ML20235Z102

Forwards Responses to Questions Re Application for Sys 80+ Design for Filing W/Project 675 Files

ML20235Z102
Person / Time
Issue date:	09/15/1987
From:	Vissing G Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
PROJECT-675A NUDOCS 8710200596
Download: ML20235Z102 (26)

Text

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l September 15,?l987-

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Prcject No. 675-

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'HEMORANDUM'FOR:

The Record f).'.,.

FROM:

Guy S.!Vissing, Project Manager i-Standardization and Non-Power Reactor -

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Project Directorate'.

,', 1 ;,

Division of Reactor Projects III, IV,

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SUBdECT:

MATERIAL WHICH WAS DELIVERED.T0 THE PROJECT MANAGER L

i" CONCERNING THE CE SYSTEM 80+ DESIGN y

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e D) q The enclosed material was hand' delivered;to Guy S. Vissing, Project o

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% nager, on Septeniber 22, 1987, by C.;Brinkman Combustion Engineering. The material consists of responses to questions posed by the staff concerning the application for the System.80+ design. The material should be filed with

/roject No. 675 files.

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original signed by je.ct Manager 1

Guy S. Vissing, Pro l

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Standardization and Non-Power Reacto*

Project Directorate Division of Reactor Projects III, IV, V and Special Projects

Enclosure:

As stated

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, DISTRIBUTION:,

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September 15, 1987 a

Project No. /J L,

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MEM0FAND H FOR:

'The Record FROM:

Guy S. Vissing, Project Manager Standardization and Non-Power Resctor Project Directorate Division of Reactor Projects III, IV, V and Special Projects

SUBJECT:

MNTERIALWHICHWASDELIVEREDTOTHEPROJECTMANAGER CONCERNING THE CE SYSTEM 80+ DESIGN s

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The enclosed matef f al was hard deliverd to Guy S. Vissing, Project Manager, on September 22. 1987, by C. Brinkman, Combustion Engineering. The

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material consists of responses to questions posed by the SP ff concerning the application for the System 80+ design. The material should be filed with Project No. 675 files.

.+-

Guy S. Vissing, f roject Manager Standard'ization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, V and Special Projects

Enclosure:

As stated 4 c

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Question 1:

Request C-E to show in greater ' detail;how the Advanced CESSAR design will differ from.the current ' design.

Idh.ntify; more specifically the major, changes being; considered..

Table., ' attached, compares System 80+TM to System 80R'.

~

Response 1:

The format of Table 1 is similar 'to.that presented.in Table 1.3-1 of CESSAR-F.

.A review..of Table 1.. demonstrates that the overwhelming number of-parameters are identical for..both System 80+ and System 80. This is consistent with l

Combustion Engineering's long-standing poucy of introducing design improve'ments through rational evolutions of proven technology.

Combustion 2ngineering believes that the best way to improve reliability and' safety of nuclear power plants is (1) to j

1 start with a proven design (i.e., one which the NRC has j

l approved with respect to current regulations and whfch has operating experience) and (2) to concentrate resources on J

those design-improvements which lead to the largest increase-in safety, while creating the smallest number of unanticipated results (and potential new problems).

i Further, we agree with the Commission's September 9,1987, l

1 Nuclear Power Plant Standardization Policy Statement which

/

declares that future reference system designs."are expected j

l-l to be evolutions of existing proven LWR designs" (emphasis -

l l

added). Modifying a proven design and concentrating on the i

most important improvements helps ensure the maximum' quality

}i 1

i

---______i_______z_.i__________._____________________________

1 in the end product..It also serves to minimize the occurrence' of unanticipated, direct or synergistic. effectsi '

which could actually decrease plant safety and reliabliity.

The design improvements, which are being implemented intoi the System 80+ design, should -- individually and collectively -- increase plant safety and improve performance.

Special emphasis is being placed on the design and analysis documentation, so that previous NRC conclusions on System 80 can remain valid whenever possible.

~

The more significant design evolutions being considered for System 80+ are summarized in Table 2.

