LD-94-036, Forwards Comments of Chapter 4 of Sys 80+ Final SER, NUREG-1462 for Technical Accuracy & Consistency w/CESSAR-DC

Forwards Comments of Chapter 4 of Sys 80+ Final SER, NUREG-1462 for Technical Accuracy & Consistency w/CESSAR-DC

ML20070R996
Person / Time
Site:	05200002
Issue date:	05/11/1994
From:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-1462 LD-94-036, LD-94-36, NUDOCS 9405230166
Download: ML20070R996 (33)

Text

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ABB j May 11,1994

- LD-94-036 4

Docket 52-002 Attn

Document Control Desk U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 j

Subject:

System 80+ Final Safety Evaluation Report Comments

Dear Sirs:

4 The enclosure to this letter provides ABB Combustion Engineering comments on Chapter 4 of the System 80+ Final Safety Evaluation Report (FSER), NUREG-1462, for technical accuracy and consistency with CESSAR-DC. The comments are provided as hand written mark-ups of

pertinent FSER pages. Bars have been added to the right hand page margin to assist in 4

locating comments.

If you have any questions please do not hesitate to contact me, or Mr. Stanley Ritterbusch of j my staff at (203)285-5206.

l Very truly yours, 3

2 COMBUSTION ENGINEERING, INC, 4

e i i

hC.B. Brinkman, Director Nuclear Systems Licensing j

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Enclosures:

As stated cc: P. Lang (DOE)

T. Wambach (NRC) 2 1

ABB Combustion Engineering Nuclear Power oUk 1

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Comtmshon Engineenng Inc. P O Box 600 Teleptum (203) 688-1911 9405230166 940511 PDR E IcNNs i

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4 Enclosure I to LD-94-036 4

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I i System 80+ Standard Plant Design l

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System 80+ Standard Plant Design

! FSER Comments I'

The following summarizes the more significant comments provided on the System 80+ Final

Safety Evaluation Report (FSER).

j 1. The FSER uses the term " poison", whereas the NRC staff has requested that in i CESSAR-DC the term " poison" should be replaced by the term " absorber" (particularly 4

in the use of the term " burnable absorber") to maintain consistency within CESSAR-

! DC. ~ For consistency with CESSAR-DC, the FSER should also use the tenn I' ' " absorber". Occurrences of " poison" which should be changed to " absorber" in the 1 FSER are noted in the attached markup.

2. The FSER (Section 4.2.2) states that the density of the UO 2 in the fuel pellets is 10.47 g/ce, which corresponds to 95.5% of theoretical density. In fact, CESSAR-DC quotes the 95.5% of theoretical density as nominal, and further states that the nominal density ')'

f of UO 2can range between 94.5% and 96.5% of theoretical density. The FSER should i be revised as indicated in the attached markup to be consistent with the statements in CESSAR-DC. -

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3. The FSER (Section 4.2.3) describes erbium oxide (Er20 3) burnable absorber rods I having natural UO2 at the top and bottom of the core. This should be modified to I 1 indicate the option of either natural or low enrichment UO2 at the top and bottom of

the core to be consistent with CESSAR-DC, which will be modined to permit either

' natural or low enrichment UO2 axial blankets in the fuel and Er20 3burnable absorber rods in Amendment W. A similar change to indicate the option of either natural or low enrichment UO2 axial blankets in the fuel and Er20 burnable 3 absorber rods should also be made in Table 4.2-1 of the FSER. The corresponding Table 4.1-1 in CESSAR-DC will be similarly modified in Amendment W.

4. The FSER (Section 4.2.7) states the permitted design changes in the initial reference design of the fuel, the CEA design or the initial core design from that presented and evaluated in CESSAR-DC which would not require prior NRC review and approval as being those listed in Table 4.2-1 and 4.2-2. The FSER does not include the changes described in Section 4.1.1 of CESSAR-DC, which also permits changes in the arrangement of the different fuel assembly batches.within the core as long as the acceptance criteria for the evaluated design parameters in Table 4.2-2 are met. The FSER should be revised to reDect permissible changes consistent with Section 4.1.1 of CESSAR-DC.

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5. Section 4.3.2 of the FSER (p. 4-10) references CENPD-169-P for the description of the COLSS algorithms and the uncertainty analysis of the COLSS calculations. This reference should be replaced by CEN-312 Rev.1-P (" Overview of the Core Operating ,

Limit Supervisory System") and CEN-356(V) P-A Rev.1-P-A (" Modified Statistical l Combination of Uncertainties"). Also, Section 4.4.2 of the FSER (p. 4-19) discusses l the use of the SCU methodology for thermal-hydraulic analysis and cites CEN-139-P-A (Reference 4 on page 4-27). In this section, SCU should be replaced by MSCU.  :

Also, Reference 4 should be replaced by CEN-356(V)-P-A Rev.1-P-A, which is the appropriate reference for the MSCU methodology.

! 6. In Section 4.3.4 of the FSER (Summary of Evaluation Findings), Item (a) uncler l Section (5) (p. 4-16) satisfies GDC 26 with respect to provision of two independent reactivity control systems of different designs. Therefore, this item should be either moved or added to Section (4) on p. 4-16.

7. The ASME code case cited in the FSER for the control element drive mechanism motor housing should be N-4-11 rather than 4-N-11.

8. In Section 4.6 of the FSER, one of the reactivity control systems mentioned is the Control Element Assembly Drive System (CEADS) (p. 4-39). For consistency with CESSAR-DC, this system should be referred to as the Control Rod Drive System (CRDS). Also, the regulating CEA groups consist only of full-strength CEAs; therefore, reference to full- and part-strength regulating CEAs (p. 4-40) should be deleted.

9. The acceptance criteria for the design features and evaluated design parameters in Tables 4.2-1 and 4.2-2 of the FSER are nominal values. Therefore, a footnote should be added to the acceptance criteria in these tables as shown in the attached markup. A similar change will be made to the corresponding Tables 4.1-1 and 4.1-2 in CESSAR-DC in Amendment W.

10. The maximum heat flux and maximum linear heat generation rate given in Section IV of Table 4.4-1 of the FSER are not the LOCA limit values and therefore not the maximum allowable values. Rather, they are maximum values based on a specified prescription used for comparative purposes only. Therefore, the word " allowable" in the " Characteristics" column should be deleted and the word " peak" replaced by

" maximum" as indicated in the attached markup. Also, several parameter values in Table 4.4-1 should be changed for consistency with Amendment V. These changes are also shown in the attached markup.

In addition to the above detailed comments, several minor editorial and typographical comments are noted. All recommended changes are included in the attached markups.

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l 4 REACTOR l

! 4.1 Introduction ,

j Criterion 10 of the general design criteria (GDC) requires that the reactor j core and associated systems be designed to assure that'specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal

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l operation, including the effects of anticipated operational occurrences (A00).

The staff reviewed the information provided in the CESSAR-DC in support of the i System 80+ standard plant design. The staff evaluation is described below.

