ML20045F387

Technical Rept Supporting Cycle 18 Operations, Haddam Neck Plant

ML20045F387
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	05/31/1993
From:	CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20045F376	List: B14517, Forwards Technical Rept Supporting Cycle 18 Operation, for Haddam Neck Plant Per TS Section 6.9.1.9.d ML20045F387
References
NFB-001, NFB-1, NUDOCS 9307070253
Download: ML20045F387 (29)

Text

...

NFB-001 May 1993 i

1 d

CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLAhT i

l l

Technical Report Supponing Cycle 18 Operation l

l Nonheast Utilities P.O. Box 270 Hartford, Connecticut s

Copy No. 2Y e_

9307070253 930629 PDR ADOCK 0500 3

P

List of Effective Pages i

Eagt Revision Date -

Eage Revision Date Title May 1993 7-1 May 1993 -

i.

June 1993 l

7-2 May 1993 ii.

May 1993 8-1 June 1993 iii.

May 1993 8-2 May 1993 1-1 May 1993 8-3 May 1993 2-1 May 1993 8-4 May 1993 3-1 May 1993 8-5

- May 1993 3-2 May 1993 9-1 May 1993

? -3 May 1993 10-1 May 1993 3-4 May 1993 10-2 May 1993 3-5 May 1993 -

4-1 May 1993 4-2 May 1993 4-3 May 1993

- t 5-1 May 1993 5-2 May 1993 5-3 May 1993 5-4 May 1993 6-1 May 1993 a

Controlled Distribution Name Cony Number M. P. Bain 1-5 l

M. S. Kai 6-10 J. R. Guerci 11-15 M. F. Bray 16-20 C. II. Wu 21-23 E. A. Perkins 24-25 i

1.

Revision 1 i

June 1993 I

i TABLE OF CONTEhTS 1

a Page 1.

Introduction and Summary

.................................... 11 2.

Operating History.

2-1 3.

General Description

............................................ 3 1 4.

Fuel System Design..

4-1 e

d 5.

NuclearDesign --

............................................... 5-1 6.

Thermal Hydraulic Design................

..................................... 6 1 f

7.

Accident and Transient Analysis.....

7-1 8.

Core Operating Limits

.......................................... 8-1 9.

Stanup Program - Physics Testing....................................................... 9 1

(

l

?

10. Re ferences....

....................................................... 10-1 List of Tables I,

Table 3-1. Cycle 17 Discharge Fuel....

.................................................. 3-2 l

1 5

4-1. Nominal Fuel Design Parameters..

............................................. 4-3 e

I 5-1. Haddam Neck Plant Physics Parameters................................................... 5-2 t

i 5-2. IhMnm Neck Plant Cycle 18 Shutdown Margin...................................... 5-3 i

)

..n.

Revision 0 f

May 1993 i

t i

I t

0 0

n.

List of Ficures Eigint P.agn 3-1. FaMam Neck Plant Cycle 18 Care I.cading Pattern.................... 3-3 3-2. Faddam Neck Plant BOC18 Bumup Distribution (MWD /MTU)............ 3-4 1

3-3. HaMam Neck Plant Cycle 18 Control Rod 1.ocations........................... 3-5 5-1. FnMam Neck Plant Cycle 18 Relative Power Distribution at t

150 MWD /MTU, HFP, ARO.

................................... 5-4 e

f' 8-1. Cycle 18 Control Group Insertion Limits f

Operating (Technical Specification 3.1.3.6.1 )...................................... 8 - 3 1

8-2. Cycle 18 Axial Offset Limits 250 EFPD Operating (Technical Specification 3.2.1.1 )......................................... 8 -4 8-3. Cycle 18 Axial Offset Limits - 250 EFPD - EOL Operating (Technical Specification 3.2.1.1 )..........................................

8 5 iii.

