LD-96-029, Forwards Finalized Package of Six Draft Changes to Sys 80+ Std Plant Design Control Document.Changes Will Be Formally Incorporated Into Sys 80+ Design Control Document at Conclusion of Design Certification Rulemaking

Forwards Finalized Package of Six Draft Changes to Sys 80+ Std Plant Design Control Document.Changes Will Be Formally Incorporated Into Sys 80+ Design Control Document at Conclusion of Design Certification Rulemaking

ML20116C422
Person / Time
Site:	05200002
Issue date:	07/25/1996
From:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LD-96-029, LD-96-29, NUDOCS 9607310161
Download: ML20116C422 (17)

Text

A ED D M ED D i l

July 25,1996 LD-96-029

)

l Docket 52-002 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

System 80+ Standard Plant Design Changes

Reference:

Letter, C.B.Brinkman to NRC," System 80+ Standard Plant Design Changes,"

LD-96-028, July 17,1996

Dear Sirs:

The Reference letter forwarded six (6) draft changes to the System 80+ Standard Plant Design for Staff review and approval.

This letter transmits a finalized package of these six changes to the System 80+* Standard Plant Design Control Document. These changes are listed on the attached table with details shown on the enclosed Design Control Document pages. None of these changes are required to meet Commission regulations or to maintain the safety conclusions of the FSER. All changes will be formally incorporated into the System 80+ DCD when it is reprinted at the conclusion of design certification rulemaking.

Please call me, Stan Ritterbusch (860-285-5206) or Virgil Paggen (860-285-4700) if you have any questions.

Very truly yours, COMBUSTION ENGINEERING,INC.

l U- I

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gNC1 /W ~

C. B. Brinkman Director, Nuclear Licensing cc: J. N. Wilson (NRC) -

8 F. A. Ross (DOE, w/o enclosure) -

/

9607310161 960725 3 PDR ADOCK 05200002 A PDR ABB Combustion Engineering Nuclear Systems Combustm Engineenng inc 2000 Day Hdi Road Telephone (B6C) 688-1911 PO Bo< 500 Fax (861) 285 5203 W:ndsor CT 06095-0500

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System 80+ DCD Changes - 7/25/96 l

Item Reference Design Change Description '

Sect 5.2.3.3+ Revise sections and table to clarify and correct material specifications, and to incorporate 1 Tbl 5.2-2 associated editorial changes.

2 Sect 5.4.2.4.1 Revise wording to permit both explosive and hydraulic tube expansion. '

3 Tbl 7.5-2 Incorporate changes in Ctmt Spray system. I 4 Sect 7.5.1.1+ Editorial clarification; System 80+ has 61 CETs arranged to display in 2 channels in the l Tbl 7.5-3 control room.

5 Fig 7.5-5 Figure corrected to be consistent with CEA locations shown in Ch 4.

6 19.11.3.3.2 Editorial correction; replace ICI

• plates" with ICI " nozzles."

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I 4

- \

4 4

1 1

1 M < ws I.

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Item: Undate Reactor Coolant System Material Snecifications I.

I DCD Section: 5.2.3.3.2.1. Table 5.2-2 l Descrintion of Chance:

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i l

. Recent System 80+ detail design work and re review of Table 5.2-2 by ABB-CE demonstrated need to update Table 5.2-2 in order to:

(1) clarify material specifications (2) . add material options for certain components '

4 (3) delete material options where no longer used in certain components (4) add omitted component parts / material specifications (5) remove inconsistencies te.g. between Table 5.2-2 and Table 3.9A-2) 1 (6) rework the Weld Mater;als section to make it better organized and more complete j (7) make editorial corrections i

Related Channes to DCD l

j DCD Section Insert # Revision

15.2.3.3.2.1 Revise as marked

Table 5.2-2 Revise as marked

3 j Table 5.2-2 5.2-2 replace current Weld Materials secuon with Insert 5.2-2 (below) 1 1

1 4 -

i 4 .

