atachment iii to indian point 3 final report ...no. mps-90-494, "indian point unit 3 draft final...

ATACHMENT III TO IPN-90-046

INDIAN POINT 3 FINAL REPORT ON

APPENDIX G REACTOR VESSEL PRESSURE

TEMPERATURE LIMITS

NEW YORK POWER AUTHORITY INDIAN POINT 3 NUCLEAR POWER PLANT

DOCKET NO. 50-286 DPR-64

,PD ," p A "I 05000286 p:rL ' ". . FIlC' -A

ALiOf 1" ASEA BROWN BOVERI

July 24, 1990 MPS-90-688

Mr. A. Decker New York Power Authority 123 Main Street White Plains, NY 10601

Subject: INDIAN POINT UNIT 3 FINAL REPORT ON APPENDIX G REACTOR

VESSEL PRESSURE-TEMPERATURE LIMITS

Reference: (1) Contract Change Order No. 2 for NYPA Agreement No. 029437-89.

(2) P. R. Kottas (CE) to Michele Ramagnoulo (NYPA), Letter PAS-90-010, "Additional Pressure Temperature Limits for Indian Point No. 3 Nuclear Power Plant," dated March 30, 1990.

(3) P. R. Kottas (CE) to Michele Ramagnoulo (NYPA), Letter No. PAS-90-011, "Pressure-Temperature Limits for Indian Point Unit 3 Nuclear Power Plant," dated April 3, 1990.

(4) C. D. Stewart (CE).to A. Decker (NYPA), Letter No. MPS-90-494, "Indian Point Unit 3 Draft Final Report on Reactor Vessel Pressure-Temperature Limits", dated May 30, 199.0.

Attachment: (1) Final Report On Pressure-Temperature Limits for Indian Point Unit 3 Nuclear Plant, July 1990.

Dear Mr. Decker:

Please find enclosed the final report (seven copies) documenting the

pressure-temperature (P-T) limits for the reactor vessel beltline

region and Low Temperature Overpressure Protection (LTOP) enable

temperatures. This report incorporates the comments made by the New

York Power Authority staff.

These limits have been calculated in accordance with 10 CRF 50 Appendix G requirements as supplemented by ASME Code Section III

Appendix G recommendations. The LTOP enable temperatures have been

determined in accordance with USNRC guidelines. The P-T limits and

LTOP enable temperatures have been independently reviewed in

accordance with CE's Quality Assurance Procedures Manual.

ABB Combustion Engineering Nuclear Power

Combustion" Engineering. Inc. 1000 Prospect Hill Road Telephone (203) 688-1911 Post Office Box 500 Fax (203) 285-9512 Windsor. Connecticut 06095-0500 Telex 99297 COMBEN WSOR

It has been a pleasure working with New York Power Authority on this reactor vessel integrity task. If there are any questions or comments regarding the attached report or if we can assist you in resolving other reactor vessel integrity issues, please feel free to contact Mr. Paul Hijeck, Supervisor, Reactor Vessel Integrity at (203) 285-3115 or the undersignedat (203) 285-2294.

. Sincerely,

COMBUSTION ENGINEERING, INC.

Craig D. Stewart

Nuclear Engineer

CDS/prr.

Enclosure

cc: F. Gumble, w/o enc. P. J. Hijeck, w/enc. K. Jacobs, w/o enc. P. R. Kottas, w/o enc. M. S. McDonald, w/o enc.

CDS008

ATTACHMENT 1

1306.doc(9033)/ch-1

FINAL REPORT

ON

PRESSURE-TEMPERATURE LIMITS FOR INDIAN POINT

UNIT 3 NUCLEAR POWER PLANT

Prepared For:

NEW YORK POWER AUTHORITY

123 MAIN STREET

WHITE PLAINS, NEW YORK 10601

By:

ABB COMBUSTION ENGINEERING NUCLEAR POWER

COMBUSTION ENGINEERING, INC.

REACTOR VESSEL INTEGRITY GROUP

1000 PROSPECT HILL ROAD

WINDSOR, CONNECTICUT 06095-0500

July 1990

1306.doc(9033)/ch-2

TABLE OF CONTENTS

SECTION TITLE PAGE

1.0 INTRODUCTION 10

2.0 ADJUSTED REFERENCE TEMPERATURE 11 PROJECTIONS

3.0 GENERAL APPROACH.FOR CALCULATING 21 PRESSURE-TEMPERATURE LIMITS

4.0 THERMAL ANALYSIS METHODOLOGY 23

5.0 COOLDOWN LIMIT ANALYSIS 25

6.0 HEATUP LIMIT ANALYSIS 27

7.0 HYDROSTATIC TEST AND CORE CRITICAL 30 LIMIT ANALYSIS

8.0 LTOP ENABLE TEMPERATURES 31

9.0 DATA 33

REFERENCES 34

APPENDIX A INDIAN POINT UNIT 3 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM SOURCE DATA - PLATE B2803-3 (TRANSVERSE)

Page 2

1306.doc(9033)/ch-3

LIST OF TABLES

NO. TITLE PAGE

1. Indian Point Unit 3 Reactor Vessel Beltline 36 Materials

2. Adjusted Reference Temperature (1/4 t) Calculation 37 for Indian Point Unit 3

3. Chemistry Factor Derivation Plant B2803-3 38 (Transverse)

4. Project Peak Neutron Fluence at Vessel Base 39 Metal-Clad Interface

5. Adjusted Reference Temperature Projections for 40 Indian Point Unit 3

6. Indian Point Unit 3 Cooldown P-T Limits 41 9 EFPY Without Correction Factors

7. Indian Point Unit 3 20-30°F/HR Heatup P-T Limits 42 9 EFPY Without Correction Factors


9. Indian Point Unit 3 60°F/HR.Heatup P-T Limits 44 9 EFPY Without Correction Factors

10. Indian Point Unit 3 Cooldown P-T Limits 45 9 EFPY With Correction Factors

11. Indian Point Unit 3 20-30°F/HR Heatup P-T Limits 46 9 EFPY With Correction Factors


13. Indian Point Unit 3 60°F/HR Heatup P-T Limits 48 9 EFPY With Correction Factors




Page 3

1306.doc(9033)/ch-4

LIST OF TABLES (Continued)

NO. TITLE PAGE

17. Indian Point Unit 3 60"F/HR Heatup P-T Limits 52 11 EFPY Without Correction Factors


19. Indian Point Unit 3 20-30"F/HR Heatup P-T Limits 54 11 EFPY With Correction Factors

20. Indian Point Unit 3 40-50*F/HR Heatup P-T Limits 55 11 EFPY With Correction Factors

21. Indian Point Unit 3 60"F/HR Heatup P-T Limits 56 11 EFPY With Correction Factors


23. Indian Point Unit 3 20-30"F/HR Heatup P-T Limits 58 13 EFPY Without Correction Factors


25. Indian Point Unit 3 60"F/HR Heatup P-T Limits 60 13 EFPY Without Correction Factors




29. Indian Point Unit 3 60"F/HR Heatup P-T Limits 64 13 EFPY With Correction Factors




Page 4

1306.doc(9033)/ch-5

LIST OF TABLES (Continued)

NO. TITLE PAGE

33. Indian Point Unit 3 60°F/HR Heatup P-T Limits 68 15 EFPY Without Correction Factors








41. Indian Point Unit 3 60°F/HR Heatup P-T Limits 76 32 EFPY Without Correction Factors





Page 5

1306.doc(9033)/ch-6

LIST OF FIGURES

DESCRIPTION

ASME Reference Fracture Toughness, KIR, Curve

Indian Point Unit 3 Cooldown P-T Limits 9 EFPY, Correction Factors: None

Indian Point Unit 3 20°F/HR Heatup 9 EFPY, Correction Factors: None





P-T Limits

P-T Limits

P-T Limits

P-T Limits

P-T Limits

Indian Point Unit 3 Cooldown P-T Limits 9 EFPY With Correction Factors

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.

14.

15.

16.

P-T Limits

P-T Limits

P-T Limits

P-T Limits

P-T Limits

Indian Point Unit 3 Cooldown P-T Limits 11 EFPY, Correction Factors: None I

Indian Point Unit 3 20°F/HR Heatup P-T Limits 11 EFPY, Correction Factors: None


PAGE

81

82

Indian Point Unit 3 20°F/HR Heatup 9 EFPY With Correction Factors

Indian Point Unit 3 30°F/HR heatup 9 EFPY With Correction Factors




Page 6

1306.doc(9033)/ch-7

NO.

17.

18.

19.

20.

21.

22.

23.

24.

25.

Indian Point Unit 3 13 EFPY, Correction


Cooldown P-T Limits Factors: None

20°F/HR Heatup P-T Limits Factors: None





P-T Limits

P-T Limits

P-T Limits

60°F/HR Heatup P-T Limits Factors: None

LIST OF FIGURES (Continued)

DESCRIPTION





Indian Point Unit 3 20°F/HR Heatup P-T Limits 11 EFPY With Correction Factors





PAGE

97

98

99

100

101

102

103

104

105

Page 7

1306.doc(9033)/ch-8


Indian Point 13 EFPY With






DESCRIPTION

Unit 3 Cooldown P-T Limits Correction Factors

Unit 3 20°F/HR Heatup P-T Correction Factors











Cooldown P-T Limits Factors: None

20°F/HR Heatup P-T Factors: None





PAGE

112

Limits

Limits

Limits

Limits

Limits

Limits

Limits

Limits

Limits

Limits




38.

39.

40.

41.

42.

43.

44.

45.

46.

120

Page 8

1306.doc(9033)/ch-9


DESCRIPTION




Indian Point Unit 3 Cooldown P-T Limits 32 EFPY, Correction Factors: None


200F/HR Heatup P-T Limits Factors: None





60°F/HR Heatup Factors: None

P-T Limits

P-T Limits

P-T Limits

P-T Limits







PAGE

56.

57.

58.

59.

60.

61.

136

137

138

139

140

141

Page 9

1306.doc(9033)/ch-10

1.0 INTRODUCTION

The following sections describe the basis for development of reactor vessel beltline pressure-temperature limitations for the Indian Point Unit 3 Nuclear Generating Station. These limits are calculated to meet the regulations of 10 CFR Part 50 AppendixA,' Design Criterion 14 and Design Criterion 31. These design criteria required that the reactor coolant pressure boundary be designed,

fabricated, erected, and tested in order to have an extremely low probability of abnormal leakage, of rapid failure, and of gross rupture. The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, and testing the boundary behaves 'in a non-brittle manner and the probability of rapidly propagating fracture is minimized..

