controls, valves, and equipment to support liquid radwaste processing and storing

3. Initiating immediate actions necessary to maintain radwaste equipment in a safe condition during normal, abnormal, and emergency operations Shift electrician, instrumentation and control (I&C) technicians, chemistry technicians and radiation protection technicians are assigned to shift schedule and report to the Operations Superintendent. These personnel perform support functions associated with electrical, I&C, chemistry and radiation monitoring disciplines. During normal operation, they are available to perform surveillance, preventive and corrective maintenance. When periods of emergency or abnormal operating conditions exist, they are available as part of the plant emergency preparedness program for emergency response and technical assistance *

SGS-UFSAR Rev£sion 17 October 16, 1998

13.1.2.3- Maintenance Department_

The Nuclear Maintenance Organization is described in Section 13. 1.1. 2. l. l.

Although the Maintenance Organization will not report directly to the Plant Manager, the Plant Manager will maintain control over those activities necessary for safe operation and maintenance of the plant.

13.1.2.4 Chemistry Department The Chemistry Superintendent re[ports to the General Manager - Salem Operations and is responsible for implementing programs to ensure plant chemistry, radiochemistry~ and plant *effluents monitoring are in accordance with the facility license and government regulations.

The Chemistry Department is responsible for the development and implementation of the chemistry, radiochemistry, environmental and liquid effluent monitoring programs. They are also responsible for operation of the condensate demineralizers, demineralized water makeup plant, service water chlorination, non-radioactive liquid waste disposal system, oil-wter separator and post accident sampling system.

The Chemistry Department is also responsible for the sampling and analysis of lant fluid systems, chemistry results reporting,_ calibration of chemistry instrumentation, evaluation of laboratory and chemisal systems operation and techniques, operation of deep bed demineralizers, plant water and chemical control systems, and maintaining the plant fluid systems and_ liquid effluents within established limits. The Chemistry Department organization is shown on Figure 13.1-Be.

SGS'-UFSAR 13.1-16 Revision-. lT October 16, 1998

13. 5 P.lant_ Procedures *

13.S.1 Administrative Procedures Administrative procedures define processes and programs that provide for the control of nuclear operations, and in turn incorporate regulatory requirements and commitments. There are three types of administrative procedures:

1) Nuclear Administrative Procedures (NAPs); 2) Station Administrative Procedures (SAPs); and 3) Department Administrative Procedures (DAPs).

Nuclear Administrative Procedures (NAPs) are written to provide direction in the areas that are common to all station departments as well as other organizations within the NBU. NAPs are prepared using a standard format and content, and a writers guide, which provides human factors and style guidance. NAPs are approved by the General Manager - Salem Operations.

Station Administrative Procedures (SAPs) are written to govern station specific programs and processes. SAPs are approved by the General Manager - Salem Operations and comply with all applicable requirements specified in the NAPs.

Department Administrative Procedures (DAPs) provide direction for the administrative control of specific activities that are within a department's functional area of responsibility or between departments with the same functional responsibility or that control administrative functions between a limited number of departments in the NBU. Department - specific procedures are approved by the individual department managers for Salem and comply with all applicable requirements specified in the NAPs.

Additional topics for administrative procedures may be addressed as required, and material may be shifted between specific procedures as needed.

A list of topics for NBU administrative procedures is listed below:

Action Request Process

Nuclear Procedure system

Nuclear Department Organization

Document Control Program

station Operations Review Committee

Corrective Action Program 13.5-1 SGS=UFSAR Revision- IT October' 16-, 1998.

Control of Design and Configuration_ Changes., Tests~ and. Experiments I

Work Control Process

Preventive Maintenance Program

Records Management Program

Technical Specification Surveillance Requirements

Control of Temporary Modifications

Training, Qualification and Certification

Safety Tagging

Monitoring the Effectiveness of Maintenance

Minor Modification Process

Material Control Program

Procurement of Materials and Services

System Cleanliness

Measuring & Test Equipment, Lifting & Rigging and Tool Control

Scaffolding Program

Radiological Protection Program

Fire Protection

Nuclear Mutual Limited/Boiler and Machinery Insurance Program

Inservice Inspection Program

Code job Packages

Commitment Management Program SGS~UESAR~-

13.5-2 Revision.11*

October 16 , 19 9.8..