7

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System 80 + Desian control Document Table b.2 2 Reactor Coolant System Materials Component l .Nf aterial Specification Reactor Vessel: e-H Forgtogs SA-508 Class.23e3 Cladding (*) Weld deposited austenitic stamless steel with 5FN 18FN delta femte or NiCrFe

~

alloyhuivalent to SB-1661 G. ease DVI Nonle Safe Ends SA-182,F3 T6 Reactor vessel head (**" CEDM Nonles l NiCrFe Alloy 690 (SB-166)

Vessel intemalst " l Austemtic Staunless Steel and NiCrFe alloy Flow skirtl 'd" l NiCrFe Alloy 690 (SB-168)

Fuel cladding;*) l Zircaioy-4 Instrument nozzles"d" l NiCrFe Alloy 690 (5B-166) c  %

Control element drive mechamsm housings: [^

U YVF2 V8 )

Lower'd Type 403 stamiess steel according to Code Case N 411 with end fittings to be SB 166 s-Alloy 690"' and SA-182 '", J~"'stamiess steel Upper'd SA-479 and SA-213 Type 316 stainless steel with f.-

g end fittingf SA 479 Typ QType 316 and vent valv,e els'eal of 316.anNnt valve Type 440C-stamless steel R./--

Closure head baior 2.#* m i l SA 540 B24 or B23 Pressuruer:

Shell SA-533 Type A or B Class 1 or SA 508 Class 3 Cladding:d Weld deposited austenitic stainless steel vath 5 FN 18FN delta ferrite or NiCrFe alloy Gequivalent to SB-1667" Forged noules y ,

SA-541xClass 3 or SA 508 Class 3 Instrument noules and heater sleeves *Xd NiCrFe Alloy 690 (SB 166f Cr- S 3 ""I G.ye dt Nonle safe ends (d SA-182 F316 or F347 gA Safety valve noule flange ld SA 182 Grode. F 316 Studs and nuts 9B +W- S B - 0 3 7 wm Demon Menenet - RCs aren commeneen syseenes  !*r* 6 N1

I System 80+ Design Control Document l Table 5.2-2 Reactor Coolant System Materials (Continued) l l

Component Material Specification Steam Generator:

Prunary Head SA 533 Type B. Class I or SA-508 Class 3 Primary Nozr.les SA 508 C' 1 Primary Side Instrument Nozzles <.xo NiCrFe Alloy 690 (SB-166)

Primary head cladding (*) Weld deposited austenitic stainless steel with 5FN-18FN delta ferrite or #Cr Fe. a ll*7 Tubesheet SA 508 Class j Tubesheet stay SA-508 Class Tubesheet claddmg(0 Weld deposited NicrFe A M#)'

Tube ('xo NiCrFe Alloy 690 (SB-163)

Tube supports ASTM A-176. Type 409 Secondary shell(d) SA-533 Type A or BxClass Igo y Class 3 Secondary head'd) SA-516 Grade 70)( SA-508 x Class Igor SA-fi3 5 Tp A or B I

Secondary nozzles (d) SA 508 3 Class h44 met.1& or .3 j 4aadary nozzle safe ends (d) SA-508 Class lg Secondary instrument nozzles (d) SA 106 Grade B Secondary studs and nuts SA 540 Grade B24, or SA-193 Grade B7 (Sl$ggy E#o d)

~

Primary studs and nuts /fAl64 Type 630H1100 637xNO7718 Reactor Coolant Pumps:

CasingM SA 508 Class and cladded with austenitic stainless steel Claddmg(4 Weld deposited austenitic stainless steel with 5FN 18FN delta ferrite Intemals(d SA-487 CA6NM, SA-336 Type 304pr austenitic 4m hus stainless steel 5kJ4- 1s- telGrode. FG No mayet# s A- wo Tye e*

S) m-

f- 1 l

System 80 + Design controt occument \

l Table 5.2-2 Reactor Coolant System Materials (Continued)  !

Component Material Specification t

Reactor Coolant Piping:

Pipe (30 in. and 42 in. ID) SA-516 Grade 70 or SA-508 Class / d.C-l Cladding (*) Weld deposited austenitic stainless steel l

., _ . . . . , . , , , , with 5FN 18FN delta femte l

~~

{ ,

__e --

) g d.

l -P'8 Nozzle forgings SA 508 Class 1 or lahme SA-182 C-ndt fi ," i sA-541C*ss i sol 5'tel Reauf 5-io l i Instrument nozzles ('" NiCrFe Alloy 690 (SB-166) l w

SA-182 G,aje F 3 lc, er F347 1Erzle safe ends (

Valvest ' SA 351 CF8M or SA-182 Cr-crde 9 3f Cs I Surge Line') t SA-312 TP347 or TP316 (Piping); )

i SA-403 WP347 or WP316 (Elbow .

I dfE182 F347 or F316 (Safe Ends [s) )'

! D v .L w d. b h 'TN n "* a C 5 o r-l frni. E" In side Co. tnin m ed- SA-312 TP316ATP3044r SA-376 TP316. TP304s Weld Materials for Reactor Coolant Pressure Boundary Components

-x -

Type of Weld Material /y Base Material Type Base Material Ty, .

