Text

Amends 153 & 148 to Licenses DPR-19 & DPR-25,respectively, Updating pressure-temperature Curves Contained in Licensee TS 22 Effective Full Power Yrs
	ML20135E150
Person / Time
Site:	Dresden, Quad Cities
Issue date:	02/28/1997
From:	Stang G; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20135E152	List: ML20135E150; ML20135E155; ML20135E430;
References
	NUDOCS 9703060297
	Download: ML20135E150 (16)
	v • d • e

-..

l fW Q U

1 UNITER, STATES g

s j

NUCLEAR REGULATORY COMMISSION l

WASHINGTf)N, D.C. m1001

%.....)

COP 990NWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.153 License No. DPR-19 1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated September 20, 1996, as supplemented January 21, 1997, 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; B.

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 (1) 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.

2.

Accordingly, the licensa is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-19 is hereby amended to read as follows:

l l

l t

i 9703060297 970228 PDR ADOCK 05000237 P

PDR

1 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.153, are hereby incorporated in the license. The licensee shall operate the facility in accordance i

with the Technical Specifications, l

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

i 1

FOR THE NUCLEAR REGULATORY COMISSION W

+.

\\

lhn F. Stang enior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical l

Specifications l

Date of Issuance: February 28', 1997 i

I l

.9 i

l 1

l

k UNITED STATES g

g

}

NUCLEAR RECULATORY COMMISSION WASHINGTON, D.C. SpeeHOM

\\...../

COMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERAT?NG LICENSE l

\\

l Amendment No.148 License No. DPR-25 1.

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

A.

The application for amendment by the Comonwealth Edison Company (the licensee) dated September 20, 1996, as supplemented January 21, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; l

B.

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 comon 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 Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B. of Facility Operating License No. DPR-25 is hereby amended to read as follows:

l l

I

i )

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 168 are hereby incorporated in the license. The licentee stall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COP 911SSION j

G d

i J n F. Stang, Seni*or Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 28, 1997

/

ATTACHMENT TO LICENSE AMENDMENT N05.153 AND 148 FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Revise the Appendia 4 Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT VIII,

VIII 3/4.6-19 3/4.6-19 3/4.6-20 3/4.6-20 3/4.6-21 3/4.6-21 3/4.6-21a 3/4.6-21b 3/4.6-21c, 3/4.6-21d B3/4.6-6 B3/4.6-6 B3/4.6-7 B3/4.6-7 B3/4.6-8 B3/4.6-8 B3/4.6-9 l

i l

l i

TABLE OF CONTENTS TOC l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1

SECTION fAGE l

3L(4 PRIMARY SYSTEM BOUNDARY.

3/4.6.A Recirculation Loops........................................ 3/4.6 1 3/4.6-3 3/4.6.B Jet Pumps 3/4.6.C Recirculation Pumps.............'........................... 3/4.6-5 3/4.6.D Idle Recirculation Loop Startup 3/4.6 6 3/4.6.E Sa fety Valves............................................ 3/4.6-7 3/4.6.F Relie f Valve s............................................. 3/4.6-8 3/4.6.G Leakage Detection Systems.................................. 3/4.6-10 3/4.6.H Operational Len! ege....................................... 3/4.6-11 3/4'.6-13 3/4.6.1 Chemistry Table 3.6.l-1, Reactor Coolant System Chemistry Limits 3/4.6.J Specific Activity............................. '............. 3/4.6-16 3/4.6.K Pressure / Temperature Limits................................. 3/4.6-19 Figure 3.6.K-1, Pressure-Temperature Limits for Pressure Testing - Valid to 18 EFPY Figure 3.6.K-2, Pressure-Temperature Limits for Pressure Testing - Valid to 20 EFPY Figure 3.6.K-3, Pressure-Temperature Limits for Pressure Testing - Valid to 22 EFPY j

Figure 3.6.K-4, Pressure-Temperature Limits for Non-Nuclear Hestup/Cooldown - Valid to 22 EFPY Figure 3.6.K 5, Pressure-Temperature Limits for Critical Core Operations - Valid to 22 EFPY l

3/4.6.L Reactor Steam Dome Pressure................................ 3/4.6-22 3/4.6.M Main Steam Line Isolation Valves.............................. 3/4.6-23 3/4.6.N Structural Integ rity........................................ 3/4.6-24 3/4.6.0 Shutdown Cooling - HOT SHUTDOWN.......................... 3/4.6-25 3/4.6.P Shutdown Cooling - COLD SHUTDOWN......................... 3/4.6 27 i

DRESDEN - UNITS 2 & 3 Vill Amendment Nos.153 and 148 l

l I

l

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K i

3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS K. Pressure / Temperature Limits K. ~ Pressure / Temperature Limits The primary' system coolant system 1.

