Text

Application for Amends to Licenses NPF-9 & NPF-17,changing Tech Specs to Ensure Plant Operation Consistent W/Design & Statements in Unit 1 Cycle 4 Reload Safety Evaluation.Fee Paid
	ML20155J567
Person / Time
Site:	McGuire, Mcguire
Issue date:	05/15/1986
From:	Tucker H; DUKE POWER CO.
To:	Harold Denton, Youngblood B; Office of Nuclear Reactor Regulation
Shared Package
ML20149C022	List: ML20155J567;
References
	TAC-56746, TAC-56747, TAC-61512, TAC-61513, NUDOCS 8605230318
	Download: ML20155J567 (5)
	v • d • e

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DUKE POWER GOMPANY l'.O.150X 33180 CitAHLOTTE, N.C. 28J42 IIAL 15. TUCKER TELE Pff 0NE

% es E ParetDEWT (704) 373-4531 aiutan reonarvion May 15, 1986 Mr. Harald R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 ATTENTION:

B.J. Youngblood, Director PWR Project Directorate #4

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370 McGuire 1/ Cycle 4 0FA Reload

Dear Mr. Denton:

Mr. H.B. Tucker's (DPC) November 14, 1983 letter to Mr. H.R. Denton (NRC/0NRR) described planned changes in the fuel design for McGuire Nuclear Station, Units 1 and 2.

Commencing with the first refueling of each of the units, the standard fuel assemblies in use were to be replaced over the next four refuelings with optimized fuel assemblies (OFA). The letter transmitted a reference safety evaluation describing the safety impact of operation with a transition core and an all 0FA core. McGuire Unit I has begun this process with the NRC naving approved the necessary license amendments, and Unit 1/ Cycle 3 is currently operating with two 0FA reload regions. The third such 0FA Reload Region is scheduled for the upcoming cycle 4 refueling.

(McGuire Unit 2 is currently in its cycle 3 refueling outage preparing for operation with its second 0FA Reload Region.)

Attached are proposed license amendments to facility operating licenses NPF-9 and NPF-17 for McGuire Nuclear Station Units 1 and 2, respectively.

The proposed amendments ensure tnat plant operation is consistent with the design and safety evaluation conclusion statements made in the McGuire Unit 1 Cycle 4 Reload Safety Evaluation (RSE) and ensure that these conclusions remain valid. Note that the McGuire Unit 2 specifications are administratively affected in that a specification currently applying to both McGuire Units 1 and 2 is split into two portions addressing the separate requirements for Units 1 and 2 created by these changes.

In addition, the increased positive moderator temperature coefficient limit of

+7 pcm/ degrees F is being requested for McGuire Unit 2 as well as Unit 1.

Although the increase in the positive MTC limit was assumed in the McGuire 1/ Cycle 4 RSE, the change is not a consequence of the reload and therefore is acceptable for immediate implementation on Unit 2 (while this increase was not assumed in the McGuire Unit 2/ Cycle 3 RSE, it will be needed (assumed) for future reloads and therefore should be approved on Unit 2 sometime prior to McGuire 2/ Cycle 4).

The McGuire FSAR will be revised to reflect this positive MTC change in the appropriate annual FSAR update k'

_ follow h app,roval._. _ _.

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Mr. Harold R. Denton, Director

(

May 15, 1986 Page 2 3

Attachment I contains the proposed technical specification changes, and Attachment.2 discusses the Justification and Safety Analysis to support the proposed changes.

Included in Attachment 2 is: A) the cycle-specific reload safety, evaluation for McGuire Unit 1/ Cycle 4; and B) the safety evaluation for operation of McGuire Units 1 and 2 with a positive moderator coefficient. The peaking factor limit report for McGuire Unit 1/ Cycle 4 which is required in accordance with McGuire Technical Specification 6.9.1.9 will be submitted at least 60 days prior to cycle initial criticality. -Pursuant to 10 CFR 50.91, Attachment 3 provides an analysis performed in accordance with the' standards contained in 10 CFR 50.92 which concludes.that the proposed amendments do not involve a significant hazards consideration. The proposed amendments have been reviewed and determined to have no adverse safety or environmental impact.

It is requested that the proposed amendments receive timely review and approval in view of the current McGuire Unit 1/ Cycle 4 startup schedule.

Unit.1 end of Cycle 3 refueling shutdown is currently scheduled for May 16,.

1986 with Cycle 4 initial criticality scheduled for July 24, 1986. Any changes to.this schedule will be provided to the NRC staff. The implications of failure to have these amendments approved by Unit 1/ Cycle 4-criticality are described in' Attachment 2.

Pursuant to 10 CFR 170.3(y), 170.12(c), and 170.21, Duke Power proposes that this application contains license amendments for McGuire Units 1 and 2 subject to fees based on the full cost of the review (to be calculated using the applicable professional staff rates shown in 10 CFR 170.20) and must be accompanied by an application fee of'4150, with the NRC to bill Duke Power at six-month intervals for all accumulated costs for the application or when review is completed, whichever is earlier.

Accordingly, please find enclosed a check in the amount of $150.00.

Should there be any questions concerning this matter or if additional information is required, please advise.

Very truly yours, a r e d i' "

x-llal B. Tucker PBN/jgm Attachments b

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Mr. Harold R. Denton, Director May 15, 1986 Page 3 xc:

(w/ attachments)

Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. Dayne Brown, Chief Radiation Protection Branch Division of Facility Services Department of Human Resources P.O. Box 12200 Raleigh, North Carolina 27605 Mr. Darl Hood Division'of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Mr. W.T. Orders Senior Resident Inspector McGuire Nuclear Station 1

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Mr. Harold R. Denton, Director May 15, 1986 Page 5 RAL B. TUCKER, being duly sworn, states that he is Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this revision to the McGuire Nuclear Station License Nos. NPF-9 and NPF-17 and that all statements and matters set forth therein are true and correct to the best of his knowledge.

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Hal B. Tucker, Vice President Subscribed and sworn to before me this 15th day of May, 1986.

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e ATTACHMENT 1 PROPOSED McGUIRE UNIT 1 AND 2 TECHNICAL SPECIFICATION CHANGES

a INDEX LIMITING CONDITIONS FOR OP RATION AND SURVEILLANCE REQUIREMENTS 1

SECTION PAGE 3/4.0 APPLICABILITY................................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T

> 200*F............................

3/4 1-1 gg Shutdown Margin - T,g <_ 200' F...........................

3/4 1-3 Moderator Temperature Coefficient.........................

3/4 1-4 FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT V 3h /-L 4 l 31N4M L m m. n m.>, n. m.,S POWER LEVEL......

a.u. n.e. v,.p z..,,. 3/4-145

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Minimum Temperature for Criticality.......................

3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown......................................

3/4 1-7 A

Flow Paths - Operating....................................

3/4 1-8 Charging Pump - Shutdown..................................

3/4 1-9 Charging Pumps - Operating................................

3/4 1-10 Borated Water Source - Shutdown...........................

3/4 1-11 Borated Water Sdurces - Operating.........................

3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height..............................................

3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00.....................

3/4 1-16 Position Indication Systems - Operating...................

3/4 1-17 Position Indication System - Shutdown.....................

3/,4 1-18 Rod Drop Time (Units 1 and 2).............................

3/4 1-19 Shutdown Rod Insertion Limit..............................

3/4 1-20

---s

' 8605230319 860515 PDR ADOCK 05000369 CP _. _m. : _ _.

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_s McQUIRE - UNITS 1 and 2 IV Amendment No.

' (Unit 1)

Amendment No.

(Unit 2)

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i INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Control Rod Insertion Limits..............................

3/4 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION...................................

3.4 1-22 FIGURE 3.1-2 (BLANK)...............................................

3/4 1-23 3/4.2 POWER DISTRIBUTION LIMITS

,s 3/4.2.1 A X I A L F LU X D I FF E RENC E BDFtt<I1............................ 3/4 2-1 S,i NWN

& M - + n W. c 0W3M h

6 FIGURE 3.2-la AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF 3

(4 b,:'

RATED THERMAL POWER (Unit 1).......................

3/42-/

FIGURE 3.2-lb AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF 9

RATED THERMAL POWER (Unit 2).......................

3/4 2-g 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z).....

3/4 2-6 9

FIGURE 3.2-2a K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT s

(Unit 1).........q 3/4 2-12

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FIGURE 3.2-2b K(Z) - NORMALIZED FQ(Z) AS /. FUNCTION OF CORE HEIGHT

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(Unit 2)...........................................

3/4 2-13 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R..................................................

3/4 2-14 c[

FIGURE 3.2-3a RCS TOTAL FLOW RATE VERSUS R (Unit 1)................

3/4 2-16

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FIGURE 3.2-3b RCS FLOW RATE VERSUS R AND R FOUR LOOPS t

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IN OPERATION (Unit 2)..............................

3/4 2-17 R

l FIGURE 3.2-4 R00 80W PENALTY AS A FUNCTION OF BURNUP (Unit 2).....

3/4 2-18 3/4.2.4 QUADRANT POWER TILT RATI0.................................

3/4 2-19 j

c 3/4.2.5 DN8 PARAMETERS.........................................

3/4 2-22 TABLE 3.2-1 DN8 PARAMETERS..............

3/4 2-23 k

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.......................

3/4 3-1 h

McGUIRE - UNITS 1 and 2 V

Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

A/ o C44mC5 FM in an nA r, ~ > ~ a REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be:

LesspositivethanthelimitsshowninFigure3.1-0,and a.

b.

Less negative than -4.1 x 10 4 delta k/k/ F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY:

Specifications 3.1.1.3a. - MODES 1 and 2* only.#

Specification 3.1.1.3b. - MODES 1, 2, and 3 only.#

ACTION:

With the MTC more positive than the limit of Specification 3.1.1.3a.

a.

above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the limits shown in Figure 3.1-0 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition d

to the insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies ~that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3.

A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.

With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l

With K,ff greater than or equal to 1.0.

See Special Test Exception 3.10.3.