Revision 0 May 1993

. _ _ - _- a

.. = - _

~

1. Introduction and Summary b

The objective of this report is to support the operation of the eighteenth cycle of the Haddam Neck Plant at its licensed core power level of 1825 MWt.

t i

Included are the analyses outlined in the USNRC document, " Guidance for Proposed License Amendments Related to Refueling". The Cycle 18 core will consist of l

approximately two-thirds Zircaloy-clad fuel assemblies and one-third stainless steel-clad fuel assemblies. Wherever possible, references are made to previously supplied q

analyses supponing the two fuel assembly designs.

This report includes the Cycle 18 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 l

Technical Specification 6.9 reporting requirements for the Technical Report

.l Supponing Cycle Operation.

l The nominal 13,700 MWD /MTU (449 Effective Full Power Days) Cycle 18, based on an assumed Cycle 17 bumup of 11,500 MWD /MTU (394 Effective Full Power Days),

is scheduled to begin in July 1993. Thc 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 a Cycle 16 burnup window of 11,500-j 13,000 MWD /MTU. The Core Operating Limits in Section 8 are also based on this burnup window.

I Based on the analyses performed and review of the proposed revisions to Technical l

Specifications and Core Operating Limits,it is concluded that the Haddam Neck Plant can be operated safely at the licensed thermal power level of 1825 MWt for Cycle 18.

i 1

S 4

I i

2 l

1-1 l

1 Revision 0 May 1993 i

~

e

2. Operating History Initial criticality for Cycle 17 occurred on March 11,1992. The plant phased online March 15,1992 and reached 100% power on March 27,1992. Cycle 17 operation was completed on May 15,1993.

The Cycle 17 core is successfully operating with 60 fuel assemblies that were reconstituted during the refueling outages at the end of Cycles 15 and 16. Evaluations of radiochemistry data indicated that there were less than ten fuel rod failures in the Cycle 17 core in high burnup fuel that will be discharged. No other operating anomalies occurred during Cycle 17 that would adversely affect fuel performance.

i l

l i

l i

!1 i

l!

~ ;

I 2-1 Revision 0 i

May 1993 j

.J.-

i 3.GeneralDescription The reactor core of the Haddam Neck Plant is described in Chapter 4 of the Updated Final Safety Analysis Report (Reference 1). The Cycle 18 core consists of 157 fuel assemblies, each of which is a 15 by 15 array containing 204 fuel rods,20 control' rod i

thimble tubes and one incore instrument sheath. The fuel rod cladding for the 52 fresh and 48 once-burned fuel assemblies is Zircaloy-4. The cladding outside diameter of 0.422 inch is identical to the stainless steel-clad design. The nominal wall thickness of the Zirealoy cladding is 0.027 inch, compared to a wall thickness of 0.0165 inch for the stainless steel cladding. All fuel pellets are bevelled, dish-end uranium dioxide.

The fuel pellets for the Zircaloy-clad rod have a 0.361 inch diameter and a 0.425 inch length compared to 0.3825 inch and 0.458 inch respectively for the stainless steel-clad l

rod. All Cycle 18 stainless steel-clad fuel assemblies have a nominal uranium loading of 411.5 kgU and an undensifled nominal active fuel length of 120.5 inches. The Zircaloy-clad fuel assemblies have a nominal uranium loading 363.8 kgU and an undensified nominal active fuel length of 118.575 inches. The minimum batch theomtical density is 94.9 percent for all Cycle 18 fuel batches.

Figure 3-1 is the core loading diagram for Cycle 18 of the Haddam Neck Plant. The nominalinitial enrichment for stainless steel-clad fuel Batches 16C,17B and 18A is 4.00 weight percent uranium-235. The 32 Batch 19A and the 40 fresh Batch 20A fuel j

assemblies have a nominal enrichment of 3.90 weight percent. The 16 Batch 19B and 12 fresh Batch 20B fuel assemblies have a nominal enrichment of 3.60 weight percent.