1. SAh SA-533 a. SFA 5.5.(b) E-8018 C3d8018-G Type B Ch. 1 Type B C1.1 b. MIL-E-18193.

! \ c. SFA 5.23,En3 * ')

4 s

! f 2. SA-508 C1.2

\

N SA-533 Type B C1.1

a. SM '5.5. E-8018-C3, E-8018-G

)>r' MIL-E-18193. B-4 l

l

}dh N

/ c. SFA-5.23. EA3 * *)

a. SFA 5.5, E-8018-C3. E-8018-G I 3. SA-508 SA 508 - ~

Cg C1.1 4 eh 4. SA-516 SA 54'

a. SFA 5.1, E 7018

.N Gr.70 / Gr.70 .s

5. SA 182

1. ~

[ SA-516 \ \

a. SFA 5.1, E-7018 F1 / Gr.70

6. SA-105 - SA-182 Filler Metal

/ F347 a. Nih Fe \

7. SA482 SA-182 a. NiCrFe Filler Me F/ F347 8 SA-105 SA-182 a. NiCrFe Filler Metal F316 ~

l Approvent Deepn Menenal RCS ed Cannocenaf Systones Pope 6.2 3.1 l

I '

(* . .

System 80 + oesian control Document Table 5.5 2 Reactor Coolant System Materials (Continued)

Weld Materials for Reactor Coolant Pressure Boundary Components

\. Base Material Type l Base Material Type I Type of Weld Material /

9. SB2166 l SA-182 a. Root Filler Metal NiCrFe /

N. I F316 Rematning Filler MetaLNiCrFe

10. SB-167 \ SA-182 a. Root Filler MetakNiCrFe

\ F304 _

Remammg Firer Metal NicrFe

! s ~

a. NictFyfiller Metal SA 351 l 11. SA 516 \ /

Gr.70 N CFBM

12. SA 182 \ SA-182 hiCrFe Filler Metal F1 \ F316 ,

13. SB-166 SA-533 a. Root filler Metal NiCrFe l ype B C1.1 Remaming Filler Metal NiCrFe

14. SA-182 l \ l67 / l a. NiCrFe Filler Metal

/

'hp 15. SA-516 Gr.70 S 5 C1.2

m. SFA 5.5.*) E-8018-C3

(,h 16. Austenitic stamless a. SFA 5.9. ER 308 7 steel cladding SFA 5.9. ER 309

17. Inconel cladding ,

! \ a NiCrFe Filler Metal

18. SA-508 SA-508 a) FA 5.5.*) E-8018-C3 E 8018-G C1. 3 C1. 3 b. . -E-18193. B-4

c. SFA .23. EA3* "

19. SA 508 SA-533 a. SFA 5. . -8018-C3 E-8018-G C1. 3 Type B Cl.1 b. MIL E-18 . B-4

c. SFA 5.23. *"

/

20. SA- SA-508 a. SFA 5.5. E-8018 b. E-8018-G Cl C1. 2 \

N

21. A-508 C1. 3 ,

SA-516 Gr.70

a. SFA 5.5.0) E-8018-C3

'-\

h. SB-16fi.162,-4f4 P-155. I5?,15*  :. 90F: Filler Me_ tal Notes: a. Matenals exposed to reactor coolant.

b. Special w 14wjgth low residual elements of copper nickel and phosphorous as specifi the reactor vessel core beltline region.

4

c. Matenal to be provided in the thermally treated condition.

. d. Steam generator secondary side pressure boundary materials including weld n:aterials r='== no greater than 0.010% by weight sulfur (S). .

,, )

g f- rwm bC r MS d'on3 arc ffe OR bbbb CO -

l 71 k /c Q.W - 4 2 2. .

Pope 6.M App, weed Dennyn M. RC3 meat r=" 3posenen

1 r-. ;m,m l

I i

INSERT 5.2-2 l  !