McGUIRE - UNITS 1 and 2 3/4 1-4 Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2)

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REACTIVITY CONTROL SYSTEMS SURVEILLANCEREQUIRE[4ENTS 4.1.1. 3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

The MTC shall be measured and compared to the BOL limit of a.

' Spec'ification 3.1.1.3a., above, prior to initial operation above 5%

l of RATED THERMAL POWER, after each fuel loading; and b.

The MTC shall be measured at any THERMAL POWER and compared to

-3.2 x 10 4 delta k/k/*F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppe.

In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 delta k/k/*F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specifica-tion 3.1.1.3b.., at least once per 14 EFPD during the remainder of the j

fuel cycle.

McGUIRE - UNITS 1 and 2 3/4 1-5 Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2) i

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stA.IVITY CONTROL SYSTEMS BASES 3/4.1.1 E0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A :,0fficient SHUTDOWN MARGIN ensures that:

(1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subtritical to preclude inadvertent criticality in the shutdown condition.

SHUTOOWN MARGIN requirements vary throughout core life as a function of fuel depletion,.RCS boron concentration, and RCS T,yg.

The most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is assdeted with a postulated steam line break accident and resulting uncon-troiled RCS cooldown.

In the analysis of this accident, a minimum SHUT 00WN MARGIN of 1.3% delta k/k is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting con-dition and is consistent with FSAR safety analysis assumptions.

With T,yg less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% delta k/k SHUTDOWN MARGIN provides adequate protection.

3/4 1 1 1 Mn0ERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MOC used in the FSAR analyses to nominal operating conditions.

These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting MTC value -4.1 x 10 4 delta k/k/*F.

The MTC value of -3.2 x 10 4 delta k/k/ F represents a conservative value (with corrections for burnup and soluble baron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10 4 k/k/ F.

McGUIU - UNITS 1 and 2 8 3/4 1-1 Amendment No. 42 (Unit 1)

Asesdaent No,23_(Unit 2)

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REACTIVITY CONTR0L S'YSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The Surveillance Requirements for measurement of the MTC at th: Rginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURI FOR CRITICALITY This specification ensures that the reactor will not be made critical i

with the Reactor Coolant System average temperature less than 551'F. This limitation is required to ensure:

(1) the moderator temperature coefficient

)

is within it analyzed temperature range, (2) the trip instrumentation is

~

within its normal operating range, (3) the pressurizer is capable of being in an OPERA 8LE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

NOT 3/4.1.2 B0 RATION SYSTEMS The Boron Injection System ensures that_ negative reactivity control is available during each mode of facility operation.

The components required to perform this function include:

(1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The boration capability of either flow path is sufficient to provide a SHUTOOWN MARGIN from expected operating conditions of 1.3% delta k/k after xenon decay

{

and cooldown to 200*F.

The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 16,321 gallons of 7000 ppm borated water from the boric acid storage tanks or 75,000 gallons of 2000 ppm borated water from the refueling water storage tank (RWST).

With'the RCS temperature below 200'F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 300*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

McGUIRE - UNITS 1 and 2 8 3/4 1-2 Amendment No. 42 (Unit 1)

Amendment No. 23 (Unit 2) l

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FsA s ~ FaAnar os o~ex s 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL ELUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

the allowed operational space defined by Figure 3.2-1 for RAOC operation, a.

or within a 15 percent target band about the target flux difference during base b.

load operation.

APPLICABILITY:

MODE I above 50% of RATED THERMAL POWER *.

ACTION:

For RA00 operation with the indicated AFD outside of the Figure 3.2-1 a.

limits, 1.

Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -

High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

For Base Load operation above APLND" with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target flux difference:

1.

Either restore the indicated AFD to within the target band limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes.

THERMAL POWER shall not be increased above 50% of RATED THERMAL c.

POWER unless the indicated AFD is within the Figure 3.2-1 limits.

See Special Test Exception 3.10.2.

ND

- APL is the minimum allowable power level for base load operation and will be provided in the Peaking Factor Limit Report per Specification 6.9.1.9.

McGUIRE - UNITS 1 and 2 3/4 2-1 Amendment No.43(Unit 1)

Amendment No.24(Unit 2)

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POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 1

4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE execre channel:

-1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitoring Alarm to OPERABLE status, b.

Monitoring and logging the indicated AFD for each OPERABLE excore j

channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable.

The logged values of the indicated AFD shall be assumed to exist duri'ng the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits when at least two OPERABLE excore channels are indicating-the AFD to be outside the limits.

4.2.1.3 When in Base L'oad operation, the target axial flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

i 4.2.1.4 When in Base Load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining l

the target flux difference in conjunction with the surveillance requirements of Specification 3/4.2.2 or by linear interpolatidn between the most recently measured value and the calculated value at the end of cycle life.

The provisions of Specification 4.0.4 are not applicable.

l McGUIRE - UNITS 1 and 2 3/4 2-la Amendment NoA2 (Unit 1) i Mai9 niRJMSUUhfifd6

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AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (w-ir2).:-

McGUIRE - UNITS 1 and 2 3/4 2-4 Amendment No.

(Unit 1)

A :endment No.

(Unit 2)

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FD A / ~ 4 A4M rro~ c~ 7 3/4.2 POWER DISTRIBUTION LIMITS BASES e

The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop-erties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

[.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

D F (Z)

Heat Flux Hot-Channel Factor, is defined as the maximum local E

O heat flux on the surface of a fuel rod at core elevation Z divided

[

by the average fuel rod heat flux, allowing for manufacturing toler-a; ances on fuel pellets and rods; b

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the integral of linear power along the rod with the highest integrated a

z power to the average rod power.

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h bound envelope of 2.26 times the normalized axial peaking factor is not exceeded p

during either normal operation or in the event of xenon redistribution following power changes.

a a.

f Target flux difference is determined at equilibrium xenon conditions.

y The full-length rods may be positioned within the core in accordance with ci their respective insertion limits and should be inserted near their normal h

position for steady-state operation at high power levels.

The value of the g

target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

f+

E.

5.

McGUIRE - UNITS 1 and 2 8 3/4 2-1 Amendment No. 43 (Unit 1)

Amendment No. 24 (Unit 2)

No CHAns n f M.

In%mpw POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)

ND At power levels below APL, the limits on AFD are defined by Figures 3.2-1, i.e. that defined by the RAOC operating procedure and limits.

These limits were calculated in a manner such that expected operational transients, e.g. load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of t'me allowed outside of the limits at reduced power levels will not result in s.]nificantxenonredistributionsuchthattheenvelopeofpeakingfactor$D would change sufficiently to prevent operation in the vicinity of 6ne APL power level.

ND At power levels greater than APL, two modes of operation are permissible; d

1) RAOC, the AFD limit of which are defined by Figure 3.2-1, and 2) Base Load operation, which is defined as the maintenance of the AFD within a t 5% band ND about a target value.

The RAOC operating procedure above APL is the same as D

that defined for operation below APL However, it is possible when following extended load following maneuvers that the AFD limits m y result in restrictions in the maximum allowed power or AFD in order to guarantee operation with F (z) q less than its limiting value.

To allow operation at the maximum permissible value, the Base Load operating procedure restricts the indicated AFD to relatively sma11g ND BL target band and power swings (AFD target band of 15%, APL 1 power 1 APL or i

l 100% Rated Thermal Power, whichever is lower).

For Base Load operation, it is expected that the plant will operate within the target band.

Operation outside of the target band for the short time period allowed will not result in signi-ficant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon redistribution impact from past operation on the Base Load operation, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months waiting period at a power level 0

above APL and allowed by RAOC is necessary.

During this time period load changes and rod motion are restricted to that allowed by the Base Load pro-cedure. After the waiting period extended Base Load operation is permissible.

The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are:

1) outside the allowed AI power operating space (for RAOC operation), or 2) outside the allowed AI target band (for Base Load operation).

These alarms are active when power is greater than:

1)'50% of RATED THERMAL POWER (for RAOC operation),

or 2) APL D (for Base Load operation).

Penalty deviation minutes for Base Load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

l 7

i McGUIRE - UNITS 1 and 2 B 3/4 2-2 Amendment No. 43(Unit 1)

Amendment No. 24(Unit 2) w rnt nu.- v i o

I ATTACHMENT 2 JUSTIFICATION AND SAFETY ANALYSIS Mr. H.B. Tucker's (DPC) November 14, 1983 letter to Mr. H.R. Denton (NRC/0NRR) described planned changes in the fuel design for McGuire Nuclear Station, Units 1 and 2.

McGuire Unit I had been operating with a Westinghouse 17x17 low parasitic (STD) fueled core.

It was planned to refuel Unit I with Westinghouse 17x17 Reconstitutable Optimized Fuel Assembly.(OFA) regions. As a result, future core loadings would range from an approximately 1/3 0FA - 2/3 STD transition core to eventually an all 0FA fueled core. Major advantages for utilizing the OFA are:

(1) increased efficiency of the core by reducing the amount of parasitic material and (2) reduced fuel cycle costs due to an optimization of the water to uranium ratio. This letter provided a Reference Safety Evaluation Report summarizing the evaluation / analysis performed on the region-by-region reload transition from the McGuire Units 1 and 2 STD fueled cores to cores with all optimized fuel. The report examined the differences between the Westinghouse OFA and STD designs and evaluated the effects of these differences for the transition to an all 0FA core. The evaluation considered the standard reload design methods described in WCAP-9272 and 9273, " Westinghouse Reload Safety Evaluation Methodology," and the transition effects described for mixed cores in Chapter 18 of WCAP-9500-A,

" Reference Core Report - 17x17 Optimized Fuel Assembly." Consistent with the Westinghouse STD reload methodology for analyzing cycle specific reloads, parameters were chosen to maximize the applicability of the transition evaluations for each reload cycle and to facilitate subsequent determination of the applicability of 10 CFR 50.59.

Subsequent cycle specific reload safety evaluations were to verify that applicable safety limits are satisfied based on the reference evaluation / analyses established in the reference report. A summary of the mechanical, nuclear, thermal and hydraulic, and accident evaluations for the McGuire Units 1 and 2 transitions to an all 0FA core were given in the reference report.

