LD-96-028, Submits Addl Draft Changes for Staff Review & Approval. Changes Listed on Attached Table W/Details Shown on Encl Design Control Document Pages

Submits Addl Draft Changes for Staff Review & Approval. Changes Listed on Attached Table W/Details Shown on Encl Design Control Document Pages

ML20115G182
Person / Time
Site:	05200002
Issue date:	07/17/1996
From:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
FACA, FACE, LD-96-028, LD-96-28, NUDOCS 9607180288
Download: ML20115G182 (17)

Text

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7%95 July 17,1996 LD-96-028 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-96-024, June 27,1996

Dear Sirs:

The Reference letter forwarded revisions to the System 80+ Standard Plant Design for Staff review and approval.

1 This letter transmits additional drafl changes for StafT review and approval. These changes are listed on j 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. On recommendation of the NRC Project Management, this letter is also being discussed with the i Advisory Committee on Reactor Safeguards to minimize the likelihood ofimpacting the Design l Certification schedule. All changes will be formally incorporated into the System 80+ DCD when it is l reprinted at the conclusion of design certification rulemaking.

Please call me or Mr. Stan Ritterbusch at 203-285-5206 if you have any questions.

l Very truly yours, COMBUSTION ENGINEERING,INC.

C. B. Brinkman Director, Nuclear Licensing l

cc: J. N. Wilson (NRC)

F. A. Ross (DOE; w/o enclosure) 10000~.a M0)

ABB Combustion Engineering Nuclear Systems Cornbusbon Engineenng Inc. 2000 Day HW Road Telephone (%0) 668-1911 t PO Box 500 Fas (860) 285-5203 i

9607180288 960717 PDR ADOCK 05200002 A PDR J

System 80+ DCD Changes - 7/17/96 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.

4 Sect 7.5.1.1 + Editorial clarification; System 80+ has 61 CETs ar.anged to display in 2 channels in the 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."

l

h ws i-Item: Update Reactor Coolant System Material Specifications DCD Section: 5.2.3.3.2.1. Table 5.1-2 Description of Change:

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

l (1) clarify material specifications (2) add material options for certain components (3) delete material options where no longer used in certain components (4) add omitted component parts / material specifications (5) remove inconsistencies (e.g. between Table 5.2-2 and Table 3.9A-2) l (6) rework the Weld Materials section to make it better organized and more complete (7) make editorial corrections Related Changes to DCD DCD Section Insert # Revision i

15.2.3.3.2.1 Revise as marked i l

Table 5.2-2 Revise as marked ,

! Table 5.2-2 5.2-2 replace current Weld Materials section with Insert 5.2-2 (below) l I

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System 80+ Desian control Document Table 5.2-2 Reactor Coolant System Materials I Component Material Specification 1

l' Reactor Vessel: y Forgings SA-508 Class 3 Cladding M Weld deposited sustenitic stainless steel with 5FN-18FN delta ferrite or NiCrFe alloyhuivalent to SB-166.7 G r A J e__

DVI Nor21e Safe Ends SA 1821F316 Reactor vessel head (*x" CEDM Nozzles NiCrFe Alloy 690 (SB-166)  ;

l Vessel internalsW Austenitic Stainless Steel and NiCrFe alloy Flow skirt ('Xd NiCrFe Alloy 690 (SB-168)

Fuel cladding

• Zircaloy-4 l l

• Instrument norzles( xo NiCrFe Alloy 690 (SB-166) s  % l Control element drive mechanism housings:

• d-[3h348 l

lower" Type 403 stainless steel according to Code .  ;

Case N-4-11 with end fittings to be SB-166 Alloy 690W and SA-182 'Egge4Wstainless steel Uppedd SA-479 and SA-213 Type 316 stainless steel with p--

and fittinggf SA 479 Typ Qf Type 316 and event seal ofvalv,ey6cFent Type 440C. valve sen 4

stainless steels' Closure head haier s Y tA d S SA-540 B24 or B23 Pressurizer:

Shell SA-533 Type A or B Class I or SA-508 Class 3 l CladdagW Weld deposited austenitic stainless steel with 5fN-18FN delta ferrite or NicrFe alloy Qequivalent to SB-166P' Forged nozzles SA-541xClass 3 or SA-508 Class 3 Instrumant nozzles and hester sleeves *X4 NiCrFe Alloy 690 (SB-1667 c r- S B - 3 07)

W Gra de.