Base Base Type of Weld Material Example of Use Material Material Type"' Type"'

P1 P1 a. SFA 5.1 E-7018. E 7016 Primary Pipmg Straignt Pipe (SA 508 Class Ia)to Ells (SA 516 Gr 70)

P-1 P3 a. SFA 5.1 E-7018, E 7016 Primary Piptng Straignt to RV Pnmary

b. SFA 5.5 E-8018 C3. E-8018-G. Nozzle (SA-508 Class 3)

E-8016-G

c. MIL-E-18193 B-4

d. SFA 5.23 EA3 P1 P-8 a. NiCrFe Filler Metal PP Surge Nozzle (SA-508 Class la) to Safe End (SA-182 Grade F347)

P-1 P-43 a. NiCrFe Filler Metal Buttermg (NiCrfe filler Metal) )

of J-Grooves in Hot Leg Pipe l (SA 508 Class la)

P-3 P-3 ru SFA 5.5 E-8018-C3 E-8018-G. Reactor Vessel Upper Shell j E-8016-G (SA-508 Class 3) to

b. MIL-E-18193 B-4*' RV Flange (SA-508 Class 3) i

c. SFA 5.23 EA3*'

P-3 P-8 a. NiCrFe filler Metal Pressurtzer Safety Valve Nozzle (SA-541 Class 3) to Safe End (SA-182 Grade F316)

P3 P-43 a. NiCrFe filler Metal Buttermg (NiCrFe Filler Metal) l of J-Grooves in RV Closure Head i (SA 508 Class 3)

P-8 P-8 a. SFA-5.4 E308. E308L. E309. E316, Surge Lme Piping (SA 312 Grace E347 TP347) to Surge Line Ells

b. SFA 5.9 ER308. ER308L. ER309. (SA-403 Class WP Grade 347) l ER 316. ER347 l P-8 P-43 a. NiCrfe Filler Metal Pressurtzer instrument Nozzaes (SB-166) to Safe Ends (SA-182 Grade F316)

P-43 P-43 a. NiCrfe filler Metal Reactor Vessel CEDM Nozzles (SB-166) to J-Groove Buttermg (NiCrFe Filler Metal)

Stamiess Steel Cladding a. SFA-5.4 E308, E308L, E309 -

f -

b. SFA 5.9 ER308. ER308L. ER309*

Nickel Alloy Cladding a. NiCrFe filler Metal -

s h

l

I System 80 + Desian control occument from these tests will be available after the required testing has been performed and may be examined upon request at the appropriate manufacturing facility.

I The reactor coolant pressure boundary also complies with 10 CFR Part 50 Appendix G. " Fracture Toughness Requirements" as enacted May 1983.

Consideration is given to the effects of irradiation on material toughness properties in the core beltline i region of the reactor vessel to assure adequate fracture toughness for the service lifetime of the vessel.

Refer to Section 5.3.1.6 for a discussion concerning prediction of irradiation effects and the material surveillance program.

Testing and measuring equipment for fracture toughness tests for the reactor vessel, steam generators.

pressurizer, piping and reactor coolant pumps are calibrated in accordance with Subarticle NB2360 of the ASME Code.Section III.

5.2.3.3.2 Control of Weiding 5.2.3.3.2.1 Avoidance of Cold Cracking C-E complies with the recome.endations of Regulatory Guide 1.50, Control of Preheat Temperature foi Welding of Low Alloy Steel May 19'l3, as discussed below.

Paragraph C.I.b implies that the qualification materials are an infinite heat sink that would instantaneously dissipate the heat input from the welding process. The qualification procedure consists of starting the welding at the minimum preheat temperature. Welding is continued until the maximum interpass temperature is reached. At this time, the test material is permitted to cool to the mimmum preheat temperature and the welding is restarted. Preheat temperatures utilized for low alloy steel are in accordance with Section III of the ASME Code. The maximum interpass temperature utilized is 500*F.

The paragraph C.2 recommendation is considered an unnecessary extension of procedures which apply to low-alloy steel welds, meeting ASME Code Seenons III and IX requirements. ((The recommendations of Regulatory Guide 1.50 are met by complying with paragraph C.4. The soundness of all welds is verified by ASME Code acceptable exammation procedures.))'

hb [

With regard to Regulatory Guide 1.43, major components are fabricated with corrosion resistant cladding on internal surfaces e,xposed to reactor coolant The major portion of the pgyrial protected by cladding f propne m reactor coolant is SA-533N Class I or SA-508 x Classt2 or'LCladding ot 5A-50h

' Class 2 forging material is performea usmg iow-neat-mput wetamg processes controlled to mmtm'"

42cating of the base metal. Low-heat-inout welding processes are not known to induce underciad cracking l 5.2.3.3.2.2 Regulatory Guide 1.34 '

f Regulatory Guide 1.34 recommends controls to be applied during welding using the electroslag process.

The electroslag process is not used in the fabrication of any RCPB components. Therefore, the recommendations of this guide are not applicable.

I -

l t COL infonnation item: see DCD Introduction Section 3.2.