Specific values of design parameters are being developed as part of the 1

l currently ongoing design process and the increased

.1 conservatism will be confirmed as part of this process. We W

believe that it would be reasonable for the NRC to perform a review of the design changes to first confirm that-compliance with NRC rules and regulations remains as stated in the existing CESSAR-F Safety Evaluation Reports and Final Design Approval.- Second, the NRC should confirm that the evolutions comply with the additional requirements of the Severe Accident and Standardization Policy Statements.

The scope of System 80+, as described in CESSAR-DC, will include features not previously part of the CESSAR-F. scope.

i 4

l The added features being considered (Emergency Feedwater, j

i Control Room, and Containment) are summarized in Table 3.

]

1 These features were identified as the most valuable j

i additions to the CESSAR scope. The scope expansions j

summarized in Table 3 will, of course, need to be reviewed; l

however, it should be recognized that each of the features l

i being considered has been reviewed and approved by the NRC I

i as parts of other submittals.

Thus, even for these systems which are now transferred to the System 80+ scope, we will I

continue to adhere to building upon a proven design to j

produce a safer, more reliable plant.

l Question 2:

Justify legally, practically and as a policy why we should I

consider a revised FDA verses a new FDA.

l Response 2:

It has been clear for some time that an important way to improve the safety and viability of nuclear power is through standardization. The essence of standardization is that a design is constructed and operated with no design changes after its review and approval by the NRC.

Standard designs l

will be safer in part because the ir.dustry and the NRC are able to concentrate more resources on fewer designs.

In addition, with accumulated construction and operating experience, operational feedback will occur faster and receive more effective NRC reviews, i

1 - _ - _ _ _ _ _ _ _ _ _ _ -

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The Commission h'as. expressed a' strong interest in 'and commitment to the concept' of s'.;ndardization. ~Among the'.

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reasons cited by' the;NRC'.in support ofistandardization areJ that it' can " facilitate high q.uality construction and safe :

operation of specific plant designs."* ' : Experience.at one

. plant is rele'vant' to ~ operating personnel,. equipment, maintenance and procedures.of other plants. :Information exchanged between utilities is more valuable and lessons; l

learned are more readily applied to other plants of the same standardized design.

"In sum,- standardized designs 'could improve plant safety and reliability."* Chairman' Zech has

.i reiterated in virtually every major speech his conviction that standardization facilitates licensing and improves

)

operational safety.

1 The Commission's commitment to standardization is reflected ~

In its decision, clearly incorporated in. theiSevere Accident Policy Statement, to permit 'the applicant for. Design o

Certification to " build".upon a previously-acquired FDA..

j

~

I One of the purposes of. the Severe Accident Policy Statement was to clarify the procedures an'd requirements related,to

.j severe accident considerations which are applicable'to licensing of new plants. The' Severe. Accident Policy Statement sets out the severe accident criteria and.

procedural requirements (Section B.2) and discusses their application to two distinct categories. The first category is " reference designs with no previous FDA'.' (Section B.2.a.)

l,

• Prepared testimony.

Hearing before Senate Committee'on Energy and Natural Resources, April 22,1986:'" Nuclear Facility Standards Act of '1986'_' pp -54, 55.

_ 1

a 9

and the second is " reference designs previously granted an FDA" (Section B.3.b. ).

In discussing the latter category; j

the Commission refers explicitly to two FDAs which had been j

issued by the Staff for two reference designs.

One of those is the FDA-2 granted to Combustion Engineering's System 80 design described in CESSAR-F.

The Severe Accident Policy Statement states the NRC's determination that the FDA could be used in any new CP and l

l OL applications if the severe accident criteria were met.

j 1

Upon completion of the severe accident review, the existing FDA could be appropriately amended. The Commission stated:

"A reference design applicant previously granted an FDA can pursue the same options of design approval or Design Certification as described... for reference designs with 1

no previous FDA" (Section B.3.b.).

1 l

l It would also appear that the limited NRC resources could best be utilized by reviewing proposed design evolutions (including their effects) rather than re-reviewing unchanged systems which have already been approved (absent, of course, a significant new safety issue).

Combustion Engineering's Design Certification Program is l

entirely consistent with the concept of standardization 1

described in the NRC's Statement of Policy on Nuclear.Powcr Plant Standardization and the Severe Accident Policy _ _ _ _ - _ - _ _ _ _ _ _.