The pressure-temperature limits are developed using the requirements of 10 CFR 50 Appendix G 2 This appendix describes the requirements for developing the pressure-temperature limits and

provides the general basis for these limitations. The margins of safety against fracture provided by the pressure-temperature limits using the requirements of 10 CFR Part 50 Appendix G are equivalent to those recommended in the ASME Boiler and Pressure Vessel Code Section III, Appendix G, "Protection Against Nonductile Failure." (3)

The general guidance provided in those procedures has been utilized to develop the Indian Point Unit 3 pressure-temperature limits with the requisite margins of safety for the heatup and cooldown conditions.

The Reactor Pressure Vessel beltline pressure-temperature limits are based upon the irradiation damage prediction methods of Regulatory

Guide 1.99 Revision 02(4 This methodology has been used to calculate the limiting material Adjusted Reference Temperatures for

Indian Point Unit 3.

Page 10

1306.doc(9033)/ch-11

This report provides reactor vessel beltline pressure-temperature

limits in accordance with 10 CFR 50 Appendix G for five

representative points in the RPV life time corresponding to 9, 11,

13, 15, and 32 Effective Full Power Years (EFPY). The events

...analyzed arethe isothermal, 20, 50, 60, 80 and 100 0F/hr cooldown

conditions and the 20, 30, 40, 50 and 60°F/hr heatup conditions.

These conditions were analyzed to provide a data base of thermal

results for use in establishing Low Temperature Overpressure

Protection (LTOP) enable temperatures. Included in events analyzed

are the inservice hydrostatic test and core crtitical conditions.

Based upon the P-T limit analyses within this report, no limiting

vessel operability issues are anticipated to exist during the 40

calendar year design life of the reactor pressure vessel. However,

based upon the projected Adjusted Reference Temperatures exceeding

the 10 CFR 50.61 PTS Screening Criteria at End of Life, a life

limiting vessel integrity issue may exist.

2.0 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS

2.1 INTRODUCTION

This section provides results of an analysis of the Indian Point

Unit 3 (IP3) reactor pressure vessel materials in accordance with

Regulatory Guide 1.99. (4) The purpose is to establish Adjusted

Reference Temperatures (ART) for the controlling reactor vessel

beltline material for use in developing operating limits for set

periods of time.

This section describes analyses performed following Reference 4.

Section 2.2 establishes the credibility of the IP3 reactor vessel

surveillance program data in order to justify their use in ART

calculations. Section 2.3 presents the basis for the values of the

standard deviation for the initial reference temperature, RTNDT, of

the beltline plate and weld materials. Section 2.4 describes the

Page 11

1306.doc(9033)/ch-12

calculation of the chemistry factor, CF, based on the credible surveillance data. Section 2.5 describes the source of data and method of projecting fast neutron fluence at the vessel inside

surface. Section 2.6 presents the prescription for deriving values of ART following Regulatory Position 2.1(4 ) and the-resultant

predicted values of ART for set periods of time.

Appendix A is a compilation of source data reproduced from the

original report for easy reference.

2.2 CREDIBILITY OF SURVEILLANCE DATA

Regulatory Guide 1.99 (4) presents five criteria by which

surveillance data are judged to be credible; i.e., acceptable for determining adjusted reference temperature (ART) following

Regulatory Position 2.1 of the guide. These criteria are addressed

individually below.

2.2.1 Controlling Material in the Capsule - The surveillance material data

is of most value if the "controlling" material is included among

those indicated in the surveillance capsule. This criterion was

addressed by calculating the ART for each of the beltline materials

following Regulatory Position 1.1(4 ). Table I lists the six plates

and three welds from the Indian Point Unit 3 reactor vessel

beltline. The initial RTNDT, copper content and nickel content from Reference 5 is provided for each material. Table 2 presents the results of ART calculations for each beltline material at the vessel quarter thickness (1/4t) location after 32 Effective Full Power

Years (EFPY) which corresponds to a neutron fluence of 1.506 x 1019

2 n/cm (E>lMev) at the vessel inside surface. The chemistry factor (CF) was obtained using the tabulated values from Reference 4. The

margin was calculated using:

Margin - 2J I2 + o A1

Page 12

1306.doc(9033)/ch-13

where a,, the standard deviation for the initial RTNDT , was taken as

O°F for measured values of plates and welds and 17'F for generic

values of initial RTNDT for submerged arc welds. (6 ) a., the standard

deviation for ARTNDT, was taken as 17°F for plates and 28°F for

welds. Adjusted Reference Temperatures (ART),were then calculated

in accordance with Regulatory Position 1.1(4 )

The Indian Point Unit 3 material exhibiting the highest ART is plate

B2803-3, and is, therefore, the controlling beltline material. This

plate is included in the surveillance capsules, (7) thus satisfying

the first credibility criterion.

2.2.2 Minimal Scatter in Charpy Test Results - Charpy impact tests are

performed before and after irradiation to develop an average curve

of impact energy versus temperature from which values of the

30-foot-pound index temperature and upper-shelf energy are obtained.

The variation of the individual data points about the mean curve

(i.e., scatter) "should be small enough to permit the determination

of the 30-foot-pound temperature and the upper-shelf energy

unambiguously." (4 )

The pre- and post-irradiation test results for the Indian Point

Unit 3 surveillance materials (7-1 0 ) were reviewed, and no

significant scatter in Charpy impact test results was observed.

Each data set was adequate for extracting reasonable values from the

mean curve, thus satisfying the seconds credibility criterion.

2.2.3 Scatter About Best-Fit Curve Less Than a. - This criterion provides

a means for judging whether the trend exhibited by the irradiated

materials is consistent with other similar vessel materials and is

within acceptable limits. Two or more irradiated data points are

available (8-1 0 ) for three surveillance materials from Indian Point

Unit 3. Each set was evaluated using Regulatory Position 2.1(4 )

(In the case of Capsule T data, an updated ("I) fluence of

Page 13

1306.doc(9033)/ch-14

3.226 x 1018 n/cm 2 was employed versus the originally reported

value (8 ) of 2.92 x i018 n/cm2 . The Reference 11 fluence update was

performed by the Hanford Engineering Development Laboratory for the

U. S. Nuclear Regulatory Commission as part of the Light Water

Reactor Pressure Vessel Surveillance Dosimetry Improvement Program.)

The computed chemistry factor (CF) and the maximum difference

between predicted and actual shift for each set is as follows:

Material CF Maximum Difference

Plate B2803-3 (transverse) 158.7 9F

Plate B2803-3 (longitudinal) 176.87 150 F

Surveillance Weld 205.3 130F

In each case, the maximum difference between the predicted and

actual shift is less than 17°F for base metal and 28°F for weld

metal, thus satisfying the third credibility criterion.

2.2.4 Irradiation Temperature - The surveillance capsule is typically

designed to maintain the temperature of the included specimens close

to that of the coolant inlet temperature; the criterion is +250F

between the encapsulated specimens and the vessel wall at the

cladding/base metal interface. For Indian Point Unit 3, the

temperature monitors did not melt (8- 10 ), demonstrating that the

capsule irradiation temperature did not exceed 579F. The vessel

inlet temperature for the most recent fuel cycles, 539.1OF(1 2),

would approximate the temperature at the vessel cladding-base metal

interface, and is consistent with the results obtained on the

correlations monitor material (see following paragraph); i.e., the

measured shift for the HSST Plate 02 material was above the mean

predicted shift as would be expected for an irradiation temperature

less than 550'F. Therefore, there is indirect evidence that the

capsule irradiation temperature and the vessel temperature were

within 25°F, thus satisfying the intent of the fourth credibility

criterion.

Page 14

1306.doc(9033)/ch-15

2.2.5 Correlation Monitor Material Data - Indian Point Unit 3, Capsule

Y(9) contained specimens from HSST Plate 02. The measured shift was

140°F corresponding to a neutron fluence of 8.05 x 1018 n/cm2

Based on a compilation (13) of similar correlation monitor material

data, the mean predicted shift for Plate 02 is 120°F with a range of

80°F to 145°F. The measured shift, therefore, falls within the

scatter band of the HSST Plate 02 data base, consistent with the

fifth credibility criterion. As noted previously, the measurement

was greater than the mean predicted shift consistent with the lower

irradiation temperature for Indian Point Unit 3 (approximately

539.1°F) versus the nominal 550'F irradiation temperature for the

overall HSST Plate 02 data base.

2.2.6 Conclusion of Credibility Criteria - All five criteria of Regulatory

Guide 1.99, Revision 2(4 ), have been addressed, and the Indian Point

Unit 3 surveillance data have been shown to satisfy those criteria

and, therefore, are credible for use in developing a plant-specific

relationship of RTNDT shift to neutron fluence in accordance with

Regulatory Position 2.1.

2.3 UNCERTAINTY IN INITIAL RTNDT

According to Position 1.1 of Regulatory Guide 1.99, Revision 2(4)

the uncertainty in the value of initial RTNDT is to be estimated

from the precision of test method when a "measured" value of initial

RTNDT is available. RTNDT is derived in accordance with NB3200 of

the ASME Boiler and Pressure Vessel Code, Section III. It involves

both a series of drop weight (ASTM E208) and Charpy impact (ASTM

E23) tests on the material. The RTNDT resulting from these two test

methods of evaluation are conservatively biased. The elements of

this conservatism include:

1) Choice for RTNDT is the higher of NDT or TCV -60°F. The

drop-weight test performed to obtain NDT and a full Charpy

impact curve is developed to obtain TCV for a given material.

Page 15

1306.doc(9033)/ch-16

The combination of the two test methods gives protection against the possibility of errors in conducting either test and, with the full Charpy curve, demonstrates that the material is typical of reactor pressure vessel steel. Choice of the more-conservative of the two (i.e., the higher of NDTT or TCV - 60°F) assures that tests at temperatures above the reference temperature will yield increasing values of toughness, and verifies the temperature dependence of the fracture toughness implicit in the KIR curve (ASME Code,

Section III, Appendix G).

2) Selection of the most adverse Charpy results for TCV. In accordance with NB2300, a temperature, TCV, is established at which three Charpy specimens exhibit at least 35 mils lateral expansion and not less than 50 ft-lb absorbed energy. The three specimens will typically exhibit a range of lateral expansion and absorbed energy consistent with the variables inherent in the test: specimen temperature, testing equipment, operator, and test specimen (e.g., dimensional tolerance and material homogeneity). All of these variables are controlled using process and procedural controls, calibration and operator training, and they are conservatively bounded by using the lowest measurement of the three specimens. Furthermore, two related criteria are used, lateral expansion and absorbed energy, where consistency between the two measurements provides further assurance that they are realistic and the material will exhibit the intended strength, ductility and toughness implicit in the KIR curve.

