L-76-347, Cycle 4 Reload Safety Evaluation Information

Cycle 4 Reload Safety Evaluation Information

ML18227B744
Person / Time
Site:	Turkey Point
Issue date:	10/04/1976
From:	Robert E. Uhrig Florida Power & Light Co
To:	Goller K Office of Nuclear Reactor Regulation
References
L-76-347
Download: ML18227B744 (32)

Text

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6 FLORIDAPOWER & LIGHTCOMPANY October 4, 1976 L-76-347 Office of Nuclear Reactor Regulation Attn:

Earl R. Goller, Assistant'irector Division of Operating Reactors U. S. Nuclear Regulatory Commission

',lashing+on,

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20555 IA

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OCT8 IgP8~

U5 ~ MICtEAR RfGU~yagg g

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Dear Nr. Goller:

Re:

Turkey Point Unit No.

3 (Docket No. 50-250),

Unit 3 C cle 4 Reload Information Attached herewith is the Safety Evaluation performed for the reload of Turkey Point Unit 3 and'he subsequent return to operation.

This report is being forwarded to you for your information.

No changes in the facility operating license or the technical specifications are required to conduct, the reload ox the return to full power following'he reload.

As discussed in our letters L-76-300 of August 19, 1976, and

'L-76-307 of August 25, 1976,'he effect of plugged stearri generator tubes was considered in tkie development of the revised Fg limit of 2.11'which is currently applicable to both Unit 3 and Unit 4.

The revised limit, incl'udes a tube plugging factor which is based on approximately 4% of the total number of-Unit 4 steam generator tubes having been plugged.

Since only approximately 2.6% of the Unit.

3 steam generator tubes have been plugged, the revised Fg limit for Unit 3 conservatively accommodates the effect of plugged steam generator tubes on ECCS performance.

In our letter L-76-300 of August 19, 1976, we also stated how rod bow affected the DNB limits of the Turkey Point units.

Rod bow for Cycle 4 will be treated in the saine manner duxing Cycle 3.

The Cycle 4 x'eload and the attacheo safety analysis have been reviewed by the Turkey Point Plant-Nuclear Safety Committee and the Florida Power 6 Light Company Nuclear Review Board.

They have concluded that the xeload and return to operation following the reload Qo not, involve an unreviewed safety question.

aors&)g PEOPLE... SERVING PEOPLE

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Karl R. Goller Re:

Turkey Point Unit No.

3 (Docket No. 50-250)

October 4, 1976 Page The attached safety evaluation considered operation at a system pressure of 2100 psia and a.core inlet temperature of 539'F, and at the FSAR design conditions of 2250 psia system pres'sure and 546.2'F core inlet temperature.

However, at this time, consideration is not being given to operating at, other than 2100 psia system pressure and. 539'F core inlet. tempera-ture, which are the current Unit 3 operating parameters.

Westinghouse has indicated that after 2000 MWD/NT of operation in Cycle 4, system operat:ing pressure may have to be increased.

Once the necessity for this change is determined, a separate review will be conducted in accordance with 10 CFR 50.59, and any xequired license amendments will be submitted to your staff.

Very.truly yours, pA.A.~

Robert: E. Uhrig Vice President REU/GDW/hlc Attachments cc:

Jack R.

Newman, Esq.

Docket 0 SO-Rg 0 Control 0 (0( g I ~

Date Recvd

]0-Q

'Regufatory Docket Fife RELOAD SAFETY EVALUATION TURKEY POINT PLANT UNIT 3,. CYCLE 4 FLORIDA POWER'& LIGHT COllPANY

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TABLE OF CONTENTS.

Title Pacae 1.0 INTPODUCTIOH AWO SUKNRY 2.0 PEACTOR DESIGN 2.1 Hechanical Design 2.2 Nuclear D sign 2.3 fhennal and Yydraulic Design 3.0 POkEP. CAPABILITY AND ACCIDENT EYALUATION 3.1 Power CapabHity 3;2 Accident Evaluation

~ 3.3 Incidents R analyzed

4.0 CONCLUSION

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5.0 REFERENCES

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s LIST OF TA6LES Table Title Fuel Assembly Design Parameters Kinetics Characteristics Shutdo~;n Requirements and Hargins

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LIST OF FIGURES

~Fi ere Title l

2 3

Loading Pattern Source and Burnable Poison Locations Trip Reactivity Versus Rod Position 1

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1.0 INTRODUCTIOH A",D

SUMMARY

Turkey Point Unit 3 is in its third cycle of operation.