The results of evaluation / analysis and tests described in the Reference Safety Evaluation Report led to the following conclusions:

a.

The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are mechanically compatible with the STD design, control rods, and reactor internals interfaces. Both fuel assemblies-satisfy the design bases for the McGuire units, b.

Changes in the nuclear characteristics due to the transition from STD to 0FA fuel will be within the range normally seen from cycle to cycle due to fuel management ef fects.

The reload 0FAs are hydraulically compatible with the STD design.

c.

d.

The accident analyses for the OFA transition core were shown to provide acceptable results by meeting the applicable criteria, such as, minimum DNBR, peak pressure, and peak clad temperature, as required. The previously reviewed and licensed safety limits were met.

Analyses in support of this safety evaluation establish a reference design on which subsequent reload safety evaluations involving 0FA reloads can be based.

(Attachment 2A of H.B. Tucker's December 12, 1983 Unit 1/ Cycle 2 OFA reload submittal presented those

Page 2 detailed non-LOCA and LOCA accident analyses of the McGuire Units 1 and 2 FSAR impacted by the changes as determined in Section 6.0 of the Reference Safety Evaluation Report).

e.

Plant operating limitations given in the Technical Specifications affected by use of the OFA design and positive MTC would be satisfied with the changes noted in Section 7.0 of the report.

McGuire Unit 1 is currently operating in Cycle 3 (including a power coastdown not part of the original cycle design which was evalcated and determined permissable under the provisions of 10CFR 50.59) with Westinghouse 17x17 low parasitic (STD) fuel assemblies and optimized fuel assemblies (OFA) following previous NRC approval of two 0FA reload regions (reference Ms. E.G. Adensam's (NRC/0NRR) April 20, 1984 and May 15, 1985 letters to H.B. Tucker), with the third such 0FA region scheduled for the upcoming Cycle 4 refueling. Subsequent McGuire Unit I cycles are also planned to be refueled with Westinghouse 17x17 0FA's.

(McGuire Unit 2 is currently in its cycle 3 refueling outage preparing for operation with its second 0FA reload region. This reload (McGuire Unit 2/ Cycle 3) is being accomplished under the provisions of 10 CFR 50.59 as indicated in a Tucker to Denton letter of February 21, 1986, with the first OFA reload having received NRC approval via Adensam to Tucker letter dated March 22, 1985). A is the cycle-specific Reload Safety Evaluation (RSE) for McGuire Unit 1/ Cycle 4.

The RSE presents an evaluation for McGuire Unit 1, Cycle 4, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was performed utilizing the methodology described in WCAP-9273, " Westinghouse Reload Safety Evaluation Methodology". As indicated above, the NRC has previously approved similar OFA reloads for McGuire Unit 1 (and 2).

The November 14, 1983 0FA transition reference safety evaluation licensing submittal (approved by the NRC) justifying the compatibility of the OFA design with the STD design in a transition core as well as a full 0FA core contained mechanical, nuclear, thermal-hydraulic, and accident evaluations which are applicable to the Cycle 4 safety evaluation.

All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 4 design.

The results of new analyses and the justification for the applicability of previous results for the remaining analyses are addressed in the cycle specific reload safety evaluation.

As in Cycles 2 and 3, this cycle will contain one region 4 demonstration assembly of an intermediate flow mixer grid fuel assembly design. This demonstration assembly has been previously discussed in Mr. H.B. Tucker's February 20, 1984 letter to Mr. H.R. Denton, in which it was concluded that the demonstration program could be implemented per the requirements of 10CFR 50.59.

This assembly will be loaded into the core in a manner which satisfies the requirements of the " Safety Evaluation for the intermediate flow mixer grid (IFM) demonstration fuel assembly in McGuire Unit 1" (Davidson, S.L. (Ed.), February 1984). During the Cycle 2/3 refueling a problem was encountered during routine inspection of removable fuel rods in this demonstration assembly, and one removable rod was not reinserted

Page 3 because of-mechanical interference (operation of-Unit 1/ Cycle 3 with one fuel. rod of this assembly removed was evaluated under the provisions of 10 CFR 50.59 and determined permissible). The safety impact for a rod removed with a water hole remaining is addressed in the Cycle 4 reload safety evaluation.

From the evaluation presented in the Cycle 4 Reload Safety Evaluation, it concluded that the Cycle 4 design does not cause the previously acceptable safety limits to be exceeded. This conclusion is based on the following:

1.

Cycle 3 burnup is.between 11000 and 12127 MWD /MTU.

2.

Cycle 4 burnup is limited to 13100 MWD /MTU including a coastdown.

3.

There is adherence to all plant operating limitations given in the Technical Specifications as revised by the proposed changes given in Appendix A of the Cycle 4 RSE.

To ensure plant. operation consistent with the design and safety evaluation conclusion statements made in the Cycle 4 RSE and to ensure that these conclusions remain valid, Technical Specification changes will be needed for Cycle 4 to incorporate RAOC and a positive moderator temperature coefficient. These changes (presented in Appendix A of the cycle-specific RSE) are discussed in the cycle-specific RSE, along with any necessa'ry.

justifications. The McGuire Unit 1 Cycle 4 reload design has been performed assuming an increase in the low power beginning of cycle moderator temperature coefficient (MTC) to +7 pcm/ degrees F (the MTC technical specification 1imit is raised to +7 pcm/ degrees F up to 70% rated

~

thermal power and ramped to O pcm/ degrees F at 100% RTP). The unit could startup and operate with the currently approved Technical Specification limit (a " step" of +5 to O pcm/ degrees F at 70% power) but the probability of entering the action statement requiring interim rod position limits and potential delays in the cycle startup process and a special report to the Commission would be increased.- Attachment 2B is the safety evaluation report presenting the evaluations and. analyses performed verifying the acceptability of operation of McGuire Unit 1 (and 2) with this increased positive MTC limit of +7.pcm/ degrees F, including a description of the proposed change's impact on the FSAR Chapter 15 transients (this report is identified'as reference no. 10 in the McGuire Unit 1/ Cycle 4 RSE). The McGuire FSAR will be revised accordingly in the appropriate annual FSAR update following approval of this change. The revised Unit 1 RAOC envelope is based upon the Fq limit of 2.26.

The McGuire Unit 1.Fq limit was increased from 2.15 to 2.26 in submittals associated with the Unit 1/ Cycle 3 reload (and for Unit 2 with the Unit 2/ Cycle 2 reload). However, the RAOC AFD envelope continued.to be based upon the 2.15 FQ limit until this cycle's RAOC analysis, although the envelope could have been expanded along with the previous limit increase. The revision to the Unit 1 RAOC AFD envelope is simply taking credit for the previously approved higher FQ limit and brings the nuclear design and LOCA analysis assumption into agreement (Note: Although credit for the 2.26 Fq limit currently in effect on Unit ~2 has not been taken in Unit 2 RAOC analyses performed to date (including the recent Unit 2/ Cycle 3 analysis as reflected in the Tucker to Denton letter of February 21, 1986), Duke anticipates similarly expanding

Page 4 the Unit 2 RAOC envelope in the near future to take credit for the higher limit). provides copies of the McGuire Units 1 and 2 Technical Specifications with the appropriate Unit 1/ Cycle 4 changes indicated. Note that although the changes given in Appendix A of the Unit 1/ Cycle 4 RSE are intended to apply only to Unit 1, Appendix A does not always reflect the fact that a change would also apply to Unit 2 unless specifically indicated otherwise since the McGuire Unit I and 2 Technical Specifications are combined into one document - this has been accounted for in Attachment 1.

Consequently, the McGuire Unit 2 specifications are administrative 1y affected in that a specification currently applying to both McGuire Units 1 and 2 is split into two portions addressing the separate requirements for Units 1 and 2 created by these changes.

In addition, although intended to apply only for Unit i via being in Appendix A, note that the increased positive MTC limit of +7 pcm/ degrees F is also being requested for McGuire Unit 2 because of its desirable effects on fuel cycle flexibility (this is justified by Attachment 2B for Unit 2 as well as Unit 1),

Although the increase in the positive MTC limit was assumed in the McGuire 1/ Cycle 4 RSE, the change is not a consequence of the reload and therefore is acceptable for immediate implementation on Unit 2.

Note that while this increase was not assumed in the McGuire Unit 2/ Cycle 3 RSE, it will be needed (assumed) for future reloads and therefore should be approved on Unit 2 sometime prior to McGuire 2/ Cycle 4.

The McGuire FSAR update will accordingly reflect this change for Unit 2 as well as unit 2.

No revisions to the Technical Specification Bases are required by the changes.

The Peaking Factor Limit Report for McGuire Unit 1/ Cycle 4 which will be submitted in accordance with Technical Specification 6.9.1.9 provides the elevation dependent W(z) values that are to be used as inputs to define the appropriate fitting coefficients for W(z) interpolations to be performed as a function of cycle burnup and axial elevation for RAOC and Base Load Operation during Cycle 4, and the value for APLND. The appropriate W(z) function is used to confirm that the Heat Flux Hot Channel Factor, FQ(z),

will be limited to the values specified in the Technical Specitications.

The peaking factor report to be submitted for Unit 1 Cycle 4 is based upon the proposed revision to the RAOC AFD envelope.

If the revised RAOC limits are not approved in time to support the Unit 1 Cycle 4 startup, confirmation of the validity of the W(z) functions with respect to the existing RAOC AFD envelope or generation of new W(z) functions would be required. An exemption to the 60 days prior to criticality submittal schedule for the peaking factor report may be necessary if revised W(z) functions are required. The use of overly conservative W(z) functions increases the probability of entering the action statement and could lead to very restrictive AFD limits and/or reductions in reactor power.