3) Inherent conservatism in the protocol used in performing the drop-weight test. The drop-weight test procedure was carefully designed to assure attainment of explicit values of deflection

and stress concentration, eliminating a specific need to account for below nominal test conditions and thereby

guaranteeing a conservative direction of these uncertainty

Page 16

1306.doc(9033)/ch-17

components. In addition, the test protocol calls for

decreasing temperature until the first failure is encountered,

followed by increasing the test temperature 10F above the

point where the last failure is encountered. This in fact

assures that one has biased the resulting estimate toward a low

failure probability region of the temperature versus failure

rate function diagrammed below. The effect of this protocol is

to conservatively accommodate the integrated uncertainty

components.

-------- -

Given the three elements of conservatism described above,

values of initial RTNDT obtained in accordance with NB2300 will

result in a conservative measure of the reference temperature.

The conservative bias of the NB2300 methodology and the

drop-weight test protocol essentially eliminate the uncertainty

which might result from the precision of an individual

drop-weight or Charpy impact test. Therefore, when measured

values of RTNDT are available, the estimate of uncertainty in

initial RTNDT is taken as zero.

Page 17

=77-

1306.doc(9033)/ch-18

2.4 CHEMISTRY FACTOR DERIVATION

Regulatory Guide 1.99 (4 ), Regulatory Position 2.1 provides a

procedure for calculating the Chemistry Factor (CF) given the

availability of three sets of credible surveillance data for the

controlling beltline material, Plate B2803-3. The irradiation data

for the transverse orientation Charpy specimens are detailed in

Table 3. The Capsule T shift measurement is from Reference 8 and

the neutron fluence is the updated value from Reference 11. The

Capsule Y and Z shift measurements and neutron fluence are from

References 9 and 10, respectively. Derivation of the Chemistry

Factor from the Table 3 data is also shown. Each ARTNDT is

multiplied by its corresponding fluence factor, and the products are

summed and divided by the sum of the squares of the fluence factors.

The resulting value is the Chemistry Factor which is then used to

predict RTNDT shift for specific time periods and vessel locations

as described in Section 2.6.

2.5 FLUENCE CALCULATION

Values of neutron fluence were calculated for the reactor vessel

base metal-clad interface at the peak flux position for operation

beyond Cycle 5. The basis was an assessment of the vessel fast

neutron exposure performed as part of the Capsule Z analysis (I0 )

The assumption was made that operation beyond Cycle 5 would be done

using the same core power distribution as Cycle 5.

The peak fluence (base metal-clad interface and 450 azimuth) at the

end of Cycle 5 was computed to be 3.13 x 1018 n/cm 2 (E>lMev),

corresponding to 5.55 Effective Full Power Years (EFPY) of

operation. The calculated fast neutron exposure rate at that same

location for Cycle 5 was 1.43 x 1010 n/cm2 .sec. Taking the neutron

fluence at 5.55 EFPY and the peak exposure rate, projections were

made to 9, 11, 13, 15 and 32 EFPY operation as shown in Table 4.

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1306.doc(9033)/ch-19

(Also indicated in the table is a sample calculation.) The Table 4

fluence projections are used with the derived Chemistry Factor

(previous section) for RTNDT shift predictions as described in

Section 2.6.

2.6 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS

In order to develop pressure-temperature limits for the reactor

vessel, Adjusted Reference Temperatures (ART) for the Indian Point

Unit 3 controlling beltline material are determined. The

controlling material was identified (see Table 2) as lower shell

plate B2803-3, and a chemistry factor of 158.7 was derived based on

three data points from the reactor vessel surveillance program as

described in Section 2.4.

The ART values have been determined using the procedures in

Regulatory Position 2.1 of Regulatory Guide 1.99 (4) The

calculative procedure for the ART values in given by the following

expression:

ART = Initial RTNDT + ARTNDT + Margin (1)

Initial RTNDT is the reference temperature for the controlling

material. 74°F (Plate B2803-3, Table 1) prior to irradiation.

ARTNDT is the mean value of the adjustment in the reference

temperature caused by irradiation and is given by the following

expression:

ARTNDT = (CF) f(O.28 - 0.10 log f) (2)

where:

CF is the derived Chemistry Factor = 158.7 (Table 3), and f is

the neutron fluence in units of 1019 n/cm2.

Page 19

1306.doc(9033)/ch-20

The reference temperature adjustment is calculated using

neutron fluence values corresponding to both the 1/4 thickness

and 3/4 thickness locations using the following expression:

f " fsurf (e -024x (3)

where:

fsurf is the neutron fluence calculated at the vessel base

metal-clad interface, and

x is the depth (in inches) into the vessel wall from

the vessel base metal-clad interface, where vessel

thickness is 8.625 inches.

Margin is the quantity that is added to obtain a conservative upper

bound value of ART, as given in the following expression:

Margin - 2J 2 + 2 (4)

where:

aI is the uncertainty in the initial reference temperature,

taken as OF as discussed in Section 2.3, and

oI Is the uncertainty (standard deviation) for the RTNDT shift prediction.

In accordance with Regulatory Position 2.1(4), o was taken as 8.5"F

(versus 170F) based on the availability of credible surveillance

data for the controlling reactor vessel material. Accordingly,

expression 4 yields the following margin:

Margin - 2(O) + (8.5)2 - 17"F

Page 20

1306.doc(9033)/ch-21

Adjusted reference temperatures were calculated for the controlling beltline material, Plate B2803-3, at the 1/4 and 3/4 thickness locations using the preceding methodology and the projected peak vessel fluence values given in Table 4. The results are given in

Table 5.

3.0 GENERAL APPROACH FOR CALCULATING PRESSURE-TEMPERATURE LIMITS

The analytical procedure for developing reactor vessel

pressure-temperature limits utilizes the methods of Linear Elastic Fracture Mechanics (LEFM) found in the ASME Boiler and Pressure

Vessel Code Section III, Appendix G (Reference 3) in accordance with

the requirements of 10 CFR Part 50 Appendix G (Reference 2). For these analyses, the Mode I (opening mode) stress intensity factors

are used for the solution basis.

The general method utilizes Linear Elastic Fracture Mechanics procedures. Linear Elastic Fracture Mechanics relates the size of a flaw with the allowable loading which precludes crack initiation.

This relation is based upon a mathematical stress analysis of the

beltline material fracture toughness properties as prescribed in

Appendix G to Section III of the ASME Code.

The reactor vessel beltline region is analyzed assuming a semi-elliptical surface flaw oriented in the axial direction with a

depth of one quarter of the reactor vessel beltline thickness and an aspect ratio of one to six. This postulated flaw is analyzed at both the inside diameter location (referred to as the 1/4t location)

and the outside diameter location (referred to as the 3/4t location)

to assure the most limiting condition is achieved. The above flaw geometry and orientation is the maximum postulated defect size

(reference flaw) described in Appendix G to Section III of the ASME

Code.

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1306.doc(9033)/ch-22

At each of the postulated flaw locations the Mode I stress intensity factor, KI, produced by each of the specified loadings is calculated and the summation of the KI values is compared to a reference stress intensity, KIR, which is the critical value of KI for the material andtemperature involved. The result of this method is a relation

of pressure versus temperature for each reactor vessel operating limits which preclude brittle fracture. KIR is obtained from a reference fracture toughness curve for low alloy reactor pressure

vessel steels as defined in Appendix G to Section III of the ASME Code. This governing curve is defined by the following expression:

[.0145(T-ART + 160)]

KIR = 26.78 + 1.223 e

where,

KIR reference stress intensity factor, Ksi-T'n

T temperature at the postulated crack tip, OF

ART adjusted reference nil ductility temperature at

the postulated crack tip, OF

A graphical representation of the KIR curve is shown in Figure 1.

For any instant during the postulated heatup or cooldown, KIR is calculated at the metal temperature at the tip of the flaw, and the value of adjusted reference temperature at that flaw location. Also

for any instant during the heatup or cooldown the temperature

gradients across the reactor vessel wall are calculated (see Section 4.0) and the corresponding thermal stress intensity factor, KIT, is determined. Through the use of superposition, the thermal stress

intensity is subtracted from the available KIR to determine the allowable pressure stress intensity factor and consequently the

allowable pressure.

Page 22

1306.doc(9033)/ch-23

In accordance with the ASME Code Section III Appendix G

requirements, the general equations for determining the allowable

pressure for any assumed rate of temperature change during Service

Level A and B operation are:

2KIM + KIT < KIR

1.SKIM + KIT < KIR (Inservice Hydrostatic Test)

where,

KIM = Allowable pressure stress intensity factor, KsiVtW

KIT = Thermal stress intensity factor, Ksi'i9

KIR = Reference stress intensity, Ksi/T-"

The pressure-temperature limits provided in this report account for

the temperature differential between the reactor vessel base metal

and the reactor coolant bulk fluid temperature. Uncertainties for

instrumentation error, are included in the development of the pressure-temperature limits. Consequently, the P-T limits are

provided on coordinates of pressurizer pressure versus indicated RCS

temperature.

THERMAL ANALYSIS METHODOLOGY

The Mode I thermal stress intensity factor is obtained through a

detailed thermal analysis of the reactor vessel beltline wall using

a computer code. In this code a one dimensional three nodded

isoparametric finite element suitable for one dimensional

axisymmetric radial conduction-convection heat transfer is used.

The vessel wall is divided into 24 elements and an accurate

distribution of temperature as a function of radial location and

Page 23

1306.doc(9033)/ch-24

transient time is calculated. The code utilizes convective boundary

on the inside wall of the vessel and an insulated boundary on the

outside wall of the vessel. Variation of material properties

through the vessel wall are permitted allowing for the change in

material thermal properties-between the cladding and the-base metal.

In general, the temperature distribution through the reactor vessel

wall is governed by a partial differential equation,

aT Kra2T PC K 2 +1 r

subject

outside

to the following boundary

wall surface locations:

At r = r i

At r = r0

conditions at the inside and

K aT = h (T-Tc) arc

where,

p = density, lb/ft3

C = specific heat, btu/lb-°F

K = thermal conductivity, btu/hr-ft-°F

T = vessel wall temperature, OF

r = radius, ft

t = time, hr

h = convective heat transfer coefficient, btu/hr-ft2-°F

Tc = RCS coolant temperature, °F

ri,r o = inside and outside radii of vessel wall, ft

The above is solved numerically using a finite element model to

determine wall temperature as a function of radius, time, and

thermal rate. Thermal stress intensity factors are determined by

the calculated temperature difference through the beltline wall

Page 24

1306.doc(9033)/ch-25

using thermal influence coefficients specifically generated for this purpose. The influence coefficients depend upon geometrical

parameters associated with the maximum postulated defect, and the geometry of the reactor vessel beltline region (i.e., r0/ri, a/c,

...a/t),, along with the assumed unit loading.