Nozzle safe ends SA-18Q316 or F347 Safety valve nozzle flange * '

SA-182 Gr o Ae. F 316 Studs and nuts ee4+e9- SE -6 3 7  !

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l ww anos ,,a- ncs e,w c -se rene s.w i

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h System 80+ Desian controlDocument l

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! Table 5.2-2 Reactor Coolant System Materials (Continued) i Component Material Specification j

Steam Generator:

Primary Head SA-533 Type B, Class I or SA-508 Class 3 Primary Nonles SA-508 Classt Primary Side Instrument Nonles(*Xd NiCrFe Alloy 690 (SB-166)

Primary head cladding (*) Weld deposited austenitic stainless steel with l 5FN-18FN delta ferrite er NiCr Fe_ a llof 1 d Tubesheet SA-508 Classt2 or 3 4 Lv Tubesheet stay SA-508 Class <2 or13 Tubesheet cladding" Weld deposited NicrFe A // )'

l Tube (*** NiCrFe Alloy 690 (SB-163)

Tube supports ASTM A-176, Type 409 Secondary shell(d) SA-533 Type A or BxClass gI or SA $083Class 3 V Secondary head'd) SA-516 Grade 70g SA-508 3 Class , 1# or S A-63 s rge.

y A or B SA-508 OUI l  ! Secondary nonles(d) 3 Class tr2w-S ia er .3 Secondary nonle safe ends (d) SA-508 Class l A' 9 l j Secondary instrument nonles(d) SA 106 Grade B l i$ M. d) l

,' Secondary studs and nuts SA-540 Grade B24, or SA-193 Grade B7

_-- # /

/ U{} 3go

(-

i

. Primary studs and nuts 8A36~4, Type 630H1100 or':SB-637f NO7718 l l

Reactor Coolant Pumps: l w  !

CasingW SA-508 Classgn3 and cladded with austenitic stainless steel Claddmg(d Weld deposited austenitic stainless steel with l 5FN-18FN delta ferrite g Internals (* SA-487 CA6NM, SA-336 Type 304 or austenitic ,

4m stainless steel l

'f $h- 18 2. 0YVJe F G NC '

u -

i _ _ _ _

s  !

7 j

! [ Divicler Plcclef SA-2efo Tyr e#S  !

C - y i

Anwered Deewn nonenner- Mcs amt conneened syseems Pere 6.2-32 l

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I System 80+ Deslan control Document Table 5.2-2 Reactor Coolant System Materials (Continued)

Component Material Specification Reactor Coolant Piping:

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

, . , . ,- - ., - - a o ., , with 5FN 18FN delta ferrite , ,

-__,___sc >

Nozzle forgings SA-508 Class I or 1s, SA-182 Gende. fi.)

S A- 541 Clus .1 or S A-ios Instrument nozzles (**) NiCrFe Alloy 690 (S0-166)

Nozzle safe ends (') SA 182 G r cJ e F 31 f, o r F347 Valves ( SA-351 CF8M or SA-182 Cerade F316 Surge Line(*) SA-312 TP347 or TP316 (Piping);

SA-403 WP347 or WP316 (Elbows) 4 DAW F347 or F316 (Safe EndsIT

~ ^ ^

DVF a n d. ="uY A 'ul" L*'i C 5 UI

-AmeGamissor-f:ine 1.n side Co.,-inin m e d- SA-312 TP316 A TP304ESM,TP316. TP3674 Weld Materials for Reactor Coolant Pressure Boundary Comeponents

\ Base Matenal Type Base Material Type Type of Weld Material [

1. SA- SA-533 a. SFA 5.5,(b) E-8018-C3d8018-G Type B Type B C1.1 b. MIL E 18193.)4' ,

c. SFA-5.23 En3 * *)

N 2. SA-508 SA-533 a. SF(5.5. E-8018-C3, E-8018-G )

!. ef C1.2 Type B Cl.1 ,1>ridIL-E-18193, B-4

@'.b / c. SFA 5.23. EA3 * *)