During non-nuclear heatup or temperature and reactor vessel metal cooldown, and pressure testing temperature and pressure shall be limited as operations, at least once per 30 specified below:

minutes, l

1.

Pressure Testing:

a. The rate of change of the primary i

system coolant temperature shall a.

The reactor vessel metal be determined to be within the j

temperature and pressure shall be hostup and cooldown rate limits, l

maintained within the Acceptable and l

Regions as shown on Figures 3.6.K 1 through 3.6.K-3 with the

b. The reactor vessel metal rats of change of the primary temperature and pressure shall be system coolant temperature determined to be within the 520*F per hour, or Acceptable Regions on Figures b.

The rate of change of the primary system coolant temperature shall 2.

For reactor critical operation, determine be $100'F per hour when reactor within 15 minutes prior to the i

vessel metal temperature and withdrawal of control rods and at least pressure is maintained within the once per 30 minutes during primary j

Acceptable Regions as shown on system heatup or cooldown, Figure 3.6.K-4.

a. The rate of change of the primary 2.

Non-Nuclear Hestup and Cooldown and system coolant temperature to be low power PHYSICS TESTS:

within the limits, and a.

The reactor vessel metal

b. The reactor vessel metal l

l temperature and prospute shall be temperature and pressure to be l

maintained within the Acceptable within the Acceptable Region on Regions as shown on Figure Figure 3.6.K-5.

3.6.K-4,, and 3.

The reactor vessel material surveillance b.

The rate of change of the primary specimens shall be removed and system coolant temperature shall examined, to determine changes in be $100'F per hour, reactor pressure vessel material properties in accordance with 10CFR Part 50, Appendix H.

?

l i

DRESDEN - UNITS 2 & 3 3/4.6-19 Amendment Nos. 153 and 148

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS 3.

Nuclear Heatup and Cooldown:

4.

The reactor vessel flange and head flange temperature shall be. verified to

a. The reactor vessel metal be k83*F:

temperature and pressure shall be

. maintained within the Acceptable

a. In OPERATIO.NAL MODE 4 when l

Region as shown on Figure 3.6.K-5, the reactor coolant temperature is:

and l,

1) 5113'F, at least once per

b. The rate of change of the primary 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

system coolant temperature shall be l

5100*F per hour.

2) 593*F, at least once per 30 minutes.

4. The reactor vessel flange and head flange temperature k83*F when

b. Within 30 minutes prior to and at i

l-reactor vessel head botting studs are least once per 30 minutes during under tension, tensioning of the reactor vessel head botting studs.

APPLICA81LITY:

1 1

At all times.

ACTION:

With ariy of the above limits exceeded,

1. Restors the reactor vessel metal temperature and/or pressure to within the limits within 30 minutes without exceeding the applicable primary system coolant temperature rate of change limit, and 2.

Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system and determine that the reactor coolant system remains acceptable for continued operations within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or

3. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pnd in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

DRESDEN - UNITS 2 & 3 3/4.6-20 Amendment Nos.153 and 148 i

l l

PRIMARY SYSTEM. BOUNDARY PT Limits 3/4.6.K FIGURE 3.6 K-1 PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VAllD TO 18 EFPY

/

i j

1400 I

N 1200

]

l I

[ UNACCEPTAnl.E IWGION l Ig 1000 I

i g

i

Tamperature change Rate l-goo 538*F/hr

[

I acCErranus naamn l

a I

noTrou 000 -

mm b

E f

/

a 1

/

I i

i 400 1

UPPER VESSEL AND f

e l

-. hTTOM UMfTS e

i 200 l

l anaa m i

I I

8

}

0 50 100, 150 200 250 300 350 MONMUM REACTOR VERSEL MAL TBFBtATtstE m i

j DRESDEN - UNITS 2 & 3 3/4.6-21 Amendment Nos. 153 and ] 48 1

i e

4

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K' FIGURE 3.6.K-2 PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VALID TO 20 EFPY 1400 I

I 1200 I

\\

4 f

I I unaccerrasta nemew l !)