ATTACHMENT 2A RELOAD SAFETY EVALUATION MCGUIRE NUCLEAR STATION UNIT 1 CYCLE 4 April, 1986 Edited by:

B. W. Gergos Contributors:

M. A. Colchaser D. S. Huegel P. J. Larouere J. R. Lesko P. Schueren C. M. Thompson Approved:

CG d w4 E.A.Dzenis,Mandr Core Operatior)s Nuclear Fuel Division ev n e-eoo m

TABLE OF CONTENTS Title be

1.0 INTRODUCTION

AND

SUMMARY

1 1.1 Introduction 1

- 1. ;2 General Description 2

1.3 Conclusions 2

2.0 REACTOR DESIGN 3

2.1 Mechanical Design 3

2.2 Nuclear Design 3

2.3 Thermal and Hydraulic Design 5

l 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 6

3.1 Power Capability 6

3.2 Accident Evaluation 6

3.2.1 Kinetic Parameters 7

3.2.2 Control Rod Worths 7

3.2.3 Core Peaking Factors 7

4.0 TECHNICAL SPECIFICATION CHANGES 8

5.0 REFERENCES

9 APPENDIX A - Technical Specification Page Changes 4

m.t..

m j

,u.._.._._.._

LIST OF TABLES Table Title P_ age a

1 Fuel Assembly Design Parameters 10 2

Kinetic Characteristics 11 3

Shutdown Requirements and Margins 12 4

Control Rod Ejection Accident Parameters 13 LIST OF FIGURES Figure Title h

1 Core Loading Pattern and Source and 14 Burnable Absorber Locations l

f-

.im e-eemi jj

1.0 INTRODUCTION

AND

SUMMARY

1.1 INTRODUCTION

This report presents an evaluation for McGuire Unit 1, Cycle 4, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was performed utilizing the methodology described in WCAP-9273, " Westinghouse Reload Safety Evaluation Methodology"(1).

McGuire Unit 1 is operating in Cycle 3 with Westinghouse 17x17 low parasitic (STD) and optimized fuel assemblies (OFA). For Cycle 4 and subsequent cycles, it is planned to refuel the McGuire Unit 1 core with Westinghouse 17x17 optimized fuel assemblies.

In the OFA transition licensing submittal (2) to the NRC, approval was requested for the transition from the STD fuel design to the OFA design and the assc.iated proposed changes to the McGuire Units 1 and 2 Technical Specifications. The licensing submittal, which has received NRC approval, justifies the compatibility of the OFA design with the STD design in a transition core as well as a full 0FA core. The OFA transition licensing submittal (2) contains mechanical, nuclear, thermal-hydraulic, and accident evaluations which are applicable to the Cycle 4 safety evaluation.

All of the accidents comprising the licensing bases (2,3) which could potentially be affected by the fuel reload have been reviewed for the Cycle 4 design described herein.

The results of new analyses and the justification for the applicability of previous results for the remaining analyses are addressed in safety evaluations for a Positive Moderator Coefficient (10) and the UHI Elimination (14) licensing submittals.

During the cycle 2/3 refueling a problem was encountered in assembly ZV-1.

One removable rod was not reinserted because of mechanical interference. This assembly will remain in the core for Cyc4 4.

The safety impact for a rod removed with a water hole remaining is presented in Reference 4.

[

.mu-ws i

1.2 GENERAL DESCRIPTION The McGuire Unit 1, Cycle 4 reactor core will be comprised of 193 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1.

During the Cycle 3/4 refueling, 64 STD fuel assemblies will be replaced with 64 Region 6 optimized fuel assemblies. A summary of the Cycle 4 fuel inventory is given in Table 1.

As in Cycles 2 and 3, this cycle will contain one Region 4 demonstration assembly, designated in Figure 1 as 4A, of an intermediate flow mixer grid fuel assembly design. This assembly will be loaded into the core in a manner which satisfies the requirements given in Reference 13.

Nominal core design parameters utilized for Cycle 4 are as follows:

Core Power (MWt) 3411 SystemPressure(psia) 2250 Core Inlet Temperature ('F) 558.9 Thermal Design Flow (gpm) 382,000 Average Linear Power Density (kw/ft) 5.43 (based on 144" active fuel length)

1.3 CONCLUSION

S From the evaluation presented in this report, it is concluded that the Cycle 4 design does not cause the previously acceptable safety limits to be exceeded.

This conclusion is based on the following:

1.

Cycle 3 burnup is between 11000 and 12127 MWD /MTU.

2.

Cycle 4 burnup is limited to 13100 MWD /MTU including a coastdown.

3.

There is adherence to plant operating limitations in the Technical Specifications.

4.

The proposed Technical Specification changes discussed in Section 4.0 of this report and provided in Appendix A are approved.

.mee-e.o'"

2

2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The Region 6 fuel assemblies are Westinghouse 0FAs. The mechanical description and justification of their compatibility with the Westinghouse STD design in a transition core is presented in the OFA transition licensing submittal.(2)

Table 1 presents a comparison of pertinent design parameters of the various fuel regions. The Region 6 fuel has been designed according to the fuel performance model(5)

The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse clad flattening I6)

For all fuel regions, the fuel rod internal pressure design model basis, which is discussed and shown acceptable in Reference 7, is satisfied.

Westinghouse has had considerable experience with Zircaloy clad fuel. This experience is described in WCAP-8183, " Operational Experience with Westinghouse Cores."(0) Operating experie.ce for Zircaloy grids has also been obtained from six demonstration 17x17 0FAs(2), four demonstration 14x14 0FAs(2) and two regions of 0FA fuel in the McGuire Unit 1 Cycle 2 and 3 designs.

2.2 NUCLEAR DESIGN The Cycle 4 core loading is designed to meet a F (z) x P ECCS limit of Q

1 2.26 x K(z).

In the event of UHI elimination, the F (z) x P ECCS limit 0

of 5 2.26* x K(z) will remain applicable to the Cycle 4 design.

Relaxed Axial Offset Control (RAOC) will be employed in Cycle 4 to enhance operational flexibility during non-steady state operation. RA0C makes use of available margin by expanding the allowable al band, particularly at reduced

Based on the LOCA analyses performed in support of the UHI elimination II4) licensing submittal

.u.c.

is 3

power. The RAOC methodology and application is fully described in Reference 9.

The-analysis for Cycle 4 indicates that no change to the safety parameters is required for RAOC operation. During operation at or near steady state equilibrium conditions, core peaking factors are significantly reduced due to the limited amount of xenon skewing possible under these operating conditions. The Cycle 4 Technical Specifications recognize this reduction in core peaking factors through the use of a Base Load Technical Specification.

Adherence to the F limit is obtained by using the F Surveillance g

g Technical Specification, also described in Reference 9.

This provides a more convenient form of assuring plant operation below the F limit while 0

retaining the intent of using a measured parameter to verify operation below Technical Specification limits. F surveillance is only a surveillance g

requirement and as such has no impact on the results of the Cycle 4 analysis or safety parameters.

Table 2 provides a summary of Cycle 4 kinetics characteristics compared with the current limits based on previously submitted accident analyses.

Table 3 provides the control rod worths and requirements at the most limiting condition during the cycle (end-of-life) for the standard burnable absorber design. The required shutdown margin is based on previously submitted accident analysis. The available shutdown margin exceeds the minimum required.

The loading pattern contains 320 burnable absorber (BA) rods located in 44 BA rod assemblies.

Location of the BA rods are shown in Figure 1.

A more Positive Moderator Coefficient as compared to the current value will be utilized during Cycle 4.

The safety evaluation is contained in Reference 10 and the associated Technical Specification changes are addressed in Section 4.0 of this report.

. m e-is 4

~

2.3 THERMAL AND HYDRAULIC DESIGN The thermal hydraulic methodology, DNBR correlation and core DNB limits used for Cycle 4, are consistent with the current licensing basis (2)

The thermal hydraulic safety analyses used for Cycle 4 are based on a reduced design flow rate (15) in comparison to Reference 2.

No significant variations in thermal margins will result from the Cycle 4 reload.

The thermal-hydraulic methods used to analyze axial power distributions generated by the RAOC methodology are similar to those used in the Constant Axial Offset Control (CAOC) methodology. Normal operation power distributions are evaluated relative to the assumed limiting normal operation power distribution used in the accident analysis. Limits on allowable ooerating axial flux difference as a function of power level from these considerations were found to be less restrictive than those resulting from LOCA Fg considerations.

l The Condition II analyses were evaluated relative to the axial power distribution assumptions used to generate DNB core limits and resultant Overtemperature Delta-T setpoints (including the f(al) function). No changes in the DNB core limits are required for RAOC operation.

.i m...**

5

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability has been evaluated considering the consequences of those incidents examined in the FSAR(3) using the previously accepted design basis.

It is concluded that the core reload will not adversely affect the ability to safely operate at the design power level (Section 1.0) during Cycle 4.

For the overpower transient, the fuel centerline temperature limit of 4700*F can be accommodated with margin in the Cycle 4 core. The time dependent densification model(11) was used for fuel temperature evaluations.

The LOCA limit at rated power can be met by maintaining F (*)

0 at or below 2.26 x K(z).

3.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the FSAR(3) were examined.

In all cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in 1) the previous applicable safety analysis, 2) the safety evaluation performed in support of the positive moderator coefficient (+7 pcm/*F) licensing submittal (10) or 3) the safety evaluation performed in support of the UHI Elimination licensing submittal (14),

i A core reload can typically affect accident analysis input parameters in the following areas: core kinetic characteristics, control rod worths, and core i

peaking factors. Cycle 4 parameters in each of these three areas were examined as discussed in the following subsections to ascertain whether new i

accident analyses were required.

[

I 6

3.2.1 KINETIC PARAMETERS Table 2 is a summary of the kinetic parameters current limits along with the associated Cycle 4 calculated values. All of the kinetic values fall within the bounds of the current limits except for the minimum moderator temperature coefficient. The safety evaluation for the Positive Moderator Coefficient is contained in Reference 10 and the associated Technical Specification changes are addressed in Section 4.0 of this report.

3.2.2 CONTROL R00 WORTHS Changes in control rod worths may affect differential rod worths, shutdown margin, ejected rod worths, and trip reactivity. Table 2 shows that the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 4 meets the current limit. Table 3 shows that the Cycle 4 shutdown margi'n requirements have been satisfied.

Table 4 is a summary of the current limit control rod ejection analysis parameters and the corresponding Cycle 4 values.

3.2.3 CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the NRC approved dropped rod methodology described in Reference 12.

