ML20087D792

Amend 74 to License DPR-25 Changing Tech Specs to Support Cycle 9 Operation W/Reload Fuel Supplied by & Associated Analyses Performed by Exxon

ML20087D792
Person / Time
Site:	Dresden
Issue date:	03/09/1984
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20087D794	List: ML20087D792 ML20087D840
References
NUDOCS 8403140143
Download: ML20087D792 (10)

Text

..

1. pn nec

'e UNITED STATES

[h

[j NUCLEAR REGULATORY COMMISSION 3g

'p WASHINGTO N, D. C. 20555 p$

w

....+

C0i'FONilEALTH EDIS0N COMPANY DOCKET NO.

50-249 DRESDEN EUCLEAR P0;!ER STATION, UNIT 3

~

AMENDriENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. DPR-25 1.

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

A.

The applications for amendment by the Commonwealth Edison Company (the licensee) dated July 18, 1983, and August 25, 1983, as supplemented by letters dated November 3, 10 and 30, 1983, two letters dated Decceber 13, 1983, and a letter dated December 16, 1983 comply 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 applications, 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, ar.d (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

i 8403140143 84d309 DR ADOCK 05000 4y

l.

j 2.

Accordingly, the license is amended by changes to the Technical i

Specifications as irdicated in the attachnent to this license anendnent and paragraph 3.B of Facility Operating License

~

No. DPR-25 is hereby amended to read as follows:

B.

Technical Specifications

'The Technical Specifications contained in Appendix A,.as revi. sed through Anendment No.

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 date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f lA t

Dennis M. Crutchfield,, Chief Operating Reactors Branch #5 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: March 9, 1984 4

4 I

l

ATTACHMENT TO LICENSE AVENDlENT NO. 74 FACILITY OPERATING LICENSE DPR-25 DOCKET i:0. 50-249 Revise the Technical Specifications by replacing the following pages with attached revised pages. These revised pages contcin the captioned amendment nunber and marginal lines to reflect the area of change.

Remove Paces Insert Pace 20 20 81C-1 81C-1 81D 81D 81E 81E 81E-1 E6A 86A 157 157 s

i

li Bases:

than 26,700 psi at an internal pressure of 1250 psig:

this is a factor of 1.5 1.2 The reactor coolant system Integrity is an important barrier in the prevention of uncontrolled release of below the yield strength of 40,100 psi at 575 F.

At that pressure limit of fission products.

It is essential that the integrity psig, the general membrane stress of this system be protected by establishing a pressure will nly be 29,400 psi, still safely limit to be observed for all operating conditions and qW the yield strength.

whenever there is irradiated fuel in the reactor vessel.

The relationships of stress 1cvels to yield strength ar'e comparajale for the The pressure safety limit of 1345 psig as measured lay, primary system piping and provide a the vessel steam space pressure indicator ensures margin to 1375 psig at the lowest elevation of the reactor similar margin of protection at the vessel. The 1375 psig value is derived from the design established safety pressure limit.

pressures of the reac';or pressure vessel and coolant system The normal operating pressure of the piping. The respective design pressures are 1250 psig reactor coolant system is 1000 psi at 575"F and 1175 psig at 5600f. The pressure safety For the turbine trip or loss of I

l limit was chosen as the lower of the pressure transients electrical load transients, the turbine permitted by the applicable design codes: ASME Boiler tr p scram or generator load rejection and pressure Vessel Code,Section III for the pressure scram, together with the turbine b ass em limit vessel and USASI 831.1 Code for the reactor coolant the pressure to approximately.

vaf2).

In addition, pressure g

system piping.

The ASME Boiler and pressure Vessel g

ves have been provided to Code permits pressure transients up to 10% over design "d"

y tyes dh bility of the safety pressure (110% X 1250 = 1375 psig), and the USASI Code harged to the drywell, permits pressure transients up to 20% over the design operatin in ti pre'ssure (120% X 1175 = 1410 psig). The Safety Limit bypass alould fail

• pressure of 1375 psig is referenced to the lowest elevation of the reactor vessel. The design pressure Finally, the safety valves are sized for the rect /c. suction line piping (1175 psig) was

["

e reactor vessel peak pressure e

ps R with no credit taken cl.osen relative to the reactor vessel design pressure.

g Demonstrating compliance of the, peak vesse'l pressure p st I te 11 l sure of all MSIV's l

"E with the ASHE overpressure protection limit (1375 psig) without di (v IV p sition switch) assures compliance of the suction piping with the scram.