Pope 6.2-14 Approved Deenyn hienenal- RCS and Coomected Systems

l ,

d(m L Item: Ernansion of Steam Generator Tubes l

! DCD Section: 5.4.2.4.1 Descrintion of Chance:

The description of the explosive expansion method is reworded for clarity and the hydraulic expansion method is added as an alternative.

l Related Changes to DCD DCD Section Insert # Revision 55.4.2.4.1 Revise as marked l

55.4.2.4.1 Insert 5.4 5 A hydraulic expansion technique may be used as an alternative to explosive expansion. For hydraulic expansion. the expansion mandrel length is set to provide full depth expansion, hydraulic pressure is accurately applied inside the tube, measured, reapplied and controlled such that the crevice between tube and ,

tubesheet is closed throughout the thickness of the l

tubesheet. J I

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l l C -

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System 80 + Design controlDocument

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At the tubesheet elevation, two 8" handholes are included to provide access for tubesheet sludge lancing as well as for inspection of the downcomer annulus. These handholes can be used to remotely inspect for, and retrieve, loose parts. Access for eddy current testing is through the primary-side manways.

l The steam generator blowdown system (Section 10.4.8) is designed for intermittent blowdown of up to

, approximately 10% of steam flow. The system is effective in removing particulate accumulations at the l tubesheet.

The pressure drop from the steam generator feedwater nozzles to the steam outlet nozzle including the economizer is approximately 40 psi.

l The steam generator supports are described in Section 5.4.14.

1 5.4.2.3 Economizer Integrity ne economizer section is designed in full consideration of operating transients. startup and standby operation, and accident conditions such as loss of feedwater flow and feedwater line creak. The structural l

design of the various parts is adequate to withstand the thermal and pressure loadings from these various conditions, consistent with the appropriate load classifications and design mies in the ASME Code.

Section III, Appendix G.

He components of the steam generator economizer section have been designed for the primary stresses which occur due to the blowdown associated with a feedline break. The divider plates, which separate l the economizer region from the evaporator region of the seconday side, are supported from the vessel i shell and the central cylindrical support welded to the tubesheet. This divider cylinder becomes an extension of the primary tubesheet stay cylinder, though less massive, and extends the full height of the i economizer. The tube support / flow baffle plates are supported from the vessel shell, the divider cylinder and the tubesheet via an array of support rods. The support rods, which also serve as support plate spacers are solid and designed for either tensile or buckling loads. Thin plates which may collapse when subjected to differential pressure are not used.

5.4.2.4 Steam Generator Materials l

l The pressure boundary materials used in the construction of the steam generator are listed in Table 5.2-2.

Rese materials are in accordance with the ASME Boiler and Pressure Vessel Code,Section III. Code cases used in the fabrication of the steam generator are discussed in Section 5.2.1.

The Class I components of the steam generator will meet the fracture toughness requirements of the ASME Code. An additional discussion of fracture toughness testing is included in Section 5.2.3.

Discussion of the techniques used to maintain cleanliness during final assembly and shipment are discussed in Section 5.2.3.

5.4.2.4.1 Steam Generator Tubes De method of fastening tubes to the tubgjheet conforms with the requirements of Sections III and IX of the ASME Code. Tube expansion into the tubegheet is total with no voids or crevices occurring along

the length of the tube in the tubejheet. After the tubing is insta!!ed in the steam generator and properly positioned, a seal weld is performed between the tube and cladding on the primary side of the tubesheet.

-"-d 5.% ... , em- v ... ._ ... ..,.,m.....  % - -Fr: u d.za-AC 'k ( ).$' D &XfM E 70 St 'ba 4cE}J dC T' by U0 GF f UVC.So Ayreved Qanopn Metenal RCS enaf Connected Systems Pope Ek9 l

l

System 80 + Desian contml Document i

/

o~ [' ,n N.w.") Tolerances on charge length and strengtn are specified so that the crevice between tube

- and tubesheet is closed through the full depth of the tubesheet and so that no crevice occurs on the

1. secondary face of the tubesheet.

Operating experience with this joint has been trouble-free from the standpoint of secondary side corrosion

' at the tube-to-tubesheet interface and from the standpoint of primary side stress corrosion cracking in the

( explanded portjnn nf the tube.

J E of h bd3 ,W_ as U up the,exp.=ded rhc ~y nsiony=t" mbm creates residual stress g in the transition zone between and ec.

. ed tbs 5-d regions of whme[ Residual stress measurements have been performed on this transition zone by X'-ray diffraction. The results verify the absence of any high residual tensile stress in the transition and fabrication I zone.