1000 i

Tampemeurs change Rate F

s 30? Air f

I AcCEPTAsL2 IEGON' l eoTTou 000 - -

esAnarr E

f i

4 i

o

~

i

/

i i

400 l

I e

UPPER VESSEL AND I m2 m l BELTUNE UMITS g

1 8

- - BOTTOM HEAD 8

UMITS ILANGE r

.rr

~

i

'l I

i o

i 0

50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL BETAL TEbrWtATURE (T) i DRESDEN - UNITS 2 & 3 3/4.6-21 a Amendment Nos.153 and 1 48 4

1 r-

i PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K i

FIGURE 3.6.K-3 PRESSURE - TEMPERATURE LIMITS FOR PRESSURE TESTING - VALID TO 22 EFPY 1400 w I

i I

i 1200 I

I unaccEmsg MGWN l I

1000 i

g l

l Temperature Change Rate i

s 500 530*FAr o

I.00

/

I AcCEmsu REGON' l l

d s

/

i.

4 i

/

l400 [

N VESSEL ANQ s

I statelo l RTUNE UMITS l-

- = sawi ivid HEAD LIMITS j

e 200 a

gag 0N E37 I

0 0

50 100 150 200 250 300 350 INNIMUM REACTOR VESSEL BETAL1BrWtATURE (T)

DRESDEN - UNITS 2 & 3 3/4.6-21 b Amendment Nos.153 and 14 8

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K FIGURE 3.6.K-4 PRESSURE - TEMPERATURE LIMITS FOR NON-NUCLEAR HEATUP/COOLDOWN - VALID TO 22 EFPY 1400 r

e l

I l

l l

1200 I

l 1 uNaccerraste neGON l i

I I-1000 i

f I

/

l

TemperSture Change Rate a

800 g

siteTthr I

800

/

E g

l

/

l.-

f 400 UPPER VESSEL AND i

t --

i _. mums a

UMITS M

l SOTTOM e

HEADENT I /

8 PUWIGE REGON EFF l i

i 0

O 50 100 150 300 E50 300 350 ISAMUM REACTOR VE5SEL IETAL TarEtATWtE PP)

DRESDEN - UNITS 2 & 3 3/4.6-21c Amendment Nos.153 and 148

PRIMARY SYSTEM BOUNDARY PT Limits 3/4.6.K FIGURE 3.6.K-5 PRESSURE - TEMPERATURE LIMITS FOR CRITICAL CORE OPERATIONS - VALID TO 22 EFPY 1400 f

1200 1000 l UNACCEMABl.2 IEEGION l 1

,00 g

stesTAw I

800 5

g i

I400

/

-5ELTUNE AND NON-BELTUNE UMITS Y

=

M ru m u m cetna m y l ACCEPTABLE IEGION

l 0

0 50 100 - 150 200 250 300 350 RENRIUM REACTOR VERSEL MAL TWrERAftNtE (*F)

DRESDEN - UNITS 2 & 3 3/4.6-21 d Amendment Nos.153 and 14U

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES 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.

The pressure-temperature limit lines are shown, for operating conditions; Pressure Testing, Figures l

3.6.K-1 through 3.6.K-3, Non-Nuclear Heatup/Cooldown, Figure 3.6.K-4, and Core Critical Operation Figure 3.6.K 5. The curves have been established to be in conformance with Appendix G to 10 CFR Part 50 and Regulatory Guide 1.99 Revision 2, and take into account the change in reference nil-ductility,trancion temperature (RTal as a result of neutron embrittlement.

l The adjusted reference temperature (ARTI of the limiting vessel material is used to account for i

irradiation effects.

Four vessel regions are considered for the development of the pressure-temperature curves: 1) the core beltline region; 2) the non-beltline region (other than the closure flange region and the bottom head region); 3) the closure flange region and 4) the bottom twed region. The beltline region is defined as that region of the reactor vessel that directly surrounds the effective height of the reactor core and is subject to an RT adjustment to account for radiation embrittlement. The non-beltline, closure flange, and bottom head regions receive insufficient fluence to necessitate an RT. adjustment. These regions contain components which include; the reactor vessel nozzles, closure flanges, top end bottom head plates, control rod drive penetrations, and shell plates that do not directly surrovT ee reactor core. Although the closure flange and bottom head regions are non-beltline regioni,, @4f are treated separately for the development of the pressure-temperature curves to address 10CFR Part 50 Appendix G requirements.