The thermal stress intensity factors are determined by the

temperature difference and temperature profile through the beltline wall using thermal influence coefficients and superposition. ASME III Appendix G recognizes the limitations of the method it provides

for calculating KIT because of the assumed temperature profile.

Since a detailed heat transfer analysis results in varying

temperature profiles (and consequently varying thermal stresses), an

alternate method for calculating KIT was employed as required by Article G-2214.3 of Reference 3. The alternate method employed used

a polynomial fit of the temperature profile and superposition using influence coefficients to calculate KIT. The influence coefficients

were calculated using a 2-dimensional finite element model of the

reactor vessel. The influence coefficients were corrected for 3 dimensional effects using the methods of ASTM Special Technical

Publication 677 (Reference 14).

5.0 COOLDOWN LIMIT ANALYSIS

During cooldown, membrane and thermal bending stresses act together

in tension at the reactor vessel inside wall. This results in the

pressure stress intensity factor, KIM, and the thermal stress intensity factor, KIT, acting in unison creating a high stress

intensity. At the reactor vessel outside wall the tensile pressure

stress and the compressive thermal stress act in opposition

resulting in a lower total stress than at the inside wall location.

Also neutron embrittlement, the shift in RTNDT and the associated

reduction in fracture toughness are less severe at the outside wall

compared to the inside wall location. Consequently, the inside flaw

location is more limiting and is analyzed for the cooldown event.

Page 25

1306.doc(9033)/ch-26

Utilizing the material metal temperature and adjusted reference

temperature at the 1/4t location, the reference stress intensity is

determined. From the method provided in Section 4.0, the through

wall temperature gradient is calculated for the assumed cooldown

rate to determine the thermal stress intensity factor. In general,

the thermal stress intensity factors are found using the temperature

difference through the wall as a function of transient time as

described in Section 4.0. They are then subtracted from the

available KIR value to find the allowable pressure stress intensity

factor and consequently the allowable pressure.

The cooldown pressure-temperature curves are thus generated by

calculating the allowable pressure on the reference flaw at the 1/4t

location based upon

K KIR - KIT

where, KIM - 2

KIM Allowable pressure stress intensity as a function of

coolant temperature, Ksi./IW

KIR = Reference stress intensity as a function of coolant

temperature, Ksi.iW

KIT = Thermal stress intensity as a function of coolant

temperature, KsiAin

To develop a composite pressure-temperature limit for the cooldown

event, the isothermal pressure-temperature limit must be calculated.

The isothermal pressure-temperature limit is then compared to the

pressure-temperature limit associated with a cooling rate and the

more restrictive allowable pressure-temperature limit is chosen

resulting in a composite limit curve for the reactor vessel

beltline.

Page 26

1306.doc(9033)/ch-27

Tables 6 through 45 provide the results for the isothermal, 20°F/hr

50°F/hr, 60°F/hr, 80°F/hr and 100°F/hr cooldown pressure-temperature

limits. Table 6 through 9, 14 through 17, 22 through 25, 30 through

33, and 38 through 41 provide pressure-temperature limits without

pressure and temperature instrument uncertainty corrections. Tables

10 through 13, 18 through 21, 26 through 29, 34 through 37, and 42

through 45 provide pressure-temperature limits which include

conservative corrections for pressure and temperature instrument

uncertainty. Table 6 through 45 provide pressure-temperature limits

for 9, 11, 13, 15 and 32 EFPY. These tables provide the allowable

pressure versus reactor coolant temperature for the various cooldown

conditions. The allowable pressure is in units of psig while the

temperature is in units of *F. Figures 2, 8, 14, 20, 26, 32, 38,

44, 50, and 56 provide a graphical presentation of the cooldown

pressure-temperature limits found in Tables 6 and 45. It is

permissible to linearly interpolate between the cooldown

pressure-temperature limits.

6.0 HEATUP LIMIT ANALYSIS

During a heatup transient, the thermal bending stress is compressive

at the reactor vessel inside wall and is tensile at the reactor

vessel outside wall. Internal pressure creates a tensile stress at

the inside wall as well as the outside wall locations.

Consequently, the outside wall location has the larger total stress

when compared to the inside wall. However, neutron embrittlement

(the shift in material RTNDT and the associated reduction in

fracture toughness) is greater at the inside location than the

outside. Therefore, both the inside and outside flaw locations must

be analyzed to assure that the most limiting condition is achieved.

As described in the cooldown case, the reference stress intensity

factor is calculated at the metal temperature at the tip of the flaw

and the adjusted reference temperature at the flaw location. For

heatup the reference stress intensity is calculated for both the

Page 27

1306.doc(9033)/ch-28

1/4t and 3/4t locations. Using the finite element method described

in Section 4.0, the temperature profile through the wall and the

metal temperatures at the tip of the flaw are calculated for the

transient history. This information is used to calculate the

thermal stress intensity factor at the 1/4t and 3/4t locations using

the calculated wall gradient and thermal influence coefficients.

The allowable pressure stress intensity is then determined by

superposition of the thermal stress intensity factor with the

available reference stress intensity at the flaw tip. The allowable

pressure is then derived from the calculated allowable pressure

stress intensity factor.

It is interesting to note that a sign change occurs in the thermal

stress through the reactor vessel beltline wall. Assuming a

reference flaw at the 1/4t location the thermal stress tends to

alleviate the pressure stress indicating the isothermal steady state

condition would represent the limiting P-T limit. However, the

isothermal condition may not always provide the limiting

pressure-temperature limit for the 1/4t location during a heatup

transient. This is due to the correction of the base metal

temperature to the Reactor Coolant System (RCS) fluid temperature at

the inside wall by accounting for clad and film temperature

differentials. For a given heatup rate (non-isothermal), the

differential temperature through the clad and film increases as a

function of thermal rate resulting in a higher RCS fluid temperature

at the inside wall than the isothermal condition for the same flaw

tip temperature and pressure. Therefore to ensure the accurate

representation of the 1/4t pressure-temperature limit during heatup,

both the isothermal and heatup rate dependent pressure-temperature

limits are calculated to ensure the limiting condition was achieved.

These limits account for clad and film differential temperatures and

for the gradual buildup of wall differential temperatures with time,

as do the cooldown limits.

At the 3/4t location the pressure stress and thermal stresses are

tensile resulting in the maximum stress at that location. Pressure

temperature limits were calculated for the 3/4t location accounting

Page 28

1306.doc(9033)/ch-29

for clad and film differential temperature and the buildup of wall

temperature gradients with time using the method described in

Section 4.0. The allowable pressure was derived based upon a flaw

at the 3/4t location by superposition of the thermal stress

intensity with the available reference stress intensity for the

metal temperature and adjusted reference temperature at that

position.

To develop composite pressure-temperature limits for the heatup

transient, the isothermal, 1/4t heatup, and 3/4t heatup pressure

temperature limits are compared for a given thermal rate. Then the

most restrictive pressure-temperature limits are combined over the

complete temperature interval resulting in a composite limit curve

for the reactor vessel beltline for the heatup event.

Tables 6 through 45 provide the results for the 20°F/hr, 30°F/hr,

40°F/hr, 50°F/hr and 60°F/hr heatup pressure-temperature limits.

Table 6 through 9, 14 through 17, 22 through 25, 30 through 33, and

38 through 41 provide the pressure-temperature limits without

pressure and temperature instrument uncertainty corrections. Table

10 through 13, 18 through 21, 26 through 29, 34 through 37, and 42

through 45 provide pressure-temperature limits which include

conservative corrections for pressure and temperature instrument

uncertainty. Table 6 through 45 provide pressure-temperature limits

for 9, 11, 13, 15, and EFPY. These tables provide the allowable

pressure versus reactor coolant temperature for the various heatup

conditions. The allowable pressure is in units of psig while the

temperature is in units of *F. Figures 3 through 7, 9, through 13,

15 through 19, 21 through 25, 27 through 31, 33 through 37, 39

through 43, 45 through 49, 51 through 55 and 57 through 61 provide a

graphical presentation of the heatup pressure-temperature limits

found in Tables 6 and 45. It is permissible to linearly interpolate

between the heatup pressure-temperature limits.

Page 29

1306.doc(9033)/ch-30

7.0 HYDROSTATIC TEST AND CORE CRITICAL LIMIT ANALYSIS

Both 10 CFR Part 50 Appendix G and the ASME Code Appendix G require

the development of pressure-temperature limits which are applicable to inservice hydrostatic-tests. For hydrostatic tests performed

subsequent to loading fuel into the reactor vessel, the minimum test

temperature is determined by evaluating KI, the mode I stress

intensity factors. The evaluation of KI is performed in the same

manner as that for normal operation heatup and cooldown conditions

except the factor of safety applied to the pressure stress intensity

factor is 1.5 versus 2.0. From this evaluation, a

pressure-temperature limit which is applicable to inservice

hydrostatic tests is established. The minimum temperature for the

inservice hydrostatic test pressure can be determined by entering

the curve at the test pressure (1.1 times normal operating pressure)

and locating the corresponding temperature. The inservice

hydrostatic test limit is provided for 9, 11, 13, 15, and 32 EFPY

and are referenced on the core critical P-T limit figure.

Appendix G to 10 CFR Part 50, specifies pressure-temperature limits

for core critical operation to provide additional margin during

actual power operation.

The pressure-temperature limit for core critical operation is based

upon two criteria. These criteria are that the reactor vessel must

be at a temperature equal to or greater than the minimum temperature

required for the inservice hydrostatic test, and be at least 40OF

higher than the minimum pressure-temperature curve for normal

operation heatup or cooldown. The core critical limit has been

developed based upon the 20°F/hr, 30°F/hr, 40°F/hr, 50°F/hr and

60°F/hr heatup P-T limit and the minimum temperature required for

inservice hydrotest for each required EFPY interval. The core

critical limits are referenced on the heatup P-T limit figure.

Page 30

1306.doc(9033)/ch-31

Note, that the core critical limits established above are solely

based upon fracture mechanics considerations, and do not consider

core reactivity safety analyses which can control the temperature at

which the core can be brought critical.