The unit will refuel and be ready for Cycle 4 startup in mid December, 1976.

The Turkey Point Unit 3 Cycle 4 core loading pattern is shown as Figure 1.

Fifty-two Region 3 assemblies will be removed and replaced by forty Region 6 assemblies; and four Region 1 ard eight Region 2 assemblies stored in the spent fuel pit during Cycle 3 (see Table 1).

Depleted borosilicate burnable poison rods will be used in Cycle 4.

The location of these rods is shown in Figure 2.

This report.presents an evaluation for Cycle 4 which demonstrates that the core reload will not adversely affect the safety of the plant. It is not the purpose of this report to present a reanalysis of all poten-.

'teal incidents".

Those incidents analyzed and. reported in the 'FSAR which could potentially be affected by fuel reload have been reviewed for the Cycle 4 des',gn "escribed herein.

The results of ne>r analyses.'have been includ d and the justification for the applicability of previous results for.the remaining analyses Iis presented.

It has been concluded that the Cycle 4 design does not cause the previously acceptable safety limits for any incident to be exceeded.

This cohclusion is based on the assumption that:

(1}.Cycle 3 operation is terminated after 8200 + 1000 N';tD/SITU and (2) there is "".

adherence to plant operating limitations in the Technical Specifications.

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Nominal design parameters for the beginning of Cycle 4 are 2192 Mwt core power,,2100 psia system pressure, 539'F core inlet temperature, and 5.56 kw/ft average linear fuel power density (based on 144" active fuel length).

However, the safety evaluation of. Cycle 4 also considered operation at the FSAR design conditions of 2250 psia system pressure and 546.2 F core inlet temperature, The conclusions presented herein are applicable to both sets of initial conditions.

2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The only fuel that has not been in the coze previously is Region 6 fuel.

The mechanical design of Region 6 fuel is dimensionally the same as

'Region 5 fuel.

Region 6 fuel has a different enrichment, as not'ed in Table l.

Other physical design aspects of Region 6 aze the same as Region 5, except that the initial pressurization level of the fuel rods has been decreased by 65 psi.

Region 6 is comprised of fuel made using U02 powder which meets Westinghouse specifications, but was fabricated by a process which differs from the standard Westinghouse powder process.

The Draft Regulatory Guide on Densification dated 'July 29, 1975, willbe used to determine the amount of densification in Region 6 fuel.

Clad flattening time is predicted to be greater than 23,500 EFPH for the limiting region (Region 2) using the current Westinghouse evaluation model Therefore, Region 2 has a nominal Cycle 4 allowed residence

'(1) time of.6400 EFPH (Cycle 2 lifetime was 6800 EFPH, and Cycle 1 lifetime was 10,300 EFPH).

Westinghouse has had considerable experience with Zizcaloy-clad fuel.

This experience is extensively described in WCAP-8183, "Operational Experience. with Westinghouse Cores" This report is updated about

~ (3) every six months.

2. 2 NUCLEAR DESIGN Table 2 provides a comparison of the range of values encompassing the Cycle 4 core kinetics parameters with the current limits based on previously submitted accident analyses.

It can be seen from the table that most of the Cycle 4 range of values fall

J within the current limits.

These parameters are evaluted in Section 3.0.

Table 3 provides the coni;ol rod worths. and requirements.

The required shutdown margin is based on previously submitted accident analysis The available shutdown margin e"ceeds the minimum required.

The reactivity insertion rate for Cycle 4 is slower than the one used in prev$ ou's cycles (see Section 3,3 and Figure 3}.

The reactivity in-sertion rate is different because the combined bank worth as a function of'ime (axial location) has changed.

The reactivity insertion rate for Cycle 4 was calculated by a very conservative method that produces a flux distribution skewed towards the bot'tom of the core.

This reduces

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the reactivity worth of'he banks at the top of the core relative to

. the total worth.

Such a calculation provides a conservative trip reac-tivity shape for accident analysis 'Since the axial flux distribution is normally distributed evenly with constant axial offset contro1.-

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2-3 THERMAL ANO HYDRAULIC DESIGN No significant variations in thermal 'rargins will results re the Cycle 4.

".reload.

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3.0 iR CAPMILITT AIID ACCIDEIIT EVATID'3 3.1 PO! JER CAPABILITY This section reviews the plant power capability considering the conse-quences

'of those incidents examined in the FSAR using the previously accepted design bases.

It is concluded that the core reload will not adversely affect the ability to safely operate at'100Ã rated power 1

during Cycle 4.