Inspection/Housekeeping- Program Nuclear Security Program Nuclear Licensing and Reporting Environmental Control

Chemical Control Program

Service Water Reliability Program

Lubricant Program

Fitness for Duty Program

Vendor Information Program

Stations Aids and Labels

Respiratory Protection Program

Station Performance and Reliability

Refueling Management

Station Testing Program

Plant Chemistry Control

Operating Experience Feedback Program

Outage Management

Action Tracking Program

10CFRS0.59 Reviews and Safety Evaluations

Repairs to Presure Relief Devices

SGS--UFSAR 13.S-2a Revisi*on 1.7 ..

October:* 16:~. 19.9.8;

Environmental Qualification Program Software and Micro-processor Based Systems (Digital Systems)

Control of On-Site Contractor Personnel Inservice Testing Program

Fuel Integrity Program

Nuclear Fuel Program

Special Nuclear Material Control Program

Valve Programs

Independent Review Program

Transient Loads

Conduct of Infrequently Performed Tests and Evolutions SGS-UFSAR 13.5-2b Revision_ 17 October 16, 1998

by low* pressurizer pressure. The *transient- isc*quite conservative ..with respect:

to cooldown, since no credit is taken for. the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators. Since the transient occurs over a period of about 10 minutes, the neglected stored energy is likely to have a significant effect in slowing the cooldown.

15.2.13.4 Conclusions The analysis has shown that the criteria stated earlier in this section are satisfied, since a DNBR less than the design DNBR limit does not occur.

15.2.14 Spurious Operation of The Safety Injection System at Power 15.2.14.1 Accident Description The Spurious Operation of the Safety Injection System (SIS) at Power is caused by either an operator error or a false electrical actuating signal.

When the SIS is actuated, charging pump suction is diverted from the Volume Control Tank to the RWST, and boric acid is pumped from the RWST to the cold leg of each reactor coolant loop. The safety injection pumps are also started automatically; but they cannot develop the head necessary to pump borated water into the reactor coolant loops when the RCS is at normal operating pressure.

The Spurious Operation of the SIS at Power is classified as a Condition II event, a fault of moderate frequency. The acceptance criteria for analysis of this event are:

1. Fuel cladding integrity shall be maintained by ensuring that the minimum DNBR remains above the applicable DNBR limit.

2. Pressure in the reactor coolant. and main steam systems should be maintained below 110% of the design values.

3. A Condition II must not escalate into, or cause a more serious fault (e.g., a Condition III or Condition IV event) without other faults occurring independently *

SGS-UFSAR 15.2-55 Re:vision. 1.7

October 1*6; 1998*

15.2.14.2 Method of Analysis The first criterion, that fuel cladding integrity be maintained, is shown to be satisfied by means of a safety evaluation (see Case 1 below). The remaining criteria, that the RCS and main steam system pressure limits are not exceeded, and that the event would not lead to a more serious event, are demonstreted by means of an accident analysis (see Case 2 below).

Case 1. Safety Evaluation to show that fuel cladding integrity is maintained.

If no reactor trip signal is assumed to be generated by the SI signal, then borated water from the SIS would cause core reactivity and power level to drop, and consequently, the calculated DNB ratio to rise. The calculated DNBR would increase throughtout the transient, without ever approaching its safety analysis limit value. Therefore, the Spurious Operation of the SIS at Power could not lead to any fuel damage.

Case 2. Accident Analysis to show that RCS and main steam system pressure limits are not exceeded, and that the event would not lead to a more serious event.

During a Spurious Operation of the SIS at Power event, the addition of borated water from the SIS, into the RCS, can fill the pressurizer and eventually lead to the discharge of water through the pressurizer safety valves. Since the pressurizer safety valves have not been qualified for water relief, one or more of the valves might fail to reseat completely, and thereby create an unisolatable leak from the RCS. Such a situation would be an escalation of a Condition II event into a more serious event (a small break LOCA), a violation of the third acceptance criterion.