The end of Cycle 17 discharge of 53 fuel assemblies are from Batches 16B,17A and 18B (Table 3-1). The 4 Batch 17B,52 Batch 18A,32 Batch 19A and 16 Batch 19B fuel assemblies will be shuffled to new locations for the Cycle 18 core. Fuel assembly l'

S22 (Batch 16C), a twice bumed recovered donor fuel assembly, will be inserted for the center fuel assembly. The 52 fresh Zircaloy-clad fuel assemblies (Batches 20A and 20B) will occupy the periphery of the core. Figure 3-2 is a quarter-core map showing the fuel assembly bumup distribution at the beginning of Cycle 18.

Reactivity control is supplied by 45 full-length Ag-In-Cd control rods and by soluble boron shim. The Cycle 18 locations of the 45 control rods and group designations are shown in Figure 3-3.

3-1 Revision 0 May 1993

f y

Table 3-1. Cycle 17DischargedFuel T

No. of Fuel Batch Assemblies Cveles Burned 16B 1

3 17A 48 3

t 18B 4

2 t:

53 i

Total discharged

=

i f

i i

d l

!i i

t i

i r

i A

k 3-2 Revision 0 May 1993

.. ~...,.

h l

-1 Figure 3-1. Haddam Neck Plant Cycle 18 Core leading Pattem 15 14 13 12 11 10 9

8 7

6 5

4-3 2

1 20A 18A 20A R

t 20A 20A 20A IBA 20A 20A 20A P

20A 20B 18A 18A 20B 18A 18A 20B 20A N

20A 19B 19B 19A 18A 18A 18A 19A 19B 19B 20A M

!OA 20B 19B 18 4

. 19A 19A 17B 19A 19A

~18A 19B 20B 20A L

20A 18A 19A 19A 18A 19A IBA 19A 18A 19A 19A 18A 20A K

r 20A 20A 18A 18A 19A 19A 18A 19B 18A 19A 19A 18A IBA 20A 20A 3

f 4

H 18A IBA 20B 18A 17B 18A 19B 16C 19B 18A 17B 18A 20B 18A 18A I

20A 20A IBA 18A 19A 19A 18A 19B 18A 19A.

19A IBA 18A 20A 20^

G 20A ISA 19A 19A 18A 19A 18A 19A 18A 19A 19A 18A 20A F

20A 20B 19B 18A 19A 19A 178 19A 19A 18A 198 20B 20A E

20A 19B 19B 19A 18A 18A 18A 19A 19B 19B 20A D

]

20A 20B 18A 18A 20B 18A 18A 20B 20A C

t 20A 20A 20A 18A 20A 20A 20A B

l l

20A 18A 20A A

~

i Batch

1. Assemblics initial w/o U235 l

16C 1

4.00 a

17B 4

4.00 Revision 0 18A 52 4.00 May 1993 19A 32 3.90 19B 16 3.60 20A 40 3.90 20B 12 3.60 3-3

+r e

e.-+-

av w-i ww-1 e

w'

- ='

wr a-e 1e m

Figure 3-2. FaMam Neck Plant Beginning Of Cycle 18 Bumup Distribution, AnVD/MTU 8

7 6

5 4

3 2

1 l

i 3

26010 14563 22863 17070 23772 0.

24292 18185 G

14010 18056 8392 10450 1990c 20968 0.

O.

f p

23686 8382 24105 12555 9624 21899 0.

E 17063 10400 12566 23566 13839 0.

O.

l D

23307 19970 9609 13821 14253 0.

l C

0.

21019 21925 0.

O.

l 23774 0.

O.

O.

B l

l A

18927 0.

i l

1 3-4 Resision 0 May 1993 l

t Figure 3-3. Haddam Neck Plant Cycle 18 Control Rod Locations j

i

)

i 15 14 13 12 11 10 9

8 7

6 5

4 3

2 1

Called Nonh R

A P

42 D

A A

D g

41 29 22 34 B

C B

C B

i M

33 21 10 14 30 b

l D

D b

i 40 35 C

D A

D C

g 20 9

2 6

15 A

A y

23 23 A

B A

A A

B A

g 45 13 5

1 3

11 43 A

A g

27 24 C

D A

D C

p 19 8

4 7

16 D

3 E

39 36 B

C B

C B

D 32 18 12 17 31 D

A A

D C

38 26 25 37 i.