Results show that DNB design basis is met for all dropped rod events initiated from full power.

The peaking factors for steamline break and control rod ejection have been evaluated and are within the bounds of the current limits.

nm s-seow 7

4.0 TECHNICAL SPECIFICATION CHANGES To ensure that plant operation is consistent with the design and safety evaluation conclusion statements made in this report and to ensure that these conclusions remain valid, Technical Specifications changes will be needed for Cycle 4 to incorporate RA0C and the Positive Moderator Temperature Coefficient. These changes are presented in Appendix A.

i 4

e.

.mt.-m.

8

5.0 REFERENCES

1. Davidson, S. L. (Ed), et. al., " Westinghouse Reload Safety Evaluation Methodology", WCAP-9272-P-A, July 1985.

2. Duke Power Company Transmittal to NRC, " Safety Evaluation for McGuire Units 1 and 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies",

December 1983.

3. "McGuire Final Safety Analysis Report."

4. " Reload Safety Evaluation McGuire - Unit 1 Cycle 3 - Revision 1," May 1985.

5. Miller, J.V., (Ed.),

Improved Analytical Model used in Westinghouse Fuel Rod Design Computations", WCAP-8785, October 1976.

6. George, R. A., (et. al.), " Revised Clad Flattening Model", WCAP-8381, July 1974.

7. Risher, D. H., (et. al.), " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964, June 1977.

8. Skaritka, J.,

lorii, J.A., " Operational Experience with Westinghouse Cores", WCAP-8183, Revision 14, July,1985.

9. Miller, R. W., (et al.'), " Relaxation of Constant Axial Offset Control-F g

Surveillance Technical Specification," WCAP-10217-A, June 1983

10. Westingbouse Transmittal to Duke Power Company, " Safety Evaluation for Operation of McGuire Units 1 and 2 with a Positive Moderator Coefficient",

January 1986.

11. Hellman, J.M. (Ed.), " Fuel Densification Experimental Results and Model for Reactor Operation", WCAP-8219-A, March 1975.

12. Morita, T., Osborne, M. P., et. al., " Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10297-P-A (Proprietary) and WCAP-10298-A (Non Proprietary), June 1983.

13. Davidson, S.

L., (Ed.), " Safety Evaluation for the Intermediate Flow Mixer Grid (IFM) Demonstration Fuel Assembly in McGuire Unit 1", February 1984.

14. Duke Power Company Transmittal to NRC, "McGuire Nuclear Station Safety Analyses for UHI Elimination", March 1986.

15. Duke Power Company Transmittal to NRC," McGuire 2 Cycle 2 0FA Reload",

November 1984.

~

estat e-seonis g

4

,---m--.

-r-7--m__ - - - --

TABLE 1 MCGUIRE UNIT 1 - CYCLE 4

~

FUEL ASSEMBLY DESIGN PARAMETERS Region 1

4*

5*

6A*

68*

Enrichment (w/o U-235)+

2.108 3.205 3.204 3.20 3.40 Density (% Theoretical)+

94.53 95.04 95.05 95.0 95.0 Number of Assemblies 9

60 60 12 52 Approximate Burnup at++

16942#

20027 14604 0

0 Beginning of Cycle 4 (MWD /MTU)

Approximate Burnup at++

28540#

33731 25240 16745 14617 End of Cycle 4 (MWD /MTV)

Optimized Fuel - Zire grid

+ All fuel region values are as-built except Region 6 values which are nominal.

++ Based on EOC3 = 11560 MWD /MTU, EOC4 = 13100 MWD /MTV (coastdown included)

The burnups noted are for the Region 1 fuel, assemblies being used and are not an average for the whole region.

.s m e-o.ie 10

TABLE 2

~

MCGUIRE UNIT 1 - CYCLE 4 KINETICS CHARACTERISTICS Cycle 4 Current Limits Design Minimum Moderator

+5 < 70% of RTP

+7 <70% of RTP Temperature Coefficient 0 > 70% of RTP

+7 ramp to 0 from 70% to 100% of RTP (pcm/*F)*

Doppler Temperature

-2.9 to -0.91

-2.9 to -0.91 Coefficient (pcm/*F)*

Least Negative Doppler-

-9.55 to -6.05

-9.55 to -6.05 Only Power Coefficient, Zero to Full Power, (pcm/% power)*

Most Negative Doppler

-19.4 to -12.6

-19.4 to -12.6 Only Power Coefficient, Zero to Full Power (pcm/%

power)*

Minimum Delayed Neutron

.44

>.44 Fraction 8,ff, (%)

Minimum Delayed Neutron

.50

>.50 Fraction B

(%)

(EjectedR8bf,t80L]

a Maximum Differential Rod 100

<100 Worth of Two Banks Moving Together(pcm/in)*

-5

pcm = 10 3,

.mt.-

oci 11

TABLE 3 END-0F-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MCGUIRE UNIT 1 - CYCLE 4 Control Rod Worth (%Ap)

Cycle 3 Cycle 4 All Rods Inserted 6.72 6.95 All Rods Inserted Less Worst Stuck Rod 5.90 5.95 (1) Less 10%

5.32 5.35 Control Rod Requirements (%ao)

Reactivity Defects (Doppler, T"V9, 3.18 3.39 Void, Redistribution)

Rod Insertion Allowance 0.50 0.50 (2) Total Requirements 3.68 3.89 Shutdown Margin ((1) - (2)] (%Ao) 1.64 1.46 Required Shutdown Margin (%ao) 1.30 1.30 e

4mt e-em.is 12

q TABLE 4 MCGUIRE UNIT 1 - CYCLE 4 CONTROL R00 EJECTION ACCIDENT PARAMETERS HZP-BOC Current Limit

Cycle 4 Maximum ejected rod 0.75

<0.75 worth, %Ap Maximum Fg (ejected) 11.0

<11.0 HFP-BOC Maximum ejected rod 0.23

<0.23 worth, %Ap Maximum Fg (ejected) 4.5

<4.5 HZP-EOC Maximum ejected rod 0.90

<0.90 worth, %ap Maximum Fg (ejected) 20.0

<20.0 HFP-E0C Maximum ejected rod 0.23

<0.23 wortn, %Ap Maximum F0 (ejected) 5.9

<5.9

Based on the safety evaluation performed in support of the Positive Moderator Coefficient licensing submittal (10),

.im e-em.

13

FIGURE 1 CORE LOADING PATTERN McGUIRE UNIT 1, CYCLE 4 l

180' R

P N

M L

K J

H G

F E

D C

B A

5 6B 5

6B 5

1 5

5 6B 4

6B 4

6B 5

5 4

4 5

5 6B 4

6B 4

5 4

6B 4

6B 5

5 3

4 8

SS 8

4 5

6B 4

5 6A 5

6A 5

4 6B 5

g y

4 8

8 4

5 68 4

5 4

6B 4

4 4

6B 4

5 4

6B 5

5 12 12 6B 4

6B 6B 4

5 4

6B 6B 4

6B 8

12 12 8

5 6B 4

6A 4

5 4

6A 4

5 4

6A 4

6B 5

7 4

8 8

8 4

6B 4

5 5

4A 4

6A 6A 4

4 5

5 4

6B 90*

8 270 i

8 8

5 6B 4

6A 4

5 4

6A 4

5 4

6A 4

6B 5

9 4

8 8

8 4

6B 4

6B 6B 4

5 4

6B 6B 4

6B 10 1

1 8

12 12 8

5 6B 4

5 4

6B 4

4 4

6B 4

5 4

6B 5

11 12 1'

5 6B 4

5 6A 5

6A 5

4 6B 5

l 1

12 4

8 8

4 5

5 6B 4

6B 4

5 4

6B 4

6B 5

5 13 4

8 SS 8

4 5

5 6B 4

6B 4

6B 5

5 14 4

4 5

6B 5

15 0*

X REGION NUMBER

Demonstration Assembly with Rod Removed Y

BA's SS SECONDARY SOURCE 14

APPENDIX A TECHNICAL SPECIFICATION PAGE CHANGES Modifications to Pages:

3/4 1-Sa 3/4 2-4 asist e-secare

)

= -.-

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be:

LesspositivethanthelimitsshowninFigure3.1-0,and a.

Less negative than -4.1 x 10

delta k/k/ F for the all rods withdrawn, b.

and of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specifications 3.1.1.3a. - MODES 1 and 2* only.#

Specification 3.1.1.3b. - MODES 1, 2, and 3 only.#

ACTION:

With the MTC imre positive than the limit of Specification 3.1.1.3a.

a.

above, operation in MODES 1 and 2 may. proceed provided:

1.

Control rod wiJhdrawal limits are established and maintained sufficient to restore the MTC to less positive than the limits shown in. Figure 3.1-0 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal ifmits established above until a subsequent calculation verifies that

~

the MTC has been restored to within its limit for the all rods withdrawn condition; and 3.

A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition, b.

With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTOOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K,ff greater than or equal to 1.0.

See Special Test Exception 3.10.3.

~

\\

McGUIRE - UNITS 1 and 2 3/4 1-4 Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2)

..m_,-,%

-.,,---r.

,.w_-,v

,-.m,,,

,.-wr-,,.e-c____.,_.,,

REACTIVITY CONTROL SYSTEMS

(

SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

The MTC shall be measured and compared to the BOL limit of a.

' Spec'ification 3.1.1.3a., above, prior to initial operation'above 5%

l of RATED THERMAL POWER, after each fuel loading; and b.

The MTC shall be measured at any THERMAL POWER and compared to

-3.2 x 10 4 delta k/k/'F (all rods withdrawn RATED THERMAL POWER condition) within 7 EFPO after reaching an eq,uilibrium baron concentration of 300 ppe.

In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 delta k/k/'F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specifica-tion 3.1.1.3b.., at least once per 14 EFPD during the remainder,of the l

fuel cycle.