. SGS-UFSAR 15.2-56 Revision 17 October 16, 1998

THIS FIGURE HAS BEEN DELETED.

REVISION 9 JULY 22, 1989 PUBLIC SERVICE ELECTRIC & GAS COMPANY

- SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.2-44

~--- - - - - - - - * -

~

L.l.J

~

J

_J C>

70 60 0::

L.l.J I-3:

0::

50 L.l.J N

u::

Cl')

4-0 Cl')

L.l.J 0::

0...

~ 30 C>

I-z:

UJ u 20 0::

L.l.J 0...

10 2300 2200 2100 C/')

0...

2000 UJ er:;

J Cl)

Cl)

L.l.J er:; 1900 0...

0::

UJ N

0:: 1800

J Cl')

C/')

UJ 0::

0... 1700 1600 0 20 4-0 60 80 100 TIME {SECONDS)

REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Spurious Actuation of Safety Injection System at Power SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.2*45

i

I. I 1.0 0.9 3:

0.8 "I 0

.....J

.....J I

LL.

~

<C

z 0.7 I STEAM FLOW

<C L..LJ i-(/')

a:z 0.6 I

..a 0 u... I 0:::

L..LJ :z 0.5 3: 0 0

Q..

I-

<.,;)

<C 0:::

<( 0:::

u...

0. 4 UJ

.....J

<.,;)

> 0.3 I

z I

0.2 I I

0.1 I I

0 0 20 40 60 TIME (SECONDS) 80 100 120

-60

-50 u...

0

- Q

-40 I-

z -30 L..LJ

<;,::l

z -20

<(

c

<.,;)

-10 0

0 20 40 60 80 100 120 TIME (SECONDS)

REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Spurious Actuation of Safety Injection System at Power SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 15.2-46

~

2.

i I I i . '

I I I I I II I I I ;1

45. I 11 I I11I ;

I v ?"--.t.,.I I'.'

1*11i1:1 1 * ':

I 40, \ 11 II It I* I I:

1 r-... .........

G 35.

I I Ul I \' Ii I'i 1 :11 rJl

~

w

~

rJl rJl 30.

I \

\

°'

~

z 25. 1,

~ \

~

a:

\ ~ .1

~ 20.

u

""" \

I' I!

15. '

J

\ 11 v . ' r-.

I .. ~

1'!l.

s.

100 1111' 112 115 . 11 4 115 106 Til'IE I SECONDS I PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION CONTAINMENT PRESSURE VERSUS TIME DOUBLE ENDED PUMP SUCTION BREAK Updated FSAR Revision 12, July 22, 1992 Figure 15.4-91

.J.

- 250

~

2DC

~

en 150 1000.~1------o-.3--------------3~-----1-0------30--------,00------300----------',ooo 1'1iiE <SEC>

50

~

~

If 40 3CI z 20 I!

u 10 0

0.1 0.3 , 3 10 30 100 , 000 TliE <SIC>

Containment Response to Steam Line Rupture -

PUBLIC SERVICE ELECTRIC AND GAS COMPANY 30% Power, 4.6 Ft2 DER, SALEM NUCLEAR GENERATING STATION Feedwater Control Valve Failure

Updated FSAR Revision 1 5 June 12, 1996 Figure 15.4-98

!+

JS()

~

""' 300 i; 250

-~

2CIO en 150 10Q , 000 0.1 0.3 3 10 JC 100 3QQ TIME (SEC)

i

~ lC II 2D I§ 1C c 1 3 10 30 100 1.00C 0.1 0.3

,,_<SIC>

Containment Response to Steam Line Rupture -

PUBLIC SERVICE ELECTRIC AND GAS COMPANY 102% Power, 0.6 Ft2 Small DER.

SALEM NUCLEAR GENERATING STATION Main Steam Isolation Valve Failure Updated FSAR Figure 1 5 .4-1 00 Revision 15 June 12, 1996

ML18106A981

Text

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