4 1

5 Statement, The basis for-~our program is the System 80.

design (described in CESSAR-F) which has been reviewed and approved by the NRC in FDA 2.

(It should be noted that the System 80 design was reviewed.in significantly greater detail'than normal because. NRC reauzed the. System 80 FDA~

could be' referenced -- at that time.-- by an unlimited number of plants.

In that review,- the NRC found that the System 80 design fully met the appucable rules and.

regulations of the Commission and could'be built and operated with reasonable assurance to the health and safety of the public.)

1 As required by the Severe Accident Policy Statement, Combustion Engineering is making further improvements to the System 80 design over and above that needed to meet the

" reasonable assurance" standard-.

One of the goals of our program, however, is to implement these design improvements without changing the conclusions of the existing regulatory approval (FDA-2).

In other words, the design changes being i

considered are denberately selected to be conservative with i

respect to safety and engineered such that the CESSAR Safety Evaluation Report will remain vaud.

In summary, our position that the Design Certification review of CESSAR should be a revised FDA (as opposed to a new FDA) is firmly supported by. the Commissions' S'evere Accident Poucy Statement and the recent Standardization 4

I Policy Statement which states thatI"the reference system designs; at"least initially, are expected !to be evolutions of existing proven LWR designs','. (emphasis. added).. To be botit existing and proven.would reasonably require,.that the design has been licensed and, operated.

An existing FDA (and-through the FDA, an Operating License) would, therefore,-

.l appear to be the Commission's anticipated. basis for designs -

submitted for certification.

Question 3:

Once' all the changes identified in ' Question 1 have been.

~

characterized, why should 50.109 ' apply.to those changes and other systems / components which would be affected by the changes?

Response 3:

The Backfit Rule states in 10CFR50.109(a)(1) that

Backfitting is defined as-the modification of or addition to...the design approval... for a facility.... which may result from a new or amended provision.in.the Commission: rules or the imposition of a regulatory staff-position-interpreting the. Commission ' rules ' that. is either new or different from a previously applicable staff position after..... [t]he date of issuance of the design approval under! Appendix M, N or O of this part."

}

b.

Thus, the requirements of the Backfit Rule explicitly apply to NRC-imposed revisions in designs which have received FDAs. -

_ _ _ _-_-_ _ __j

4 t.

The Be.ckfit Rule provides the NRC reviewer with criteria and a process for determining whether to. require additional-improvements in safety. over and above those features of the design which have been found to be necessary to meet the Commission's rules and regulations and provide adequate protection to public health and safety. By issuing an FDA to C-F and Operating Licenses to Palo Verde, the NRC. staff-recorded its determination that the System 80 design meets the rules and regulations' of the Commission and provides adequate protection to the health and safety of the public.

There are other NRC requirements, however, which go beyond

~

adequate protection. The Severe Accident Policy, for example, requires that the applicant resolve the applicable Unresolved Safety Insues.

Anything offered by the applicant.

in that area is, therefore, necessarily beyond that required i

1 to meet the adequate protection standard.

The staff will j

benefit by_ the Backfit Rule being applied b'ecause, without f

the Backfit Rule there are no Commission approved criteria by which the staff can require further design improvements than the applicunt is.willing to make.

It is for just such a case as this that the Backfit Rule was provided.

By rule, the Commission requires that the Staff's proposed changes be considered in light of the factors listed ' n 50.109(c).

The FDA holder can be required to make 3

a change if analysis shows that its costs are justified by I

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. the increase.in overall protection of public health andL safety.

.1

.i

.I Question 4:

Taking.into consideration all relevant documents and material. relating toLthe development and' application of

50.109, from all legal ~ and practical perspectives, why shouldn't all issues be ' reviewed? Is the staff wUling= nota to revisit all issues? We have a duty to revisit everything in light of the extensive changes.

4

~

Response 4:

The System 80 design was exhaustively. reviewed by' the NRC with the full understanding (at the time)' that.it was ao standard design which would be referenced'in. applications for operating licenses for an unlimited number of plants'at.

an unlimited number of sites over a finite period of time.