8.0 LTOP ENABLE TEMPERATURES

Standard Review Plan 5.2.2, Overpressure Protection (18), has

defined the temperature at which the Low Temperature Overpressure

Protection (LTOP) system should be operable during startup and

shutdown conditions. This temperature know as the LTOP enable

temperature is defined as the water temperature corresponding to a

metal temperature of at least RTNDT + 90*F at the beltline location

(1/4t or 3/4t) that is controlling in the Appendix G calculations.

Below the LTOP enable temperature the LTOP system must be aligned to

the RCS to prevent exceeding the applicable technical specification

and Appendix G limits in the event of a transient.

The LTOP enable temperature for a cooldown is based upon the

isothermal P-T limit. Consequently the LTOP enable temperature is

equal to the 1/4t adjusted reference temperature plus ninety (900F)

degrees Fahrenheit. Therefore, the LTOP enable temperatures,

neglecting temperature instrumentation uncertainties are 284°F,

292-F, 298-F, 304°F and 335°F for 9, 11, 13, 15 and 32 EFPY

respectively.

The LTOP enable temperatures for heatup are given below for each

time period (EFPY) based on the analyzed heatup rate conditions.

These temperatures do not include temperature instrumentation

uncertainties.

Page 31

1306.doc(9033)/ch-32

Heatup LTOP Enable Temperatures, (*F)

Rate Limiting

°F/hr Location 9 EFPY 11 EFPY 13 EFPY 15 EFPY 32 EFPY

20 1/4t 292 300 306 312 343

30 1/4t 296 304 310 316 347

40 1/4t 300 308 314 320 351

50 1/4t 304 312 318 324 355

60 1/4t 308 316 322 328 359

Page 32

1306.doc(9033)/ch-33

DATA

Reactor Vessel Data

Design Pressure

Design Temperature

Operating Pressure

Beltline Thickness

Inside Radius

Outside Radius

Cladding Thickness

Material-SA 302 Grade B

Thermal Conductivity

Youngs Modulus

Coefficient of Thermal

Expansion

Specific Heat

Density

Stainless Steel Cladding

= 2500 psia

- 650°F

= 2250 psia

= 8.625 in

- 86.906 in

= 95.53 in

= .2187 in

= 24.7 BTU/hr-ft-°F

= 28 x 106 psi

= 7.77 x 10- 6in/in/°F

- .12 BTU/lb-°F

- .283 lb/in3

Thermal Conductivity - 10 BTU/hr-ft-°F

Adjusted Reference Temperature Values

EFPY

1/4t

3/4t

1940F 1570F

11

2020F

1630F

13

2080F

1680F

15

214°F

172 0F

32

2450F

200°F

Film coefficient on inside surface = 1000 BTU/hr-ft2-°F

Page 33

Reference

16

16

15

16

16

16

16

Reference

1306.doc(9033)/ch-34

REFERENCES

(1) Code of Federal Regulations, 10 CFR Part 50, Appendix A, "General

Design Criteria for Nuclear Power Plants", January 1988.

(2) Code of Federal Regulations, 10 CFR Part 50, Appendix G "Fracture

Toughness Requirements", January 1988.

(3) ASME Boiler and Pressure Vessel Code Section III, Appendix G,

"Protection Against Nonductile Failure", 1986 Edition.

(4) Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel

Materials", U.S. Nuclear Regulatory Commission, Revision 2, May

1988.

(5) V. A. Perone et. al., "Indian Point Unit 3 Reactor Vessel Fluence

and RTPTS Evaluations", Westinghouse Report WCAP-11045, January

1986.

(6) "Evaluation of Pressurized Thermal Shock Effects Due to Small Break

LOCA's with Loss of Feedwater for the Combustion Engineering NSSS",

Combustion Engineering Report CEN-189, December 1981.

(7) S. E. Yanichko and J. A. Davison, "Consolidated Edison Company of

New York, Indian Point Unit No. 3 Reactor Vessel Irradiation

Surveillance Program", Westinghouse Report WCAP-8475, January 1975.

(8) J. A. Davison, et. al., "Analysis of Capsule T from the Indian Point

Unit No. 3 Reactor Vessel Radiation Surveillance Program",

Westinghouse Report WCAP-9491, April 1979.

(9) S. E. Yanichko and S. L. Anderson, "Analysis of Capsule Y from the

Power Authority of the State of New York, Indian Point Unit 3

Reactor Vessel Radiation Surveillance Program", Westinghouse Report

WCAP-10300, Volume 1, March 1983.

Page 34

1306.doc(9033)/ch-35

(10) S. E. Yanichko, et. al., "Analysis of Capsule Z from the New York

Power Authority, Indian Point Unit 3 Reactor Vessel Irradiation

Surveillance Program", Westinghouse Report WCAP-11815, March 1988.

(11) "LWR-Pressure Vessel Surveillance Dosimetry Improvement Program:

LWR Power Reactor Surveillance Physics - Dosimetry Data Base

Compendium", Hanford Engineering Development Laboratory Report

HEDL-TIME 85-3 (NUREG/CR-3319), August 1985.

(12) Letter to Mr. Paul Hijeck (C-E) from Michele Romagnuolo (NYPA)",

Indian Point Unit 3 Nuclear Power Plant Heatup and Cooldowm

Limitation Curves for the Reactor Coolant System Agreement No.

029437-89, Change Order No. 2", dated March 27, 1990.

(13) F. W. Stallmann, "Analysis of the A302B and A533B Standard Reference

Materials in Surveillance Capsules of Commercial Power Reactors",

Oak Ridge National Laboratory Report ORNL/TM-10459 (NUREG/CR-4947),

January 1988.

(14) "Semi-Elliptical Cracks in a Cylinder Subjected to Stress

Gradients", J. Hellot, R. C. Labbens and Pellisser - Tanon ASTM

Special Technical Publication 677, August 1979.

(15) Indian Point Unit 3 Final Safety Analysis Report

(16) General Arrangement - Elevation for Westinghouse Electric

Corporation, 173" I.D. Reactor Vessel, Drawing E-234-040-5, dated

August 5, 1969.

(17) ASME Boiler and Pressure Vessel Code Section III, Appendix I,

"Design Stress Intensity Values, Allowable Stresses, Material

Properties, and Design Fatigue Curves", 1986 Edition.

(18) USNRC Standard Review Plan 5.2.2, Overpressure Protection, Revision

02, dated November 1988.

Page 35

1306.doc(9033)/ch-36

TABLE 1

INDIAN POINT UNIT 3 REACTOR

VESSEL BELTLINE MATERIALS

Material

Plate

Plate

Plate

Plate

Plate

Plate

Welds

Welds

Welds

B2802-1

B2802-2

B2802-3

B2803-1

B2803-2

B2803-3

2-042 A/C

3-042 A/C

9-042

Cu

(W/O)

0.20

0.22

0.20

0.19

0.22

0.24

0.19 0.19

0.27

Ni

(w/o)

0.50

0.53

0.49

0.47

0.52

0.52

1.00

1.00

0.74

Initial

RTNDT

5

-4

17

49

-5

74

-56(a)

-56(a)

(a) Generic mean value per Reference 6.

Page 36

1306.doc(9033)/ch-37

TABLE 2

ADJUSTED REFERENCE TEMPERATURE (1/4t) CALCULATION FOR INDIAN POINT UNIT 3

1/4t Fluence

(101 9n/cm2)

B-2802-1

B-2802-2

B-2802-3

B-2803-1

B-2803-2

B-2803-3

- 2-042 A/C

3-042 A/C

9-042

125

141

130

128

150

160

220

220

206

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

RTNDT(i)

(OF)

Margin

(OF)

ART (OF)(0F~

5

-4

17

49

-5

74

-56

-56

-70

160

167

177

Page 37

1306.doc(9033)/ch-38

TABLE 3

CHEMISTRY FACTOR DERIVATION

PLATE B2803-3 (TRANSVERSE)

Irradiation ARTNDT(a),

Capsule "A"('F)

T 118

Y .150

z .155

Neutron Fluence

(1019 n/cm2 )

0.3226

0.805

1.07

Fl uence

Factor(b) "B" "A" X "B"

0.689 81.31

0.939 140.87

1.019 157.93

380.11

zAX B

z (B)

380.11 2.3951

= 158.70. Chemistry Factor

(a) Shift in reference temperature measured at 30-foot-pound level

(b) Fluence Factor = f(O.28 - 0.10 logf)

where f = neutron fluence in units of 1019 n/cm 2, E>lMev

Page 38

(B) 2

0.4749

0.8810

1. 0382

2.3951

1306.doc(9033)/ch-39

TABLE 4

PROJECTED PEAK NEUTRON FLUENCE

AT VESSEL BASE METAL-CLAD INTERFACE

Exposure Time

(EFPY)

5.55

9.0

11.0

13.0

15.0

32.0

Projected(a) Peak Neutron

Fluence(n/cm ,E>lMev)

3.13

4.69

5.59

6.49

7.39

1.506

1018 10 18

10 18

1018

1018

1019

Projection based on peak fluence of 3.13 x 1018 n/cm2 after

5.55 EFPY (end-of-cycle 5) and a neutron exposure fluence rate of

1.43 x 1010 n/cm 2.sec. Sample calculation for 15.0 EFPY:

i) (15.0 - 5.55) EFPY = 9.45 EFPY = 2.9802 x 108 sec.

ii) (1.43 x 1010 n/cm 2.sec) x (2.9802 x 108 sec) =

4.26 x 1018 n/cm2, fluence for additional 9.45 EFPY

iii) (3.13 + 4.26) x 1018 n/cm2 . 7.39 x 1018 n/cm2

total fast fluence after 15.0 EFPY

Page 39

1306.doc(9033)/ch-40

TABLE 5

ADJUSTED REFERENCE TEMPERATURE

PROJECTIONS FOR INDIAN POINT UNIT 3

(Lower Shell Plate B2803-3)

Vessel Wall

Location(a)

1/4 T

3/4 T

1/4 T

3/4 T

1/4 T

3/4 T

1/4 T

3/4 T

1/4 T

3/4 T

Neutron Fluence

n/cm 2, 1018 n/cm 2)

Adjusted

Reference

Temperature (OF)

2.795

0.993

3.332

1.184

3.868

1.374

4.405

1.565

8.976

3.188

202

163

208

168

(a) Fraction of vessel wall thickness, T = 8.625 inches

Page 40

Time

(EFPY)

9.0

9.0

11.0

11.0

13.0

13.0

15.0

15.0

32.0

32.0

TABLE 6

INDIAN POINT UNIT 3 9 EFPY

COOLDOWN P-ALLOWdABLE (PSIG)

