ML20043A483

Rev 1 to Technical Rept Supporting Cycle 16 Operation

ML20043A483
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/30/1990
From:	CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20043A482	List: B13520, Proposed Tech Specs Supporting Cycle 16 Operation ML20043A481 ML20043A483
References
NUSCO-167, NUSCO-167-R01, NUSCO-167-R1, NUDOCS 9005220125
Download: ML20043A483 (37)

Text

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' NUSCO-167, Revision -1.

April 1990 l

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CONNECTICUT YANKEE ATOMIC POWER COMPANY '

HADDAM NECK PLANT Technical Report Supporting Cycle 16 Operation.

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. ',1 Northeast Utilities 1

P.O. Box 270 Hartford, Connecticut l

i 9005220125 900509 POR ADOCK 05000213 p

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TABLE OF CONTENTS -

L i

D L1.

In trod u c tion and S u mmary.......................................................... 1-1 i

t-I-

2.

Op e ra t i n g H i s t ory.................................................................... 2-1 3.

. G e n e r al D e s c ri p ti o n................................................................... 3 1 4.

F u el t S y s t e m D e s i g n.................................................................. 4-1

-t De'ign........................................................................5-1.

5.

Nucleat:

s F

4 6.

. Therm al H ydraulic De sig n.......................................................... 6-l '

7.

Acciden t and Transient 'A n aly sis...................................................

7-1 8.

Core Ope rati n g Limit s............................................................... 8-1

9. - S tattup Program - Physics Testing................................................. 9-1

10. - References

.............................................................................10-1 List of Tables Table 1

3-1. Cycle 15 Di s c h arge Fu el '............................................................ 3-2 i;

4-1. Nominal Fuel Design Parameters.................................................. 4-3 1

i-5-1. Haddam Neck Plant Physics Parameters......................................... 5-2 5-2. Haddam Neck Plant Cycle ' 16 Shutdown Margin............................... 5-3

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List of Finnes Figure 1

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.3-1. Haddam Neck Plant Cycle 16 Core Loading Pattern....................... 3-3 3 2. Haddam Neck Plant BOC16 Burnup Distribution (MWD /MTU)........... 3 '

3-3. Haddam Neck Plant Cycle 16 Control Rod Locations....................... 3-5 5-1. Haddam Neck Plant Cycle 16 Relative Power Distribution at 15 0 MWD /MTU, HFP, A R O.................................................... 5-4.

8-1. Cycle.16 Control Group Insertion Limits - Four loops Operating (Technical Specification 3.1. 3. 6.1 )............................... 8 - 3 8 2. Cycle 16 Contml Group Insertion Limits - Three Loops Operating (Technical Specification. 3.1. 3. 6.2)................................ 8-4

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!l 8-3. Cycle 16 Axial Offset Limits 250 EFPD Four Loops Operating (Technical Specification : 3.2.1.1).................................. 8-5 8-4. Cycle 16 Axial Offset Limits - 250 EFPD - EOL Four Loops Operating (Technical Specification 3.2.1.1 ).................................. 8 - 6 1

l 8-5. Cycle 16 Axial Offset Limits 250 EFPD Three Loops

<(

Operating (Technical Specification. 3.2.1.2).................................. ' 8-7 f

t 8-6 Cycle 16 Axial Offset Limits - 250 EFPD - EOL Four Loops Operating (Technical Specification 3.2.1.2)..................................

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1. Introduction and Summary o

The objective of this report is to support the operation of the sixteenth cycle of the Haddam Neck Plant at its licensed core power level of 1825 MWt. This revision is required due to the fuel assembly dama' e that was discovemd during the. refueling g

outage and the subsequent recovery program to mpair the fuel and redesign the Cycle -

' 16 core.

Included am the analyses outlined in the USNRC document, " Guidance for Proposed License' Amendmenis Related to Refueling". Since it is the licensee's intention to replace expended fuel with fuel of similar design, references are made to previously.

supplied analyses wherever possible.

This report includes the Cycle 16 specific core operating limits. Cycle specific

- operating limits have been removed from Technical Specifications following the guidance provided in Generic Letter 88-16. These limits have been developed using NRC approved methodologies. This report is submitted in accordance with the Technical-Specification 6.9 reporting requirements for -the Technical Report Supponing Cycle Operation.

The nominal 9900 MWD /MTU (351 Effective Full Power Days) Cycle 16, based on the Cycle 15 burnup of 13000 MWD /MTU (459 Effective Full Power Days) is scheduled to begin in June 1990. The reviews of the fuel mechanical performance in Section 4, the thermal hydraulic performance in Section 6, and the accident and transient analysis in Section 7 were based on this Cycle 16 burnup. The Core Operating Limits in Section 8 are also based on this burnup.