. m, Material specincations3such as the use of thertnally treated tubing, weldin procedures, preclude the need for complete-bundle stress relief after assembly.

' 4gf ) Operating C E steam generators have expertenced the following corrosion degradation mechan d / phosphate wastage, sulfate wastage. intergranular attack. secondary side stress corrosion cracking, and h pitting and denting resulting tram tube support corrosion. With respect to these phenomena. the most ,

/ important design feature of the System 30- Standard Design steam generators is the selection of tubing l and tubing support materials. For the System 80, steam generators. Ni-Cr-Fe Alloy 690 in a thermally treated (TT) condition is specined for the tubes. For tne tube supports, Stainless Steel 409 material is specified (see Table 5.2-2).

Volatile chemistry has been successfully used to minimize corrosion in all C-E steam generators that have gone into operation since 1972. Secondary water chemistry and operating chemistry limits for secondary water and feedwater are discussed in Section 10.3.5. Removal of solids from the secondary side of the steam generator is discussed in Section 10.4.8.

Here are three types of structures within the System 80t Standard Design steam generators which l i

support the tubes. Rese are the horizontal grid or "eggerate", the vertical supports, and the diagonal I

supports, all of which are fabricated from Stainless Steel 409. One of the design considerations for these supports is prevention of dryout at support locations. With one exception. all tube supports in the System 80+ steam generator are constructed of flat strips which present a fat surface to the tube. The one j exception is the flow distribution plate just above the entrance to the economizer section of the tube bundle. At this location secondary water is subcooled and, therefore, dryout will not be experienced.

He eggerate may have three different configurations depending upon its location in the evaporator-a full circular structure, a, half circular structure, or a structure bounded by the circumference and a chord.

An eggerate is composed of strips intersecting at an angle of 60 degrees and joined together at the outer and inner perimeters with a pair of square bars on top and bottom. The strips alternate between a 2-inch '

slotted type and a 1-inch unslotted type; both are 0.090 inches thick (see Figure 5.4.2-4). The eggerates themselves are supported and spaced by tie rods located throughout the tube bundle and by weldment to the tube bundle shroud. l The eggerates form an open lattice and thus minimize the potential for local dryout conditions. He number of eggerates is selected to maintain the natural frequency of the tubes significantly higher than the exciting frequencies induced by cross flow at the fluid entrances to the bundle. Both analysis and test results have been applied to define spacing which precludes vibration induced damage (fretting and wear). l' In addition, careful attention is paid to localized flow path details where velocities may be higher than nominal. The vertical supports (see Figure 5.4.2-5) are assembled concurrently with tube installation and are comprised of vertical, slotted 2-inch strips intersecting with horizontal 0.5 inch strips; both 0.090 I

I Page 6.4-10 Appmved Deno9n Motenal- RCS and Connected Systems

I h3 System 80+

vesian control Document 7

Table 7.5-2 Engineered Safety Feature System Monitoring (Cont'd.)

Number Number of of IE Mimmurn Parameter Channels Channels Indicated Range:1 l Loc,ggo,p1 Safety Depressurization System Rapid Depressunzation 1 --

Open/Closs Control Room Valve Position pair / valve 0-100 %

and Control Room 1 per valve RDS Line Temperature 2 2 50-700*F Control Room j SDS Pressure 2 2 0-2500 esi.t Control Room 1 Containment Spray System Qff CS Pump Suct:en Pressure

~'

CS Pump Discharge 2 2 0-1000 psi;; Control Room I CS Pump Flow 2 2 0-7500 gpct Control Room Spray Hender Isolation 1 -

Open/Close Control Room Valve Position pair / valve and Control Room

-n' I 1 per 0-100 %

valve CS Pump Status 1 pair / -

On/Off Control Room l pump ()7k l CS P:nnp Motor Current 2 2 0,MVAmps Control Room CS Heat Exchanger 2 2 40-400'F Control Room Outlet Tempera:ure Control Building / Control Room Ventilation Systen Inlet Radiation Monitor 2/ inlet 2/ inlet 10-3 10*R/hr Control Room Inlet Chemical Monitot 2/ inlet 2/ inlet Site Dependent Control Room Notes:

[1] See Chapter 18 for type of readout.

[2] Display channel inaccuracies negligible due to digital processing and display.

{ [3] All indication on electrically actuated valves in the Safety Injection, Shutdown Cooling, and Contemnant

Spray Symems with eaception of SI-661, receive IE power.

l (4] . Valves which are required to bring the plant to cold shutdown also have open/ clos.: position indicated outside i

r the Control Room.