The reviewers often' asked for, and'. received,' more information than normal,'specifically on th'e basis that this was a standard design,- not a one-of-a-kind plant. ;When that exhaustive review was complete, the' NuC found,.in December 3 1983, that the' System 80 design met the applicable rules and regulations 'of the Commission and could be built and operated without undue risk-to th'e health and safety of the S

public. The NRC continued to' reinforce that finding by-issuing Operating Licenses (based on FDA-2) to Palo Verde -

Unit 1 on December 31, 1984, Palo Verde Unit 2 on December 9,1985, and Palo Verde Unit 3 on s. arch 25,.1987.

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- _ - _ - _ - _ - _ = _ _ _ _ _ _ - _ _ _ _ _ _ _ -

4 9

'1 In addition ~, as discussed in.our' Responses.to Question 1 &

2, there are numerous reasons' to concentrate the review 'on-ll the l design improvements and not squander valuable resources I

retracing previous reviews.of design features already found to be in compliance with the rules 'and regulations 'of. the Commission.

1 lj

~

As stated in our Response to' Question 3, the Backfit Rule (50.109) provides the' reviewer with a process for deciding what additional improvements are. required of the System. 80.-

l design. This is consistent with the Commission's Policy

{

l Statements on Standardization, the Severe Accident' Policy.

j k

l Statement, the Backfit Rule and the recent decision of the

'l 1

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D. C. Circuit Court of Appeals.

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C___________._______________n___.________

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' TABLE =1-l (Sheet 1 of 12)

-COMPARISON OF DESIGN CHARACTERISTICS-

?

SYSTEM 80 SYSTEM 80+:

n Nuclear Design Dag Structural Characteristics

-Core Diameter, in. (Equivalent) 143.6 Same-Core Height', in. (Active Fuel)

.150 Same-

.i H/U, Unlimited Assembly (Hot) 4.23 Same Number of Fuel Assemblies 241-Same UO Fuel Rod Locations ~ Per Assembly 236 (*)

Same' 2

i Performance Characteristics Loading Technique 3 Batch, Mixed S ame'.

Central Zone

,I l

l Fuel Discharge Burnup Mwd /MTU Average First Cycle 13,740 Same First Core Average 24,144-Same Fuel Enrichments W/0 U-235 Region 1 1.83-Same Region 2 2.49 Same Region 3

-2.95-Same J

Region 4 Control Characteristics j

Critical Boron Concentrations, PPM (beginning of life, Rods out)

-Cold, Zero: Power, Clean 902 Same-Hot, Zero Puer, Clean 882

.Same Hot, Equilibrium Xe, Full Power 516 Same Hot, Full Power, Clean 764

.Same l

rods may-be replaced by burnable poison rods ~

L V Some U0 2

,/96((82H3)bh-2 TABLE 1 (Cont'd')

(Sheet 2 of-12)

COMPARISON OF DESIGN CHARACTERISTICS ~

q

.I SYSTEM 80 SYSTEM'80+.

l Nuclear Design Data' (Cont'd.)

j Control Element Assemblies Material (Full /Part-Length)

B C/B C-Inconel.

Same j

4 4

Number of Control Assemblies

j (Full /Part Length /Part Strength) 76/13/0 (*)

68/0/25 (*).

1 1

Number of Absorber Rods Per CEA 1

l (or RCC) Assembly 4 or 12 Same-TotalRodWorth(Hot)%A@

16.8 16.0 Kinetic Characteristics Range Over First Cycle Moderator Temperature Coefficient

~4

-0.7 x 10 /

.Same Ag/F(Hot,FullPower)BOL/EOL

-4

-2.5 x 10

-5 Moderator Pressure Coefficient (BOL)

+0.69 x 10 Same l

A{/ psi (Hot, Operating)

-3 Moderator Void Coefficient (BOL)

-0.24 x 10 Same Ap/% Void (Hot Operating)

U

-5

-1.18 x 10 Same DopplerCoefficient/g/F(Hot Operating Range) to

-5

-1.66 x 10 i

i

• Locations are provided for 8 additional CEAs.

j l

L

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' 964(.82H3)bh-3 '

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s

. TABLE 1.(Cont'd.)1 (Sheet 3 of 12).