Based on the analyses performed and review of the proposed revisions to Technical Specifications,it is concluded that the Haddam Neck Plant can be operated safely at the licensed thermal power level of 1825 MWt for Cycle 16.

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2. Operating History a

L Initial criticality for Cycle 15 occurred on March 19,1988. The plant phased online -

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March 26,1988 and reached 100% power on April 9,198S. A 1000 MWD /MTU 4

coastdown was performed at the end of Cycle 15. Cycle 15 operation was completed on September 2,1989. During reactor shutdown and Reactor Coolant System (RCS) -

q depressurization and cooldown, radiochemistry data spiked significantly higher than -

levels that were observed during previous shutdowns. The magnitude of these spikes indicated that the nurnber of failed fuel rods may be higher than previously estimated.

I The entire Cycle 15 core (157 fuel assemblies)_w'as inspected using ultrasome techniques (UT) to identify failed fuel rods. An initial evaluation of the data indicated up to-20 failed fuel rods and many questionable-signals in the re-insert fuel assemblies. The first two fuel assemblies selected for reconstitution wem visually inspected to attempt to determine the cause of the failures. Metallic debris was observed on the top of the lower nozzle between the rows of fuel rods? Initial; reconstitution effo.ts showed that rods with questionable UT signals were failed and that a number of rods surrounding the failed rods were severly damaged. Based on -

these developments, the recovery effort was expanded to include debris removal from the 109 reinsert fuel assemblies and debris maps were developed prior to restarting -

j reconstitution.' Concurrent with the cleaning activities, the UT results were re-i evaluated. When combined with the debris maps, it was determined that 104'of the

}

109 re-insert fuel assemblies, with over 300 failed fuel rods, needed repair. Five of the re insert fuel assemblies had'no UT indications and no observed debris. An i

l independent UT campaign confirmed the extent of the fiel damage.

The recovery effort was further expanded to include an Eddy Current (ECT) examination of rods surrounding the failed rods, rods at' debris sites and rods on a -

random basis to estimate the number of unknown damaged rods remaining in the re-insert fuel. Due to the number of rods that would require replacement, a donor fuel rod program was developed using twelve fuel assemblies that were scheduled for re-l use in the original Cycle 16 core.

?

The fuel recovery program was successfully completed by repairing 92 re-insert fuel -

assemblies. Over 3300 fuel rods were removed from fuel assemblies and inspected, resulting in 474 rods discharged for cladding failure and damage. All discharged fuel rods were replaced with donor fuel rods at approximately the same bumup. Only two i

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8-solid stainless steel dummy fuel rods were used due to suspected spacer grid damage.

Eight new fuel assemblies were purchased'and four previously ' discharged fuel-assemblies were selected from the fuel pool to replace the twelve fuel assemblies used

-as donors. The four previously discharged fuel assemblies were not in' the core when

. the debris damage took place and no failed rods were detected during a previous UT.

campaign. After completing the recovery of the Cycle 16 reinsert fuel, six of the twelve donor fuel assemblies were rebuilt for use in future cycles.

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3. General Description -

s The reactor core of the Haddam Neck Plant is described in Chapter 4 of the Updated Final Safety Analysis Report (Reference 1). The Cycle 16 core consists of 157 fuel

.. assemblies, each of which is a 15 by 15 array containing 204 fuel rods,20 control rod thimble tubes and one incore instrument sheath. Two reconstituted fuel assemblies each contain one solid stainless steel dummy fuel rod. The fuel rod cladding is i

stainless steel (Type 304)lwith an outside diameter of 0.422 inch and a nominal wall thickness of 0.0165 inch. All fuel pellets are bevelled, dish-end uranium dioxide. The _

fuel pellets are 0.3825 inch in diameter and 0.458 inch in length. All Cycle 16 fuel i

assemblies have an' average nominal fuel loading of 411.5 kg of uranium 'and an -

undensified nominal active fuel length of 120.5 inches. The minimum batch theoretical-3 density is 94.9 percent for all Cycle 16 fuel batches.