RCS --------------------------------------------------TEMP ISO 20 F/ 50 F/ 60 F/ 80 F/ 100 F/ DEG F THERMAL HOUR HOUR HOUR HOUR HOUR .... .... --- -- ----- -----. ..... ----- -----

50 527.6 452.9 341.7 304.9 231.7 159.3 60 532.1 457.9 347.7 311.3 239.0 167.5 70 537.2 463.8 354.7 318.7 247.2 176.9 80 543.2 470.5 362.7 327.2 256.9 187.8 90 550.2 478.3 372.0 337.0 267.9 200.3

100 558.2 487.3 382.8 348.4 280.9 214.6 110 567.4 497.7 395.2 361.6 295.6 231.5 120 578.1 509.7 409.6 376.9 312.9 251.0 130 590.5 523.6 426.2 394.5 332.6 273.3 140 604.8 539.7 445.4 414.8 355.7 299.1 150 621.3 558.3 67.5 438.4 382.0 328.7 160 640.4 579.8 493.3 465.6 412.9 363.6 170 662.5 604.6 522.9 497.1 448.1 403.7 180 688.0 633.4 557.2 533.4 489.4 449.8 190 717.5 666.6 596.9 575.5 536.3 502.9 200 751.6 705.0 642.6 624.1 591.5 564.0 210 791.0 749.3 695.7 680.3 654.3 635.7 220 836.6 800.6 756.9 745.2 728.0 718.3 230 889.3 859.9 827.7 820.3 811.8 813.2 240 950.2 928.5 909.7 907.1 910.3 922.5 250 1020.7 1007.7 1004.0 1007.5 1020.7 1020.7 260 1102.1 1099.4 1102.1 1102.1 1102.1 1102.1 270 1196.2 1196.2 1196.2 1196.2 1196.2 1196.2 280 1305.0 1305.0 1305.0 1305.0 1305.0 1305.0 290 1430.8 1430.8 1430.8 1430.8 1430.8 1430.8 300 1576.2 1576.2 1576.2 1576.2 1576.2 1576.2 310 1744.3 1744.3 1744.3 1744.3 1744.3 1744.3 320 1938.7 1938.7 1938.7 1938.7 1938.7 1938.7 330 2163.3 2163.3 2163.3 2163.3 2163.3 2163.3 340 2423.1 2423.1 2423.1 2423.1 2423.1 2423.1

342.56 2500.0 2500.0 2500.0 2500.0 2500.0 2500.0 350 2723.3 2723.3 2723.3 2723.3 2723.3 2723.3 360 3000.0 3000.0 3000.0 3000.0 3000.0 3000.0

CORRECTION FACTORS: NONE

Page 41

TABLE 7


RCS TEMP DEG F

50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 310

315.80 315.80 316.72

320 330 340

347.98 350 360 370 380

387.98 390 400 410

HEATUP 20 F/HR P-ALLOWABLE (PSIG)

-------.-----.------.--.

HYDRO 20 F/ CORE STATIC HOUR CRITICAL

703.4 527.6 709.4 532.1 716.3 537.2 724.3 543.2 733.5 550.2 744.2 558.2 756.5 567.4 770.8 578.1 787.3 590.5 806.3 604.8 828.4 621.3 853.8 640.4 883.3 662.5 917.3 688.0 956.6 717.5

1002.1 751.6 1054.7 791.0 1115.5 836.6 1185.7 889.3 1267.0 950.2 1360.9 1020.7 1469.4 1102.1 1594.9 1196.2 1740.0 1293.6 1907.7 1405.4 2101.6 1534.6 2325.8 1684.0

- - 0.0 - - 1252.7

2500.0 2584.9 1856.7 1293.6 2884.5 2056.3 1405.4 3000.0 2287.2 1534.6

2500.0 2554.0 1684.0

- 2862.4 1856.7 3000.0 2056.3

S - 2287.2 - 2500.0

2554.0 2862.4

- 3000.0

RCS TEMP DEG F

50 60 70 8o 90

100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 310

315.80 315.80 316.72

320 330 340 350

350.69 360 370 380 390

390.69 400 410


Page 42


---.-.------.-..--------

HYDRO .30 F/ CORE STATIC HOUR CRITICAL

703.4 527.6 709.4 532.1 716.3 537.2 724.3 543.2 733.5 550.2 744.2 558.2 756.5 562.2 770.8 565.6 787.3 575.6 806.3 590.7 828.4 610.3 853.8 634.3 883.3 662.5 917.3 688.0 956.6 717.5

1002.1 751.6 1054.7 791.0 1115.5 836.6 1185.7 889.3 1267.0 950.2 1360.9 1020.7 1469.4 1102.1 1594.9 1196.2 1740.0 1291.7 1907.7 1397.1 2101.6 1518.9 2325.8 1659.7

- - 0.0 S - 1251.6

2500.0 2584.9 1822.5 1291.7 2884.5 2010.7 1397.1 3000.0 2228.3 1518.9

- 2479.9 1659.7 - 2500.0

2770.6 1822.5 3000.0 2010.7

S - 2228.3 S - 2479.9 - 2500.0 - 2770.6 S - 3000.0

TABLE 8 INDIAN POINT UNIT 3

9 EFPY


RCS ------------------------TEMP HYDRO 40 F/ CORE DEG F STATIC HOUR CRITICAL ..... ...... .....- ........

50 703.4 527.6 60 709.4 532.1 70 716.3 537.2 80 724.3 543.2 90 733.5 550.2 100 744.2 558.2 110 756.5 548.0 120 770.8 544.5 130 787.3 548.1 140 806.3 557.5 150 828.4 572.1 160 853.8 591.3 170 883.3 615.3 180 917.3 644.3 190 956.6 678.5 200 1002.1 718.7 210 1054.7 765.6 220 1115.5 820.0 230 1185.7 883.2 240 1267.0 950.2 250 1360.9 1020.7 260 1469.4 1102.1 270 1594.9 1196.2 280 1740.0 1292.1 290 1907.7 1391.5 300 2101.6 1506.3 310 2325.8 1639.1

315.80 0.0 315.80 - 1251.8 316.72 2500.0

320 2584.9 1792.6 1292.1 330 2884.5 1970.0 1391.5 340 3000.0 2175.1 1506.3 350 - 2412.2 1639.1

353.20 - 2500.0 360 - 2686.3 1792.6 370 - 3000.0 1970.0 380 - 2175.1 390 - 2412.2

393.20 2500.0 400 - 2686.3 410 - 3000.0

RCS TEMP DEG F

so 60 70 8o 90

100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 310

315.80 315.80 316.72

320 330 340 350

355.76 360 370 380 390

395.76 400 410 420


.........................


703.4 527.6 709.4 532.1 716.3 537.2 724.3 543.2 733.5 550.2 744.2 558.2 756.5 539.6 770.8 530.1 787.3 527.7 806.3 531.8 828.4 541.1 853.8 555.4 883.3 574.6 917.3 598.6 956.6 628.0

1002.1 662.9 1054.7 704.1 1115.5 752.4 1185.7 808.5 1267.0 874.1 1360.9 949.9 1469.4 1037.8 1594.9 1139.7 1740.0 1257.1 1907.7 1388.7 2101.6 1496.9 2325.8 1622.2

- -0.0 1207.8

2500.0 2584.9 1767.0 1257.1 2884.5 1933.8 1388.7 3000.0 2127.8 1496.9

- 2351.2 1622.2 2500.0

- 2609.7 1767.0 - 2908.8 1933.8

3000.0 2127.8 - 2351.2 - 2500.0 - 2609.7 - 2908.8 - 3000.0


Page 43


9 EFPY


RCS .. . . . . . . . . . . . TEMP HYDRO 60 F/ CORE DEG F STATIC HOUR CRITICAL ..... ...... .....--- ........

50 703.4 527.6 60 709.4 532.1 70 716.3 537.2 80 724.3 543.2 90 733.5 550.2 100 744.2 557.0 110 756.5 534.1 120 770.8 519.6 130 787.3 512.4 140 806.3 511.4 150 828.4 515.9 160 853.8 525.4 170 883.3 539.9 180 917.3 559.2 190 956.6 583.6 200 1002.1 613.4 210 1054.7 649.1 220 1115.5 691.4 230 1185.7 741.1 240 1267.0 799.2 250 1360.9 866.8 260 1469.4 945.4 270 1594.9 1036.6 280 1740.0 1142.2 290 1907.7 1264.6 300 2101.6 1406.2 310 2325.8 1570.1

315.80 - 0.0 315.80 - 1097.8 316.72 2500.0 -

320 2584.9 1744.4 1142.2 330 2884.5 1902.1 1264.6 340 3000.0 2084.4 1406.2 350 - 2295.1 1570.1

358.41 - 2500.0 360 - 2538.8 1744.4 370 - 2820.4 1902.1 380 - 3000.0 2084.4 390 - - 2295.1

398.41 - - 2500.0 400 - 2538.8 410 - - 2820.4 420 - - 3000.0


Page 44

TABLE 10


COOLDOWIN P-ALLO ABLE (PSIG)

RCS --------------------------------------------------TEMP ISO 20 F/ 50 F/ 60 F/ 80 F/ 100 F/ DEG F THERMAL HOUR HOUR HOUR HOUR HOUR