(

McGUIRE - UNITS 1 and 2 3/4 1-5 Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2)

I 1.0' w

b 0.9 ss T

0.8

-f 0.7

&cceptable Unacceotable v

C 0.6 Operation Oceration 8

0.5 E

3 j

0.4 y

E 0.3

~

5 1

1 0.2 k

E 0.1

.l 0

10 20 30 40 50 60

70 30 90 100

% of Rated Thermal Power FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL 3/4 1-Sa

i 3

12C 110

(-20,100)

(10,100)

~~

~-

100

]

i LMACCEPTABLE L

90 l

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( 'A,50 )

(21.50)

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~50

-40

-30

-10

-10 0

10 20 30 40 50 Axial Flux Difference (% Delta-1)

FIGURE 3.2-l AFD Limits as a Function of Rated Thermal Power McGuire Unit 1 Cycle 4

1 I

ATTACHENT 2B SAFETY EVALUATION FOR OPERATION OF Ms,UIRE UNITS 1 & 2 WITH A POSITIVE 10DERA10R CODTICIENT 6

.,,,------v,-.

,,,.,,y.,

,n

-, - - - -,a,n..

n-

SAFETY EVALUATION FOR OPERATION OF McGUIRE UNITS 1 & 2 WITH A POSITIVE MODERATOR COEFFICIENT P. N. Kirby G. H. Heberle January 1986 Approved:

M Tid M. P. Osborne, Manager Transient Analysis II WESTINGHOUSE ELECTRIC CORPORATION Water Reactor Divisions P. O. Box 355 Pittsburgh, Pennsylvania 15230 9124Q 10/0DD003

CONTENTS 1

~

1.

INTRODUCTION 3

2.

ACCIDENT EVALUATIONS 4

I.

Introduction 4

II. Transients Not Affected By a Positive Moderator Coefficient 5

A.

RCCA Misoperation 5

8.

Startup of an Inactive Reactor Coolant Loop 5

C.

Excessive Heat Removal Due to Feedwater System Malfunctions 6

D.

Excessive Load Increase 6

E.

Spurious Actuation of Safety Injection 6

F.

Main Steam Line Depressurization/ Rupture of a Main Steam Pipe 7

G.

Loss of Coolant Accident (LOCA) 7 III. Transients Sensitive to a Positive Moderator Coefficient 7

A.

Boron Dilution 7

B.

Control Rod Bank Withdrawal From a Subcritical Condition 8

C.

Uncontrolled Control Rod Bank Withdrawal at Power 9

D.

Loss of Reactor Coolant Flow 10 E.

Locked Rotor 11 F.

Turbine Trip 12 G.

Loss of Normal Feedwater/ Loss of Offsite Power 14 H.

Rupture of a Main Feedwater Pipe 15 I.

Control Rod Ejection 16 J.

Accidental Depressurization of the Reactor Coolant System 17 3.

CONCLUSIONS 19 Figure 1 20 Table 1 21 9124Q:10/020386 2

SECTION 1 INTRODUCTION Safety analyses and evaluations have been performed to support the proposed Technical Specification change for McGuire Units 1 & 2 which would allow a more positive moderator temperature coefficient to exist during power operation. The results of the analysis, which are presented in the following section, show that the proposed change can be acconnedated with margin to applicable FSAR safety limits.

The present McGuire Technical Specifications require the moderator temperature coefficient (MTC) to be +5 pcm/*F* or less at all times while the reactor is critical. A positive coefficient at reduced power levels results in a significant increase in fuel cycle flexibility, but has only a minor effect on the safety analysis of the accident events presented in the FSAR.

The proposed Technical Specification change would allow a +7 pcm/*F MTC below 70 percent of rated power, ramping down to O pcm/'F at 100 percent power. This MTC is diagrammed in Figure 1.

A power-level dependent MTC was chosen to minimize the ef fect of the specification on postulated accidents at high power levels.

Moreover, as the power level is raised, the average core water temperature becomes higher as allowed by the programmed average temperature for the plant, tending to bring the moderator coefficient more negative. Also, the boron concentration can be reduced as xenon builds into the core. Thus, there is less need to allow a positive coefficient as full power is approached. As fuel burnup is achieved, boron is further reduced and the moderator coefficient will become negative over the entire operating power range.

-5

1 pcm = 10 ak/k i

9124Q:10/011486 3

SECTION 2 ACCIDENT EVALUATIONS 1.

Introduction i

The impact of a positive moderator temperature coefficient on the accident analyses presented in Chapter 15 of the McGuire Units 1 &

2 Final Safety Analysis Report (FSAR) has been assessed. Those

~

incidents which were found to be sensitive to positive or minimum moderator coefficients were reanalyzed.

In general, these incidents are limited to transients which cause reactor coolant temperature to increase. The analyses presented herein were based on a +7 pcm/*F moderator temperature coef ficient, which was assumed to remain constant for variations in temperature.

The control rod ejection and rod withdrawal from subcritical analyses were based on a coefficient which was at least +7 pcm/'F at zero power nominal average temperature, and which became less positive for higher temperatures. This was necessary since the TWINKLE computer code, on which the analyses are based, is a diffusion-theory code rather than a point-kinetics approximation and the moderator temperature feedback cannot be artificially held constant with temperature.

For all accidents which were reanalyzed, the assumption of a positive moderator temperature coefficient existing at full power is conservative since as diagramed in Figure 1, the proposed Technical Specification requires that the coefficient be linearly ramped to zero above 70 percent power.

In general, reanalysis was based on the identical analysis methods, computer codes, and assumptions employed in the FSAR; any exceptions are noted in the discussion of each incident. Accidents j

not reanalyzed included those resulting in excessive heat removal from the reactor coolant system for which a large negative l

moderator coefficient is conservative. Table 1 gives a list of l

9124Q:10/011486 4

l l

accidents presented in the McGuire Units 1 & 2 FSAR, and denotes those events reanalyzed for a positive coef ficient. The following sections provide discussions for each of the FSAR events.

II. Transients Not Af fected By a Pcsitive Moderator Coef ficient The following transients were not reanalyzed since they either result in a reduction in reactor coolant system temperature, and are therefore sensitive to a negative moderator temperature coefficient, or are otherwise not affected by a positive moderator temperature coefficient.

A.

RCCA MisoDeration Only the RCCA drop case presented in Section 15.4.3 of the FSAR is potentially affected by a positive moderator temperature coefficient. Use of a positive coefficient in the analysis would result in a larger reduction in core power level following the RCCA drop, thereby increasing the probability of a reactor trip. For the return to power automatic rod control case, a positive coefficient would result in a small increase in the power overshoot. Since the limiting conditions for this accident are at or near 100 percent power where the moderator temperature coefficient must be close to zero or negative, this accident is unaffected by the proposed Technical Specification and thus the analysis was not repeated.

8.

StartuD of an Inactive Reactor Coolant looD An inadvertent startup of an idle reactor coolant loop at an incorrect temperature results in a decrease in core average temperature. As the most negative values of moderator reactivity coefficient produce the greatest reactivity addition, the most limiting case is represented by the analysis reported in the FSAR, Section 15.4.4.

9124Q:10/011486 5

C.

Excessive Heat Removal Due to Feedwater System Malfunctions The addition of excessive feedwater or the reduction of feedwater temperature are excessive heat removal incidents, and are consequently most sensitive to a negative moderator temperature coefficient. Results presented in Section 15.1.1 and 15.1.2 of the FSAR, based on a negative coefficient, represent the limiting case. Therefore, this incident was not reanalyzed.

D.

Excessive load Increase An excessive load increase event, in which the steam load exceeds the core power, results in a decrease in reactor coolant system temperature. With the reactor in manual control, the analysis presented in Section 15.1.3 of the FSAR shows that the limiting case is with a large negative moderator coefficient.

If the reactor is in automatic control, the control rods are withdrawn to increase power and restore the average temperature to the programmed value. The analysis of this case in the FSAR show that the minimum DN8R is not sensitive to moderator temperature coefficient. Therefore, the results presented in the FSAR are still applicable to this incident.

E.

Sourious Actuation of Safety Injection Analysis of a spurious actuation of safety injection at power is presented in Section 15.5.1 of the FSAR. This transient results in a decrease in average coolant temperature and core power. and the results are not sensitive to moderator temperature coefficient. Therefore, this incident was not reanalyzed with a positive moderator coefficient.

9124Q:10/011486 6

F.

Main Steam Line Depressurization/Ruoture of a Main Steam Pine Since the steam line depressurization or rupture of a main steam pipe is a temperature reduction transient, minimum core shutdown margin is associated with a strong negative moderator temperature coefficient. The worst conditions for a steamline break are therefore those analyzed in the FSAR (Section 15.1.4 and 15.1.5).

G.

Loss of Coolant Accident (LOCA)

The loss of coolant accident (Section 15.6.5 of the FSAR) is analyzed to determine the core heatup consequences caused by a rupture of the reactor coolant system boundary. The event results in a depressurization of the RCS and a reactor shutdown at the beginning of the transient. This accident was not reanalyzed since the Technical Specification requirement that the temperature coef ficient be zero or negative at 100 percent power ensures that the previous analysis basis for this event is not affected.

III. Transients Sensitive to a Positive Moderator Coefficient A.

Boron Dilution i

l l

As stated in Section 15.4.6 cf the FSAR, a boron dilution incident cannot occur during refueling due to administrative controls which isolate the RCS from potential sources of diluted water.

If a boron dilution incident occurs during cold shutdown, hot standby, or startup, the operator must take action to terminate the dilution before the reactor returns l

critical. Therefore, since a return to criticality is l

prevented by the operator, the value of the moderator coefficient has no effect during a boron dilution incident in these operating modes. The reactivity addition due to a boron 9124Q:10/011486 7

l l

l

dilution at power will cause an increase in power and reactor coolant system temperature. Due to the temperature increase, a positivt moderator coefficient would add additional reactivity and increase the severity of the transient. With the reactor in automatic control, however, the rod insertion alarms provide the operator with adequate time to terminate the dilution before shutdown margin is lost. A boron dilution incident with the reactor in manual control is no more severe than a rod withdrawal at power, which is discussed in Section III.C. and therefore this case was not specifically analyzed.

Following reactor trip, the amount of time available before shutdown margin is lost is not affected by the moderator coefficient.

8.