' Figure 3-1 is the revised core loading diagram for Cycle 16 of the 'Haddam Neck 3

- Plant. The nominal initial enrichment for fuel Batches 14B,15C,16,17 and 18A is--

4.00 weight percent uranium-235. The four fresh'. Batch 18B fuel assemblies have a'

'i nominal initial enrichment of 3.00 weight percent. The normal end of Cycle '15

- discharge of 48 fuel assemblies are from Batches 15A and ISB. Twelve additional 3

fuel assemblies (8 from Batch 17 and 4 from Batch 16) were prematurely discharged to supply donor fuel rods for the fuel recovery program (see Table 3-1). The j

remaining 48 Batch 16 and 48 Batch 17 fuel assemblies will be shuffled to new i

locations at the beginning of Cycle 16. Twice burned fuel assembly R40 (Batch 15C) -

will be retained as the center fuel assembly in Cycle 16. The 52 fresh Batch 18A fuel assemblies will occupy the periphery of the core and the 4 fresh Batch 18B fuel-assemblies will be located in the core interior on the major axes. Four Batch 14B fuel '

y assemblies discharged at the end of Cycle 14 will be re-inserted on the core periphery l

on the major axes (see Figure 3-1). Figure 3-2 is a quarter-core map showing the fuel assembly burnup distribution at the beginning of Cycle 16.

- Reactivity control is supplied by 45 full-length Ag-In-Cd control rods 'and by soluble g

boron shim. The Cycle 16 locations of the 45 control rods and group designations are shown in Figure 3-3.

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4 Table 3-1: Ovele 15 Dischareed Fuel:

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No. of fuel i

L-Batch assemblies Cveles burned 15A -

44 3

1 15B 4

3 I

u l-1 16 4

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i ls 17 1-1

' Total discharged 60.

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Figure 3-1. Haddam Neck Plant Cycle 16 Core Loading Pattern i

3 15 14 13-12 11-10 9

8 7-6 5

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18A 149 18A R^

18A.

18A

' 18 A 17

-18A 18A 18A P-i 18A 18A 16

~ 16 17.

16 16 IBA '

18A N.

18A 18A 16.

17 17 16

- 17.

- 17 16 18A

' 18A M

j 18A 18 A -

16' 16 17'-

16 18B 16 i 17 ~

16 16 18A

-18A L

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' 17.

16' 17 16 17

' 16 17 '-

17 -

- 16 18A lK Lj i

18A 18 A '

16 17 "16 17 -

17 17 ~

4 17 17 16 17 16 :

-18A 18 A.

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14B 17 -

17 16 ISB '

16 '

' 17 15C 17 16 -

188 l 16 '

- 17 '

17 14B '

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.i 18A.

18A 16 17

' 16 17 17 17 17 17-16 17 '

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' 18A 18A

~ (3 18A 16 17 17 16 17 16

-17 16 :

17

~ 17 16 ;

18A F

18A 18A 16

- 16.

17 16.

18B 16 17 16 t16'-

18A 18A E:

18A 18A 16 17

. 17 16 17 17

'16 18A 18A -

D 18A 18A 16 16 -

17

.16 16 18A-18 A -

C.

i 18A 18A 18A 17 -

18A 18A 18A B

18A 14B 18^

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Batch

1. assemblies Initial w/o U235 j

14B 4

'4.00 t

15C 1

4.00 16 48 4.00 17' 48 4.00 18A

'52 4.00 18B 4

3.00 3-3

q Figure 3-2. Haddam Neck Plant Beginning Of Cycle 16 Burnup Distribution, MWD /MTU

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25461 13985:

22694 0.'

23294 11903 13976 32980

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13985 14716 88071 21332' 8301 22214.'

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21389-10183 25926-25539' O.-

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D 23294 8303

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C 11903 22301 24510 0.

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. 3 Figure 3-3. Haddam Neck Plant Cycle 16 Control Rod Locations -

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14 30 M.-

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~ 35 L-C

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- 28 A

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A B'

A A

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11 43 A

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A D-38 26-25 37 '

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A NO.0F BANK RODS PUNCTION LEGEND B

B CONTROL

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W-ROD BANK DESIGNATION A

17 COMROL l

UU ROD LOCATION NUMBER D

12 SIRTIDOWN -

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'8 SHUTDOWN 45 35

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4. Fuel System Design -

l

'Ihe revised Cycle 16 core consists of reinserted fuel assemblies of Batches 14B,15C, -

1 16 and 17 and the fresh fuel assemblies of Batches 18A and 18B. The pertinent design j

parameters for all six fuel batches are listed in Table 4-1. All fuel assemblies are identical in design (except for enrichment) and mechanically interchangeable.

The fuel reconstitu, tion was performed using procedures, tooling and services provided and approved by the fuel supplier. All reconstituted fuel assemblies continue n

to meet all mechanical design criteria. '

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The fuel rods for all batches are 304 SS clad ofidentical design. The fuel stack length is 120.5 inches. All fuel assemblies' have the same uranium loading. The mechanical l

evaluation of the Cycle 16 fuel rods is provided below:

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Claddine Collaose

>j The Batch 14B fuel rods are the most limiting in terms of creep. collapse due to -

their having the highest previous incore exposure time and burnup. The power 0

- histories of the Cycle 16 fuel assemblies were analyzed to determine a bounding i

power history for creep collapse. This bounding power history was used to analyze a fuel rod operating under conservative conditions for creep collapse.