Cr t f USASI limit (1410 psig).

Evaluation methodology used flux scram, howeve i

to assure that this safety limit pressure is not The indirect flux scram and cafety valve exceeded for any reload as documented in lieference cruation provide adequate margin XN-NF-79-71.

The design basis for the reactor yl cak allowable vessel pressure he pressure vessel makes evident the substantial margin p

g.

of protection against failure at the safety pressure limit of 1375 psig.

The vensel has been designed Icactor pressure is continuously monitored n

for a general membrane stress no greater control room during operation on psi full scale pressure recorder.

T4) SAR. Sect ion fl~7.2 -

also:

"Dresden 3 Second Rel,oad License ftmendmentNO.,4',)

submittal." 9-14-73 20

,,Dresden Station Special Report a so:

No. 29 Supplement H."

d g

DPR-25

~

6

.,.1..

g

...,6 4

..A a.r.a.. :...

.l

,i.. ;.. e a-

2....;..;a

.. p.. t...... a. ;._ a. p......,......

.. ; ~..._. :.. u...:

.3.:,

.;r.

r=

.,n

_:2 m. 4......a.. : n.....

.. x. 2.. a....j..........a. u.

. 2.... _ =..u..n.

... a.. u.

..n. g

_4_.....

~

...... ~..

."8..n 5:*

..... ll.:.... l".u.._..

• =8

~ :t. =..........J:.-

nla. :

. ;.. (......-.

..g..

....4.-

. _. i..

r.. j...

..r.. :3...c..n.

n...=.....=n...

n 13 0

....'.."....~..r:n..=.r,~~

g

" ~.*"..

~

d^

t ~i~

=J..._.._

u ".

"-:.:. n. "r

": ___--.".-"."r"n ns:

= u..

. u : n: : -

e l c 4.3 wN 12' '

'[-[

Sfi* :!. 7-Q:jhh"!!!j.-h- "(:);d. "*;' :il,:Iij [ih

• d

• 'h.!

f;'73:;..,. "".h C L-e

.. n..U. :.1..:.a... l :.....G. J.... l... T.;.... _.,

...'...................l."..a.."..*..........

. -.... :. 8_. 14.

.... f *..

• . 8;

......._.....l....

...l...

~.

.......--..."3..

....,..l.._.

N

... s. a.. h.=......_n. in. _.. 4.. n.. l a..... =_.::. n...a.. :::

. _::.c..r._........._

n..-..._:I. n..:....

2: =. I.

_..a.. E........ :l. :

.nt.:=...u.4 :a.. 2.._;._

k-

...........a.........._.......

.I 5

12.O

.2.. a u... a U

. =.;:=.f.... = a a p n a. = = j :=7-

=- r -- =;,. =:= 24_.. u.. j.

.i.. :i.. :.1. c =

==

ns:-

n-a :== ::. = a.--- -

=.

..=; u.;. :- -

~. u,

= =.;.

a

..,i:....:

_=i= r.n-

.= n.....==.

. m a::= -

a f_=_:~:.+:. : :=

g

=:- -

:- - r :2==. =.. 3 =... -,- - r = n...nt:= t =_: --- = : :: :_

c":

e-4 a

- =:

i

.'.z.:.._..._.....u...=.u..

-. : a.

_ a.. : _._..._. s.. a:.=..

....."t.=.=..=.. ;.. u...: : r..a.. =._.a.=.__=. - _. : =._.. t _......u.s. _._. _u __u.

C C0

..::..~a _: :==_a-

.2= nr:.::-=:.:a ::t.:--..~_ ~.. _~.. ~.. 1. - _E-.:.. r -- ~.t"..=..n :.

=. =a _.r =-

~.~\\..=_-..

..........t.~..~.-

...~..~..~..~.-.".-e..

4*

3.

~

_. a.... a.._..=..a. r.:._=. =.