A g

i 44 A

NO. OP BANK RODS FUNCllON ImEND e

B B

CONTROL h

h W

ROD B ANK DESIGNATION S

'N UU ROD IDCATION NIMBER C

_B SHLTfDOWN 45 i

3-5 Revision 0 i

May 1993

4. Fuel System Design The Cycle 18 core consists of reinserted fuel assemblies of Batches 16C,17B,18A, 19A, and 19B, and the fresh fuel assemblies of Batches 20A and 20B. The pertinent design parameters forall seven fuel batches are listed in Table 4-1. All fuel assemblies are mechanir-ally interchangeable and hydraulically similar.

A summary of the mechanical evaluations of the Cycle 18 fuel rods is provided below:

Claddine Collapse The Batch 18A stainless steel-clad fuel rods are the most limiting in terms of creep collapse due to their having the highest previous incore exposure time and bumup. The power histories of the Cycle 18 fuel assemblies were analyzed to determine a bounding power history for creep collapse. This bounding power history was used to analyze a fuel rod operating under conservative conditions for creep collapse. The results of this analysis are applicable to all stainless steel-clad fuel batches. The predicted creep collapse times and burnup exceed the maximum expected residence times and burnups of all fuel batches.

The cladding collapse evaluation for Zircaloy-clad fuel rods provided in Reference 2 demonstrates a creep collapse life of greater than 50,000 MWD /MTU, using an assumed bounding power history. This evaluation was also based on limiting core operational and fuel manufacturing parameters.

Cladding Stress

~Ihe Cycle 18 fuel rods were analyzed by conservative stress analyses following ASME guidelines for pressure vessels. For the design evaluation, the primary membrane stress intensity and any single mss must be less than two-thirds of the mimmum specified unirradiated yield strength of the claddmg. The trargin is in excess of 13% for the stamless seel-clad fuel rods and in excess of 40% for the Zucaloy-clad fuel rods.

Claddine Strain R

The fuel design criteria specify a limit of 1% on cladding plastic tensile circumferential strain. The pellet is designed to assure that cladding plastic strain 4-1 Revision 0 May 1993

-l

is less than 1% at design local pellet bumup and linear heat generation rate. The design bumup and linear heat generation rates are high::r than the limiting values that any of the Cycle 18 fuel batches are expected to experience. The strain analyses are based on the upper tolerance valuer for the fuel pellet diameter and density, and on the lower tolerance value for the cladding inside diameter.

Claddine Faticuj:

Fatigue analyses were performed using conservative conditions to find the cumulative fatigue usage factor. The fatigue usage factor for the Cycle 18 fuel rods was calculated following the ASME Pressure Vessel design code and compared to the maximum allowed factor of 0.9. The cumulative fatigue factor was found to be 0.2 for the stainless steel-clad fuel rods and 0.48 for the Zucaloy-clad fuelmds.

All fuel in Cycle 18 is thermally similar. Analyses for all fuel batches were performed with the TACO 2 code (Reference 3), using the analysis methodology consistent with -

Reference 4. The maximum fuel rod pressure for all Cycle 18 fuel batches will remain below nominal system pressure.

I I

The Zircaloy-clad fresh fuel assemblies (Batches 20A and 20B) are equivalent to the other fuel assemblies in the Cycle 18 core. The stainless steel-clad design has performed well in previous cycles with no adverse materials effects. Zircaloy cladding has been used r'

- ilusively in Pressurized Water Reactors and the behavior of the material is. 31! vn.. rstood in the range of operating conditions (eg. burnup, chemistry,pressurt,

' fattue, etc.) for the Haddam Neck Plant. Thus, all possible fuel-cladding-assembly-c olant material interactions have been proven to have no adverse effects on fuel pe2formance when operated under the conditions expected in Cycle 18.

i 4-2 Revision 0 May 1993

Table 4-1. Fuel Design Parameters Bl1Xh 15C lla 18A 19A (B) 20A (B) 8 Manufacttner B&W B&W B&W B&W B&W Numberof Assemblies 1