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. i The CESSAR-DC nuclear steam supply system is designed to operate at a maximus j core thermal output of 3,914 megawatts (MWt), with sufficient margin to allow i for transient operation and instrument error, without causing damage to the core and without exceeding the pressure settings of the safety valves in the .

coolant system.

i The reactor will be cooled and moderated by light water at a pressure of ,

1.55 x 10' kPa (2250 psia). The reactor coolant will contain soluble baron for neutron absorption. The concentration of the boron will be varied, as I

required, to control relatively slow reactivity changes, inc{uding the effects j of fuel burnup. Additional boron, in the form of burnable M rods, may also be emplov to establish the desired initial reactivity. Two other types d her i of burnable rod designs consist'of gadolinium oxide or erbium oxide 4

mixed with 002 . Both full-strength and part-strength control element assen- ,

blies (PSCEAs) may be used to compensate for changes in reactivity associated with changes in power level, power distribution, moderator temperature, and; i:

boron concentration. The PSCEAs are used primarily fo'r. power shape i:ontrol' during power operations. Both types of CEAs are osed for reactor shutdown.

T ABB-CE System 80+ FSER 4-1 February 1994-

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4.2.2 Fuel Rod Description The fuel rods consist of slightly enriched UO cylindrical ceramic pellets, a i round wire Type 302 stainless steel compression spring, and an alumina spacer j disc located at each end of the fuel column, all encapsulated within a l

Zircaloy-4 tube. The fuel rods are internally pressurized with helium during I

assembly. This internal pressurization, by reducing the stresses from differential pressure, extends the time required to produce creep collapse

! beyond the required service life of the fuel and also improves the thermal j conductivity of the pellet-to-cladding gap.

1 l The fuel cladding is cold-worked and stress-relief-annealed Zircaloy-4 tubing j . with a wall thickness no less than 0.0584 cm (0.023 in.)

l- The U0, pellets are dished at both ends in order to better accomodate thermal expansion and fuel swelling. Th de'ns ty of the UO in the pellets is 10.47 g/ce, which corresponds to 95.5spercent of the 10.96 g/cc theoretical density. However, because the pellet dishes and chamfers constitute.about

3-percent of the volume of the pellet stack, the average density of the' pellet stack is reduced to.10.315 g/cc3 (noN%l), t lhe 4W'W\ AeMih i  % ,5 ofhe OOw in Yhe tehet tq% csce4PQMe u be.ta)een in e. homina\perCeck CN.S vomotic%

op o the fuel pel et column maintainsteRetl shck i

o'r TheM. theoretta\ spring compression demk015tted at the keMmf

the column in its proper position during handling and shipping. The alumina l spacer disc at the lower end of the fuel rod reduces the lower end cap i temperature, while the upper spacer disc prevents U0 2 chips, if present, from entering the plenum region. The fuel rod plenum, which is located above the

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i pellet column, provides space for axial thermal differential expansion of the ,

t fuel column and accommodates the initial helium loading and evolved fission ,,

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gases.

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4.2.3 Burnable pe4 son Rod Description

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j Fixed burnable neutron absorber (pi:=) rods are included in selected fuel j assemblies to reduce the beginning-of-life moderator coefficient. They will l

i l ABB-CE System 80+ FSER 4-3 February 1994 1

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V replace or auament fuel rods at selected locations. Three alternative  !

ab soder i burnable peison rod designs, each mechanically similar to fuel rods, are i

! proposed for the System 80+ standard design.

i i chwber abwber  ;

The first burnable peisen rod design contains a column of burnable peisen anorber

pellets instead of fuel pellets. The petsen material is alumina (A10 23 ) with j uniformly dispersed boron carbide (B 4 C) particles. The balance of the column consist of Zircaloy-4 spacers; the total column length is the same as the column length in fuel rods.

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! The second burnable poison rod design consists of gadolinium oxide (Gd23 0) l admixed in natural UO in the central rod portion (axially) and natural d0, at I l g

the top and bottom. The total column length is also the same as the column

! length in fuel rods.

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, abwher /cc lcu eAnthred l The third burnable poisen rod design consists of erbium oxide (Er 0 3) admixed 3

in enriched UO g in the central rod portion (axially) and natural)02 at the j j top and bottom. The total column length is also the same as the column length l

} in fuel rods. l i

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' Each of these types of burnable poisen rods have been used previously in ABB-CE-designed reactors and have been approved by the NRC. Therefore, the l

j staff considers any of-these designs acceptable.

4.2.4 Control Element Assembly Description The control element assemblies (CEAs) consist of either 4 or 12 neutron j absorber elements arranged to engage the peripheral guide tubes of fuel l assemblies. These elements are connected by a spider structure which couples l to the control element drive mechanism (CEDM) drive-shaft extension. The j neutron absorber elements of a 4-element CEA engage ths 4 corner guide tubes in a single fuel assembly. The 12-element CEAs engage the 4 corner guide I tubes in one fuel assembly and the 2 nearest corner guide tubes in adjacent

[ fuel assemblies. A total of 93 CEAs exist, 25 of which are PSCEAs.

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i ABB-CE System 80+ FSER 4-4 February 1994 5

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(3) ABB-CE has provided for testing and inspection of new fuel to ensure 4

that it is within design tolerances at the time of core loading. ABB-CE i

! has agreed that this is the responsibility of a COL applicant and is, I therefore, a COL action item. Each COL applicant referencing the System 80+ design certification must address in their facility-specific

! safety analysis report the performance of on-line fuel failure monitor-ing'and post-irradiation surveillance to detect anomalies or confirm

' that the fuel has performed as expected. This is a COL Action l Item 4.2.7-1.

(4) ABB-CE has described methods of adequately predicting fuel rod failures-during postulated accidents so that radioactivity releases are not f- underestbated and thereby satisfy the related requirements of 10 CFR l Part NO. In meeting these requirements, ABB-CE has (a) used the fission-product release assumptions of RGs 1.25 and 1.77, and an accept-l able (more conservative) alternative to RG 1.4.

On the basis of its review, the staff concludes that the System 80+ fuel j system design, including the fuel design, the CEA design, and the initial core l, design, satisfies all the requirements of the applicable regulations, RGs, and l

current regulatory positions. The fuel design and the CEA design have been j specified and the associated analyses results have been presented in

!. CESSAR-DC. Startup tests to confirm specified nuclear and' thermal-hydraulic i

design parameters are described in CESSAR-DC of Chapter 14. l Any changes to the initial reference design of the fuel, the CEA design, or the initial core design from that presented and evaluated in CESSAR-DC would l

i involve an unreviewed safety question and requires prior NRC review and I

approval prior to implementation, with the exception of the following: e t i j (1) , changes to the design features and evaluated design parameters listed in l Tables 4.2-1 and 4.2-2 within the boundaries of the indicated acceptance criteria; and (2.)

ch4%eA lh he 4rrygenegt et the different fue\ CWeQl] batCAe5

. uth %e Core 90% %ch hun m Sechon 93 wth twt the l occeptance cnkna for ihe esaboted ciegn prometeN 'hsted j in TQ Me 'i.1 -1 QT thei) and ABB-CE System 80+ FSER 4-7 February 1994-5 i

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(2) changes covered by applicable NRC-approved topical reports.

Furthermore, prior NRC review and approval will be required to change the

NRC-approved analysis methods used to demonstrate conformance of the fuel design, the CEA design, and the initial core design to the design limits given in CESSAR-DC, with the exception of those changes to analysis methods which are covered by applicable NRC-approved topical reports.

Any requested change to the above information that requires NRC review and l approval prior to implementation shall either be specifically described.in the COL application or submitted for license amendment after COL issuance. ABB-CE has agreed to these commitments and, therefore, draft safety evaluation report )

(DSER) Confirmatory Item 4.2.7-1 is resolved. l l

4.3 Nuclear Desian l 4

The staff based its review of the nuclear design on information in the CESSAR-DC, ABB-CE's responses to staff requests for additional information, i and the referenced topical reports. The staff conducted its review in accordance with the guidelines provided by SRP Section 4.3.