----- ~ ~ ~ . .-- - -- - - - - - .- - - - -.. . . 66 490.6 415.9 304.7 267.9 194.7 122.3 76 495.1 420.9 310.7 274.3 202.0 130.5 86 500.2 426.8 317.7 281.7 210.2 139.9 96 506.2 433.5 325.7 290.2 219.9 150.8 106 513.2 441.3 335.0 300.0 230.9 163.3 116 521.2 450.3 345.8 311.4 243.9 177.6 126 530.4 460.7 358.2 324.6 258.6 194.5 136 541.1 472.7 372.6 339.9 275.9 214.0 146 553.5 486.6 389.2 357.5 295.6 236.3 156 567.8 502.7 408.4 377.8 318.7 262.1 166 584.3 521.3 430.5 401.4 345.0 291.7 176 603.4 542.8 456.3 428.6 375.9 326.6 186 625.5 567.6 485.9 460.1 411.1 366.7 196 651.0 596.4 520.2 496.4 452.4 412.8 206 680.5 629.6 559.9 538.5 499.3 465.9 216 714.6 668.0 605.6 587.1 554.5 527.0 226 754.0 712.3 658.7 643.3 617.3 598.7 236 799.6 763.6 719.9 708.2 691.0 681.3 246 852.3 822.9 790.7 783.3 774.8 776.2 256 913.2 891.5 872.7 870.1 873.3 885.5 266 983.7 970.7 967.0 970.5 983.7 983.7 276 1065.1 1062.4 1065.1 1065.1 1065.1 1065.1 286 1159.2 1159.2 1159.2 1159.2 1159.2 1159.2 296 1268.0 1268.0 1268.0 1268.0 1268.0 1268.0 306 1393.8 1393.8 1393.8 1393.8 1393.8 1393.8 310.76 1463.0 1463.0 1463.0 1463.0 1463.0 1463.0 310.76 1380.0 1380.0 1380.0 1380.0 1380.0 1380.0 316 1456.2 1456.2 1456.2 1456.2 1456.2 1456.2 323.36 1580.0 1580.0 1580.0 1580.0 1580.0 1580.0 323.36 1679.0 1679.0 1679.0 1679.0 1679.0 1679.0 326 1723.3 1723.3 1723.3 1723.3 1723.3 1723.3 336 1917.7 1917.7 1917.7 1917.7 1917.7 1917.7 346 2142.3 2142.3 2142.3 2142.3 2142.3 2142.3 356 2402.1 2402.1 2402.1 2402.1 2402.1 2402.1 358.56 2479.0 2479.0 2479.0 2479.0 2479.0 2479.0 358.56 2380.0 2380.0 2380.0 2380.0 2380.0 2380.0 362.56 2500.0 2500.0 2500.0 2500.0 2500.0 2500.0 366 2603.3 2603.3 2603.3 2603.3 2603.3 2603.3 376 2880.0 2880.0 2880.0 2880.0 2880.0 2880.0

CORRECTION FACTORS: TEMPERATURE 16 DEGREES F

PRESSURE 0 cu P (u 1500 -37 PSI 1500 < P 4x 1700 -120 PSI 1700 < P

RCS TEMP DEG F

66 76 86 96

106 116 126 136 146 156 166 176 186 196 206 216 226 236 246 256 266 276

278.44 278.44

286 293.24 293.24

296 306

313.32 313.32

316 326

326.93 326.93 332.72 332.72

336 336.3 336.3

337.17 346

353.32 353.32

356 363.98 363.98

366 366.93 366.93 368.14

376 386 396

403-98 403.98

406 408.14

416 426


---..--.--------.--.-

HYDRO 20 F/ CORE STATIC HOUR CRITI

666.4 490.6 672.4 495.1 679.3 500.2 687.3 506.2 696.5 513.2 707.2 521.2 719.5 530.4 733.8 541.1 750.3 553.5 769.3 567.8 791.4 584.3 816.8 603.4 846.3 625.5 880.3 651.0 919.6 680.5 965.1 714.6

1017.7 754.0 1078.5 799.6 1148.7 852.3 1230.0 913.2 1323.9 983.7 1432.4 1065.1 1463.0 1380.0 1474.9 1159.2 1580.0 1679.0 1719.0 1256.6 1886.7 1368.4

1463.0 - 1380.0

2080.6 1414.6 2304.8 1564.0

- 1580.0 - 1679.0

2479.0 2380.0 2464.9 1835.7

S - 0 - 1260

2500.0 -2764.5 2035.3 1368

- - 1463 - - 1380

2880.0 2266.2 1414 - 2479.0 - 2380.0

2434.0 1564 - 1580

- * 1679, 2500.0 1 2742.4 1835 2880.0 2035,

- - 2266. - 2479,

2380. 2434, 2500. 2742, 2880

TABLE 11


.... RCS

TEMP CAL DEG F

66 76 86 96 106 116 126 136 146 156 166 176 186 196 206 216 226 236 246 256 266 276

278.44 278.44

286 293.24 293.24

296 306

314.45 314.45

316 326

328.48 328.48 332.72 332.72

336 .0 336.3 .0 336.3

337.17 .4 346 .0 354.45 .0 354.45 .6 356

366 366.69

.0 366.69

.0 368.48

.0 368.48 370.82

.7 376

.3 386

.2 396

.0 406

.0 406.69

.0 406.69

.0 410.82

.4 416

.0 426


PRESSURE 0 4z P


9 EFPY

RCS TEMP DEG F

66 76 86 96

106 116 126 136 146 156 166 176 186 196 206 216 226 236 246 256 266 276

278.44 278.44

286 293.24 293.24

296 306

315.45 315.45

316 326

329.97 329.97 332.72 332.72

336 336.3 336.3

337.17 346

355.45 355.45

356 366

369.2 369.2

369.97 369.97 373.58

376 386 396 406

409.2 409.2

413.58 416 426


.........................


666.4 490.6 672.4 495.1 679.3 500.2 687.3 506.2 696.5 513.2 707.2 521.2 719.5 511.0 733.8 507.5 750.3 511.1 769.3 520.5 791.4 535.1 816.8 554.3 846.3 578.3 880.3 607.3 919.6 641.5 965.1 681.7

1017.7 728.6 1078.5 783.0 1148.7 846.2 1230.0 913.2 1323.9 983.7 1432.4 1065.1 1463.0 1380.0 1474.9 1159.2 1580.0 1679.0 1719.0 1255.1 1886.7 1354.5

- 1463.0 1380.0

2080.6 1386.3 2304.8 1519.1

- 1580.0 - 1679.0

2479.0 -2380.0 2464.9 1771.6

- - 0.0 1258.1

2500.0 2764.5 1949.0 1354.5

1463.0 S - 1380.0

2880.0 2154.1 1386.3 2391.2 1519.1

- 2479.0 - 2380.0

- 1580.0 - - 1679.0 - 2500.0 - 2566.3 1771.6 - 2880.0 1949.0

- 2154.1 - - 2391.2 - - 2479.0 - - 2380.0 - - 2500.0 - - 2566.3 - - 2880.0

RCS TEMP DEG F

66 76 86 96

106 116 126 136 146 156 166 176 186 196 206 216 226 236 246 256 266 276

278.44 278.44

286 293.24 293.24

296 306 316

316.25 316.25

326 331.37 331.37 332.72 332.72

336 336.3 336.3

337.17 346 356

356.25 356.25

366 371.37 371.37 371.76 371.76

376 376.34

386 396 406

411.76 411.76

416 416.34

426 436


PRESSURE 0


9 EFPY


RCS ------------------------TEMP HYDRO 60 F/ CORE DEG F STATIC HOUR CRITICAL -. ---. ...... .--.-- ........

66 666.4 490.6 76 672.4 495.1 86 679.3 500.2 96 687.3 506.2 106 696.5 513.2 116 707.2 520.0 126 719.5 497.1 136 733.8 482.6 146 750.3 475.4 156 769.3 474.4 166 791.4 478.9 176 816.8 488.4 186 846.3 502.9 196 880.3 522.2 206 919.6 546.6 216 965.1 576.4 226 1017.7 612.1 236 1078.5 654.4 246 1148.7 704.1 256 1230.0 762.2 266 1323.9 829.8 276 1432.4 908.4

278.44 1463.0 278.44 1380.0

286 1474.9 999.6 293.24 1580.0 293.24 1679.0

296 1719.0 1105.2 306 1886.7 1227.6 316 2080.6 1369.2

321.72 - 1463.0 321.72 - 1380.0

326 2304.8 1450.1 332.72 2479.0 332.72 2380.0 333.45 - 1580.0 333.45 - 1679.0

336 2464.9 1723.4 336.3 - 0.0 336.3 - 1108.9 337.17 2500.0

346 2764.5 1881.1 1227.6 356 2880.0 2063.4 1369.2

361.72 - 1463.0 361.72 - 1380.0

366 - 2274.1 1450.1 373.45 - - 1580.0 373.45 - - 1679.0 374.41 - 2479.0 374.41 2380.0

376 2418.8 1723.4 378.88 2500.0

386 - 2700.4 1881.1 396 - 2880.0 2063.4 406 - - 2274.1

414.41 - 2479.0 414.41 - - 2380.0

416 - - 2418.8 418.88 - 2500.0

426 - - 2700.4 436 - - 2880.0


PRESSURE 0 c" P = 1500 -37 PSI 1500 P - 1700 -120 PSI 1700 P 2500 -21 PSI 2500 P c- 3000 -120 PSI

Page 48

TABLE 14


COOLDOWN P-ALLOWABLE (PSIG)

RCS -- - - - - - - - - - - - - - - - - - - - - - - -TEMP ISO 20 F/ 50 F/ 60 F/ 80 Ff 100 F/ DEG F THERMAL HOUR HOUR HOUR HOUR HOUR

50 524.4 4"9.3 337.5 300.4 226.7 153.7 60 528.4 453.8 342.8 306.1 233.1 160.9 70 533.0 459.0 349.0 312.7 240.5 169.3 80 538.4 465.0 356.2 320.3 249.1 179.0 90 544.5 472.0 364.5 329.0 258.9 190.1

100 551.7 480.0 374.0 339.2 270.5 202.9 110 559.9 489.2 385.1 350.9 283.6 218.0 120 569.4 500.0 397.9 364.5 299.0 235.3 130 580.4 512.3 412.7 380.2 316.5 255.2 140 593.2 526.7 429.8 398.3 337.1 278.2 150 607.9 543.2 449.5 419.3 360.5 304.6 160 624.9 562.4 472.4 443.5 388.0 335.6 170 644.5 584.5 498.8 471.5 419.4 371.3 180 667.3 610.1 529.4 503.9 456.1 412.4 190 693.5 639.6 564.7 541.4 498.0 459.8 200 723.9 673.8 605.4 584.6 547.1 514.2 210 759.0 713.3 652.7 634.7 603.0 578.1 220 799.6 759.0 707.2 692.5 668.6 651.6

20 846.6 811.8 770.3 759.4 743.3 736.2 240 900.8 872.9 843.2 836.7 831.0 833.5 250 963.5 943.4 927.2 926.0 930.7 945.4 260 1036.0 1025.0 1024.9 1029.3 1036.0 1036.0 270 1119.8 1119.3 1119.8 1119.8 1119.8 1119.8 280 1216.7 1216.7 1216.7 1216.7 1216.7 1216.7 290 1328.7 1328.7 1328.7 1328.7 1328.7 1328.7 300 1458.2 1458.2 1458.2 1458.2 1458.2 1458.2 310 1607.9 1607.9 1607.9 1607.9 1607.9 1607.9 320 1781.0 1781.0 1781.0 1781.0 1781.0 1781.0 330 1981.0 1981.0 1981.0 1981.0 1981.0 1981.0 340 2212.3 2212.3 2212.3 2212.3 2212.3 2212.3 350 2479.7 2479.7 2479.7 2479.7 2479.7 2479.7

350.66 2500.0 2500.0 2500.0 2500.0 2500.0 2500.0 360 2788.8 2788.8 2788.8 2788.8 2788.8 2788.8 370 3000.0 3000.0 3000.0 3000.0 3000.0 3000.0

CORRECTION FACTORS: NOAE

Page 49

TABLE 15


HEATUP 20 F/HR P-ALLOMABLE (PSIG)

RCS ------------------------TEMP HYDRO 20 F/ CORE DEG F STATIC HOUR CRITICAL --. -.. ...... -- - --------..