4 52 32 (16) 40 (12)

Previous Irradiation, Cycles 2

2 2

1 0

Initial Fuct Ergiciurent, wt%

4.0 4.0 4.0 3.9 (3.6) 3.9 (3.6)

InitialfuelDensity

% Theon:tical(minimum) 94.9 94.9 94.9 95.0 95.0 FuelPellet Diameter, inch 0.3825 0.3825 0.3825 0.361 0.361 i

Active FuelStack, inch 120.5 120.5 120.5 118.575 118.575 l

Cladding Material 3J4-SS 304-SS 304-SS Zr-4 Zr-4 CladdingThickness, inch 0.0165 0.0165 0.0165 0.027 0.027 Fuel Rod Length, inch 126.68 126.68 126.68 126.00

~126.00 Initial Gas Pressure, psia 54.7 54.7 54.7 269.7 269.7 E @2 x

s s-S$

. m, -. -... - _

5. Nuclear Design The physics methodology used to support the Haddam Neck Plant for the Cycle 18 I

reload design is provided in Reference 5. Fifty-two (52) fresh Zircaloy-clad fuel assemblies are required as feed for Cycle 18. The Cycle 18 core will consist of approximately two-thirds (100) Zircaloy-clad fuel assemblies and one-third (57) sininless steel-clad fuel assemblies.

lU The fuel cycle was designed using standard reload design procedures. Table 5-1 provides a summary of the Cycle 18 physics characteristics compared with the current limits based on the reference safety analysis (Reference 6). Table 5-2 shows the e

shutdown margin calculations at beginning and end of cycle conditions. Figure 5-1 illustrates the Cycle 18 relative power distribution at 150 MWD /MTU, hot full power, all rods out and equilibrium xenon conditions. Power distribution and bumup data ll b

presented are based on the end of Cycle 17 burnup of 11,500 MWD /MTU.

i:

i 5-1 Revision 0 May 1993

Table 5-1. Haddam Neck Physics Parameters Current Limit Cycle 18 Cycle Design length, (EFPD) 449 Cycle Design Bumup,(MWD /MTU) 13,700 Average Core Burnup, EOC,(MWD /MTU) 25,444 Design Core Loading,(MTU) 59.87 Most Positive Moderator Temperature 0.0 at HFP 0.0 at HFP l

Coefficient,(pcm/F)

+5.0 H7P

+5.0 at HZP Most Negative Moderator Temperature Coefficient,(pcm/F)

-32 s HFP

-32 s HFP t

Doppler Temperature Coefficient at HFP, (pem/F)

-1.18 to -1.81

-1.18 to -1.81

[

t Delayed Neutron Fmetion, Beff, (%)

0.47 - 0.67 0.47 - 0.67 Maximum Differential Rod Worth 150 150 l

at Subcritical, (pcm/ inch) i t

i i

i i

i

~

5-2 Revision 0 l

May 1993

^

i Table 5-2. FaMam Neck Plant Cycle 18 Shutdown Margin Calculations With Maximum Stuck Rod P

i Hot Full Power. ocm l

Available Rod Wonh HDL EOL l

Totalrod worth less max. stuckrod, HZP 5620 5900 (1) Less 10% uncenainty 5060 5310 i

5 Er auired Rod Worth t

DopplerDefect 1010 940 ModeratorDefect 220 810 RodInsenion Allowance 460 340 Flux Redistribution 260 700 VoidEfrect 50 50 (2) Total RodWonh Required 2000 2840 i

Shutdown Marcin (1)-(2) 3060 2490 1

Recuired Shutdown Marcin

> 2000

> 2000 i

i 4

5-3 Resision 0 May 1993

Figure 5-1. Haddam Neck P1 tnt Cycle 18 Relative Power Distribution At 1.50 MWD /MTU,IIFP, ARO 8

7 6

5 4

3 2

1

))