The tesults of this analysis are applicable to all fuel batches. The predicted creep.

j collapse times and burnups exceed the maximum ex' ected residence times andL l

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. bumups of all fuel batches. The results are shown in Table 4-1 Claddine Stress 1'

The Cycle 16 fuel rods were analyzed by conservative stress analys,es following.

ASME guidelines for pressure vessels. For the design evaluation, the primary 3

membrane stress intensity and any single stress must be less than two-thirds of.

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the minimum specified unirradiated yield strength of the cladding. In all cases the i

margin is in excess of 13.8%

Claddine Stmin 1

The fuel design criteria specify a limit of 1% on cladding plastic tensile 4-1 i

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circumferential strain. The pellet is designed to assure that cladding plastic strain.

j is less than 1% at design local pellet burnup and linear heat generation rate. The'-

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. design burnup and linear heat generation rates are higher than the limiting values :

.l that any of the Cycle 16 fuel batches are expected to experience. The strain.

analyses are based on the upper tolerance values for the fuel pellet diameter and' density, and on the lower tolerence value for the cladding inside diameter. :

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Cladding Fatigue :

L Fatigue analyses were performed using conservative conditions to find the cumulative fatigue usage factor. The fatigue usage factor for the Cycle 16 fuel?

rods was calculated following the ASME Pressure Vessel design code and' 4

. compared to the maximum allowed factor of 0.9. The cumulative fatigue factor.

was found to be 0.2.

All fuelin Cycle 16 is thermally similar. Analyses for all fuel bitches were perfonned with the TACO 2 code (Reference 2), using the analysis methodology consistent with' Reference 3. The maximum fuel rod pressure for all Cycle 16 fuel batches will remain ;

i below nominal system pressure.'

The fresh fuel assemblies-(Batches 18A and B) are identical to the other fuel-assemblies in the Cycle 16 core. This design has performed wellin. previous cycles with no' adverse materials effects. Thus, all possible fuel-cladding-assembly-coolant L

material interactions have been proven to have no adverse effects on fuel performance i

when operated under the conditions expected in Cycle 16.

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o Table 4-1. Fuel Design Parameters -

1 1

L L

Rauh l'

IdB =

-llc M:

,12

.IBAJ

=18R i

Manufactunr B&W

- B&W.

-B&W. iB&W B&W B&W.

Numberof Asseinblies 4

-1'

48-L48' 52 4'

- Previous 1rradiation, Cycles 3.

2 2'

1 0

1 0 Initial Fuel Enrichment, wt% '

4.0.

4.0 -

4.0 4.0 -

4.0 :-

' 3.0 -

-Initial fuel Density.

% Theoretical (minimum) 94.9 '

- 94.9 94.9-94.9' 94.9 l 94.9f r

D Fue1 Pellet Diameter, inch -

0.3825 0.3825. 0.3825 0.3825 0.3825: l 0.3825

- Active Fuel Stack, inch

- 120.5 120.5 120.5 -

120.5 120.5:

120.5 l

Cladding Matenal 304-SS ~ _304 SS 304 SS J 304.SS - 304 SS. : 304-SS Cladding 'Ihickness, inch 0.0165 0.01'65 - 0.0165~ 0.0165 0.0165 0.0165.

Fuel Rod Length, inch --

126.68-- :126.68 [126.68 126.68-126.'68 126.68

}

Initial Gas Pressure, psia -

54.7 154.7

~ 54.7 I 54.7

. 54.7- -

' 54,7 -

Cladding Collapse Time,(EFPH) ~

.38300-32900 32900-32900:

32900 32900

)

Design ResidenceTime EOC 16 i

EFPH Maximum 37272 26232 27960' -19440:

8400 8400-Cladding Collapse Burnup (MWD /MTU) 45120 39485 39485-39485L 39485 39485 1

Design Bumup EOC 16, Peak ~

- Pin, MWD /MTU Maximum 37800 35300

'31900: -20500 11500 11500 v

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5. Nuclear Design:

7 I

- The physics methodology used to support the Haddam Neck Plant for the Cycle 16 reload design is provided in Reference 4.' Fifty-six (56) fresh fuel assemblies are' i

required as feed for the revised Cycle 16. The Cycle 16 redesign accounts for the o

1000 MWD /MTU Cycle 15 coastdown and the effects of replacing failed and damaged -

g fuel rods with fuel rods of approximately the same burnup or solid stainless steel-dummy rods.