! A Agnevest Dempo Asseewer awmswenreelen saar conoot esse 7.s.2s

1. .

'N e m 'i System 80 + Desian control Doc.ument As pictured in Figure 7.5-2 the HJTC sensor consists of a Chromel Alumel thermocouple near a heater (or heate'd junction) and another Chromel-Alumel thermocouple positioned away from the heater (or unheated junction). In a fluid with relatively good heat transfer properties, the temperature difference between the adjacent thermocouples is small. In a 11uid with relatively poor heat transfer properties, the temperature difference between the thermocouples is large.

1 Two probe assemblies are provided to allow two channels of HJTC instruments. Each HJTC probe assembly includes eight (8) HJTC sensors, a separator tube, a seal plug, and electrical connectors (Figure 7.5-3). The eight (8) HJTC sensors are electrically independent.

Two design features ensure proper operation under all thermal-hydraulic conditions. First, each HJTC is shielded to avoid overcooling due to direct water contact during two phase fluid conditions. The HJTC with the splash shield is referred to as the HJTC sensor (see Figure 7.5-2). Second, a string of HJTC sensors is enclosed in a tube that separates the liquid and gas phases that surround it.

The separator tube (see Figure 7.5-4) creates a collapsed liquid level that the HJTC sensors measure.

This collapsed liquid level is directly related to the average liquid fraction of the fluid in the reactor head volume above the fuel alignment plate. This mode of direct m-vessel sensing reduces spurious effects I due to pressure,11oid properties, and heterogeneities of the riuid medium. The string of HJTC sensors l and the separator tube are referred to as the probe assembly.

The probe assembly is housed in a stainless steel structure that protects it from flow loads.

7.5.1.1.7.1.3 Core Exit Thermocouples (CET)N* ',t"so c6w wds 9 MA .C M-one.40 awa p Thes Core Exit Thermocouples,g prov ide a measure of core heatup via measurement of core exit fluid temperature.

The design of the neutron flux In-Core instrumentation (ICI) system includes Type K (Chromel-Alumel) thermocouples within each of the ICI detector assemblies. These Core Exit Thermocouples (CET) monitor the temperature of the reactor coolant as it exits the fuel assemblies. The core locations of the ICI detector assemblies are shown in Figure 7.5-5.

'Be CETs have a usable temperature range from 32*F to 2300'F.

7.5.1.1.7.2 Description of ICC Sensor Signal Processing The following secti,ons provide a description of the processing control and display functions associated with each of the ICC detection instruments. The sensor inputs for the major ICC parameters (reactor vessel inventory / temperature above the core, and core exit temperature) are signal conditioned by the l two-channel PAMI processors and transmitted to the DIAS and DPS for primary display and trending, i 7.5.1.1.7.2.1 Heated Junction Thermocouple The signal conditioning equipment performs the following functions for the HJTC:

• Determine collapsed liquid level above core.

The heated and unheated thermocouples in the HJTC are connected in such a way that absolute and differential temperature signals are available. This is shown in Figure 7.5-6. When liquid Approved Design Matenal- Instrumentation and Control  !*9* I S*l I

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Thm 4 System 80 + Des /an contrat Document Table 7.5-3 Post-Accident Monitoring Instrumeniation l Number of Reg. Guide Sensed Minimum Minimum 1.97 Parameter Channelsl0 Sensor Ranges!' 31 Indicated Range Location tt.21 Category RCS Pressure 2 0-4000 psig l 0-4000 psig l Control Room 1.2 l

Primary Safety Valve Position 1/ Valve N/A Closed /Not Control Room 2 )

(Acoustic Leak Detector) Closed j In-contamment RWST Level 2 0-100 % l 0-100 % l Control Room: 2 In-containment RWST 2 50-250'F 50-250*F Control Room 2 l

l Temperature Coolant Temperature (Hot) 4 50-750'F 50-750* F l Control Room 1 l

Coolant Temperature (Cold) l 4 l 50-750* F 50-750* F Control Room 1.3 Containment Pressure (Wide 2 -5 psig to 4 times -5 psig to 4 times Control Room 1 l Range) design psig design psig Con:r_mment Pressure 4 5 psig to 1 ttmes 5 psig to I times Control Room I i I

(Narrow Range) design psig design psig l Steam Generator Pressure 2/SG 15-1500 psia l 0-1485 psig ' Control Room 1.2 l Steam Generator Level (Wide 2/SG 0-100 % 0-100 % Control Room 1 j Range)