0.893 0.984 1.056 1.093 0.94 R 1.179 0.858 0.489 G

0.991 1.093 1.251 1.228 1.019 0.980 1.W2 0.556 F

1.N9 1.253 1.106 1.2N 1.191 0.989 0.952 E

1.093 1.229 1.2 m 1.037 1.065 1.141 0.702 D

0.952 1.019 1.191 1.065 0.921 0.757 C

1.179 0.9S0 0.989 1.142 0.758 B

0.862 1.093 0.952 0.702 A

0.486 0.555 l

i 5-4 Revisi 7 0 May IkT

T' 6.ThermalHydraulic Design The thermal hydraulic design evaluation of the Zircaloy-clad fuel assembly and comparisons with the thermal hydraulic performance of the stainless steel-clad fuel N

Ali, Linear assembly were provided in Reference 2. Since the design values for F Heat Generation Rate, Reactor Coolant System flow rate, pressurizer pressure and core inlet temperature are equivalent to, or bound the design values, the steady state minimum DNBR and maximum UO temperature results provided in Reference 2 2

remain applicable.

< i I :

1!

ii l

i 4:

I 6-1 i

Revision 0 -

May 1993 1

o

a

7. Accident andTrarNra Analysis

• Ihe accident and transient analysis design basis was reviewed for potential impact due to changes in the Cycle 18 reload physics parameters contained in the Reload Safety Analysis Checklist (RSAC). The significant changes in the RSAC parameters jvere the inverse boron worth and the maximum critical boron concentration, which affected the Baron Dilution accident. The minor change in the prompt neutron lifetime had no i

significant impact on any accident. The review of the LOCA and non-LOCA transient design bases for Cycle 17 is discussed below:

8 Small Break LOCA The small break LOCA design basis (Reference 7 for stainless steel-clad fuel assemblies and Reference 8 for zircaloy-clad fuel assemblies) was reviewed. These analysis results remain bounding for Cycle 18 operation.

l Laree Break LOCA The large break LOCA design basis (Reference 9 for stainless steel-clad fuel and Reference 10 for zircaloy-clad fuel) was reviewed. These analyses remain bounding for Cycle 18 operation.

Non-LOCA Transients The design basis non-LOCA transients were reviewed due to the RSAC changes for the Cycle 18 design. The boron dilution accident was the only accident that required reanalysis. All other accidents were not significantly impacted by the changes in the RSAC.

Boron Dilotion The Cycle 18 inverse boron worth has decreased as a result of the increased number of zircaloy-clad fuel assemblies in the core. The increase in the amount of zircaloy in the core changes the neutron spectrum. Since there is now a more thermalized neutron spectrum, the boron in the reactor coolant is a more effective i

absorber. Since the inverse boron worth has decreased, any dilution will result in a higher reactivity insertion, and a shorter time to criticality. The shutdown margin requirements for MODES 1-5 have increased. The shutdown margin requirements for MODES 1,2, and 3 (four loops operating) have increased from i

1800 to 2000 pcm. The MODE 3 (less than four loops operating) shutdown margin requirement has increased from 2600 to 2800 pcm. The shutdown margin requirements for MODES 4 and 5 have increased from 3100 to 3400 7-1 Revision 0 May 1993 l

pcm. The higher shutdown margin requuements preserve the operator action time of 15 minutes.

Based on the analyses performed and review of the proposed revisions to Technical Specifications and Core Operating Limits,it is concluded that the Haddam Neck Plant can be operated safely at the licensed thermal power level of 1825 MWt for Cycle 18.

s t

1 P

t t

7-2 Revision 0 May 1993 i.

[

8.0 Core Operating Limits i

The Cycle 18 Core Operating Limits have been developed using the methodologies approved by the NRC in References 8 and 10-17. The operating limits have been determined so that all applicable design limits (e.g. fuel and core thermal-hydraulic I

limits, ECCS limits, nuclear limits and transient and accident analysis limits) of the safety analysis are met. There were no changes in the Operating Limits for Cycle 18.