3. SA-508 SA-508

1. a. SFA 5.5, E-8018-C3, E-8018-G f '

C1.1 Cl<

4. SA-516 S)$ a. SFA 5.1, E-7018

, Gr.70 Gr.70 s

5. SA 182

/[ SA-516 h '

a. SFA 5.1, E-7018 '

F1 Gr.70 \

6. SA 105 S M 82 a.Nb Metal i 7. SA-J82' SA-182 a. NicrFe Filler M Fy F347 8 SA-105 SA-182 a. NiCrFe Filler Metal

- Fn6 -

1 w w nenannel- Mcs anat Conneneenf Evennene nine 5.3-33

'i __

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

System 80+ Desian Control Document y

Table 5.2-2 Reactor Coolant System Materials (Continued)

Weld Materials for Reactor Coolant Pressure Boundary Components k Base Material Type Base Material Type . Type of Weld Material [

9. Shl66 SA-182 m. Root Filler Metal NiCrFe/

1

\ F316 Famammg Filler MetalMicrFe

/

10. SB-167 SA-182 a. Root Filler MetakNiCrFe 4

F304 Ra==mmg FiHsr Metal NiCrFe

. 11. SA-516 SA-351 a. NiCrFe Filler Metal Gr.70 CF8M /

12. SA-182 SA 182 adhiCrFe Filler Metal F1 F316 /

\gL b 13. SB-166 .

SA-533 a. Root filler Metal NiCrFe p Type B C1.1 Ramammg Filler Metal NiCrFe N-167/ a. NicrFe Filler Metal l

14. SA-182 rhnd

15. SA-516 SA-5'08 a. SFA 5.5,*) E-8018-C3 p Gr.70 /C1.2 ' ,

16. Austenitic stainless a. SFA 5.9, ER-308

'[ .

[D steel claddag SFA 5.9, ER-309  ;

17.1; x>nel cladding , \ a. NiCrFe Filler Metal

18. SA-508 Cl. 3 /

/ SA-508 C1. 3 h

s SFA 5.5,*) E-8018-C3, E-8018-G

b. MIL-E-18193, B-4

c. SFX 5.23. EA3* d  !

19. SA-508 / SA-533 a. SFA 5.k,'E-8018-C3 E-8018-G

b. MlleE-18193, B-4 f

C1. 3 - Type B C1.1

/ c. SFA-5.23. EA39 "

20. SA-508' SA-508 a. SFA 5.5, E-8018-C3. E-8018-G Cl. f Cl. 2 \ '

21.Sh508 SA-516 a. SFA 5.5,*) E-8018-C3 \

/Cl.3 Gr.70 \ l

/ k2dB-166 167tl68-- 41M.167r15" 2. NiCrFe-Filler Metal )

Notes: a. Materials exposed to reactor coolant.

b. Special wg' ys'th low residual elements of copper, nickel and phosphorous as specifi he readfor vessel core beltline region. ,

c. Matenal to provided in the thermally treated condition.

d. Steam generator secondary side pressure boundary materials including weld materials contam no greater than 0.010% by weight sulfur (S). .

g, f. nu mth r kSi y N on$ N C l>U Y >

~

) '

71 k/c G W - 4 2. '2. .

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,a- , . _ _ , , , - .

O INSERT 5.2-2 f Base Base Type of Weld Material Example of Use Material Material Type) Type)

P-1 P-1 a. SFA 5.1 E-7018, E-7016 Primary Piping Straight Pipe (SA-508 Class la) to Ells (SA-516 Gr 70)

P-1 P-3 a. SFA 5.1 E-7018, E-7016 Primary Piping Straight to RV Primary

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

E-8016-G l c. MIL-E-18193 B-4 i d. SFA 5.23 EA3 P-1 P-8 a. NiCrFe Filler Metal PP Surge Nozzle (SA-508 Class Ia) l to Safe End (SA-182 Grade F347)

! P-1 P-43 a. NiCrFe Filler Metal Buttering (NiCrFe Filler Metal) of J-Grooves in Hot Leg Pipe (SA-508 Class la)

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

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

c. SFA 5.23 EA3*)

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

P-3 P-43 a. NiCrfe Filler Metal Buttering (NiCrFe Filler Metal) ofJ-Grooves in RV Closure Head (SA-508 Class 3)