. The fuel cycle was designed using standard reload design procedures and the revised

' i Cycle 16 fuel inventory assuming that no fuel rod replacements were perfonned.;At

. the completion of the fuel recovery program, the characteristics of each replaced fuel.-

H rod (or dummy rod) were modeled in a discrete full core, two dimensional analysis to.

l determine correction factors for the unperturbed core. In general, the correction factors

- were insignificant, due to the ability to closely match fuel rod burnup during the repair -

program.-

L Table 5-1 provides a summary of the Cycle 16 kinetics characteristics compared with the current limits based on the reference safety analysis (Reference 5'. The values

)

identified as Current Limits bound the results for the Cycle 16' design and therefore L

continue to be treated as Current Limits, with the exception of the most negative:

Moderator' Temperature Coefficient, the maximum differential l rod w' orth from i

subcritical and the Hot Full Power Doppler Temperature Coefficient. The Cycle '16

[

values for these parameters become the new design values and have been selected to -

provide additional cycle design margin. The impact of these changes is discussed'in Section 7.0 - Accident Analysis. Table 5-2 shows the shutdown margin' calculations for 4-loop and 3-loop bperation at beginning and end of cycle conditions. Figure 5 i illustrates the Cycle 16 relative power distribution at 150 MWD /MTU, hot full power, all rods out and equilibrium xenon conditions. Power distribution and burnup data presented are based on the end of Cycle 15 burnup of 13000 MWD /MTU.

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-g Table So1. Haddam Neck Physics Parameters

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('unent I imit Cvele 16

-l 351 Cycle Design I.ength,(EFPD) 9900 l

Cycle Design Burnup,(MWD /MTU) t, 21695 l

Average Core Burnup, EOC,(MWD /MTU)

Design Com I.cading,(MTU) 64.70 i

Most Positive Moderator Temperatum 0.0 at HFP 0.0 at HFP l

Coefficient,(pcm/F)

+5.0 at 65%

+5.0 at 65%-

l power and HZP power and HZP Most Negative Moderator Temperatum Coefficient,(pcm/F)

-29 s HFP

-32 s HFP j

Doppler Temperatum Coefficient at HFP,(pem/F)

-1.07 to -1,64

-1.18 to 1.81 Delayed Neutron Fraction Beff 0.47 to 0.67 0A7 to 0.67

[

Prompt Neutron Lifetime,(psec)

' 20 to 10 20 to 10 MaximumDifferential Rod Worth 135 150 at Suberitical,(pcm/mch) l I

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52 e

4 a

Table 5-2. Haddam Neck Plant Cycle 16' Shutdown Margin Calculations With Maximum Stuck Rod i

l Hot Full Power 65% Power 4 Imp. trm 3 Imn. nem

}

Available Rod Worth BQL EQL BQL-EQL

= !

Total rod worth less max, stuck rod, HZP '

'5600 5970 5600 5970

?

(1) 1.ess 10% uncertainty 5040 5370 5040 5370 Requimd Rnd Worth Doppler Defect 1040

.970 700 650 f

Moderator Defect 320-790 10 20 i

RodInsertion Allowance 540 440 610 890

}

Flux Redistribution 260

.640 190 440 Void Effect 50 50 50 50

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(2). Total Rod Worth Required' 2210 2890 1560 2050 Shutdown Margin (1)-(2) 2830 2480 3480 3320

- i Requimd Shutdown Margin

> 1800

> 1800

> 2600

> 2600 i

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1 5-3

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a Figure 5-1.- HmMmm Neck Plant Cycle 16 Relative Power i

Distribution At 150 MWDMTU,HFP, ARO 8

7 6

5 4

3 2

1 l

r i

H 1.018

1.153 1.082 1.188 1.044 1.103 0.976 -

0.435 G

1.153 1.168 1.215 1.0%

1.193' 1.002 1.118 0.558 t

F.

1.082 1.214 0.991 1.130

-1.082 0.946 0.947 :

E 1.188 1.084 1.129 0.914

. 0.943

' t.151 0.6 % '

D 1.044 1.193 1.082 ~

0.943 1.179.

0.816 1

C 1.103 1.003 0.946 -

1.151 0.816 l

B 0.976 1.118 0.948 -

0.696 l.

5 A

0.435 0.558 e

k 5-4 4

e

6.'Ihermal Hydraulic Design All fuel assemblies in the Cycle 16 core are hydraulically similar. Since the design N

AH, Linear Heat Generation Rate, Reactor Coolant System flow rate,-

values for F pressurizer pressure and core inlet temperature are equivalent to, or bound the Design values, the steady state minimum DNBR and maximum UO temperature results 2

t provided in Reference 6 remain applicable.