'l COMPARISON OF DESIGN CHARACTERISTIC 1.

SYSTEM 80 SYSTEM 80+;

]

Nuclear Design Data (Cont'd.)

J l

I Hydraulic and Thermal Design Parameters Total Core Heat Output, Mwt 3800 Same H

7 Total Core Heat Output, Btu /hr.

1300 x 10 Same Heat Ger.erated in Fuel, %

97.5 Same System Pressure, Nominal, psia 2250 Same System Pressure, Minimum Steady State, psia 2200 Same Hot Channel Factors, Overall Heat Flux, F

2.35 Same-q

~

Enthalpy Rise, F_H, 1.56.

Same DNB Ratio at NominP Conditions 1.79 (CE-1)*

1.91'(CE-1)*

i i

Coolant Flow 6

Total Flow Rate, ib/hr.

164 x'10 33,,

Effective Flow Rate for Heat 6

-Transfer,1b/hr.

157.4 x 10 3,,,

Effective Flow Area for Heat 2

Transfer, ft 60.8 Same Average Velocity Along Fuel l

l Rods, ft/sec.

16.4 Same 2

6 Average Mass Velocity, 1b/hr-ft 2.59.x 10 33,,

9

• Minimum DNBR at nominal conditions.

J I

I

.' D 964182H3)bh-4 j

.j s,

3

' TABLE l'(Cont'd.)

L (Sheet:4 of 12)

+

. COMPARISON OF DESIGN CHARACTERISTICS 7

SYSTEM 80-SYSTEM 80'+-

lj Nuclear Design Data-(Cont'd.)_

i 0

Reactor Temperatures,- F 565 557 Nominal Inlet Average Rise in Vessel F

56 58 ij Average Temperature--in Vessel 593-586 Average Film Coefficient, Btu /hr-ft op 6300 Same 2

1 Heat Transfer at 100% Power 2

Active Heat Transfer Area, ft 68,600 Same 2

Average Heat Flux, Btu /hr-ft.

184,400 Same:

2 Maximum Heat Flux', Btu /hr-ft 433,000 Same Average Thermal' Output,.kw/ft 5.40 Same-Maximum Thermal Output, kw/ft

'12.7' SameL

o Maximum Clad Surface Temperature at Nominal Pressure, F

656'-

Same Fuel Centerline Temperature, OF Maximum at 100%_ Power 3,290 Same i

4

.-_m_

i 94408,2H3 ) bh-5 '

. c. -.

TABLE 1 (Cont'd.)

(Sheet 5 ofi12)

COMPARISON OF DESIGN CHARACTERISTICS SYSTEM 80 SYSTEM 86+*

Core Mechanical Design Parameters Fuel-Assemblies Rod Bundle Arrangement 16 x 16 Same Design CEA-Same

.)

~

Rod Pitch, in.

'O.506' Same l

Cross Section' Dimensions, in.

7.972 x 7.972' Same Fuel Weight (as UO ), 16.

257,245 Same Jj 2

Number of Grids Per Assembly 11 Same vuel Rods Number of Locations 56,876*

Same

('

Outside Diameter, in.

0.382 Same Diametral Gap, in.

0.007 Same Clad Thickness, in.

0.025 Same-l Clad Material Zircaloy-4 Same Fuel Pellets Material 00 Sintered Same 2

Diameter, in.

0.325 Same Length, in.

0.390 Same Control Assemblies i

Cladding Material NiCrFe Alloy Same Clad Thickness, in.

0.035 Same l

Core Structure Core Barrel 10/00, in.

157/162 25 Same 4

i

• Some of the rod locations are occupied by burnable poison rods.

i l'

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J'.964('8'2H3)bh 6'-

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- l TABLE 1 (Cont'd.);

j

,(Sheet 6 of. 12)

L

' COMPARISON OF-DESIGN CHARACTERISTICS 8-)

SYSTEM 80 SYSTEM 80+

' Reactor Coolant -System Code Requirements Component Reactor Vessel ASME III Same Class 1 l

Steam Generator Tube Side ASME'III, Same

~

. Class 1 Shell Side ASME III, Same Class 2 Pressurizer ASME III,'

Same Class 1 s

Pressurizer Safety Valves ASME III, Same 1

Class 1 Reactor Coolant Piping ASME III, Same Class 1 1

-]

Principal Design Parameters of the 3

Reactor Vessel

'1 Material Low alloy

' Same j

I steel clad with austen-itic SS.