Bottuo Tomoerature The limiting initial RT of the main closure flanges, the shell and head materials connecting to these flanges, connecting welds and the vertical electroslag welds which terminate immediately below the vesselflange is 23T. Therefore, the minimum allowable boltup temperature is esur fished as'83T (RT

+ 60T) which includes a 607 conservatism required by the original ASME Code of construction.

Fiaures 3.6.K-1 throuah 3.6.K Pressure Testina As indicated in Figure 3.6.K-1 through 3.6.K-3 for pressure testing, the minimum metal temperature of the reactor vessel shellis 837 for reactor pressures less than 312 psig. This 83T minimum boltup temperature is based on a RT, of 237 for the electrostag weld immediately below the vessel flange and a 607 conservatism required by the original ASME Code of construction. The bottom head region limit is established as 68T, based on moderator temperature assumptions for shutdown margin analyses. At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 113T. The 113T minimum temperature is based on a ciosure flange region RT, of 237 and a 907 conservatism required by 10CFR Put 50 Appendix G. Beltline curves as a function of vessel i

' exposure for 18,20 and 22 ef'ective full power years (EFPY) are presented to allow the use of l

the appropriate curve up to 22 EFPY of operation.

DRESDEN UNITS 2 & 3 8 3/4.6-6 Amendment Nos.153 and 148

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES Figures 3.6.K-1 through 3.6.K-3 are governing for applicable pressure testing with a maximum heatup/cooldown rate of 20*F/ hour.

Fioure 3.6.K Non-Nuclear Heatun/Cooldown Figure 3.6.K-4 applies during heatups with non-nuclear heat (e.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram). The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress. The maximum heatup/cooldown rate of 100*F/ hour is applicable.

Flame 3.6.K Core Critin at Onoration The core critical operation curve shown in Figure 3.6.K 5, is generated in accordance with 10CFR Port 50 Appendix G which requires cora critical pressure-temperature limits to be 40*F above any pressure testing or non-nuclear heatup/cooldown limits. Since Figure 3.6.K 4 is more limiting, Figure 3.6.K 5 is Figure 3.6.K-4 plus 40*F. The maximum heatup/cooldown rate of 100*F/ hour is applicable.

The actual shift in RT,, of the vessel material w!!! be established periodically during operation by removing and evaluating, in accordance with ASTM E185-82 and 10CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The irradiated specimens are used in predicting reactor vessel material embrittlement. The operating limit curves of Figures 3.6.K-1 through 3.6.K-5 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

314.6.L Reactor Steam Dome Pressure The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria. The reactor steam dome pressure of :s1005 psig is an initial condition of the vessel overpressure protection analysis. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

3/4.6.M Main Steam Line Isolation Valves Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two DRESDEN - UNITS 2 & 3 B 3/4.6 7 Amendment Nos.153 and 148

l i

PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES valves be OPERABLE. The surveillance requirements are based on the operating history of this type of valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent i

with the assumptions in the safety analyses to prevent pressure surges.

3/4.6.N Structural intearity j

The inspection programs for ASME Code Class 1,2 and 3 components ensure that the structural integrity of these components will be maintained at en acceptable level throughout the life of the plant.

The inservice inspection program for ASME Code Class 1,2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by thc, NRC pursuant to 10 CFR Part 50.55alg)(6)(i).

3/4.6,0 Shutdown Ceulina - HOT SHUTDOWN 3/4.6.P Shutdown Coolina - COLD SHUTDOWN j

Irradiated fuelin the reactor pressure vessel generates decay heat during normal and abnormal i

shutdown conditions, potentially resulting in an increase in the temperature of the reactor coolant.

This decay heat is required to be removed such that the reactor coolant temperature can be reduced in preparation for performing refueling, maintenance operations or for maintaining the reactor in cold shutdown conditions. Systems capable of removing decay heat are therefore required to perform these functions.

l A single shutdown cooling mode Ibop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure i

considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

l a

DRESDEN - UNITS 2 & 3 8 3/4.6-8 Amendment Nos.153 and 148 1

-1

ML20135E150

Text

Attachment:

Attachment:

Navigation menu

Search