4.3.1 Design Basis ABB-CE presented design bases that comply with the applicable GDC. Acceptable fuel design limits are specified (GDC 10), a negative prompt feedback coeffi- {

cient is specified (GDC 11), and tendency toward divergent operation (power oscillation) is not permitted (GDC 12). ABB-CE presented design bases that require a control and monitoring system (GDC 13) that automatically initiates a rapid reactivity insertion to prevent exceeding fuel design limits in normal' operation or anticipated transients (GDC 20). The control system is required to be designed so that a single malfunction or single 6perator error will.

cause no violation of fuel design limits (GDC 25). A reactor coolant boration system is provided that is capable of bringing the reactor to cold shutdown conditions (GDC 26) and 'the control system is required to control reactivity changes during accident conditions when combined with the engineered safety.

features (GDC 27).

ABB-CE System 80+ FSER 4-8 February 1994

Reactivity accident conditions are required to be limited so that no damage to the reactor coolant system (RCS) boundary occurs (GDC 28).

The design bases presented in the CESSAR-DC comply with the GDC and, there-fore, are. acceptable.

4.3.2 Dei;ign Description In the CESSAR-DC, ABB-CE aescribes the first-cycle fuel loading which consists of a three-batch loading scheme ~in which the type B and C fuel assemblies contain rods of erbia burnable pN. Fuel enrichment and burnable M l distributions tre shown. Assembly enrichments, core burnup, critical soluble boron concentrations and worths, and plutonium buildup are also presented.

Values presented for the delayed neutron fraction and prompt neutron lifetime at beginning and end of cycle are consistent with those normally used and are acceptable.

Power Distribution

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The design bases affecting power distribution are:

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• The peaking factor in the core will not be greater than 2.28 during normal operation at full power in order to meet the initial conditions assumed in the LOCA analysis.

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. Under normal conditions (including maximum overpower) the peak fuel ,

power will not produce fuel centerline melting.

• The core will not operate during normal operation or A00s, with a. power N distribution that will cause the departure from nucleate boiling ratio!

(DNBR) to fall below 1.24 using the CE-1 DNB correlation and the statis '

tical combination of uncertainties (SCU) methodology discussed in Sec-tion 4.4. ,

The[re /perating/lmit supervisoryjhstem f (COLSS) is employed to continu-ously monitor important reactor characteristics and establish margins to ABB-CE System 80+ FSER 4-9 February 1994

operating limits. This system, which consists of software executed on the plant computer, utilizes the output of the incore detector system to synthe-size the core average axial power distribution. Rod positions taken from the control rod position indication system, together with precalculated radial peaking factors, are used to construct axially dependent, radial power distributions. By using this information, together with measured primary-coolant flow, pressure, and temperature, the COLSS establishes the margin to l l

the operating limits on maximum linear heat generation rate and minimum DNBR.

The system also monitors azimuthal flux tilt and total power level and gene- l rates an alarm if any of these limits are exceeded. The margins to all of these limits except azimuthal tilt are continuously displayed to the opera-tors; the tilt can be displayed at the request of the operator. The operator i l

monitors these margins and takes corrective action if the limits are approached. These actions include improving the power distribution by moving full-strength or part-strength rods, reducing power, or changing thermal-hydraulic conditions, that is, coolant inlet temperature and primary system pressure.

A description of the COLSS algorithms and an uncertainty analysis of the calculations performed by the COLSS is presented in CFTopica4-Report CENPO4C P, "E0tSS=Asseasment ui the Accuracyjf'Pcessurized-water-Reactor (pWR}-Opent4n944mits-as-Determined by tim C6re Operating t:tmit SupervisorT Systems ^- The staff reviewed tMs-r4po d found the methods employed in COLSS to determine power distributions /are acceptable. The COLSS is currently used at AND (Unit 2), San OnofreUd ( j .s 2 and 3), Waterford (Unit 3), and Palo Verde (Units 1, 2, and 3). /

C.E Topgai Qpts CEN qil, Regtston 1-9, "O9erveu dr Mc Ccre Ope @ Lak Weervwy Reactivity Coefficients g a g g _ g y _p4 g pp "Modahed 5%hhui (cmbmahon cf 0%erbhes."

The reactivity coefficients are expressions of the effect on core reactivity of changes in such core conditions as power, fuel and' moderator temperature, moderator density, and boron concentration. These coefficients vary with fuel burnup and power level. ABB-CE presents calculated values of the coefficients in the CESSAR-DC and has also evaluated the accuracy of these calculations.

The staff reviewed the calculated values of reactivity coefficients and concludes that they adequately represent the full range of expected values.

ABB-CE System 80+ FSER 4-10 February 1994

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The staff reviewed the reactivity coefficients used in the transient and-accident analyses and concludes that they conservatively bound the expected values, including uncertainties. Further, moderator and power Doppler coefficients along with boron worth are measured as part of the startup physics testing to ensure that actual values are _within those used in these-analyses. Although applicant-predicted moderator temperature coefficient .

(MTC) valu'es for the full range of expected operating conditions during the l initial cycle are negative, the technical specifications (TS) initially  !

proposed / allowed a positive MTC below 100-percent rated thermal power, presumably to encompass future reload cycles. However, the-Electric Power Research Institute (EPRI) Advanced Light Water Reactor (ALWR) Evolutionary Plant Utility Requirements Document (URD) states that the fuel-cycle design shall have a non-positive MTC over the entire fuel cycle when the reactor is critical . The staff agrees with this requirement and considers it prudent to upgrade the safety characteristics of advanced reactors to enable reactivity increases that cause a moderator temperature increase to be self-limiting.

resulting in stable power operation. Therefore, the staff will require the l

MTC to be non-positive at all operating conditions. This was designated as l

l DSER Open Item 4.3.2-1. In accordance with this staff requirement, ABB-CE has I i

revised the MTC TS to ensure a non-positive value at all operating conditions (CESSAR-DC to Chapter 16). DSER Open Item 4.3.2-1 is, therefore, resolved.

Control To allow for changes 'of reactivity due to reactor heatup, changes in operating conditions, fuel burnup, and fission-product buildup, a significant amount of f excess reactivity is built into the core. A88-CE has provided sufficient information relating to core reactivity balance for the first core and has shown that means are incorporated into the design to control excess reactivity.

at all times. ,

Both excess reactivity and power level are controlled with movable CEAs and by varying boron concentration in the reactor coolant. In addition, the chemical' and volume control system (CVCS) is capable of shutting down the reactor by ABB-CE System 80+ FSER 4-11 February 1994

q % cber adding soluble boron pehen and maintaining the reactor at least 5-percent subcritical when refueling. The combination of control systems satisfies the requirement of GDC 26.

l System 80+ plants will operate at steady-state full power with only one bank of the full-strength CEAs slightly inserted. Limited insertion of the full-CEAg strength cer,tre. rods permits compensating for fast reactivity changes (e.g.,

that required for power level changes and for the effects of minor variations in moderator temperature and boron concentrations) without impairing shutdown capability.

The PSCEAs are provided primarily to assist in the control of core power distribution, including the suppression of xenon induced axial power oscilla-tions during power operations, and the control of axial power shape during load-following transients. They can also provide reactivity control to compensate for minor variations in moderator temperature and boron concentra-tion during power operations, and assist in compensating for changes in reactivity due to power level and xenon during load-following transients. The ,

total reactivity worth of the PSCEAs is sufficient to-enable control strate-

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gies for load-following which can reduce or even remove the need for changes  :

in boron concentration during these transients.