==__:=t - _ _,.. n_ n2._u..: n.=~ - -- _: =.: = : =: x. :

n 2 ---

u_

-_=_.._-;,._.__.,

._a =....._e._; :.a. _=.. :._

m u

0 6

. =: a.n.=....

_._.4_

f- - p= =.- n =_u

_ _ _ =..

= =.d. ' _" ; "t=.: =-. --

=

1. U":

>b

=--

=

.::; o

! -..=i=i: = *=e Dresden Unit 3

" n =.".=i==GE- - -

c o

= =. =n :

.. = =

a

=t.--

1==

-===n

.== =. __ v = r=

..=-=======

_... _ -__.--=n Fuel e: XN8D2.6 -

...=___=.4._.u._a._._._=..__.

=_..

Puel T;pe: XN6D2.83-5

-_--_._i===g===.=.=_.=.==g_=r=.-

Ka

====>=.:.=.;==-

= - -.

= =.

n. *o

.....-..L..

.4

=.._=_:.:=...=._=.._=_=....__._=..=._.._._._._________ _. _..__._._ m 7.

.. a.. n_ -....=_.. r. -. u_ =..

......r............_...__.__.__..._,___:___2_._._a.....

. _.._......_.._::.__.._...._ _._ =.._2_. __ _ _ _.

<.0 Q

= J~l :. u...a _ _ _=.4,. = = _2.= = = =-- --

.._a a

_: _= :. 'l.:_=. = _==. : :. = = + _==.u,. = _

. =..y-... -. n i.

=

.:~.u-

-s====: -

. - - -.._--.:==

=.: =..

v-,-_a_.__=_.>__._.....__

c_=.._.1_._.._._

d _. -...... _. a =_..

.1,000 20,000.

...-..___....._.._....u......

.=....=.a..._. _... _

.s. _.. _...

40,000 30,000 Bundle Anrage Exposure (MWD /MT)

J s

J Maximum Average Planar Linear Figure 3.5-1 He.at Generation Rate (MAPLHGR)

(Sheet 1 of 5) vs.. Bundle Average Ex.oosure L.

r Amendment No.,7, f6 74,

(i

l. [ )

i J

i OPR-25 I

3.5 LlHITING CONDITIONS FOR UPERATION 4.5 StlRVEILIANCE REllOIREMENTS 96 K.

Minimum Critical Power Ratio (HCpR)

K.

Minimune Critical Power Ratio (MCPRJ During steady state operatton at rated core i

Mcln shall be determined dalTy during a flow HCPR shalt Ise greater titan or estiset tes -

reactor power operation at 3.251 rated 1 34 fir r.C 8 x 8R fuel thermal power and Intlowirg any change in p.,,er level or distritsut1o3 that would 1.33 for f.NC and CC 8 x 8 fuel enuse operatAun with n ilma ting control rod pattern se described in tiie bases for For core flows other than rated, the MCPR SpectfIcatinen 3.3.5.5.

Operating Limit shall be as follows:

1.

Manual Flow Control - the HCPR Operating Limit shall be the value from Figure 3.5-2 sheet 1 or the above rated flow value, whichever is greater.

2.

Automatic Flow Control - the MCPR Operating Limit shall te the value from Figure 3.5-2 cheet 1. sheet 2. or the above rated flow value, whichever is greatest.

e if at any time aloring steady state power olie rat ion, it in determlncd that the Ilmiting value for MCI'M la licing exceeded, net.lon al al l lac init iated within 15 minutes to restore operat ion to wis liin t iie prescrlhed limits.

I f a lie st eady ne at e HCI'M is nos returned to willoin a he preserlhed limle a within two (2) leisu ra, the reactor shall he brought to a he Cold Shus shown condit ton i..t hin M bourai.

Surveillance and correspondling net inn wha t I cont lause uneii renceor esperallosi is withlei the prescribed limits.

In the event the average 901 scram insertion time detennined by Spec. 3.3.C for all operald e control.

rods eacceds 2.58 secnnds, the MCP81 liselt shall he Increased by the anount equal in [0.0544T - 0.14]

where i equals the average 90% scram insertion time for~ the most recent halt-core or full core surveillance Amendment [, g M SID data from Spec. 4.3.C.

l I

s

.r T, s,

?

DPR-25.