50 699.3 524.4 60 704.6 528.4 70 710.7 533.0 80 717.8 538.4 90 726.1 544.5

100 735.6 551.7 110 746.5 559.9 120 759.2 569.4 130 773.9 580.4 140 790.9 593.2 150 810.5 607.9 160 833.2 624.9 170 859.4 644.5 180 889.7 667.3 190 924.7 693.5 200 965.2 723.9 210 1012.0 759.0 220 1066.2 799.6 230 1128.7 846.6 240 1201.1 900.8 250 1284.7 963.5 260 1381.4 1036.0 270 1493.1 1119.8 280 1622.3 1215.1 290 1771.7 1314.7 300 1944.3 1429.7 310 2143.9 1562.8 320 2374.7 1716.6

323.8 - 0.0 323.8 1252.9 324.7 2500.0

330 2641.4 1894.3 1314.7 340 2949.8 2099.9 1429.7 350 3000.0 2337.5 1562.8

355.92 - 2500.0 360 - 2612.1 1716.6 370 - 2929.7 1894.3 380 - 3000.0 2099.9 390 - 2337.5

395.92 - - 2500.0 400 - - 2612.1 410 - 2929.7 420 - 3000.0

RCS TEMP DEG F

50 60 70 80 90

100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 310 320

323.8 323.8 324.7 330 340 350

358.66 360 370 380 390

398.66 400 410 420

HEATUP 30 F/HR P-ALLOUASLE (PSIG)

-..------------..-------


699.3 524.4 704.6 528.4 710.7 533.0 717.8 538.4 726.1 544.5 735.6 551.7 746.5 554.0 759.2 556.5 773.9 565.1 790.9 578.7 810.5 596.5 833.2 618.4 859.4 644.5 889.7 667.3 924.7 693.5 965.2 723.9 1012.0 759.0 1066.2 799.6 1128.7 846.6 1201.1 900.8 1284.7 963.5 1381.4 1036.0 1493.1 1119.8 1622.3 1216.7 1771.7 1311.5 1944.3 1420.0 2143.9 1545.4 2374.7 1690.4

- 0.0 1252.7

2500.0 2641.4 1858.0 1311.5 2949.8 2051.8 1420.0 3000.0 2275.7 1545.4

- 2500.0 - 2534.7 1690.4 - 2834.0 1858.0 - 3000.0 2051.8

- 2275.7 - - 2500.0 - - 2534.7

- 2834.0 - - 3000.0


Page 50

TABLE 16


RCS TEMP DEG F

50 60 70 80 90

100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 310 320

323.8 323.8 324.7

330 340 350 360

361.28 370 380 390 400

401.28 410 420

HEATUP 40 F/HR P-ALLOdABLE (PSIG)

----------------.--.----.


699.3 524.4 704.6 528.4 710.7 533.0 717.8 538.4 726.1 544.5 735.6 551.7 746.5 540.2 759.2 535.8 773.9 538.3 790.9 546.5 810.5 559.4 833.2 576.9 859.4 598.7 889.7 625.1 924.7 656.5 965.2 693.3

1012.0 736.2 1066.2 786.1 1128.7 843.9 1201.1 900.8 1284.7 963.5 1381.4 1036.0 1493.1 1119.8 1622.3 1216.7 1771.7 1310.8 1944.3 1413.1 2143.9 1531.3 2374.7 1668.0

- - 0.0 - - 1252.5

2500.0 2641.4 1826.0 1310.8 2949.8 2008.7 1413.1 3000.0 2219.8 1531.3

- 2463.9 1668.0 2500.0 2746.1 1826.0 3000.0 2008.7

- 2219.8 - 2463.9

2500.0 S - 2746.1 S - 3000.0


Page 51

HEATUP 50 F/HR P-ALLOUABLE (PSIG)

RCS ------------------------TEMP HYDRO 50 F/ CORE DEG F STATIC HOUR CRITICAL -. ---. .. °.... ..... ........

50 699.3 524.4 60 704.6 528.4 70 710.7 533.0 80 717.8 538.4 90 726.1 544.5

100 735.6 551.2 110 746.5 532.0 120 759.2 521.7 130 773.9 518.5 140 790.9 521.4 150 810.5 529.4 160 833.2 542.1 170 859.4 559.3 180 889.7 581.1 190 924.7 607.9 200 965.2 639.8 210 1012.0 677.5 220 1066.2 721.7 230 1128.7 773.0 240 1201.1 833.1 250 1284.7 902.6 260 1381.4 983.1 270 1493.1 1076.5 280 1622.3 1184.2 290 1771.7 1309.5 300 1944.3 1409.0 310 2143.9 1520.6 320 2374.7 1649.6

323.8 - 0.0 323.8 - 1231.8 324.7 2500.0 -

330 2641.4 1798.1 1309.5 340 2949.8 1970.9 1409.0 350 3000.0 2169.8 1520.6 360 - 2400.0 1649.6

363.76 - 2500.0 370 - 2666.3 1798.1 380 2973.0 1970.9 390 * 3000.0 2169.8 400 - 2400.0

403.76 - 2500.0 410 - - 2666.3 420 - 2973.0 430 - 3000.0

TABLE 17



RCS ------------------------TEMP HYDRO 60 F/ CORE DEG F STATIC HOUR CRITICAL

50 699.3 524.4 60 704.6 528.4 70 710.7 533.0 80 717.8 538.4 90 726.1 544.5 100 735.6 550.0 110 746.5 526.6 120 759.2 511.6 130 773.9 503.6 140 790.9 501.6 150 810.5 504.9 160 833.2 513.0 170 859.4 525.7 180 889.7 543.1 190 924.7 565.1 200 965.2 592.2 210 1012.0 624.8 220 1066.2 663.5 230 1128.7 708.9 240 1201.1 762.1 250 1284.7 824.0 260 1381.4 896.0 270 1493.1 979.6 280 1622.3 1076.4 290 1771.7 1188.5 300 1944.3 1318.3 310 2143.9 1468.5 320 2374.7 1633.7

323.8 - - 0.0 323.8 - 1119.0 324.7 2500.0 -

330 2641.4 1774.1 1188.5 340 2949.8 1936.5 1318.3 350 3000.0 2124.1 1468.5 360 2341.0 1633.7

366.34 - 2500.0 370 2591.8 1774.1 380 - 2881.7 1936.5 390 - 3000.0 2124.1 400 - 2341.0

406.34 - - 2500.0 410 - - 2591.8 420 - 2881.7 430 - - 3000.0


Page 52

TABLE 18


RCS -------TEMP ISO DEG F THERMAL -----........ °

66 487.4 76 491.4 86 496.0 96 501.4 106 507.5 116 514.7 126 522.9 136 532.4 146 543.4 156 556.2 166 570.9 176 587.9 186 607.5 196 630.3 206 656.5 216 686.9 226 722.0 236 762.6 246 809.6 256 863.8 266 926.5 276 999.0 286 1082.8 296 1179.7 306 1291.7 316 1421.2

318.79 1463.0 318.79 1380.0

326 1487.9 331.32 1580.0 331.32 1679.0

336 1760.0 346 1960.0 356 2191.3 366 2458.7

366.66 2479.0 366.66 2380.0 370.54 2500.0

376 2668.8 386 2880.0

COOLDOWN P-ALLOUABLE (PSIG)

....................................

20 F/ HOUR

412.3 416.8 422.0 428.0 435.0 443 .0 452.2 463.0 475.3 489.7 506.2 525.4 547.5 573.1 602.6 636.8 676.3 722.0 774.8 835.9 906.4 988.0

1082.3 1179.7 1291.7 1380.1 1463.0 1380.0 1487.9 1580.0 1679.0 1760.0 1960.0 2191.3 2458.7 2479.0 2380.0 2500.0 2668.8 2880.0

50 F/ HOUR

300.5 305.8 312.0 319.2 327.5 337.0 348.1 360.9 375.7 392.8 412.5 435.4 461.8 492.4 527.7 568.4 615.7 670.2 733.3 806.2 890.2 987.9 1082.8 1179.7 1291.7 1421.2 1463.0 1380.0 1487.9 1580.0 1679.0 1760.0 1960.0 2191.3 2458.7 2479.0 2380.0 2500.0 2668.8 2880.0

60 F/ HOUR

263.4 269.1 275.7 283.3 292.0 302.2 313.9 327.5 343.2 361.3 382.3 406.5 434.5 466.9 504.4 547.6 597.7 655.5 722.4 799.7 889.0 992.3

1082.8 1179.7 1291.7 1421.2 1463.0 1380.0 1487.9 1580.0 1679.0 1760.0 1960.0 2191.3 2458.7 2479.0 2380.0 2500.0 2668.8 2880.0

80 F/ HOUR

189.7 196.1 203.5 212.1 221.9 233.5 246.6 262.0 279.5 300.1 323.5 351.0 382.4 419.1 461.0 510.1 566.0 631.6 706.3 794.0 893.7 999.0

1082.8 1179.7 1291.7 1421.2 1463.0 1380.0 1487.9 1580.0 1679.0 1760.0 1960.0 2191.3 2458.7 2479.0 2380.0 2500.0 2668.8 2880.0


PRESSURE 0 . P '- 1500 -37 PSI 1500 < P

TABLE 19


RCS TEMP DEG F

66 76 86 96

106 116 126 136 146 156 166 176 186 196 206 216 226 236 246 256 266 276 286

286.53 286.53

296 301.2 301.2

306 316

321.28 321.28

326 334.92 334.92

336 340.7 340.7

344.26 344.26 345.20

346 356

361.28 361.28

366 371.92 371.92 374.92 374.92

376 376.25

386 396 406

411.92 411.92

416 416.25

426 436


--------.----.-.---.-....


662.3 487.4 667.6 491.4 673.7 496.0 680.8 501.4 689.1 507.5 698.6 514.7 709.5 522.9 722.2 532.4 7

atachment iii to indian point 3 final report ...no. mps-90-494, "indian point unit 3 draft final...

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