(

Design Pressure, psig 2485 Same Design Temperature, F

650 Same Operating Pressure, psig' 2235 Same.

jrq 984l('82H3)bh-7 M

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1 1

TABLE lj(Cont'd.).

q

'(Sheet 7 of.12) i COMPARISON OF DESIGN CHARACTERISTICS ~

SYSTEM 80 SYSTEM 80Fs l

.I i

Principal Design Parameters of the Reactor Vessel (Cont'd.)

/

i

.Inside Diameter at.shell, in.

182-1/4 I

Same:

y/

Outside Diameter Across Nozzles, in.

267-1/4 Same.

Overall Height of Vessel

/

and Enclosure Head, ft-in. to top 50' 1-5/8" Same (including bottom

~

instrumentation nozzles)

Minimum Clad Thickness, in.

1/8-Same' Principal Design Parameters of the Reacter Coolant Pipino Material Carbon steel Same internally. clad with stainless steel Hot Leg - I.D., in.

42

'Same

' Cold Leg - 10., in.

30 Same l

Between Pump and Steam Generator -

I.D., in.

30 Same Design Pressure, psia 2500 Same s

L s

~

'9p4482H3)b' h 5 o

TABLE 1 (Cont'd.)'

(Sheet 8 of 12)

,OMpARISON OF DESIGN' CHARACTERISTICS

(

SYSTEM 80 SYSTEM' 80+'

~

Principal Design par ameters' of the Reactor Coolant System Operating Pressure, psia

'2250 Same-Number of Loops

'2 Same Design Pressure, psig 2485 Same Design Temperature,- F 650 Same Hydrostatic Test Pressure (cold), psig 3110.

Same.

3 Total Coolant Volume, ft 11',643 Same Total Reactor Flow, gal / min.

445,600 Same Principal Design Parameters of the Reactor Coolant Pumps-Number of Units 4

Same Type

Vertical, Same-single stage l

. centrifugal s'

with bottom suction'and horizontal discharge Design Pressure, psig 2485 Same Design Temperature,. F 650'

'Same Operating Pressure, nominal psig 2235.

Same-Suction. Temperature, F

564.5 556.5

-_.u

I 964tB2H3)bh-9 i

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l l

TABLE 1(Cont'd.)

]

(Sheet 9 of 12)

COMPARISON OF REACTOR-CHARACTERISTICS SYSTEM 80 SYSTEM 80+

f Principal Design Parameters of the Reactor Coolant Pumps (Cont'd.)

1

{

Design Capacity, gal / min.

111,400 Same Design Head, ft 365 Same Hydrostatic Test Pressure, (cold) psig 3110 Same Motor Type AC Induction Same f

Single Speed Motor Rating, hp. (cold) 12,000 Same Principal Design Parameters of the Steam Generators Number of Units 2

Same l

Type Vertical U-tube Same with integral j

moisture sep-arator and economizer Tube Material SB-163 NiCrFe inconel 690 alloy l

Primary side -

low alloy steel clad with aus-tenitic stain-less steel Secondary side - carbon steel l

l L-

• ' 964l82H3)bh-10I 4

TABLE l'(Cont'd.)

(Sheet:10'of 12)

COMPARISON OF REACTOR CHARACTERISTICS

)

' 1 i

SYSTEM 80' SYSTEM,80+-

i 1

Principal Design parameters of the Steam Generators (Cont'd.)-

J

'ube Side Design Pressure, psig 2485 Same T

Tube Side' Design Temperature, F'

650 Same Tube Side Design Flow, 1b/hr per steam 6

generator 82 x 10 Same Shell Side Design Pressure, psig 1255-Same "Shell Side Design Temperature, F

575 Same Operating Pressure, Tube Side, Nominal, psig 2235 Same-Operating Pressure, Shell Side, Maximum psig 1155 1085 Maximum Moisture at Outlet at Full Load, %

0.25 Same Hydrostatic Test Pressure, Tube Side (cold), psig.

3110 Same Steam Pressure, psig, at Full Power 1070 1000 Stera Temperature, F at Full Power 552.9 544.6 Steam Flow, at Full Power, ib/hr per 6

steam generator 8.59 x 10

.Samt i

a.x

.904(82H3)bh-11 i

i TABLE 1 (Cont'd.)