1. ! -

1.4

.t' 5

?

% 1.3 b

e ce S-x t

1.2 1.1 a

e a

1.0

^

^

30 40 50 60 70 80 90 100 Total Core Recirculating Flow (1 Rated 98 alb/hr)

~

l Figure 3 5-2 (Sheet 1 of 2) MCPR Limit' Fdr ' Reduc'ed Core Flow n.>

r..

a 81E Amendment No.

74 l

~

DPR-25'-

1.7 1.6

%, 'Ing

.t g '*f t

• j* f *t m3

% 1.5 k

1. 1. etr l

%"at i

fm* l1,I ts% )'

3 b

t.13 g,4

, WPg PatIng gI 1.3 Mit

• j

%s}-

2 et Yet 1.2 30 40 50 60 70 80 90 100'~

Total Core Recirculating Flow (X Rated, 98 alb/hr)

F16ure 3 5-2 (Sheet 2 of 2) MCPR Limit For Automatic Flow Control

~

~

Amendment No. 74

II DPR-25 I

4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd. )

1.

Average Planar LHGR K.

Minimum Critical Power Ratio (MCPR)

At core thermal power levels less than or equal to 25 per cent, operating pl&nt experience and thennal hydraulic At core thermal power levels less than or equal to 25 per analyses indicate that the resulting average planar LHGR is cent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

below the maximum average planar LHGR by a considerable For all designated control rod patterns which may be employe margin; therefore, evaluation of the average planar LHGR below at this point operating plant experience and thermal this power level is not necessary. The daily requirement for hydraulic analysis indicates that the resulting MCPR value is (

calculating average planar LHGR above 25 per cent rated in excess of requirements by a considerable margin thermal power is sufficient since power distribution shif ts low void content, any inadvertent core flow increase would With this are slow when there have not been significant power or control rod changes.

place operation in a more conservative mode relative to The daily requirement for calculating HCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

l J.

Local LHGR

- The LHGR for G.E. fuel shall be cheded daily during reactor operation at greater than or equal to 9.5 per cent power to determine if fuel burnup or control rod :novement has caused changes in power distribution. A l'imiting LHGR value is precluded by a considerable' margin when employing a

, permissible control rod pattern below 257, rated thermal power.

9 8

AmendmentNo.,4I, 74 86A

DPR-25 5.0 htSIG?1 FEATimES 5.1 Site 5.4 Containnent Dresden Unit 3 is located at the Dresden A.

The pr incipal design parameter s and '

riuclear Power Station which consists of applicable design codes for the primary a tract of land of approximately 953 acres containment shall be as given in Table located in the northeast quarter of the 5.2.1 of the SAR.

Morris 15-minute quadrangle (as designated by the tinited States Geological Survey),

8.

The secondary containment shall be Goose Lake Township, Grundy County, IL.

as described in Section 5.3.2 of the SAR The tract is situated in portions of and the applicable codes shall be as Sections 25, 26, 27, 34, 35 and 36 of described in Section 12.~1.1.3 of the of Township 34 fiorth, Range 8 East of the SAR.

Third Principal Meridian.

C.

Penetrations to the primary contain-

5. 2 Reactor ment and piping passing through such penetrations shall be designed in A.

The core shall consist of not more than accordance with standards set forth 724 fuel assemblies.

in Section 5.2.2 of the SAR.

B.

The reactor core shall contain 177 5.5 Fuel Storage cruciformshaped control rods.

The control material shall be boron carbide A. The new fuel storage facility shall powder (B C) compacted to approximately be such that the Keff dry is less 4

70% of theoretical density, or Hafnium than 0.90 and flooded is less than metal.

0.95.

5.3 Reactor Vessel B. The Keff of the spent fuel storage pool shall be less than or equal to The reactor vessel shall be as described in 0.95.

Table 4.1.1 of the SAR:

The applicable design codes shall be as described in 5.6 Seismic Design Table 4.1.1 of the SAR.

The reactor building and all contained engineered safeguards are design for t'he maximum credible earthquake ground motion with an acceleration of 20 percent of gravity.

Dynamic analysis was used to determine the earthquake acceleration, applicable to the various elevations in the reactor building.

/cendment tio.,56 74

.157 6312ti

__ _