'Ihe following Core Operating Limits have been established for Cycle 18 i

Moderator Temnerature Coefficient Limit (Technical Soccification 3.1.1.5)

The Moderator Temperatum 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 condi ion; and t

i t

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

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

P I

Linear IIcat Generaton Rate I imits j

IIechnical Soecification 3.2.2.1)

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

9 Stainless Steel Zircaloy Cladding Cladding

a. Less than 250 EFPD 13.7 kW/ft 14.5 kW/ft l

b. Greater than 250 EFPD but less than END-OF-CORE LIFE 13.7 kW/ft 14.5 kW/ft t

8-1 Revision 1 June 1993

l

)

Nuclear Enthalov Rise Hot Channel Factor (Technlent Snecification 313.1) 1 The NUCLEAR ENTHALPY RISE HOT CHANNEL FACIOR, FAH, shall f

N be hmiteA by the following mlationship:

N i

AH g 1,c,0 [ 1.0 + 0.3 (1-P) ]

F I

whem: P = Thermal Power / Rated Thermal Power l

'i I

i r

I 5

i i

i I

t l

i i

i i

I

)

8-2 Revision 0 May 1993

Ficure 8-1. Cvele 18 Control Groun Insertion Limits (Technient Soecification 3.1.3.6.1) 200D

!f f

i 1

il 6

i.

(240.1625) 3750 Il

i

/I I

II I

l i

fi l

i V

I

'lll 1

i t

ll t

!! /e iiii 11 l ill Ifi ll, iifl l

i ilI

/;l ili ilit 1500 l'

I I

'II l

Ill/

liI IIII

'III

!!I4 l lii

!/

sisw.1473) t i

il i

l (l

l/

il til I

i i

i

/ i lli iIi ll1 I

i i

I fi l l

l ti lill i

1250

!iI I

I I

i

'l Iiii I'l i!i b

-y i

il i

i i

I l/ I l

!I i !i iii' I

!l

/

I i

i

!*I' e

i

/!

L 1

14 iI t: I!

4 y

l

/I I j!

l

!l'l l

E 1000 I

I

'/'II

II i

i 11 111 lil' I

iI

/II ill i

ll

!i!:

ii!

p i'l

!!jl li /!!

l11 i

ii:

'il i

y si!!

11ll lil/

li I l ll 1: 1

il w

I'r i

1, V, I

til il i

illi il;!

i c

750

!I I

I I

I i

I I/ l I

l II'l II'I lI I

fI!!

I i

l/li il I

fili

(!!t I

t !

eil1 l

j y;ti ii i

!!i

,ji:

1e i

i i

/! i l

I IfI Itti i

l l<

/

lI I

Iil

!ii!

500 1;'

'i

8/

8

' i ii 8

II

/t l'

i

'i1 I

l il!

'?-

i fil l

l l

1ii I

/ II i

/-

l

'Il 250

'i8

/

l l

1i ll

/

/

/

e

/

-1 I

FULLY INSERTED -

20 70 120 170 220 270 320 GROUP B STEPS WITHDRAWN 150 200 250 300 317 - FULLY WITHDRAWN GROUP A STEPS %TTHDRAWN 8-3 Revision 0 May 1993

Ficure 8-2. Creie 18 Arint Offset Limits 250 EFPD (Technical Soecification 3.2.L1) 320 200 lt-15.0.1001 j

g 110.100)

/

\\

/

T

/

\\

\\

/

/

\\

/

(

l 80

/

\\'

l g

/

)

/

(

C

/

\\

o

(-30.701 (20.70)

((

ts 50 a

sa

.J C

$o A-(-30.101 (20.40)

~

40 20 1

0

-40

-20 0

to 40 RXIRL OFFSET t%)

)

l l

8-4 Revision 0 May 1993

u Ficure 8-3. Cycle 18 Arial Offset Limits - 250 EFPD - EOL (Technical Snecification 3.2.1.1) l 320 100 t10.1001 g.20.1001 g

j

/

\\

/

y

/

\\

/

\\

/

)

\\

)

1

\\

80

/

if Y

b

/

(

e

/i

\\

E t-30,701 I

(20,701 I

O t'

60 a

w I

>>w

.J C

E O

8-(-30.403 (20.401

[

40 20 0

-40

-20 0

20 to RXIRL OFFSET (%)

8-5 Revision 0 May 1993

- +

,u..