.(Sheet 11'of:12)

COMPARISON OF REACTOR CHARACTERISTICS SYSTEM 80 SYSTEM 80+

Engineered Safety Features Safety Injection' System No. of High Head Pumps 2

4 No. of Low Head Pumps 2

. 0 Safety Injection Tanks, No.

4 Same Emergency power Standby Generator Units

(*)'

Same l

l Instrun entation and Control Systems l

i' Reactor pectective System Initiating Reactor Trip Nurber of Manual Switches 2 Sets of JSame 2 each' Automatic Initiation 4 channels Same Parameter Channels /

'provided, Logic coincidence of 2 required for trip-

• See Site Specific SAR.

4 1

954T82H3)bh-12

}

TABLE 1 (Cont'd.)

(Sheet 12 of 12)

COMPARISON OF REACTOR CHARACTERISTICS SYSTEM 80 SYSTEM 80+-

1 Instrumentation and Control Systems (Cont'd.)

ESFAS j

Initiating ESFAS f

Number of Manual Switches 2 Sets of Same 2 each 4 Channels Same l

Automatic Initiation Parameter Channels / Logic provided coincidence of 2 received for

\\

1 each function i

1 l

1 l

i l

1 i

1 l

.n..i..,

o TABLE 2 l

SUMMARY

OF IMPROVEMENTS BEING CON 5IDERED FOR SYSTEM 80+

SYSTEM INVOLVED DESCRIPTION OF CHANGE-1.

Reactor Part length CEAs have been eliminated and a new CEA pattern using part strength rods employed to provide a rodded maneu'ver-

)

ing capability 2.

Reactor Coolant System The Reactor Coolant System will be operated at a lower hot leg temperature to increase thermal margin.

3.

Pr,essurizer The pressurizer volume has been increased to accommodate larger fluctuation in 3

primary plant pressure without challenging the safety systems.

l 4.

Steam Generator The steam generator has a number of design improvements including improved steam dryers to increase steam quality, a larger heat transfer area and reduced steam pressure to enhance thermal margin and greater feedwater inventory to improve NSSS transient response.

5.

Safety Injection System The Safety Injection System (SIS) design has been; modified to permit direct vessel injection.

Also, the SIS has four trains using 50%

l capacity pumps.

An in-containment RWFT has been included, eliminating the need to switch to a recirculation line up follow-l ing a LOCA.

6.

Shutdown Cooling System The Shutdown Cooling System design pressure nas been increased to 900 psig.

This in-l creased design pressure will reduce the probability of inadvertent system over-pressurization and enhance NSSS operational l

flexibility.

7.

Safety Depressurization The Safety Depressurization System has been j

System added to provide a safety grade, manual depressurization capability.

]

8.

Chemical and Volume The Chemical and Volume Control System is no Control System longer being credited for any safety func-l tions, thereby allowing a simpler design;

]

centrifugal charging pumps have been pro-j vided, along with two-stage letdown.

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TABLE 3 T

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SUMMARY

'0F ADDITIONS-T0 SCOPE BEING CONSIDERED FOR' SYSTEM 80+c l

1

-SYSTEM INVOLVED;.

DESCRIPTION 0F CHANGE-

.i 1.

Emergency Feedwater A dedicated four-trainLEmergency Feed-System '

water System has-been included as part of the Engineered Safety Features systems to enhance overall. safety.

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Instrumentation and-The Nuplex 80 Advanced Control Complex Control has been included to provide"an enhanced-operator control capability.-

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Containment An. enhanced containment design'~has been included to resolve;the remaining severe:

accident issues, inclu' ding. heat;and' fission

product removal, pressure capability,,and leakage: limitations.

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