P-8 P-8 a. SFA-5.4 E308, E308L, E309, E316, Surge Line Piping (SA 312 Grade l E347 TP347) to Surge Line Ells l b. SFA-5.9 ER308, ER308L, ER309, (SA-403 Class WP Grade 347) i ER 316, ER347 P-8 P-43 a. NiCrFe Filler Metal Pressurizer instrument Nozzles (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 Buttering (NiCrFe Filler Metal)

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

b. SFA-5.9 ER308, ER308L, ER309)

Nickel Alloy Cladding a. NiCrFe Filler Metal -

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i l System 80+ Design ControlDocument l

from these tests will be available after the required testing has been performed and may be examined upon request at the appropriate manufacturing facility.

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 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 Welding 5.2.3.3.2.1 Avoidance of Cold Cracking C-E complies with the recommendations of Regulatory Guide 1.50, Control of Preheat Temperature for )

Welding of Low Alloy Steel, May 1973, as discussed below.  !

Paragraph C.l.b implies that the qualification materials are an infinite heat sink that would instantaneously dissipate the heat input from the welding process. The qualification procedura .onsists 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 minimum 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. I The paragraph C.2 recommendation is considered an unnecessary extension of procedures which apply to low-alloy steel welds, meeting ASME Code Sections 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 examination rocedures.))! ,

T~ne. 6D Y With regard to Regulatory Guide 1.43, major c ponents are fabricated with corrosion resistant cladding l on internal surfaces exposed to reactor coolant The major portion of the gateri'al protected by cladding F from_erp-ra MLr_ eactor coolant is SA-533 Class 1 or SA-508 x Class (2 orLCladding of YA-508W t

(Class 2 forging material is performed usi ow-neat-input welding processes controlled to minimize 4 eating of the base metal. Low-heat-input welding processes am net known to induce underclad cracking l 5.2.3.3.2.2 Regulatory Guide 1.34 '

l 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.

1 COL information item; see DCD Introduction Section 3.2.

Approved Design Meterial- RCS and Connected Systems Page 62-14 I

1

Tk4.m L Item: Ernannion of Steam Generator Tubes DCD Section: 5.4.2.4.1 Descrintion of Change:

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

I Related Changes to DCD l

DCD Section Insert # Revision 15.4.2.4.1 Revise as marked 15.4.2.4.1 Insert 5.4 1 A hydraulic expansion technique may be used as an I

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 l tubesheet is closed throughout the thickness of the

! tubesheet.

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. GD l System 80+ Design ControlDocument 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. 1 The steam generator blowdown system (Section 10.4.8) is designed for intermittent blowdown of up to j approximately 10% of steam flow. The system is effective in removing particulate accumulations at the tubesheet.

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

The steam generator supports are described in Section 5.4.14.

5.4.2.3 Economizer Integrity The 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 break. The structural 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 rules in the ASME Code, Svetion III, Appendix G.

The 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 the economizer region from the evaporator region of the secondary side, are supported from the vessel 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 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 i subjected to differential pressure are not used.

5.4.2.4 Steam Generator Materials The pressure boundary materials used in the construction of the steam generator are listed in Table 5.2-2. l The '. materials are in accordance with the ASME Boiler and Pressure Vessel Code,Section III. Code l casu. used in the fabrication of the steam generator are discussed in Section 5.2.1. l l

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

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 l l

The method of fastening tubes to the tubgjheet conforms with the requirements of Sections III and IX j of the ASME Code. Tube expansion into the tubq)heet is total with no voids or crevices occurring along the length of the tube in the tubejheet. After the tubing is installed in the steam generator and properly positioned, a seal weld is performed between the tube and cladding on the primary side of the tubesheet, e t---- " "

- _ cm moA ..,6,. on o m1 ~ :- ' - - -

k ( / t' QM / WD u h h )I C C f b y N C- bf S VC,3 o Approved Design Meterial . RCS and Connected Systems Page 6.4 9

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System 80+ Design ControlDocument 1

l ' / ff"ay 2 "E Tolerances on charge length and strength 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

secondary face of the tubesheet.

y '

/ Operating experience with thisjoint 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 l texplanded portinn of the tube. _