~. -. -. -...

9.0 Startup Program - Physics Testing i

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 safe plant operation. -

Prr-Critical Tests I

1. Control rod drag test n

2. Hot controlrod dmp-time testing.

Zero-PowerTests

1. Critical boron concentration

2. Temperature reactivity coefficient

a. All rods out

b. Banks B, A,and D inserted

3. Control rod group worths for Banks B, A, and D l

Power Tests j

1. Core power distribution mapping at 580 and 100% full power, normal contml bank configuration.

2. Excore / incere correlation verification.

i 0

e I

k a

r 9-1 Revision 0 l

4 May 1993 l

i

I

10. References

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

2. E. J. Mroczka to USNRC, 'Zircaloy. Clad Fuel Mechanical Design Repon, NUSCO-166", December 4,1989.

3. Y. H. Hsii, et al., TACO 2 - Fuel Pin Performance Analysis. B AW-10141-PA.

Babcock & Wilcox Company, Lynchburg, Virginia.

4. J. H. Taylor (B&W) to J.S. Berggren (NRC), Letter, "B&W's Responses to L

TACO 2 Questions, April 8,1982.

l-

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

6. Nonheast Utilities Service Company, Haddam Neck Plant - Reanalysis of Non-LOCA Design Basis Accidents, NUSCO-151, June 30,1986.

1

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

8. E. J. Mroczka to USNRC, Zircaloy Clad Fuel Topical Repon "Small Break LOCA Analysis - Zircaloy Clad Fuel, NUSCO-163, December 1988", December 30,1988.

I

9. E. J. Mroczka to USNRC, Revised Technical Repon Supporting Cycle Operations, " Technical Repon Supponing Cycle 16 Operation, NUSCO-167.

Revision 1. April 1990", May 9,1990.

4 10.E. J. Mroczka to USNRC, Large Break Loss of Coolant Accident Analysis -

Zircaloy Conversion, " Application of the WCOBRA/ TRAC Best Estimate UPI Model to the Haddam Neck PWR, WCAP-12766, November 1990", November 30,1990.

i 10-1 Revision 0 May 1993

11.F. M. Akstulewicz to E. J. Mroczka, Review of NUSCO Topical Report on Physics Methodology for PWR Reload Design, NUSCO-152. August 3,1987.

12. A. B. Wang to E. J. Mroczka, Safety Evaluation for Nonheast Utilities Topical Repon 140-1, "NUSCO Thermal Hydraulic Qualification, Volume I (RETRAN),

July 26,1988.

13.F. M. Akstulewicz to J. F. Opeka, "NUSCO Thermal Hydraulic Model Qualification, Volume II (VIPRE)", Topical Repon NUSCO 140-2, October 16, 1986.

i'

14. A. B. Wang to E. J. Mroczka, Safety Evaluation of Nonheast Utilities Topical Repon 151, "Haddam Neck Non-LOCA Transient Analysis", October 18,1988.

15. Supplement to the Safety Evaluation by the Directorate of Licensing, U.S. Atomic Energy Commission, Docket No. 50-213, Connecticut Yankee Atomic Power Company, Haddam Neck Plant, December 27,1974.

16. A. B. Wang to E. 3. Mroczka, Safety Evaluation for NULAP5 Code and its Use in Haddam Neck Small Break LOCA Analyses (NUREG-0737 Items II.K.3.30 and II.K.3.31), August 3,1988.

17. A. C. Thadani to W. J. Johnson, Acceptance for Referencing of Licensing Topical Report WCAP-10924, " Westinghouse Large Break LOCA Best Estimate Methodology", August 29,1988.

i h

i i

i i

I 10-2 Revision 0 May 1993