Rod insertion is controlled by the power-dependent insertion limits that are given in the TS. These limits ensure that (1) there is sufficient negative reactivity available to permit the rapid shutdown of the reactor with ample margin, and (2) the worth of a control rod that might be ejected in the unlikely event of an ejected rod accident is no worse than that assumed in the

accident analysis.

l abohr Soluble boron poison is used to compensate for slow reactivity changes,

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including changes associated with fuel burnup, changes 'in xenon and samarium'.

concentration, buildup of long-life fission products, burnable peNrod depletion, and the large moderatgr tyrature change from cold shutdown to.

hot standby. The soluble. boron aa system provides the capability to take the reactor at least 10-percent suberitical in the cold shutdown condition.

ABB-CE System 80+ FSER 4-12 February 1994 l .. . .-. -. .. . . . . - - , - - - _ . . . -. _ . - . - . - -

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(a) instrumentation and systems to monitor the core power distribu-

! tion, control rod positions and patterns, and other process variables such as temperature and pressure, and i

l suitable alanns or control room indications or both for these l J (b)

!, monitored variables. .

i i ABB-CE. satisfied the requirements'of GDC 26 with respect to provision of l

? (4). '

two independent reactivity control systems of different designs by

] 1 l having a system that can: 1

! i i (a) reliably co..- s1 A00s,

- (b) hold the core subcritical under cold' conditions, and j i

l (c). control planned, norr , t -

er changes. l

! (5) ABB-CE satisfied the requirements _of GDC 27 with respect to reactivity

! control systems that have a combined capability in conjunction with l' kdaddition by the emergency core cooling system of reliably -

l_ controlling reactivity changes under postulated accident conditions by:

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! abserber D'C " (a) providing a movable (..itrol rod system and a liquid po4 son system,

!- 4M to and (M abo 9e.

(b) performing calculations to demonstrate that the core has suffi-l cient shutdown margin with the highest worth stuck rod.

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(6) ABB-CE has satisfied the requirements of GDC 28 with respect to postu- ]-

f j lated reactivity accidents by:

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j (a) meeting the regulatory position in RG 1.77, 1

1 (b) meeting the' fuel enthalpy limit of 280 cal /ge, 1

j (c) meeting the criteria on the capability to cool the core, and.

j j ABB-CE System 80+ FSER 4-16 February 1994-4

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(d) using calculational methods that have been found acceptable for {

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! reactivity insertion accidents. l l

(7) ABB-CE has satisfied the requirements of GDC 10, 20, and 25 with re:,pect to SAFDLs by providing analyses demonstrating that:

(a) normal operation, including the effects of A00s, have met fuel design criteria, (b) the automatic initiation of the reactivity control system ensures 1

that fuel design criteria are not exceeded as a result of A00s and assures the automatic operation of systems and components impor-tant to safety under accident conditions, and (c) no single malfunction of the _ reactivity control system causes violation of the fuel design limits.

4.4 Thermal-Hydraulic Desian The scope of the staff's review of the thermal-hydraulic design of the core for the System 80+ design includes the design-basis and steady-state analysis  ;

of the core thermal-hydraulic performance. The acceptance criteria used as the bases of the staff's evaluation'are given in SRP_Section 4.4. ,

4.4.1 Thermal-Hydraulic Design Bases ,

l The principal thermal-hydraulic design 8 basis for the System 80+ design core is the avoidance of thennal-hydraulic induced fuel damage during normal steady-state operation and A00s. In order to satisfy the designedasis, the digital core protection calculator (CPC) provides for automatic trip or other.. correc-tive action to prevent violation of design limits. Th'e design analysis was:

performed and design limits were established based on the criteria in the-sections that follow. .

i ABC-CE System 80+ FSER 4-17 February 1994-

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4.4.1.1 Departure From Nucleate Boiling

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The margin to DNB at any point in the core is expressed in terms of the departure from DNBR. The DNBR is defined as the ratio of the heat flux required to produce DNB at the calculated local conditions to the actual local heat flux.

Thethermal-hydraulicdesign6tIasisinCESSAR-DCSection4.4.1.1fortheDNBR is as follows: "The minimum DNBR shall be such as to provide at least a 95-percent probability with 95-percent confidence that DNB does not occur on a fuel rod having that minimum DNBR during steady-state operation and A00s'."

Thisdesigni4bsisisconsistentwiththeguidelines'inSRPSection4.4.II.1 ,

and 'is acceptable. In CESSAR-DC Section 15.0.4,. ABB-CE stated that for CESSAR-DC Chapter 15 designdasis events resulting in a violation of the 95/95 DNBR safety limit, the statistical convolution method was used to calculate -

the number of failed rods. The staff evaluation of the statistical convolu-tion approach is discussed in Section 15.1 of this report.

i i 4.4.1.2 Hydraulic Stability l Thehydraulicstabilitydesign6dasisinCESSAR-DCSection4.4.1.2isasfol-I lows: " Operating conditions shall not lead to flow instability during steady-state operation or during anticipated operational occurrences." -As discussed in Section 4.4.2.3 below, this designdasis is acceptable.

j 4.4.1.3 Core Flow-l The minimum allowable reactor coolant flow less a maxinum bypass flow V (3.0 percent) is the design-basis used in the themal nargin analysis. The- ~

minimum allowable reactor coolant flow is the total de' sign flow with the-four reactor coolant pumps in operation. This is a connonly used. definition of' core flow design-basis and is acceptable.

ABB-CE System 80+ FSER 4-18 February 1994-

4.4.2 Thermal-Hydraulic Design Methodology 4.4.2.1 Thermal-Hydraulic Analysis Methods Steady-state thermal-hydraulic analysis for the System 80+ design was per-formed using the approved thermal-hydraulic code, thermal-hydraulic analytical code (TORC), and the CE-1 critical heat flux correlation. The design thermal-hydraulic margin analysis was performed with the NRC-approved methods: TORC (Ref. 1) and the fast-running version of the TORC code, CETOP-D (Refs. 2 and 3).

Mo&heA %tuhai Ce@when et Unrtanhet (MKU)

ABB-CE also used the NRC-approved hmethodology (Ref. 4) for the thermal-hydraulic analysis. Using this methodology, the engineering hot channel factors for heat flux, heat input, fuel rod pitch, and cladding diameter are combined statistically with other uncertainty factors to arrive at overall uncertainty penalty factors to be applied to the DNBR calculations performed by the CPCs and the COLSS.

l The NRC-approved methods (Refs. I through 4) were used for the thercal- l hydraulic analysis and this approach is acceptable. l 4.4.2.2 Departure From Nucleate Boiling i l

The correlation used to determine the DNBR is the "CE-1" critical heat flux correlation (Refs. 5 'and 6). The safety limit DNBR of 1.24 is used for the I

System 80+ design to provide assurance with a 95/95 confidence / probability that the hottest fuel rod will not experience DNB.

M SC U The 1.24 value incorporates all applicable penalties, such as f r rod bow, the 0.01 DNBR for HID-1 grids, and the penalties specified in the SCU The rod.

bow value used in the analysis is 1.75-percent DNBR fdr burnups up to ,

30,000 - MWD /MTU. For burnups higher than 30,000 MWD /MTU, sufficient margins exist to offset the rod bow penalty due to lower radial power peaks in these higher burnup assemblie's and rods.