A maximum local rod power of 23 kw/ft corresponds to the fuel centerline temperature limit of 4700'F for Region 5 or 6 fuel.

This I'an be accommodated with margin in the Cycle 4 core.

The time dependent.

(2) densification model was used for this evaluation.

No significant variation in the LOCA limitwill result from the Cycle 4 reload.

3. 2 ACC IDENT EVALUATION,'

3 The effects. of 'the reload on the design basis and postulated incidents

. analyzed in, the FSAR~ 'ave'een examined.

In most cases it was found that the effects can be accommodated within'the conservatism of the initial assumptions used in the.previous applicable safety analysis.

For those incidents which were reanalyzed, it was determined that th

'applicable design basis limits are not exceeded, and therefore, the

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• conclusions presented in the FSAR are still valid.

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V A reload can typically affect accident anIalysis input'parameters in

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three major areas:

kinetics characteristics, control rod worths, and core peaking factors.

Cycle 4 parameters in each of these three areas were examined as discussed below to ascertain whether new accident analyses are required.

Kinetics Parameters A comparison of the range of values encompassing the Cycle 4 kinetics parameters with.the current limits is given in Table 2.

t.ost of the

11

range of values remain 'srithin the bounds of current limits.

The small changes in core physics parameters have.a negligible effect on transient analysis.

Therefore, no additional accident analysis is required due to changes in these parameters.

Control Rod lforths Changes in control rod worths may affect shutdown margin, differential rod wor ths, ejected rod wor ths, and trip reactivity.

Table 3 shows that the Cycle 4 shutdown margin requirements are satisfied.

As shown in Table 2, the maximum differential rod worth of trio RCCA control banks moving'together in their highest worth region for Cycle 4 is less'-

than or equal to the current limit.

.. 'Ejected rod vrorths for 'Cycle 4 are within the bounds of the current limits.

Cycle 4 has a slower trip reactivity insertion rate than Cycle 1; -..

ho('revec the.tota'i trip reactivity is significantly greater than the value assumed in Cycle 1.

The effects of this reduced reactivity trip rate,has been'evaluted for those accidents affected and compared with the Cycle 1 analyses.

Slower transients are relatively insensitive to changes in trip reactivity insertion rate, and therefore need not be

.. reanalyzed due to the change in trip reactivity versus rod position.

Fast tra'nsients such as rod ejection ard rod withdra>ral from svbcritical,

.'-.:in which negative reactivity insertion. is due primarily to Dopp1er feedback, will be unaffected'y the change in trip reactivity since the

'transient is essentially'turned around before 'rod insei tion starts."

. The effect of variations in trip reactivity insertion rates for the r'od withdrawal at power incident has been investigated.

- The results of

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this analysis show that the minimum OHBR is unaffected,.since the minimum GNBR for the transient occurs at. relatively low reactivity insertion rates.

For the loss'of flow and locked rotor transients, the. change in trip reactivity versus rod insertion will result in a slightly higher tran-sient heat flux.

Since the minimum O.tB ratios for these transients are sensitive to heat flux relative to flow, these incidents were reanalyzed.

. The results of these calculations are discussed in Section 3.3.

Core Peakinq Factors Peaking factors folloiiing control rod ejection are within the bounds of the current limits.

Evaluation of peaking factors for the rod out of position and dropped RCCA incidents shows that DflBR is maintained above 1.3.

For the dropped bank incident, the minimum ONBR criteria of 1.3 is satisfied without taking credit for a turbine runback.

A peaking

-factor evaluation for the hypothetical steamline break transient sho,ved that the 0"(BR is maintained. above 1,30.

3. 3 INCIDENTS REANALYZED Certain incidents were reanalyzed at a system pressure of 2250 psia.

The results described below are also conservative for a system pressure.

of 2100 psia.

The complete loss of flow and locked rotor transients wer'e reanalyzed due to a slower trip reactivity insertion rate.

A comparison of the PSAR and Cycle 4 trip reactivity versus rod position values are shown in Figure 3 along with the conservative values assumed in the reanalysis.

As noted in Section 3.2, the assumed total trip worth has been increased from a value of 2,8% to 4.0% bk/k which is conservative for Cycle 4.

The calculations vere performed using the-same methods and assumptions used for Cycle l(5)

For the complete 3/3 pump loss of flow in ident, the minimum OHB ratio does not fall below the limiting value of 1.30.