Ema nsion_a of +u W es, e adeA g a creates residual stress m the transition zone between thep.eq: id and g

l jpc,M-aphmd=4 regions of mobing[ Residual stress measurements have been performed on this transition z

i by X-ray diffraction. The results verify the absence of any high residual tensile stress i,n the transition

! zone. Material specifications)such as the use of thermally treated tubing, weldinggand fab ication f 0 C' dro S procedures; preclude the need for complete-bundle stress relief after assembly, Operating C-E steam generators have experienced the following corrosion degradation mechanisms:

4 4 phosphate wastage, sulfate wastage, intergranular attack, secondary side stress corrosion cracking, and kb/

V pitting and denting resulting from tube support corrosion. With respect to these phenomena, the most important design feature of the System 80+ 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 (TI) condition is specified for the tubes. For the 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 I 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.

l There are three types of structures within the System 80+ Standard Design steam generators which I support the tubes. These are the horizontal grid or "eggerate", the vertical supports, and the diagonal supports, all of whi:h are fabricated from Stainless Steel 409. One of the design considerations for these l

supports is prevention of dryout at support locations. With one exception, all tube supports in the System i 80+ steam generator are constructed of flat strips which present a flat surface to the tube. The one 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.

The 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.

l 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 l

l 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.

The eggerates form an open lattice and thus minimize the potential for local dryout conditions. The 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 i results have been applied to define spacing which precludet vibration induced damage (fretting and wear).

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 I are comprised of vertical, slotted 2-inch strips intersecting with horizontal 0.5 inch strips; both 0.090 merm o irn wra,.i acs amt com, cses syms.ma noe wo

. System 80+ Deslan ConkelD~> cwt i

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

Number Nanber of of IE Mininuart i Parameter Channels Channels Indicated Range 21 Locationli l Safdy Depressurization Systeen l Rapid Depressurization 1 --

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

, and Control Room I per valve RDS Line Temperature

( 2 2 50-700*F Control Room SDS Pressure 2 2 0-2500_ psia Control Room i Containment Spray Systesu @f/

CS Pump Suction Pressure CS Pump Discharge 2 2 0-1000 psig Control Room CS Pump Flow 2 2 0-7500 gpe Control Room j Spray Header Isolation 1 -

Open/Closo Control Room l Valve Position pair / valve .

and '

i Control Room 1 per 0-100 %

i valve CS Pump Status 1 pair / -

On/Off Control Room Pump (j7 CS Pump Motor Current 2 2 0-MAmps Control Room

, CS Heat Exchanger 2 2 40-400*F Control Room Outlet Temperature Control Building / Control Roosn Ventilation System l Inlet Radiation Monitor 2/ inlet 2/ inlet 10 10*R/hr Control Room I Inlet Chemical Monitor 2/ inlet 2/inld 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] Allindication on electrically actuated valves in the Safay Injection, Shutdown Cooling, and Containment Spray Systems with exception of SI-661, receive IE power.

, [4] Valves which are required to bring the plant to cold shutdown also have open/closs position indicated outside the Control Room.

1 O

Approved w neeeerter - hwownenenaien and coneet Peen 1.5-26

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_Tbm i System 80+ Design control Document l

As pictured in Figure 7.5-2 the HJTC sensor consists of a Chromel-Alumel thermocouple near a heater (or heated 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 i

between the adjacent thermocouples is small. In a fluid with relatively poor heat transfer properties, the temperature difference between the thermocouples is large.

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 in-vessel sensing reduces spurious effects due to pressure, fluid properties, and heterogeneities of the fluid medium. The string of HJTC sensors 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) p4eb\.{.od-one.g6 a r*6d * ""

Thes Core Exit Thermocouples,{ provide a measure of core heatup via measurement o 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.

The 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 sections 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 two-channel PAMI processors and transmitted to the DIAS and DPS for primary display and trending.