ABB-CE System 80+ FSER 4-19 February 1994

. - . . -- .. - ._. = .

j.

i i ',

enclosure, " Assessment of Core Flow Stability for Applicant PWRs," CEN-64(A)-P (proprietary) and CEN-64(A) (non-proprietary), July 1977. This document presents an assessment of core flow stability for a typical AB8-CE PWR using.

the CE-HYDNA code (Currin, H.B., et al., "HYDNA-Digital Computer Program for Hydrodynamic Transients in a Pressure Tube Reactor or a Closed Channel Core,"

report CVNA-77, 1961). It was found that, for nominal coolant conditions; the flow is stable throughout the range of reactor power levels examined (100-percent to 250-percent rated power).

Past operating experience, flow stability experiments, and inherent themal-hydraulic characteristics of ABB-CE PWRs provide a basis for accepting the 1 l

System 80+ design stability evaluation for justification of an adequate design.

4.4.3 Loose Parts Monitoring System The presence of a loose object in the primary coolant system can be indicative of degraded reactor safety resulting from failure or weakening of a safety-related component. A loose part may be from a failed or weakened component or from an item inadvertently left in the primary system during construction, refueling, or maintenance and can contribute to component damage and material wear by frequently hitting other parts in the system. A loose part can pose a seriougthgt of partial flow blockage with fuel experiencing DNB, which in turn rc=lts in a fuel failure. In addition, a loose part increases the potential for control rod januning and for accumulation of increased level'of radioactive crud in the primary system. Paragraph 3.5.2.2 in the EPRI URD ,

requires instrumentation to be provided to detect the presence of loose parts in the RCS. The purpose of the detection system is the early detection of- ,

loose metallic parts in the primary system. Early. detection can provide the -

time required to avoid or mitigate safety-related damage to, or malfunction of ~

primary system components. As specified in Paragraph'II.7 of SRP.Section 4'4, .

~

the design description and proposed procedures for use of the loose parts.

monitoring system (LPMS) should be consistent with the guidance in RG 1.133, Revision 1, " Loose-Part Detection Program for Primary System of Light-Water-Cooled Reactor."

ABB-CE System 80+ FSER 4-21 February 1994

l.

. l ABB-CE described the LPMS in CESSAR-DC, Section 7.7.1.6.3 and in responses to l

j generic safety issues B-60 and C-12 in CESSAR-DC, Appendix A. The LPMS has two sensors at each natural collection region. The LPMS design will comply

! with RG 1.133, Revision 1. To be consistent with the CE standard TS, the LPMS

?

is not included in the TS for System 80+. However, as' stated in CESSAR-DC >

l j Section 7.7.1.6.3 (Amendment R), ABB-CE will provide operating guidelines .

which call for performing a channel check at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a channel functional check at least once per 31 days, a background noise check j at least once 91 days, and a channel calibration at least once per 18 months.

l These operating guidelines for the surveillance requirements are consistent

! with the guidelines of RG 1.133, Revision 1. The staff has evaluated the l System 80+ LPMS by comparing it with the systems used at other plants, taking l f into account pertinent differences.

I i

f On the basis of its review of the LPMS design and ABB-CE's commitment to t

2 comply with the guidelines in RG 1.133, Revision 1, the staff concludes that i

i an acceptable LPMS will be implemented for the System 80+ plants. ,

i l

4.4.4 Digital Core Protection Calculator and Control Element Assembly l ' Calculator ,

i l f The CPC and , control /lement $sembly [alculator (CEAC) system is designed to l provide the low departure from DNBR and high local' power density (LPD) reactor. j I

l trips to ensure that the SAFDLs on DNB and centerline fuel melt are not-I ceeded during A00s. In addition, these trips assist the h .$3 N g  ;-

L.au sw.= q ste m i In limiting the consequences of certain postulated accidents.

! In CESSAR-DC Section 7.2.1.1.2.5, ABB-CE described the hardware design of-the digital CPC and CEAC. However, ABB-CE did'not initially provide the staff l

i with a description of the software design and additional information was.

1 necessary for the staff to complete its review of the CPC/CEAC design. In' l order to complete its review, the staff requested the'following information:

! (1) description of the CPC/CEAC software design and protection algorithm; (2) identification of all the differences from the previously approved designs and evaluation of the impact of the differences on the CPC/CEAC performance-l and safety functions, and (3) discussion of the verification program for implementation of CPC/CEAC software consistent with the certificate schedule.

1 i

ABB-CE System 80+ FSER 4-22 February 1994 4

- m.-.~,-....--._m-_m-,,,4%.m._- ...- . . . ~ , .,--,..,-----_.----.---_+_.-..,,.m x ..~mc..,v.,,. - - . , ..._,_m.,

j. -

i-i As an alternative, ABS-CE could discuss the' applicability of the previously i '

submitted and approved documents for the System 80 CPC/CEAC design to the i

i System 80+ design. This was identified as DSER Open Item 4.4.4-1. I'

]

l In response, ABB-CE revised CESSAR-DC'7.2.1.1.2.5 (Amendment R) to indicate that the software design of the CPC/CEAC system is described in References 10 l

4 through 16, and has been reviewed and approved by the NRC in References 15 through 21. The COL application and ABB-CE will follow the procedures i

described in References 22 and 23 for all changes to the algorithms, data base

! constants and data block constants for the CPCs and CEACs. The staff finds

) that the procedures documented in References 22 and 23 were previously approved by the NRC. The overall CPC/CEAC. software implementation, which is to translate the system functional requirements into modules of machine executable code and to integrate these modules into a real time software

~

i system, is verified through the Phase I and Phase II software verification test. The scope of testing will include generation of plant-specific data f base document, generation of appropriate test cases and acceptable criteria, j ~

i and test reports. Phase I testing is to be performed on the DNBR/LPD calcula-tion systems to veri y that CPC/CEAC system software modifications have been properly . Phase II testing is performed on the CPC/CEAC system to l

j verify that CPC and CEAC software modifications have been properly integrated with the CPC and CEAC software and system hardware, and to provide confirma-tion that the static and dynamic operation and the integrated system as

) modified is consistent with that predicted by design analyses.

Testing of the CPC/CEAC software for each license applicant referencing the ~

i System 80+ design certification will be considered complete with the formal

! issuance of (1) CPC/CEAC data base document, (2) the Phase I test report, and (3) the Phase II test report. These documents are plant specific and will be-

reviewed individually for each license application referencing the System 80+'

f design certification. DSER Open Item 4.4.4-1 ~is, ther'efore, reclassified to COL Action Item 4.4.4-1.

l i .

i-i i

ABB-CE System 80+ FSER 4-23 February 1994

]s

d

i. .

thermal-hydraulic analyses using analytical methods, DNBR correlations, and the ' safety limit DNBR that the staff previously approved. Therefore, the staff concludes that the thermal-hydraulic design of the System 80+ design.

core provides appropriate thermal margin to assure that SAFDLs are not exceeded during any conditions of normal operation and A00s, and thus, _

conforms to the requirements of GDC 10 and is acceptable. The staff's -

evaluation of the calculated DNBRs during A00s is included in Sections 15.1

(

l through 15.3 of this report.