Control Rod Bank Withdrawal From a Subcritical Condition Introduction A control rod assembly bank withdrawal incident when the reactor is suberitical results in an uncontrolled addition of reactivity leading to a power excursion (Section 15.4.1 of the FSAR). The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature coefficient. The power excursion causes a heatup of the moderator and fuel. The reactivity addition due to a positive moderator coefficient results in increases in peak heat flux and peak fuel and clad temperatures.

Method of Analysis The analysis was performed in the FSAR for a reactivity insertion rate of 75 x 10~ ak/sec. This reactivity insertion rate was used in this analysis and is greater than that for the simultaneous withdrawal of the combination of the two sequential control banks having the greatest combined worth at maximum speed (45 inches / minute). The analysis used a 9124Q:10/011486 8

moderator temperature coef ficient more conservative than a +7 pcm/*F for all appropriate temperature values.

Results and Conclusions Reanalysis of this event assuming a 75 pcm/sec insertion rate, coupled with a positive moderator temperature coefficient of +7 pcm/*F, yields a peak heat flux which does not exceed the nominal full power value.

In the event of a RCCA bank withdrawal accident from a subcritical condition, the core and the RCS are not adversely affected, since the combination of thermal power and the coolant temperature result in a ONBR greater than the limit value and thus, no fuel or clad damage is predicted. Therefore, the conclusions presented in the FSAR remain valid.

C.

Uncontrolled Control Rod Bank Withdrawal at Power Introduction An uncontrolled control rod bank withdrawal at power produces a mismatch in steam flow and core power, resulting in an increase in reactor coolant temperature. A positive moderator coefficient would increase the power mismatch resulting in a faster heatup of the reactor coolant. However, this effect is offset by the fact that the faster heatup and reactivity addition result in an earlier reactor trip from either overtemperature delta-T or high neutron flux. A discussion of th'is incident is presented in Section 15.4.2 of the FSAR.

9124Q:10/020386 9

Method of Analysis The transient was reanalyzed employing the same digital computer code and assumptions regarding instrumentation and setpoint errors used for the FSAR. This transient was analyzed at 100, 60 and 10 percent power with a positive moderator coefficient. A constant moderator coefficient of +7 pcm/*F was used in the analysis.

Results and Conclusions For each initial power level the full range of reactivity insertion rates wa,s reanalyzed. The limiting case for DN8 margin remains abo've tae limit DNBR value. These results demonstrate that the conclusions presented in the FSAR are still valid.

That is, the core and reactor coolant system are not adversely affected since nuclear flux and overtemperature AT trips prevent the core minimum DNB ratio from falling below the limit value for this incident.

D.

Loss of Reactor Coolant Flow Introduction The loss of flow events presented in FSAR Sections 15.3.1 and 15.3.2 were reanalyzed to determine the effect of a +7 pcm/*F moderator temperature coefficient on the nuclear power transient and the resultant minimum DNBR reached during the incident. The effect on the nuclear power transient would be limited to the initial stages of the incident during which re" actor coolant temperature increases; this increase is terminated shortly after reactor trip.

9124Q:10/020386 10

1 Method of Analysis l

Analysis methods and assumptions used in the reevaluation were consistent with those employed in the FSAR.

The digital computer codes used to calculate the flow coastdown and resulting system transient were the same as those used to perform the analysis described in the FSAR.

The analysis as'sumed a constant moderator coef ficient of +7 pcm/*F.

Results and Conclusions Results of the analysis show that the minimum DNBR remains above the limit value for these transients.

Therefore, the conclusions of the FSAR analyses remain valid.

E.

Locked Rotor Introduction The case presented in the FSAR (Section 15.3.3) for this transient was reanalyzed. Following a locked rotor incident, reactor coolant system temperature rises until shortly after reactor trip. A positive moderator coefficient will not affect the time to DNS since DNB is conservatively assumed to occur at the beginning of the incident. The transient was reanalyzed, however, due to the effect on the nuclear power transient and

-thus on the peak reactor coolant system pressure and fuel and clad temperatures.

Method of Analysis The digital computer codes used in the reanalysis to evaluate the pressure transient and thermal transient were the same as 9124Q:10/020386 11

those used in the FSAR. The analysis employed a constant moderator coef ficient of +7 pcm/*F. Other assumptions used were consistent with those employed in the FSAR.

Results and Conclusions Analysis of the locked rotor incident with a +7 pcm/'F moderator temperature coefficient shows that the peak reactor coolant system pressure remains below that which would cause stresses to exceed the faulted condition stress limits. The peak clad temperature for the hot spot during the worst transient remains much less than 2700'F and the amount of Zirconium - water reaction is small. Therefore, the conclusions presented in the FSAR remain valid.

F.

Turbine Trio Introduction Two cases, analyzed for both beginning and end of life conditions, are presented in Sections 15.2.3 of the FSAR:

1.

Full credit is taken for the effect of the pressurizer spray and the pressurizer power operated relief valves.

Safety valves are also available.

2.

No credit is taken for the effect of the pressurizer spray or power operated relief valves. Safety valves are operable.

9124Q:10/020386 12 l

l

Although the moderator temperature coefficient will be negative at end of life, all cases were repeated. The result of a loss of load ij a core power level which momentarily exceeds the secondary system power extraction causing an increase in core water temperature. The consequences of the reactivity addition due to a positive moderator coefficient are increases in both peak nuclear power and pressurizer pressure.

Method of Analysis A constant moderator temperature coefficient of +7 pcm/*F was assumed.

fhe method of analysis and assumptions used were otherwise in accordance with those presented in the FSAR.

Results and Conclusions The beginning of life case system transient response to a total loss of load from 100 percent power assuming pressurizer relief and spray valves was calculated.

Peak pressurizer pressure reaches 2531 psia following a reactor trip on high pressurizer pressure. A minimum DNBR well above the limit value is reached shortly af ter reactor trip.

The transient response to a loss of load assuming no credit for pressure control was also calculated.

Peak pressurizer pressure reaches 2572 psia following reactor trip on high pressurizer pressure. The ONBR increases throughout the transient.

Th,e analysis of the beginning of life cases demonstrates that the integrity of the core and the reactor coolant system pressure boundary during a loss of load or turbine trip transient will not be impacted by a +7 pcm/*F moderator reactivity coefficient since the minimum DNS ratio remains well i

91240:10/020386 13

l l

above the limit value, and the peak reactor coolant pressure is less than 110 percent of the design value of 2500 psia.

I Therefore, the conclusions presented in the FSAR remain valid.

G.

Loss of Normal Feedwater/ Loss of Off site Power Introduction The loss of normal feedwater and loss of offsite power accidents (Sections 15.2.7 and 15.2.6 of the FSAR) are analyzed to demonstrate the ability of the secondary system auxiliary feedwater to remove decay heat from the reactor coolant system.

Following initiation of the event the reactor coolant temperature rises prior to reactor trip due to the reduced heat transfer in the steam generators. Thus, the assumption of a positive moderator temperature coefficient results in a reactivity insertion and resultant increase in core power prior to reactor trip. This is turn increases the amount of heat that must be removed following reactor trip, resulting in a more severe transient.

Method of Analysis A constant moderator temperature coefficient of +7 pcm/'F was assumed. A conservative core residual heat generation based on the 1979 version of ANS-5.1 was used. The method of analysis and assumptions used were otherwise in accordance with those presented in the FSAR.

I Results and Conclusions The transient results show that the capacity of the auxiliary feedwater system is adequate to provide suf ficient heat removal from the RCS. The pressurizer does not fill with water, assuring that the integrity of the core is not adversely affected. For the case without offsite power, the results 9124Q:10/011486 14

.=

=

verify the natural circulation capacity of the RCS to provide sufficient heat removal capability to prevent fuel or clad damage following reactor coolant pump coastdown.

H.

Ruoture of a Main Feedwater Pipe Introduction The main feedwater pipe rupture accident (Section 15.2.8 of the FSAR) is analyzed to demonstrate the ability of the secondary system auxiliary feedwater to remove decay heat from the reactor coolant system.

Following initiation of the event the reactor coolant temperature rises prior to reactor trip due to the reduced heat transfer in the steam generators. Thus, the assumption of a positive moderator temperature coefficient results in a reactivity insertion and resultant increase in core power prior to reactor trip. This is turn increases the amount of heat that must be removed following reactor trip, resulting in a more severe transient.

t Method of Analysis A constant moderator temperature coefficient of +7 pcm/*F was

[

assumed. A conservative core residual heat generation based on l

the 1979 version of ANS-5.1 was used. The method of analysis and assumptions used were otherwise in accordance with those i

presented in the FSAR.

Results and Conclusions The transient results show that the capacity of the auxiliary t

feedwater system is adequate to provide sufficient heat removal from the RCS to prevent overpressurization of the RCS or core uncovery. The reactor coolant remains subcooled, assuring that l

the core remains covered with water.

For the case without l

t 9124Q:10/011486 15

offsite power, the results verify the natural circulation capacity of the RCS to provide sufficient heat removal

. capability to prevent RCS overpressurization and fuel or clad damage fo'11owing reactor coolant pump coastdown.

I.

Control Rod Election Introduction The rod ejection transient is analyzed at full power and hot standby for both beginning and end of life conditions in the FSAR. Since the moderator temperature coef ficient is negative at end of life, only the beginning of life cases are affected by a positive MTC. The high nuclear power levels and hot spot fuel temperatures resulting from a rod ejection are increased by a positive moderator coefficient. A discussion of this transient is presented in Section 15.4.8 of the FSAR.

Method of Analysis i

The digital computer codes for analysis of the nuclear power transient and hot spot heat transfer are the same as those used in the FSAR. The ejected rod worths and transient peaking factors assumed are conservative with respect to the actual calculated values for current fuel cycles. The analysis used a moderator temperature coef ficient more conservative than a +7 pcm/*F for all appropriate temperature values and power levels. This is a conservative assumption since the moderator coefficient actually decreases to zero from 70 percent to 100 percent power.

Results and Conclusions A peak clad average temperature of 2683*F was reached in the beginning of life hot zero power case. Maximum fuel temperatures were associated with the full power case.

l l

l 9124Q:10/020386 16 i

Although the peak hot spot fuel centerline temperature for this transient exceeded the melting point, melting was restricted to less than the innermost 10 percent of the pellet.