The other cases des-cribed in the FSAR incident show larger C.';B ratios than the 3/3 pump incident.

Thus it is concluded that all cases would remain above the

.1.30 limit and the conclusions as presented in the FSAR for this inci-dent are still valid.

For the two and three 'loop locked rotor cases, no rods are expected to exper ience DiHB; however, 'the hot channel quality at the location of the minimum DNB ratio exceeds the range of quality over which the DNB cor-relation was derived.

For this reason, D!(B was conservatively assumed A

to occur.

The three loop case is most limiting.from a peak clad tem-

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perature standpoint.

The results show a peak clad tet.perature of 1600'F which is well below the limiting value of 2700'F.

The amount of Zr-H20 reaction is small '(less than 1X by weight) and contributes less'than 10'F to the. peak cladding temperature..

The two loop case is most limit-ing from a peak pressure standpoint.

The results sho'~t a 'peak pressure d%--le~s Vea222lLosia Nhich. is below the reactor vessel stress 1hxit.

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Vous, cire m--~e Md in the FSAR are still valid for Cycle'.

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4.0 CONCLUSION

S No changes in the Technical Specifications are required for the upcoming Unit 3 refueling or for the return to power operation following the r e fueling.

The refueling and the subsequent return to full power operation have been evaluated to verify that they do not involve an unreviewed safety question.

Title 10 of the Code. of Federal Regulations,, Part 50.59(a),

specifies that an unreviewed safety question exists if:

(1)

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or

, (2)

A possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created; or (3)

The margin of safety as defined in the basis for any technical specification is reduced.

The next refueling of Unit 3, which is scheduled for the Fall of 1976, is the third refueling of this Unit.

There are no significant differences between this refueling operation and previous refueling operations.

The

~refueling operation has been'reviously analyzed as reported in the Turkey Point Units 3 and 4 Final=.Safety Analysis Report.

The NRC Staff. Safety Evaluation Report. dated March, 15, 1972 found this refueling operation to be-acceptable.

'Therefore, there. is no unreviewed safety question associated with the refueling operation.

Reactor. design, power capability, and postulated incidents have also been. evaluated.to determine if they involve an unreviewed safety question.

The conclusion presented above in Section 2.0 and 3.0 of this safety evaluation show that the Unit 3 Cycle 4 reload satisfies the criteria of 10 CFR Part 50.59(a),

and that al.l limits previously found to be acceptable by the NRC Staff will still apply.

In summary, we have concluded that:

(1)

Since the c'ore refueling does not incr ease the probability or consequences of a previously analyzed accident, decrease safety margin, or create the possibility of an accident not previously analyzed, the refueling does not involve an un-reviewed safety question; (2)

The existing Limiting Conditions for Operation incorporated in the Technical Specifications are appropriate for use during. Unit 3 Cycle 4 operation; (3)

There is reasonable assurance that the health and safety of the public will not be endangered by the operation as described herein; and (4)

Such activities will be conducted, in compliance with the Commission's regulations and will not be inimical to the common defense and security or to the health and safety of II'he public.

5. 0 REFERENCES

'I.

George, R. A., et al "Revised C1ad Flattening llodel", 4'CAP 8377 (Proprietary) and tlCAP 8381 (tior Proprietary),

Quly 1974.

Helln;an, J. fI.'(Ed.), "Fuel Densifica'tion Experimental Results and Model for Peactor Operation",

HCAP 8218-P-A, March 1975 (Proprietary) and hCAP 8219-A, March 1975 (Non Proprietary).

3.

Schreiber, R. E., Iorii, J. A., Plocido, Y, J., "Operational'xper-ience >lith Hestinghouse Cores",

MCAp 8183",. Reyiston 4'; f.arch 1976.

4.

Final Safety Analysis Peport, Turkey Point Units No. 3 and 4.

."Fuel Densification, Turkey Point Plant Unit halo. 3",

>fCAP 80?4

,/Proprietary).and MCAP 8075 (I'ion Proprietary}, February 1973.-

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~ABLE 1

TURKEY POI>lT 3 - CYCLE 4 FUEL ASSEHBLY DESI(iH PARAHETERS RBQIOn Enrichment (w/o U 235) 1.86

2. 56 2.56

2. 60

2. 90 3.10 Density (5 Theoretical)*

~ Number of Assemblies 93.8 5

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92.8

94. 6..