I l ~

Each COL applicant referencing the System 80+ design certification has overall l

j responsibility for the startup test program. However, the CESSAR-DC defines

( ABB-CE's participation and provides guidelines to the reference plants for i

! pre-operational and initial startup test program in accordance with RG 1.68 to i

measure and confirm thermal-hydraulic design aspects. .The evaluation of startup testing program is included in Section 14 of this report.

References for Section 4.4

1. " TORC Code: A Computer Code for Determining the Thermal Margin of a Reactor Core," CENPD-161-A, April 1986.

2. " TORC Code - Verification and Simplified Modeling Methods," CENPD-206-P-A, June 1981.

l

3. "CETOP-D Code Structure and Modelling Methods for Arkansas Nuclear One .

Unit 2," CEN-214-A-P, July 1982.

" Medthed Shtuhmi Combmahon ct Oncertatnhes," CEN t5L(,0- 9-k ,

4. "Statistice Cu.iuination of Unceri.aini.i :5," CEN-130-A-F, Hv  ;,er 1000. '

ketHon 1 h e ' Mo3 MM,

5. " Critical Heat Flux Correlations for CE Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," CENPD-162-A-P, September 1976.

6. " Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Spacer Grids, Part 2, Non-Uniform Power Distribution," CEN-207-A-P, June 1976.

ABB-CE System 80+ FSER 4-27 February 1994 L , . __ ~ u.. _ _ . _ . . _ _ _ . . _ ___ _ _ _ _ _ . _ _ _ _ _ _ . .

18. " Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station Unit No. 3," Docket No. 50-382, Louisiana Power and Light Company,' July 1981.

19. " Safety Evaluation Report Related to the Operation of Palo Verde Nuclear Generating Station, Units 1, 2 and 3," Docket-Nos. STN 50-528, STN 50,529, and STN 50-530, Arizona Public Service Company, October 1984.

i

20. " Safety Evaluation Related to Amendment No. 32 to Nuclear Power Facility (NPF)-10 and Amendment No. 21 to NPF-15 for San Onofre Nuclear Generating l

Station, Units 2 and 3," Docket Nos. 50-361 and 50-362, Southern Califor-nia Edison Company, March 1985.

21. " Safety Evaluation Related to Amendment No. 66 of Facility Operating License No. NPF-6, Arkansas Power & Light Company, Arkansas Nuclear One Unit 2," Docket No. 50-368, May 1985.

22. "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P, Revision 3-P-A, November 1986.

23. " Reload Data Block Constant Installation Guidelines," CEN-323-P-A, f Revision 1-P, December 1986.

4.5 Reactor Materials 4.5.1 Control Element Drive Mechanism (CEDM) Structural Materials The staff's review was performed in accordance with SRP Section 4.5.1 and l

(

i consisted of an evaluation of CESSAR-DC Section 4.5.1. Areas reviewed were material specifications, austenitic stainless steel components and their-welding, fabrication, and inspection in plant systems,' heat treatment of other materials, and cleaning and cleanliness control. Resolution of DSER open items is as follows: ,

In Amendment Q to CESSAR-DC Section 4.5.1.1, ABB-CE indicates that Inconel 690 is to be used in lieu of 600 materials in the fabrication of th mo$orhousing i

4-29 February 1994' ABB-CE System 80+ FSER

l-l- assembly. The staff views the Inconel 690 alloy as the preferred nickel base i l alloy in the primary and secondary coolant loops because of its improved l corrosion resistance compared to Inconel 600. The use of Inconel 690 and its l equivalent weld metals (Types 52 and 152) will provide reasonable assurance of

l. the material integrity of the components and tubing in contact with reactor I coolant and most secondary water chemistries. On this basis, DSER Open -

1 l

l Item 4.5.1-1 is resolved. ,I

, 1 l ABB-CE proposes to use American Society for Testing and Materials. (ASTM) A-708 l in lieu of ASTM A-262, Practice A or E (recommended in RG 1.44) for verifying-

non-sensitization of austenitic stainless steel materials. The proposed

l. alternative (ASTM A-708) is a more stringent test than ASTM A-262. The ASTM A-708 test uses the same test solution and same temperature as ASTM A-262, f

Practice E; however, A-708 requires 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in the test solution while A-262,

! Practice E requires only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (with an option for longer test times upon agreement between supplier and user). Note 26 of A-262, Practice E acknowl-l j edges that results in the literature indicate that Practice E is more sensi-

! tive when run for longer times. Accordingly, as acknowledged in SRP 5.4.1, paragraph III.2, ASTM A-708 is an acceptable alternative test for ASTM A-262, Practice A or E. On this basis, DSER Open Item 4.5.1-2 is resolved.

i i

ABB-CEproposestouseStegtej which is a cobalt-based alloy, for pins and

! latches in the CEDM. R:db-Ntt$1= of cobalt is a concern relating to the i

i radioactivity in current nuclear plants. Therefore, cobalt application should l be avoided in the Sys' tem 80+ design for as low as reasonably achievable cons-

) iderations. In CESSAR-DC Section 5.2.3.2.2, " Materials of Construction

{ Compatibility with Reactor Coolant," ABB-CE states that cobalt-based alloys j will be avoided-except in cases where no proven alternative exists. Cobalt-

! free alloys with the wear and corrosion properties of the Stellite (cobalt l base) type alloys, although under development, have not been fully demon-l strated to have the usability of the Ste111tes at this' time. Accord'ingly, the-staff acknowledges that, at this time, there is ' limited availability of i alternative materials for Ste111tes. On this basis, DSER Open Item 4.5.1-3 is 3

3 resolved.

i i

j ABB-CE System 80+ FSER 4-30 February 1994-I . . . _ . . . - _ _ . _ _ . , _ . . . _ . . _ . ,_ - , , _ _ , - , . .

ABB-CE proposes to use Types 304 and 316 austenitic stainless steel. However, these materials are susceptible to intergranular stress corrosion cracking when the oxygen content of the reactor coolant exceeds 0.010 ppm at tempera-tures above 200 *F during normal operations. During start-up and operation of the System 80+ plant, these temperature and chemical conditions are maintained through specified chemistry control. ABB-CE has taken alternative mitigating approache's as allowed in RG 1.44, thus providing reasonable. assurance of the integrity of austenitic stainless steel components in contact with reactor cool ant. On this basis, DSER Open Item 4.5.1-4 is resolved.

The ferrite content limits for austenitic stainless weld metal in the CESSAR-DC were broader than those in industry guidelines (Ref. 1) and staff guidance (Ref. 2). In CESSAR-DC Amendment L, ABB-CE modified the CESSAR-DC to be consistent with~ industry guidelines and staff guidance. This should provide reasonable assurance of an acceptable level of structural integrity for stainless steel welds over the life of the plant. On this basis, DSER Open Item 4.5.1-5 is resolved.

In CESSAR-DC Section'4.5.1.1, ABB-CE indicates that martensitic stainless steel (Type 403) will be used for the control element drive motor housing.