! Pressunzer Level 2 0-100 % 0-100 % l Control Room I i Pressunzer Heater Status I pairl N/A On/Off Control Room 2 l

heater bank Pressurizer Pressure (High 4 1500-2500 psiat" Note 4 Control Room 1 Range)

! Pressunzer Pressure (Mid 4 600-1650 psia to Note 4 Control Room 1 Range)

Pressunzer Pressure (Low 4 0-750 psiato Note 4 Control Room 1 Range)

Degree of Subcooling 2 Note (4) 200'F suocooling Control Room 1 to 35'F superheat Neutron Flux Power Level 2 2x104 to 200% 2x104 to 200% Control Room I i (Safety Channels)  ;

j Reactor Cavity Level 2 0-100 % 0-100 % Control Room 1.2 l Contamment Area F,adiation 2 1R/hr - 10*R/hr IR/hr - 10*R/hr Control Room 1.3 Containment Hydrogen 2 0-15% by volume 0-15 % Cc arol Room 1 Concentration Containment Spray Flow 2 0-7500 gpm 0-7500 gpm Co grol Room 2 Containment Atmosphere 2 0-400* F 0-400*F Contrc.l Room 2 Temperature i Containment Isolation Valve 1 pair / N/A Closed /Not Control Room I i Position jalve Closed i

Core Exit Temperature (#ZJ 32-2300*F 32-2300*F Control Room 1.3 Reactor Vessel Coolant Level Y 0-370 inches 0-100 % Control Room 1

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g t .I- = . =n - n which time reactor cavity flooding commences. Flooding ceases when water levels in the IRWST, HVT l and reactor tavity equalize. Fully flooded, the reactor cavity will fill to near the 17 foot elevation. The time to fill the reactor cavity is dependent on the number of operational flood valves. The CFS has been

designed to effectively flood the reactor cavity to the 5 foot level in about 30 minutes. The time to l completely fill the reactor cavity to the equilibrium elevat. ion was calculated to be about 72 min 6tes with two HVT rpillway valves and one reactor cavity spillway valve open and about E8 minutes with one HVT l spillway valve and one reactor cavity spillway valve operational (See Figures 19.11.3.3-2 and 19.11.3.3 3).

To ensure a rapid water delivery to the reactor cavity, while simultaneously protecting the valves from direct corium attack, the HVT spillways are located approximately 5 feet above the basemat (approximate elevation: 67.0 ft). The HVT spillways and the reactor cavity spillways are equipped with remote manual motor operated valves. The reactor cavity spillways are located low enough to ensure sufficient flooding of the reactor cavity when the IRWST water level is at its minimum. The valven are rated for submerged cperation, since valve operation is typically not expected prior to submergence.

Minimum equilibration levels were established to ensure that operation of the CFS does not compromise the ininimum ECCS pump NPSH requirements. The maximum equilibration level was established so as to avoid contact between 6

This latterthe cavitywas flood watertoand thecons In-Core Instrumentatio reactor vessel lower head. requirement established minimize inadvertent operation of the cavity flooding system.

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Afo22.lis Flooding of the reactor cavity is an EPRI URD evolutionary plant design requirement and serves several O

i purposes in the overall strategy to mitigate the consequences of a severe accident. These include:

e Minimize or eliminate corium-concrete attack.

• Minimize or eliminate the generation of combustible gases (hydrogen and carbon monoxide).

• Reduce fission products released due to carium-concrete interaction.

o Scrub fission products released from the vapped core debris, nese features are discussed in detail in Section 19.11.4.3.2.

19.11.3.3.3 Role of the CFS in Accident Management ne CFS is designed as a manually actuated system. The manual operation of the CFS provides a mechanism for the operator to most efficiently use plant resources and protect the general welfare of the public and atows flexibility in the incorporation of new severe accident info:marion into the accident management process. Based on the current state of knowledge it is envisioned that the CFS will be actuated once a potential core melt condition is imminent or has been diagnosed as being in progress.

Typical indications of core uncovery include (1) core exit thermocouple (CET) temperatures in excess of 1200*F, (2) reactor vessel level monitoring system (RVLMS) readings indicative of no liquid above i

the fuel alignment plate, and (3) significant changes in readings of self-pcwered neutron detectors

! (SPND).. . .

i It is understood that steam explosions may pose a non-negligible threat to the cavity and contammant

! integrity. Thus, there may be an incentive to delay actuation of the CFS ur.til vessel breach (VB) is imminent or when the reactor vessel lower head has failed. This issue is considered tdo premature to Annemt centen nennener - rmsaunne a ask Auene=== rare is.11 1o l

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