ML17328A256

Amend 123 to License DPR-74,changing Tech Spec 3/4.4.9, Pressure/Temp Limits

ML17328A256
Person / Time
Site:	Cook
Issue date:	05/24/1990
From:	Diianni D Office of Nuclear Reactor Regulation
To:
Shared Package
ML17328A257	List: ML17328A256 ML17328A258
References
NUDOCS 9006130278
Download: ML17328A256 (15)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA V(ICHIGAN POllER COHPANY DOCKET NO. 50-316 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

2 AIlENDHENT TO FACILITY OPERATING LICENSE Amendment No. 123 License No.

DPR-74 1.

The Nuclear Regulatory Commission (the Commission) has found that:

-A.

The application for amendment by Irdiana Ilichigan Power Company (the licensee) dated October 25, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended

( the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9006i3027S 900524 PDR ADOCK 050003i6 P

PDC 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.

DPR-74 is hereby amended to read as follows:

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

123,.are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the beginning of Cycle 8.

FOR THE NUCLEAR REGULATORY COYit'.ISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

)lay 24, 1990 Dominic C. DiIanni, Acting Director Project Directorate III-1 Division of Reactor Projects - III, IV, V 5 Special Projects Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO.

123 FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Revise Appendix A Technical.Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

RENOVE 3/4 4-24 3/4 4-25 3/4 4-26 B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-9a B 3/4 4-10 INSERT 3/4 4-23*

3/4 4-24 3/4 4-25 3/4 4-26 B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-9a B 3/4 4-10

• Overleaf page provided to maintain document completeness.

No change contained on this page.

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ACCEPTABLE OPERATION UNACCEPTABLE OPERATION 20, 30 40 50 60 70 80 SO 100 PERCENT OF RATED THERMALPOWER FIGURE 3.4-1 DOSE EQUIVALENTI-131 Primary Coolant Specific ActivityLimitVersus Percept of RATED THERMALPOWER with the Primary Cooiant Specific Activity>1.0pCilgram Dose EquivaIent I-13'I D.C.

COOK - UNIT 2 3/4 4-23

REACTOR COOLANT SYSTEM 3 4.4.9 PRESSURE TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 60 F in any one hour period.

0 b.

A maximum cooldown of 100 F in any one hour period.

0 0

c.

A maximum temperature of less than or equal to 5 F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and 0

ave pressure to less than 200 F and 500 psig, respectively, within the foHowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup,

cooldown, and inservice leak and hydrostatic testing operations',4.9.1,2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5.

The results of these examinations shall be used to update Figures 3.4-2 and 3 '-3.

COOK NUCLEAR PLANT - UNIT 2 3/4 4-24 AMENDMENT NO ~ 123

2600 2400 2200 2000 1800 1600 1400 1200 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS. ( MARGINS OF 60 PSIG AND IOQF ARE INCLUDED'OR POSSIBLE INSTRUMENT ERROR.)

- LEAK TEST LIMIT 1IATERIAL PROPERTY BASIS BASE HETAL CU ~ 0.15X NI ~ 057K IIIITIALRTIIgT ~ 58 F

12 EFPY RTIIPT (1/4T) 178 F

3/4T) 150 F

UNACCEPTABLE OPERATION ACCEPTABLE OPERATION 1000 800 PRESSURE-TEMPERATURE LIMITFOR HEATUP RATES Ur TO eooFSHR RITICALITY LIMIT 600 400 200 50 100 150 '00 250 300 350 400 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (dag F)

FIGURE 3.4-2 raaCVOR COOLANT SVSTE14 PRESSea Ve~RA~ t.ITS VaaSuS 6O Ileum'aVE.

CRITICALITYLI14IT At% HYDROSTATIC TEST LIHIT 450

1 I

C5M CO CL hl lK CO CO a:

CL IllI-CO CO I-C3oD IX C)I-D lll G.

2600 2400 2200 2000

%BOO 1600 1400 4200 1000 800 600 400 200 MATERIAL PROPERTY BASIS BASK METAL CU ~ 0.15K Nl i 0.57X INITIAL RTNBT ~ 5B F 12 EFPY RTIIBT (I 4T)

> 178 F

(3/47) 150 F

UNACCEPTABLE OPERATION PRESSURE-TEMPERATURE LIMITS COOLDOWN RATE F/HR 0

20 40 60 100 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE FOR FIRST 12 EFFECTIVE FULL POWER YEARS. ( MARGINS OF 80 PSIG AND 10oF ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR.)

ACCEPTABLE OPERATION

~I 50 100 150 200 250 300 350 400 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (deg F)

FIGURE 3.4-3 FACTOR COOLANT SYSTEM RKSStNE - Tee~A~ LINE'S VOUS COOLDOMH RATES 450

REACTOR COOLANT SYSTEM BASES 3 4.4.9 PRESSURE TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to vith-stand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 4.1.4 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation, An ID or OD one-quarter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case.

There are several factors which influence the postulated location.

The thermal induced bending stress during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall.

During cooldown the bending stress profile is reversed.

In addition, the material tough-ness is dependent upon irradiation and temperature and therefore the fluence profile through the reactor vessel wall, the rate of heatup and also the rate of cooldown influence the postulated flaw location.

The heatup limit curve, Figure 3 '-2, is a composite curve which was prepared by determining the most conservative

case, with either the inside or outside wall controlling, for any heatup rate up to 60 F per hour.

0 I

The cooldown limit curves of Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall.