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7. Accident and Transient Analysis i

The accident and transient analysis design basis was reviewed for potential impact due to changes in the Cycle 16 reload physics parameters contained in the Reload Safety Analysis Checklist (RSAC) and plant design modifications to the Reactor Protection System (RPS) and thermal shield removal. Additionally, the assumed steam generator tube plugging level was conservatively increased from 500 to 600 tubes per steam

+

generator. The only revised reload physics design parameters that impacted any of the j

core operating limits,were the Moderator Temperature Coefficient and the Axial Offset limits for four loop operation (see Section 8.0 Core Operating Limits). The Linear

{

Heat Generation Rate limits for four loop operation have been revised due to a i

modeling error that was discovered during Cycle 15 operation.

The design basis non-LOCA transients that required re evaluation due to these changes are:

Boron Dilution l

Uncontmiled Rod Withdrawal from Suberidcal Uncontrolled Rod Withdrawal from Power Idled / Isolated Loop Startup Control Rod Ejection l

Reactor Coolant Pump Rotor Seizure / Shaft Break Loss of Forced Reactor Coolant Flow i

i Steamline Break Steam Generator Tube Rupture Dropped Rod i

The re-evaluation results for each of these transients is provided below:

l Boron Dilution The Cycle 16 critical boron concentration and inverse boron worth have f

decreased slightly from the values used in the Reference 5 analysis.

Additionally, the dilution rate coefficient has increased slightly to more conservatively model the dilution volume. Although these changes result in less time available for operator actiori than the Reference 5 analysis, the results continue to meet the operator action acceptance criteria of 15 minutes -

for Modes 1 through 5 and 30 minutes for Mode 6.

7-1

Uncontm11ed Rnd Withdrawal from Subcritical The Cycle 16 transient values of FQ and FAHN were slightly higher than the RSAC values used in the Reference 5 analysis. Additionally, the maximum reactivity insenion rate was increased from 135 to 150 pcmrmch. 'Ihe impact of the higher peaking factors was more than offset by the design changes in L

the RPS. The Overpower Trip analysis serpoint was redrced and the Stanup Rate trip.is' now operable over a wider range of power levels. The combination of the revised peaking factors and design changes to the RPS yields fuel centerline temperature and minimum DNBR results that are bounded by the Reference 5 results.

i Uncontrolled Rod Withdrawal From Power i

The design change to the RPS increased the total delay and response time of the Overpower Trip from 0.55 to 1.0 second. The increased delay and response time has an insignificant impact due to the slow power ramp that occurs during the rod withdrawal. Additionally, the changes in the Doppler l

Temperature Coefficient and RCS volume increase due to thermal shield removal were assessed. The minimum DNBR and maximum fuel centerline temperature remain within the acceptable fuel design limits.

Isolated / Idled Loon Startun The total RPS delay and response time for the Overpower Trip has increased frem 0.55 to 1.0 second. The increased delay and response time was offset i

by less severe radial power peaking factors consistent with the FAHN values provided in Section 8 Core Operating Limits. The combination of revised RPS response and delay time and power peaking factors yields a minimum DNBR of 1.52, which is insignificantly lower than the Reference 5 value of 1.55.

l Control Rod Eiection The Cycle 16 design axial peaking factors were slightly higher than the values used in the Reference 5 analysis. In addition, the design change to the i

RPS has increased the total delay and response time of the Overpower Trip l

l from 0.55 to 1.0 second. The higher axial peaking factors and longer RPS l

delay time were offset by a less severe Cycle 16 radial pin peaking census i

used in the DNB evaluation. The combination of the revised peaking factors, delay times and pin census yields minimum DNBR and radiological results I

that are bounded by the Reference 5 results. The limiting fuel centerline temperature case is for the four loop, Hot Zero Power (HZP) case. The i

7-2 L

i i

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maximum fuel centerline temperature increased by less than 200 F for this l

l case. The maximum fuel centerline temperature for the three loop, HZP case increased by approximately 470 F. However, this case is not the limiting case

[

and therefore the impact is insignificant. The increase in the maximum fuel stored energy for the four and three-loop HZP cases is also insignificant.

l Reactor Coolant Pumn Rotor Seizure / Sh' aft Break

[

Imss of Forced Reactor Coolant Flow The total RPS delay and response time for the Low Reactor Coolant Flow '

l l

trip has inorcased from 1.65 to 2.15 seconds. The increased delay and response time was offset by less severe radial power peaking factors for the -

four loop analysis, consistent with the FAHN values provided in Section 8, Core Operating Limits. The combination of revised RPS response and delay time and power peaking factors yields minimum DNBR results for the four

' loop case that are bounded by the Reference 5 results. The increased delay time causes a slight reduction in the minimum DNBR for the three loop analysis. However, the three loop results remain bounded by the four loop results, thus the impact of the increased delay time on the three loop analyses l

is insignificant.