As fuel and clad temperatures do not exceed the fuel and clad limits specified in the FSAR, there is no danger of sudden fuel dispersal into the coolant, or consequential damage to the primary coolant loop.

J.

Accidental Depressurization of the Reactor Coolant System Introduction An accidental depressurization of the reactor coolant system results from an inadvertent opening of a pressurizer relief or safety valve (FSAR Section 15.6.1).

Since a safety valve is sized to relieve at a much greater flow rate than a relief valve and will therefore allow a much more rapid depressuriza-tion, the case of a safety valve opening is analyzed. This situation initially results in a rapidly decreasing reactor coolant system pressure until the hot leg saturation pressure is reached. With a negative moderator density coefficient (positive MTC), the decrease in pressure results in an increase in core reactivity because the coolant density decreases as the pressure decreases. The most limiting case therefore assumes the reactor is in manual control, such that the increase in core reactivity causes nuclear power and average coolant temperature to increase until the reactor trips. Therefore, the consequence of the reactivity addition due to the +7 pcm/*F i

moderator coefficient is an increase in peak nuclear power.

Method of Analysis The method of analysis and assumptions used were the same as those presented in the FSAR except for the following:

91240:10/020386 17

I 1.

A constant moderator temperature coef ficient of +7 pcm/*F was assumed.

2.

The reactor was assumed to operate in the manual mode of operation to prevent rod insertion prior to reactor trip.

3.

A least negative Doppler-only power coefficient of reactivity was assumed to augment any power increase due to moderator reactivity.

Results and Conclusions The system transient response to the inadvertent opening of a pressurizer safety valve with the reactor in manual rod control was calculated. The reactor trips on overtemperature delta-T and the minimum DN8R occurs shortly after control rods begin to drop into the core.

I The analysis demonstrates that the integrity of the core during a reactor coolant system depressurization transient is not j

adversely affected by a positive moderator reactivity l

coefficient since the minimum DN8 ratio remains above the limit value. Therefore, the conclusions presented in the FSAR remain valid.

l l

l i

I 91240:10/020386 18

e SECTION 3 CONCLUSIONS To assess the effect on accident analysis of operation of McGuire Units 1 & 2 with a positive moderator temperature coef ficient of +7 pcm/*F safety analyses of transients sensitive to a minimum or positive moderator coefficient were performed. These transients included control rod assembly withdrawal from suberitical, control rod assembly withdrawal at power, loss of reactor coolant flow, locked rotor, turbine trip, loss of normal feedwater, rupture of a main feedwater pipe, control rod ejection, and RCS depressurization. This study indicates that a +7 pcm/'F moderator coef ficient does not result in the violation of safety limits for any of the transients analyzed.

Except as noted, the analyses employed a constant moderator coefficient of +7 pcm/*F, independent of power level. The results of this study are conservative for the accidents investigated at full power, since the proposed Technical Specification diagrammed in Figure 1 requires that the coefficient linearly decrease from +7 pcm/*F to O pcm/*F from 70 percent to 100 percent of rated power.

9124Q:10/011486 19

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Figure 1 M00 ERAT 09 TEWERATIM

+7 CMFFICIENT

\\

(pcm/*F) l 70 100 POWER

(%)

{

I 9124Q:10/011486 20

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TABLE 1 FSAR ACCIDENTS EVALUATED FOR POSITIVE MODERATOR COEFFICIENT EFFECTS FSAR Accident Time in Life 15.1.1/1.2 Feedwater Malfunction EOC 1;.1.3 Excessive Load Increase BOC/EOC 15.1.4/1.5 Steam Line Depressurization/ Break EOC 15.2.2/2.3 Loss of Load / Turbine Trip BOC/EOC 15.2.6 Station Blackout BOC 15.2.7 Loss of Feedwater BOC 15.2.8 Feed Line Break BOC 15.3.1/3.2 Loss of Flow BOC 15.3.3 Locked Rotor BOC 15.4.1 RCCA Withdrawal from Subcritical BOC 15.4.2 RCCA Withdrawal at Power BOC/EOC 15.4.3 RCCA Misoperation BOC 15.4.4 Startup of an Inactive Loop EOC 15.4.6 Boron Dilution BOC 15.4.8 RCCA Ejection BOC/EOC 15.5.1 Spurious Actuation of SI BOC 15.6.1 Accidental Depressurization of RCS BOC 15.6.5 LOCA BOC l

t

Accidents Explicitly Addressed BOC - Beginnjng of Cycle EOC - End of Cycle i

9124Q:10/011486 21

4 C

ATTACHMENT 3 ANALYSIS OF SIGNIFICANT RAZARDS CONSIDERATION As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10 CFR 50.92.

Standards for determination that a proposed amendment involves no significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not:

1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.

I.

McGuire 1/ Cycle 4 Reload Related Technical Specification Changes:

The proposed amendments to incorporate RAOC and a positive moderator temperature coefficient ensure that plant operation is consistent with the design and safety evaluation conclusion statements made in the McGuire Unit 1 Cycle 4 reload safety evaluation and ensure that those conclusions remain 4

valid. The reference safety evaluation report submitted by Mr. H.B.

l Tucker's November 14, 1983 letter to Mr. H.R. Denton summarized the evaluation performed on the region-by-region reload transition from the McGuire Units 1 and 2 standard (STD) fueled cores to cores with all optimized fuel (OFA). The report examined the dif ferences between the Westinghouse STD design and 0FA desiFn and evaluated the effects of these differences for the transition to an all 0FA core. The report (approved by the NRC) justifies the compatibility of the OFA design with the STD design in a transition core as well as a full 0FA core, and contains summaries of the mechanical, nuclear, thermal-hydraulic, and accident evaluations which are applicable to the Cycle 4 safety evaluation.

Subsequent cycle specific reload safety evaluations were to verify that applicable safety limits are satisfied based on the reference evaluation / analyses established in the reference report.

The McGuire Unit 1/ Cycle 4 reload safety evaluation (Attachment 2A) presents an evaluation which demonstrates that the core reload will not adversely affect the safety of the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload were reviewed for the Unit 1 Cycle 4 design. The results of new analyses and the justification for the applicability of previous results for the remaining analyses is presented in the cycle specific reload safety evaluation. The results of these evaluation / analysis and tests lead to the following conclusions:

a.

The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are mechanically compatible with the STD design, control rods, and reactor internals interfaces. Both fuel assemblies satisfy the design bases for the McGuire units.

b.

Changes in the nuclear characteristics due to the transition from STD to 0FA fuel will be within the range normally seen from cycle to cycle due to fuel management effects.

c.

The reload 0FAs are hydraulically compatible with the STD design.

r o

Page 2 d.

The accident analyses for the OFA transition core were shown to provide acceptable results by meeting the applicable criteria, such as, minimum DNBR, peak pressure, and peak clad temperature, as required. The previously reviewed and licensed safety limits are met.

Plant operating limitations given in the Technical Specifications e.

affected by the reload will be satisfied with the proposed changes.

From these evaluations, it is concluded that the Unit 1 Cycle 4 design does not cause the previously acceptable safety limits to be exceeded. Further, the reload and associated changes to Unit l's operating limitations have no effects not previously evaluated and approved on accident causal mechanisms or probabilities.

The commission has provided examples of amendments likely to involve no significant hazards considerations (48 FR 14870). One example of this type is (vi), "A change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where results of the change are clearly within all acceptable criteria with respect to the system or component specified in the standard review plan:

for example, a change resulting from the application of a small refinement of a previously used calculational model or design method".

Because the evaluations previously discussed show that all of the accidents comprising the licensing bases which could potentially be affected by the fuel reload were reviewed for the Unit 1 Cycle 4 design and conclude that the reload design does not cause the previously acceptable safety limits to be exceeded, the above example can be applied to this situation.

In addition, the NRC has previously issued no significant hazards consideration determinations for similar McGuire Unit 1 (and 2) reload amendments. Consequently, example (iii) which states "For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.", also applies.

Another example of actions not likely to-involve a significant hazards consideration is (i), "A purely administrative change to technical specifications: For example, a change to achieve consistently throughout the technical specifications, correction of an error, or a change in nomenclature". Accordingly the changes to Unit 2 specifications which do not change the content for Unit 2 but which preserve the distinctions between units within the common document are administrative in nature and involve no significant hazards considerations.

an%

Page 3

[

II.

Unit 2 Increased Positive Moderator Temperature Coefficient ii The proposed amendments would allow a more positive moderator temperature coefficient (MTC) to exist during power operation on McGuire Unit 2.

Safety analyses and evaluations have been performed to support operation of McGuire Units 1 and 2 with a positive MTC of +7 pcm/ degrees F below 70% of rated tharmal power, ramping down to O pcm/drgrees F at 100% power. These are described in a safety evaluation report (Attachment 2B) verifying the acceptability of operation of McGuire units 1 and 2 with this increased positive MTC. To assess the effect on accident analysis transients sensitive to a minimum or positive moderator coefficient were analyzed.

This study indicates that a +7 pcm/ degrees F moderator coefficient does not result in the violation of safety limits for any of these transients.

Since the results of the analysis show that the proposed change can be accomplished with margin to applicable FSAk safety limits, and the e:hange is to an operating limitation involving no changes to hardware or other accident causal mechanisms and can have no effect on accident l

proh2bi.lities, the three standards for determination that no significant hazards considerations are involved are met.

In addition, example (vi) cited in Part I above (which covered, among other changes, an identical change on Unit 1) is also applicable to this Unit 2 change.

Based upon the preceding analyses, Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration.

f l

L l

_.

v t e Letter Sequence Request
TAC:56746, (Open) TAC:56747, (Open) TAC:61512, (Open) TAC:61513, (Open)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval Other: ML20113H622, ML20113H637, ML20128M056
MONTHYEAR1986-05-23 [Table View]

ML20155J567

Text

Subject:

Dear Mr. Denton:

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

1.3 CONCLUSION

5.0 REFERENCES

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