94.'8

94. 5 52

.."24 24 94.5

.'Approximate Burnup at Beginning of Cycle 4 (W>Dll.rru)

.. 12000

=237PO:

16700 8100.=.. 6600'0 Change.in Internal Pod Pressure compared to Region 1 (psi)

. 0

+1 30

+1 00

+1 00

+35

.~ All.regions. except Region 6 are as-built values.

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TURKEY POIl(T UHIT 3 KIt(ETICS CHAPS.CTERISTICS Current Limit 2250 sia)

Cyc1e 3 2100 sia Cyc1e 4 ~

2250 sia tloderator Temperature

-3.5 to +0.3*.

. Coefficient, (hp/'F)x10

-3.5 to 0

- -3.5 to 0 Dopp1er Coefficient, (hp/ F)x105

-1.6 to -1.0

-2.6 to -1.0

-2.6 to -1 0-

..Delayed Neutron Fraction,

.50 to.72

.50 to.72 v

.50 to.72 Prompt Neutron Lifetime

':(pseud) 14 to 18.

20.'

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20 tlaximum Differentia1 Rod '

80 Morth of Tv~o 8anks Hoving Together at;.

HZP {pcm/in)**

80

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The positive coefficient does not occur at operating conditions

• ~. pcm = 10 hp The Cycle 4 range of values is valid for both the nominal design conditions of

'he plant (2250 psia pressure, 546.2'F inlet temperature) and for the planned Cycle 4 operating conditions (2100 psia pressure,

'539 P ihlet temperature).

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TABLE 3 TURKEY POItlT 3 - CYCLE 3 Ai(D 4 SflUTDOllif REQUIREHEHTS A'.aD tQPGIHS Control Rod 'rth hp CYCLE 3

-BOC EOC CYCLE 4

• BOC EOC All Rods Inserted Less'llorst Stuck Rod 5.87 6.19 "P ) Less 10Ã 5.28:

5.57 6.34 6.39 5.70 '.75 Control Rod Re uirements

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Peactivity Defects (Doppler, Tavg, Yoid, Redi'stribution) 1.64 2.52'.62 2.57

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'od Insertion Allo<rance (2) Total Requirements

'hutdown Mar in 1 - 2

..50

'.14

.50

3. 02 3.1L 2.55

.60

.60 2.22

'.17

~3.48 "

2.58 r

Required Shutdo".n Hargin

(~ hp) 1.00

.1.77 1'. 00 1.77

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• The Cycle 4 range'f values is valid. fo'r both.the nominal:.design"conditions of

.the plant (2250 psia pressure, 546.2 F inlet temperature) and for the planned Cycle 4 operating conditions (2100 psi'a pressure, 539 F inlet temperature).

III 4

FIGURE 1

t CORE LOADING PATTERN TURKEY POII'lT U"tIT 3 CYCLE 4 P

N N

L

.K J

H.

6 F

E D

C 6";

6.

4 6

6 2

4 5A 6

6

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58 5A 4

6 2,

6 5A 58

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4" 5A 4

58 4

5A 6

6 5A 4-6.

6

'-'4

.58 58 58 58 2 '-

58 1

58 4

.5A 4

5A 4

58 4

58 4

5 6

5A 1

5A 2

5A 4

4 6

6 4

58 58 4

'5A 4

58

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6

'6 5A 4

2 6

5A 58 58 2

58 1

58 4

4 5A 4

58 4

.5A.

58'8 4

5A 4

6

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5A 4.

4 4

5A 6

6 6

4 6

6 2

~p X

Fuel Region P

Removable Rod Assembly

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t FIBVRE 2 SOURCE AKD TURKEY POI '7 U'kIT 3 CYCLE 4 BURHABLZ POI SO!l LOCATIONS R

P N

N L

K J

H G

F.

E 0

C B

12 BP's 12 BP's S

12 pl 12 P's 12 BP's 1

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12.

BP's 2

BP's 12 BP's 12 BP's

~ 12 P's 12 BP's 12 BP's

.12-BP's 12 BP's 12 BP's 12 BP's 12 BP's 12 BP's 12 P's 12 P's 12 BP's 12 BP's 12 BP's 12 BP's 12 BP'S S

12 BP's 12 BQ I s 12 P's S

BP'5 Source Locat;ion Depleted Burnable Poisons

figure 3 TURKEY POINT UNIT 3 CYCLE 4 TRIP REACTIYITY YERSUS ROD POSITIO'A 5.0 4.0 3.0

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2.0 1.0 I

..0 10, 20 30

. 40, 50 60 70,':. 80 SO

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~ l I Rod Insertion Current Limit Cycle 4 Yalue R

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