ABB-CE has specified the following material and the particular heat treatment for this component in the reactor coolant pressure boundary:

Material SA-182, Type 403, as modified by ASME Code Case (4-N-111 4-%-\\

Heat treat Heat to 1800 *F (982 *C) +/- 25 *F (14 *C), air cool and temper at 1125 *F (607 *C) minimum for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (Code Case .

k -% - W This is the only component in the reactor coolant pressure boundary using martensitic stainless steel. For this component, ABB-CE has. elected to use ASME Code Case N-4-11 material which is a modified Type 403 martensitic stainless steel, with the additional requirements of ASME SA-182. This heat treatment should ' provide a Rockwell "C" hardness of the material in the ABB-CE System 80+ FSER 4-31 February 1994 4

, - - . . , . . . . . - , . -% sm,-,.., -, ,-, , .-......_.#_m... . . . . , , . - . .~.-,,.,,-m.,s..m . , ,w , , - .. v..e

e ABB-CE has evaluated CEDM materials in an extensively tested CEDM assembly that exceeded lifetime requirements, as described in CESSAR-DC Sec-tion 3.9.4.4.1 (Amendments E, K, and Q). The design duty or lifetime require-ment as defined in CESSAR-DC Section 3.9.4.1 is a total cumulative CEA travel of 100,000 feet of operation without loss of function and not the 60-year plant life. As indicated in CESSAR-DC Section 3.9.4.1, the CEDM is designed to operate without maintenance for a minimum of 1-1/2 years and without replacing components for a minimum of 3 years. Therefore, the test results of the extensively tested CEDM do not need to be verified to the 60-year plant life.

The operational requirement for the System 80+ CEAr/s, with the possible exception of the lead regulating CEA group, is expected to be less than the 100,000 feet of travel (the tested life) over the 60 year plant life. If plants institute daily load cycle operation on a regular basis, the lead regulating CEA group may exceed 100,000 feet of travel.

The regulating CEA are much lighter than the CEA weight used during accelerated CEDM motor life tests, and it is expected that, when operating a regulating CEA, the System 80+ CEDM motors are capable of operation in excess of 100,000 feet of cumulative travel. Depending on the extent that the lead regulating CEA group is utilized, a one time CEDM motor replacement for this bank of CEAk may be required during the 60 year plant life. Replacement of the CEDM motor in case of extreme usage can be accomplished, and is an acceptable approach.

As indicated in CESSAR-DC Section 3.9.4.1, all CEDM pressure boundary compo-nents have a design life of 60 years. On this basis, DSER Open Item 4.5.1-9 is resolved.

The staff concludes from the review that the control rod drive mechanism structural materials are acceptable and meet the applicable portions of the requirements of GOC 1, 14, and 26 of Appendix A and Section 50.55a of 10 CFR Part 50.

ABB-CE System 80+ FSER 4-33 February 1994

l Fabrication and heat treat practices specified in accordance with these I recommendations provide added assurance that stress corrosion cracking will l not occur during the design life of the components. The compatibility of all materials used in the control rod system in contact with the reactor coolant, satisfies the criterion of Subarticles NB-2120 and N8-3120 of the Code. .

Cleaning and cleanliness control are in accordance with American National Standards Institute (ANSI)/ASME NQA-2-1983, " Quality Assurance Requirements l

for Nuclear Power Plants," and RG 1.37, " Quality Assurance Requirements for Cleaning Fluid Systems and Associated Components of Water-Cooled Nuclear Power Pl ants . " The staff has previously reviewed ANSI /ASME NQA-2-1983 and finds it acceptable. The cleaning and cleanliness control specified will provide adequate contamination control of components during fabrication, shipment, and storage.

Conformance with the codes, standards, and RGs, conformance with our positions on the allowable maximum yield strength of cold worked @tenitxip stainless steel, and generally the tempering temperatures of martensitic stainless steels, constitute an acceptable basis for meeting the requirements of GDC 1, 14, and 26 and Appendix A and Section 50.55a of 10 CFR Part 50.

l l References .

i (1) EPRI, " Advanced Light Water Reactor Utility Requirements Document,"' '

NP-6780-L, Volume 2, ALWR Evolutionary Plant, Chapter 1, Overall Require-ments, Revision 3, November 1991. ,

4

%h

/

4.5.2 Reactor Internals Materials MUNk$

wM

' The staff's review was perfonned in accordance with SRP Section 4.52a consisted of an evaluation of CESSAR-DC Section 4.5~ 4 material specifications, control on welding of materials used for rea v4 c ' N internals, non-destructive examination of wrought seamless tubular products- %

and fittings, austenitic stainless steel components and the welding,, fNica-tion, and inspection of reactor internals, and heat treatment of other -['

materials. Resolution of DSER open items is as follows: M 1

4-35 Februaryr1994:

ABB-CE System 80+ FSER

4 I

adequately protected during operation from conditions that could lead to stress corrosion of the materials and loss of component structural integrity.

The material selection, fabrication practices, examination and testing pro- l cedures, and control practices performed in accordance with these recommen- j dations offer reasonable assurance that the materials used for the reactor l internals'and core support structures will be in a metallurgical condition to preclude inservice deterioration. Conformance with requirements of the ASME Code and the recomendations of the above RGs constitutes an acceptable basis ,

for meeting, in part, the requirements of GDC 1 and 10 CFR 50.55a. l l

4.6 Functional Desian of Reactivity Control Systems l The functional design of the reactivity control systems for the System 80+

standard plant is within the scope of design certification, and the staff l reviewed CESSAR-DC to confirm that the design has the capability to satisfy l 1

the various reactivity control conditions for all modes of plant operations.

These are:

(1) the capability to operate in the unrodded, critical, full-power mode throughout plant life (2) the capability to vary power level from full power to hot shutdown and ensure control of power distributions M within acceptable limits at-3 any power level i

J (3) the capability to shut down the reactor in'a manner sufficient to miti-

gate the effects of postulated events discussed in Chapter 15 of~ this c nn b

j report. W g.g .

g W QDnd YNY j Thg v ntrol' systems for the facility are the ::-tn! el

. , _. .= sistes; (CE".00), the safety injection system (SIS), and the-

.t

[Q l

i[

CVCS.  ;

1 C.Ab3 l The GEABS contains magnetic jack CEDMs. When electrical power is removedifrom-the coils of the CEDM, the armature springs automatically cause the holding.

I AB8-CE System 80+ FSER' 4-39; February 1994; f

5

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

s.u... - .. - . - . + ....n n..- . . . _

l' l latches to be disengaged from the CEDM drive shaft, allowing insertion of the 1 CEAs and the PSCEAs by gravity. There are 68 full-strength CEAs and l

l 25 PSCEAs. The regulating CEA groups (T.11 =d p t tra.Jh) may be used to  !

compensate for changes in reactivity associated with power-level changes and t

power distribution, variations in moderator temperature, or changes in baron concentration. Refer to Sections 3.9.4 and 4.3 of this FSER for further .

f j discussion of this feature. The PSCEA consists of an Inconel 625 tube loaded a with'Inconel 625 bars over the full active length. The PSCEAs, which have '

j lower worth in comparison to the full-strength CEAs, are provided for reactiv-ity and axial power shape control during power operations.

The SIS is automatically actuated to inject borated water into the RCS upon 3

i receipt of a safety injection actuation signal. The SIS pumps take suction

! from the in-containment refueling water storage tank. The SIS is discussed s

further in Section 6.3 of this report. ,

t The CVCS is a non-safety-grade system designed to control slow or long-ters reactivity changes such as those caused by variation in coolant temperature, l  !

I fuel burnup, or variations in the xenon concentration. The CVCS controls

~

reactivity by adjusting the dissolved boron concentration in the RCS. The l

boron concentration is controlled to obtain optimum CEA positioning,'to compensate for reactivity changes during startup, load following (changes in l

reactor power level), and shutdown, 'and to provide shutdown margin for main-i tenance and refueling operations or emergencies. The boric acid concentration s

j in the RCS is controlled by the charging and letdown portions of.the CVCS.