Steamline Break The Cycle 16 evaluation of the the reactor physics statepoints using the RETRAN system results yields a slightly higher return to power lesel.

However, the statepoint power peaking values are less severe and more than

~

offset the effects of the higher power level. The combination of these changes j

yield fuel temperature and minimum DNBR results that are bounded by the

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Reference 5 results.

1 Steam Generator Tube Rupture l

The RETRAN system response was re-evaluated to assess the impact of the j

thermal shield removal. The thermal shield removal only impacts the atmospheric releases from the intact steam generator during RCS cooldown due to the increased fluid volume. *Ihe re-evaluation shows that there is less than a 1% increase in the atmospheric releases and therefore an insignificant impact on the radiological consequences.

Dropoed Rod The change in the Cycle 16 most negative Moderator Temperature Coefficient i

s 7-3 j

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was assessed by re-evaluating the limiting dropped rod transient (Hot Full

Power, four loop operation). The change in MTC yielded an insignificant reduction in the minimum DNBR.

l

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The small break LOCA design basis (Reference 7) was reviewed. The Reference 7 results remain bounding for Cycle 16 operation.

l The large break LOCA design basis (Reference 6) was reviewed. Revisions to the

~

Safety Injection flotv rate and distribution of the downcomer/ lower plenum nodal volumes have required a reduction in the maximum allowable Linear Heat Generation Rate (LHGR). The LUGR has been reduced from 13.3 kw/ft to 12.9 kw/ft for the 0-250 EFPD burnup window and from 14.6 kw/ft to 13.7 kw/ft for the 250 EFPD to End of Life burnup window. This change restores the resulting Peak Cladding j

Temperature (PCT) to 2295 F, which is less than the 2300 F Interim Acceptance Criteria requirement.-

v The large break LOCA was further reanalyzed to determine the effects of the thermal shield removal. This reanalysis also conservatively assumed a steam generator tube plugging level of 600 tubes per steam generator. Removal of the thermal shield reduced the downcomer resistance, but increased the the Reactor Coolant System (RCS) inventory. These are offsetting effects with respect to RCS blowdown time.

]

Removal of the shield resulted in a 1.3 second increase in End of Blowdown (EOB),

and coincidently a 1.3 second increase in the the refill time to Bottom of Core (BOC) recovery such that the adiabatic period between EOB and BOC titnes were unchanged.

However, when compared to the previous analysis, the heatup rate was reduced because the heatup rate at the extended EOB time was reduced. The later EOB time means that the initial temperature at the onset of the adiabatic is slightly higher (approximately 10 F). However, this is compensated for by the reduced heatup rate.

The resulting PCT was reduced to 2283 F.

}

The Axial Offset alarm setpoints have also been adjusted to account for the change in

. i the LHGR and the Cycle 16 design. The Cycle 16 limits for LHGR and Axial Offset are provided in Section 8, Core Operating Limits.

l The ten non LOCA transients that required re-evaluation due to changes in the plant design and design reload physics parameters have been shown to yield results that are bounded by the Reference 5 results or yield an insignificant change. The revisions 7-4 5

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made to the LHGR and Axial Offset alarm setpoints have restond the LOCA PCT to ~

less than the 2300 F requirement. 'Iherefore, the Cycle 16 design does not adversly i

' affect the accident and transient analysis design basis.

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8.0 CoreOperatingLimits j

i The Cycle 16 Core Operating Limits have been developed using the methodologies approved by the NRC in References 8 13. The operating limits have been determined so that all applicable design limits (e.g. fuel and core thermal. hydraulic limits, ECCS limits, nuclear limits and transient and accident analysis limits) of the safety analysis are met. The changes in the Operating Limits for Cycle 16 include the most negative

- f Moderator Temperature Coefficient (3.1.1.5), the four loop Linear Heat Generation Rate limits (3.2.2.1) and the four loop Axial Offset limits (3.2.1.1). The following Core Operating Limits have been established for Cycle 16:

Moderator Temnerature Coefficient Limit (Technical Snecification 3.1.1.5) i The Cycle 16 ModeratorTemperature Coefficient (MTC) limit shall be:

a. Less positive than 5 pcm/F for the all rods withdrawn, Beginning of Cycle Life (BOL), hot zero THERMAL POWER condition; and t

b. Less positive than 0 pcm/F for the all mds withdrawn, BOL, RATED THERMAL POWER condition; and -

c. Less negadve than -32 pcm/F for the all rods withdrawn, End of Cycle Life (EOL), RATED THERMAL POWER condition.