The heatup and cooldown curves were prepared based'n the most limiting value of the predicted adjusted reference temperature at the end of 12 EFPY.

The reactor vessel materials have been tested to determine their initial RTNDT',

The results of these tests are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E > 1 MeV) irradiation will cause an increase in the RT>

Therefore, an adjusted reference tem-perature must be predicted Xn accordance with Regulatory Guide 1.99, HDT'Revision 2.

This prediction is based on the fluence and a chemistry factor determined from one of two Positions presented in the Regulatory Guide.

Position (1) determines the chemistry factor from the copper and nickel content of the material.

Position (2) utilizes surveillance data sets which relate the shift in reference temperature of surveillance specimens to the fluence.

The selection of Position (1) or (2) is made based on the availability of credible surveillance

data, and the results achieved in applying the two Positions.

COOK NUCLEAR PLANT- - UNIT 2 B 3/4 4-6 AMENDMENT NO. P),12

INTENTIONALLY LEFT BLANK COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-7 AMENDMENT NO. 12~

INTENTIONALLY LEFT BLANK COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-8

, AMENDMENT NO. i~3

TABLE B REACTOR VESSEL TOUGHNESS 50 FT-LB 35 MIL USE COMPONENT CL. HD.

DOME CL.

HD.

SEG.

CL.

HD.

SEG.

CL.

HEAD FLG.

VESSEL FLANGE INLET NOZZLE INLET NOZZLE INLET NOZZLE INLET NOZZLE OUTLET NOZZLE OUTLET NOZZLE OUTLET NOZZLE OUTLET NOZZLE UPPER SHELL UPPER SHELL UPPER SHELL INTER SHELL INTER SHELL LOWER SHELL LOVER SHELL BOT. HD. SEG.

BOT. HD. SEG.

BOT.

HD. SEG.

INTER. & LOWER SHELL LONG. and GIRTH VELD SEAM CODE NO.

B0048-2 B9883-2 A5189-2 4437-V-1 4436-V-2 269T-2 270T-1 269T-1 270T-2 271T-1 271T-2 272T-1 272T-2 C5518-2 C5521-1 C5518-1 C5556-2 C5521-2 C5540-2 C5592-1 C5823-2 A4957-3 B0019-18 (HT S3986 Linde 124 Flux Lot No. 0934)

MATERIAL TYPE A533B CL. 1 A533B CL. 1 A533B CL.1 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A508 CL.2 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 A533B CL.1 SAV CU

~t NA NA NA NA NA NA NA NA NA NA NA NA NA

.12

.14

.12

.15

.14

.11

.14 NA NA NA

.06 0.64 0.66 0.63 0.70 0.70 0.85 0.91 NA NA 0.80 0.80 NA NA 0.61 0.59 0.57 0.57 0.58 0.64 0.59 0.57 0.51 0.61 0.97

-20

-20 10

-20 30

-20

-20

-10

-10 0

-10 0

10 0

10 0

10

-20

-20

-10

-10

-50

-40 Temp(a)

~F 30

-3 72 5

15

-15

-3 NA NA 12

-15 NA NA 88 93 66 118(b) 98(b) 35(b) 25(b) 45 20 0

25

-20

-20 12

-20 30

-20

-20

-10

-10 0

0

-10 0

28 33 10 58(b) 38(b)

-20(b)

-20(b)

-10

-10

-50

-35(b)

MWD

~ft-1b 148 143.5 140.5 239 161 201.5 239.5 NA NA

>179 181 NA NA 107.5 112

> 82.5 109.5 111.5 113 107 129 149 177 Nh NMWD(a)

~ft-1b 96 93 91 155 105 131 156 NA NA NA 117.5 Nh 70 73 90(b) 86(b) 110(b) 103(b) 84 97 115 97(b) a) Estimated per NRC Standard Review Plan b) Actual values Nh - Not available or not applicable, as appropriate COOK NUCLEAR PLANT - UNIT 2 3/4 MWD - Major Working Direction NMWD - Normal to MWD 4-9 AMENDMENT N03.23

4

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INTENTIONALLY LEFT BLANK COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-9a AMENDMENT NO. PP',$ 23

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vV REACTOR COOLANT SYSTEM BASES The actual shift in the reference temperature of surveillance specimens and neutron fluence is established periodically by removing and evaluating reactor vessel material irradiation surveillance specimens and dosimetry installed near the inside wall of the reactor vessel in the core area.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted ad5ustments for this shift in RTNDT at the end of 12 EFPY, as well as ad5ustments for possible errors in the pressure and temperature sensing instruments, The 12 EFPY heatup and cooldown curves were developed based on the following:

1.

The pro)ected fluence values established by specimen analysis.

2.

Intermediate shell plate C5556-2 being the limiting material as determined by Position 1 of Regulatory Guide 1.99, Revision 2, with a copper and nickel content of 0.15% and 0.57%, respectively.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G

to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, one PORV and the RHR safety valve, or an RCS vent opening of greater than or equal to 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 152 F.

Either PORV or RHR safety valve has 0

adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal

" to 50 F above the RCS cold leg temperatures or (2) the start of a charging pump and its in)ection into a water solid RCS.

3 4.4.10 STRUCTURAL INTEGRITY The inspection and testing programs for ASME Code Class 1,

2 and 3

components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant

~

To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

COOK NUCLEAR PLANT - UNIT 2, B 3/4 4-10 AMENDMENT NO ~ $g,123