1 i The CVCS can be used to maintain reactivity within the required bounds by. l means of the automatic makeup system which replaces minor coolant leakage N(

)

without significantly changing the boron concentration in the RCS system.; f<

j Dilution of'the RCS boron concentration is required to compensate for that 4 f ' e& 3 mb o i reactivity losses occurring as a result of fuel' and barnable depletion. .. %

This is accomplished by manual' operation of the CVCS. The CVCS is discuss b h h o$

i further in Section 9.3.4 of this report. %gv i a j n i  %

ABB-CE System 80+ FSER 4-40' February 1994:

_=

l

~ l The concentration of boron in the RCS is changed manually under the following  :

)

operating conditions ,

(1) Startup - boron concentration is decreased to compensate for moderator temperature and power increase.

(2) Load' follow-boronconcegntationisincreajedordecreasedtoco sateYo nom [NItsTolEgioa'dNhanIjh Y*

(3) Fuel burnup - boron concentration is decreased to compensate for burnup.

1 (4) Cold shutdown - boron concentration is increased to compensate for increased moderator density due to cooldown.

ch m qEs are_ I boccA Soluble pe44on concentration is used to control slow operating reactivity changes. If necessary, CEA movement can also be used to accommodate such changes, but assembly insertion is used rainly to mitigate ACOs (the analysis.

. In either case, feal assumes design limitsawill single not bemalfunction exceeded. Thesuch as a stuck rod)M control is capa soluble maintaining the core subcritical under conditions of cold shutdown which conforms to the requirements of GDC 26, " Reactivity Control System Redundancy and Capability."  !

The CEAs are the primary shutdown mechanism for normal operation, accidents, and transients. They'are inserted automatically irt accident and transient conditions. Concentrated boric acid solution is injected by the SIS in the-

?

event of a LOCA, steamline break, loss of normal feed-water flow, steam generator tube rupture, or CEA ejection, thereby complying with GDC 20, ' A

" Protection System Functions," which requires that autonatic protective ${

systemsbeprovided(1)toensurethat-SAFDLsarenot,exceededand(2)to{^%%fq ' '

sense accident conditions and to initiate operation of* safety related systems-

- and components. 3r Iyk '

up'L; g ,

The ability of'the CEAs"to have their position changed,is tested quarterlyt }.

during power operation. At every refueling shutdown, each CEA is stepped over its entire range of movement, and drop tests are performed to demonstrate the .

ABB-CE System 80+ FSER 4-41 February 1994-

~ . - - - _ - . - .

, . 1 CEh or.

e l ability of the assemblies to meet required drop times. The GEA-is-designed j such that a single failure will not result in loss of the protection system, nor will a loss of redundancy occur as a result of removal of a channel or l component from service. This is discussed further in Section 7.2 of this report. The foregoing periodic testing, reliability, and redundancy conform to the requirements of GDC 21, " Protection System Reliability and Test- -

ability." '

Failure of electrical power to any CEDM will result in insertion of that assembly. Analysis of accidental withdrawal of a CEA was found to have acceptable results as discussed in Section 15.2.4 of this report. This f conforms to the requirements of GDC.23, " Protection System Failure Modes," and GDC 25, " Protection System Requirements for Reactivity Control Malfunctions."

The reactivity control system functional design meets the requirements of 1 GDC 21, 23, 25, 26, and 27 with respect to its reliability and testability, fail-safe design, malfunction protection design, redundancy and capability, and combined systems capability, and is, therefore,' acceptable.  !

l l

ABB-CE System 80+ FSER 4-42 February 1994

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• Table 4.2-1. Design Features And Acceptance Criteria For Fuel And Initial Core Design Desian Feature Accentance Criterion Fuel Desian Maximum Fuel Rod Burnup s60,000 MWD /MTU Fuel Rod U-235 Enrichment s3.7 wt. %

Ababen Er230 Burnable Pci:= Concentration s2.0 wt. %

Number of Integral Er 03 Burnable s124

.Woner Peiten%s Per Fuel kssembly Ws Natural or Lcu Enncheneck Uranium Axial Blanketing th %,0s turnqbb $7.5 inches Absocbec (Top an9 Bottom of Fuel and4P Rods)

Er2 30 Burnable Petson Rod Cutback s15.0 inches Length (Top and Botto h g l Initial Core Desian l

Core Power Level $3914 MWt l Cycle Length s16,000 MWD /MTU ActephWe CSder\Q Ore. nownol %\ue.g.

l .

ABB-CE System 80+ FSER 4-43 February 1994

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l Table 4.2-2. Evaluated Design Parameters And Acceptance Criteria For Fuel And Initial Core Design l .

Evaluated Desion Parameter Accentance Criterion Core Average U-235 Enrichment 82.6 wt. %

Maximum Unrodded 3-Dimensional s2.28 i Peaking Factor. (F,)

Maximum Unrodded Integrated sl.55 f

l Radial Peaking Factor (F,)

i Minimum DNBR hl.24 Net CEA Shutdown Worth h8.86% v hot full power (HFP)

NTC (HFP, all rods out) s-0.1 E-4 g/*F, h-3.5 E-4 g/*F Power Coefficient <0.0 g/(kW/ft) l

! Critical Baron Concentration s1056 ppe l

Ctephve. CNterio Qre nommq) Natues, i

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l i

ABB-CE System 80+ FSER 4-44 February 1994

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Table 4.4-1. Reactor Design Comparison Characteristics System 80+ System 80

!. Performance Characteristics:

Reactor Core Heat Output (Mwt) 3,914 3,800 System Pressure, kPa (psia) 15.5 x 103 (2,250) 15.5 x 103(2,250)

Minimum DNBR at Steady-State 2.00 1.98 Full Power Minimum DNBR Limit 1.24 (SCU) 1.24 (SCU)

Critical Heat Flux Correlation CE-1 CE-1 II. Coolant Flow:

Tet::1 Flow Rate L per min (gpm) 1.68 x 10' 1.69 x 10' (444,650) (445,600)

Effective Flow Rate for 1.63 x 10' 1.64 x 10' Heat Transfer (gpm) (431,300) (432,200)

Average Velocity Along Fuel 5.1 (16.7) 5.1 (W S)

Rods m/s (ft/s) 6.3 QA Average Mass Veloc'ity lih0(2.65) 12.8 (2.62) 10' kg/hr-ma (10 lb/hr-ft')

III. Coolant Temoerature:

Nominal Reactor . Inlet *C (*F) 291 (556) 296 (565)

Average Rise in Core *C (*F) 33.9 (61) 32.2 (58)

IV. Heat Transfer. 100-Percent Power:

6.6 x 103 (70,960) 6.4 x 103(68,320)

Active Area Hpat m' (ft T[)ansfer Surface ST),%c Pn,Too Average Heat Flux w/m2 (BTU /hr-ftz) 577,000 (183,300) 59Er700(184,800) 1.W '

i.%

2 Maximum Alls ; bis Heat Flux w/m h Sf x 10' h3P x:10'

' F (434,3M) N%,100)

(BTU /hr-ft ) (482400) 9%,ted) 17-n.c (7.m -(s.3(,)

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Average Linear Heat Rate kw/m . 0 17.7 (+c42)

(kw/ft) es.33) hemmurrs Peait-Alle.;;ble Linear Heat 4ht-1WS) +h (12.7)

Generation Rate kw/m (kw/ft)- 91.1 Og.Q 'h.5 ABB-CE System 80+ FSER 4-45 February 1994

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