Linear Heat Generaton Rate Limits. Four Loons Oneratine (Technical Snecification 3.2.2.1)

The Linear Heat Generation Rates (LHGRs) shall not exceed the following limits for the following cycle residency times:

a. Less than 250 EFPD 12.9 kW/ft t

b. Greater than 250 EFPD but less than END-OF-CORE LIFE 13.7 kW/ft

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The Linear Heat Generation Rates (LHGRs) shall not exceed the following limits for the following cycle residency times:

a. Less than 125 EFPD 9.295 kW/ft b.125 to 250 EFPD 9A25 kW/ft

c. Greater than 250 EFPD but less than END-OF-CORE LIFE 10.075 kW/ft i

Nuclear Enthalov Rise Hot Channel Factor Four Loons Oneratino (Technical Snecification 3.2.3.1)

N The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, FAH,

shall be limited by the following relationship:

f N

FAH s 1.60 [ 1.0 + 0.3 (1-P) ]

where: P = Thermal Power / Rated Thermal Power Nuclear Enthalov Rise Hot Channel Factor Th ree Loons Oneratino (Technical Snecification 3.2.3.2)

The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, FAH.

shall be limited by the following relationship:

N FAH s 1.64 [ 1.0 + 0.3 (0.65 P) )

where: P = Thermal Power / Rated Thermal Power 8-2

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L 9.0 Startup Program Physics Testing -

The planned startup tests associated with core performance are outlined below.

These tests verify that core performance is within the assumptions of the safety analysis and provide the necessary data for continued hafe plant opera' tion.

4 Pre-Critical Tests

+

1. Control rod drag test

2. Ilot control rod drop time testing.

t Zero-Power Tests

1. Critical boron concentration

2. Temperature reactivity coefficient

a. All rods out c

b. Banks B,' A, and D inserted

3. Control rod coupling verification

4. Control rod group worths for Banks B, A, and D Power Tests l

1. Core power distribution mapping at s80 and 100% full power, normal control bank configuration.

2. Excore / incore correlation verification.

l-t 91 f

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10. References

1. Updated Final Safety Analysis Report, Connecticut Yankee Atomic Power Company, Haddam Neck Plant.

't

2. Y. H. Hsil, et al., TACO 2. Fuel Pin Perfonnance Analysis. B AW-10141 PA Babcock & Wilcox Company, Lynchburg, Virginia.

I

3. J. H. Taylor (B&W) to J.S. Berggren (NRC), Letter, "B&W's Responses to TACO 2 Questions, April 8,1982.

4. Nonheast Utilities Service Company, Physics Methodology for PWR Reload Design, NUSCO-152, August,1986.

5. Nonheast Utilities Service Company, Haddam Neck Plant - Remalysis of Non-LOCA Design Basis Accidents, NUSCO-151, June'30,1985.

6. Connecticut Yankee Atomic Power Company - Haddam Neck Plant, Technical Repon Supponing Cycle 15 Operation, NUSCO-155. June 1987, r

7. Nonheast Utilities Service Company, Haddam Neck Plant - Small Break LOCA Topical Report TMI Action Plan Items II.K.3.5, II.K.3.30 and II.K.3.31, December,1984.

8. F. M. Akstulewicz to E.J. Mroczka, Review of NUSCO Topical Report on i

Physics Methodology for PWR Reload Design (NUSCO-152), August 3,1987.

l l

L

9. A. B. Wang to E.J. Mroczka, Safety Evaluation for Northeast Utilities Topical i

Report 140-1, "NUSCO Thermal Hydraulle Qualification, Volume I (RETRAN),

1 July 26,1988.

I

10. F. M. Akstulewicz to J.F. Opeka, "NUSCO Thermal Hydraulic Model Qualification, Volume II (VIPRE)" Topical Repon NUSCO 140-2, October 16,-

1986.

11. A. B. Wang to E.J. Mroczka, Safety Evaluation of Nonheast Utilities Topical.

Report 151, "Haddam Neck Non LOCA Transient Analysis", October 18,1988.

i 10 ;

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12. Supplement to the Safety Evaluation by the Directorate of Licensing, U.S. Atomic j

Energy Commission, Docket No. 50-213, Connecticut Yankee' Atomic Power i

Company, Haddam Neck Plant, December 27,1974.

l l

13. A. B. Wang (USNRC) to E. J. Mroczka (CYAPCO), Safety Evaluation for l

l NULAP5 Code and Its Use in Haddam Neck Small Break IDCA Analyses

.(NUREG 0737 Items II.K.3.30 and II.K.3.31), August 3,1988.

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