7.5.1.1.7.2.1 Heated Junction Thermocouple The signal conditioning equipment performs the following functions for the HJTC:

e 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 approves o aign uarenas- instrumentation and contrat rose 7.s-1 t

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Table 7.5-3 Post-Accident Monitoring Instrumentation Number of Reg. Guide Sensed Minimum Minimum 1.97 ChannelsI51 Sensor Ranges'*31 I Indicated Range 1,ocation tt.21 Category Parameter RCS Pressure 2 0-4000 psig 0-4000 psig Control Room 1,2 Primary Safety Valve Position 1/ Valve N/A Closed /Not Control Room 2 (Acoustic Leak Detector) Closed In-containment RWST Level 2 0-100 % 0-100 % Control Room 2 In-containment RWST 2 50-250*F 50-250*F Control Room 2 Temperature i

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

, Range) design psig design psig Containment Pressure 4 -5 psig to I times -5 psig to I times Control Room 1 1 (Narrow Range) design psig design psig I Steam Generator Pressure 2/SG 15-1500 psia 0-1485 psig Control Room 1,2 Steam Generator Level (Wide 2/SG 0-100 % 0-100 % Control Room 1 i

Range)

Pressurizer Level 2 0-100 % 0-100 % Control Room 1 Pressurizer Heater Status 1 pair / N/A On/Off Control Room 2 heater bank 4

Pressurizer Pressure (High 4 1500-2500 psia") Note 4 Control Room 1 Range)

Pressurizer Pressure (Mid 4 600-1650 psia") Note 4 Control Room 1 Range)

Pressurizer Pressure (Low 4 0-750 psia'4 Note 4 Control Room 1 Range)

Degree of Subcooling 2 Note (4) 200*F subcooling Control Room 1 to 35'F superheat Neutron Flux Power Level 2 2x10-7 to 200% 2x10-7 to 200% Control Room I (Safety Channels)

Reactor Cavity Level 2 0-100 % 0-100 % Control Room 1,2 Containment Area Radiation 2 1R/hr - 10'R/hr "IR/hr - 10 R/hr Control Room 8

1,3 Containment Hydrogen 2 0-15% by volume 0-15 % Control Room 1 Concentration Containment Spray Flow 2 0-7500 gpm 0-7500 gpm Control Room 2 Containment Atmosphere 2 0-400*F 0-400*F Control Room 2 Temperature Containment isolation Valve 1 pair / N/A Closed /Not Control Room 1 Position valve Closed Core Exit Temperature pH E 32-2300*F 32-2300*F Control Room 1,3 Reactor Vessel Coolant Level 2 0-370 inches 0-100 % Control Room 1 above core support surface Anwoved Design Meterial- butmmentation and Control Page 7.5-27

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which time reactor cavity flooding commences. Flooding ceases when water levels in the IRWST, HVT and reactor cavity 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 v Jves. The CFS has been  !

designed to effectively flood the reactor cavity to the 5 foot level in about 30 minutes. 'Ihe time to  !

completely fill the reactor cavity to the equilibrium elevation was calculated to be about 72 minutes with two HVT spillway valves and one reactor cavity spillway valve open and about 88 minutes with one HVT spillway valve and one reactor cavity spillway valve operational (See Figures 19.11.3.3-2 and  !

l 19.I1.3.3-3).

1 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 v.rith 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 valves are rated for submerged  ;

operation, 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 minimum ECCS pump NPSH requirements. The maximum equilibrationlevel was established so as i to avoid contact between the cavity flood water and the In-Core Instrumentatie (ICl)@ ow the l reactor vessel lower head. This latter requirement was established to minimize coWquences of l inadvertent operation of the cavity flooding system.

Afozz./E5 Flooding of the reactor cavity is an EPRI URD evolutionary plant design requirement and serves several purposes in the overall strategy to mitigate the consequences of a severe accident. These include: p '

o Minimize or eliminate corium-concrete attack.

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

e Reduce fission products released due to corium-concrete interaction.

e Scrub fission products released from the trapped core debris.

These features are discussed in detail in Section 19.11.4.3.2.

19.11.3.3.3 Role of the CFS in Accident Management The 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 1he general welfare of the public and allows flexibility in the incorporation of new severe accident information 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 the fuel alignment plate, and (3) significant changes in readings of self-pcwered neutron detectors (SPND).

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

integrity. Thus, there may be an incentive to delay actuation of the CFS urtil vessel breach (VB) is imminent or when the reactor vessel lower head has failed. This issue is considered too premature to Anume conen senauw - nosasawanc muk Ameneene rmee 1s.1110

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