ML20032D515
| ML20032D515 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png (DPR-45-A-026, DPR-45-A-26) |
| Issue date: | 11/06/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20032D512 | List: |
| References | |
| NUDOCS 8111170204 | |
| Download: ML20032D515 (7) | |
Text
'i f""%,
UNITED STATES
- r J't '
NUCLEAR REGULATORY COMMISSION g*
(l WASHINGTON, D. C. 20555 g
j
+
- 1 0 la$
DAIRYLAND POWER COOPERATIVE DOCKET NO. 50-409 LA CROSSE BOILING WATER REACTOR AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 26 License No. DPR-45
, 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Dairyland Power Cooperative (the licensee) dated June 1,1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as emended (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 c~ompliance 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.
P
~
2.
Accordingly, the license is amended by changes to this Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C(2) of Provisional Operating License No. DPR-45 is hereby amended to read as follows:
~~ '
(2) Technical Specifications
~ ~ ' ~
'" ~
,i s i.-
,7_r
.::: ci n:: C:
uL
=~- =-
~~The Technical Specifications contained in Appendix A
~'
issued October 31, 1969, with Authorization No. DPRA-6, as revised through Amendment No. 26, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical
' Specifications.
-- a c r.i--
- . 2 1 - <_
.<1 m cr.biv c:PCsure ci L' Y C00-ii w
e
- 3. ' This license amendment.is ~ effective-as.of.tfie, date.of 1ts.dssuance.
t
. = = =
^
FORTHE NUCLEAR REGULATORY COMMISSION 2.
w Dennis M. Crutchfield, Ch_ief Operating Reactors Branch '#5 Division of Licensing Y
L -- -
-L
~ -~ ~ ~~
~
. L.
L - :-
Attachment:
Changes to the Technical Speci fications Date of Issuance:
November 6,1981 t
l I
e k.
9
. ATTACHMENT TO LICENSE AMENDMENT NO. 26 i.A CROSSE B0ILIN'G WATER REACTOR (LACBWR).
PROVISIONAL OPERATING LICENSE NO.'DPR-45 Revise Appendix A by replacing the following pages with the enclosed'-
1 pages. The revised pages are identified by the captioned ameroment number and contain marginal 1'nes indicating the areas of chariye.
PAGES
~
- r
... = 32aa =~-
=
-= - - - '
1 4
32gg '
~
i
~
32hh 1
- - - - 32i f -
- ~ - ~
=.
a m.e.
1
-.-e.
g.
9 i
i 9
i e
O
1
- 32aa -
T POWER DISTRIBUTION LIMITS MAXIMUM AVERAGE FUEL ASSEMBLY EXPOSURE LIMITING CONDITION FOR OPERATION' 4.2.4.2.5 The maximum average exposure of any fuel assembly not on the periphery of the core shall be limited to 16,8'00 MWD /MTU.
l APPLICABILITY:
OPERATIONAL CONDITION 1.
ACTION:
.i
}- - With the maximum average, fuel assembly -exposure of-any non- -- -- --
.. peripheral assembly greater than' 16,8 00 MWD /MTU, be(in-at least l
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
)
HOT SHUTDOWN with the main, steam line isolation valve closed m
- '..- ~
y _:, -
^
SURVEILLANCE REQUIREMENT 5"2.17.5 The maximum average exposure of each fudl assembly not on the periphery of the core shall be determined to be les5 than 1E,200 MWD /MTU by calculation ar least once per 31 EFPD.
l Amendment No. JT, g, J#, 26
-g-
3
- 32gg -
K.
1 POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17
~~ '
LINEAR HEAT GENERATION RATE - (Continued)
~,
For Type I and Type II (A-C) fuel, the original design LINEAR HEAT GENERATION RATE specified by the fuel manufacturer was conserva-tively reduced to 11.94 kw/ft to account for the effects of dens-ification, power spikes and manufacturing-factors.
For Type III (ENC) fuel, the design LINEAR HEAT GENERATION RATE of 11.52 kw/ft is also calculated with design conservatisms that are larger than the calculated axial densification effects plus manufacturing tolerances : and -power' ~spik.e effects, Reference 6 and 7.
The daily requi'r'ement for' surveillance of th'e'co're LHGR:above:25%
of RATED THERMAL POWER is 'ufficient since power-distribution s
shifts are very' slow when there have not been significant power or control rod changes.
The surveillance of core LHGR after power increases > 15% of RATED TEERMAL POWER will assure that significant incr. eases in LEGR'are determined.
4.2.4.2.5 and 5.2.17.5 Maximum' Average Fuel' Assembly E:@osure Fuel cladding integrity is a. f' unction of many parameters including fuel exposure, pellet clad interaction, THERMAL POWER, ratesof change in power density, coolant chemistry,.etc.
Therefore, liniting fuel exposure to 16,800 MWD /MTU in the non-peripheral
[
fuel assemblies which experience higher than average power densities and rates of change of power will give additional assurance that the condition of the fuel during operation will be satisfactory.
It is not necessary to limit exposure in the peripheral ccre locations since operating experience at LACBWR has shown that the '8 peripheral fuel assemblies have a much lower L
rate of failure tt the 44 interior fuel assemblies.
This trend has been attributec te the icwer pcwer density at these locatiens, i
and the minimal ef fects of control rod movements which cause local power peaking in the fuel rods near the tips of the control rods.
The outer control rods e.re fully withdrawn at the beginning of
' cycle (BOC) and remain withdrawn during normal cycle operations.
Minor clad defects that may occur in the peripheral core positions would be expected to develop very slowly, and the consequence.s of such failures would be-minimal.
During previous operation with l
a number of fuel assemblies have exceeded 15,600 MWD /MTU A-C fuel, j
without any indication of failure and at the end of Cycle 3,CEOC-3),
four assemblies had exceeded 18,000 MWD /MTU without failure.
The averein exposure of the 25 assemblies discharged. at EOC-3 was 15,530 MWD /MTU~and the peak exposure was 21,532 MWD /MTU.
The aw nge exposure of the 32 assemblies discharged at EOC-4 was 16,459 MWD /MTU.
~
Amendment No. W. M, ?4 2o
~ '.
~
u 32hh -
POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4. 2.'4. 2 and 5. 2.17 Maximum Average Fuel Assembly Exposure * (Continued)
Pellet-clad interaction is_a well known'and documented contrib-The presence'of pellet uting factor to fuel rod failures.
cladding interaction has been identified in post-irradiation examinations of fuel rods removed from LACBWR fuel assemblies.
Fuel rods removed from fuel assemblies with average exposure up to 14,700 MWD /MTU have been examined.
The strength, ductility, and condition of-the cladding in.these rods was found to be adequate.as det' ermined 'by mechanical' t;ests'.r 'The.. examination.further confirmed thati power history of the' rods is of. prime importance',?_y
~
though not the 'only factor in contributing to fuel rod f ailiire.'.~
l
~
MWD /MTU fuel-element ~ average exposure is con
- A limit of 16,8 00 sistant with the results obtained from examinations conducted on fuel assemblies with similar exposure history.(Reference 8,9)
(
During future operation the rate of withdrawal of control rods when the THERMAL POWER is above 25% of RATED THERMAL POWER will
~
be reduced from that experienced;during operation prior to.
Cycle-5 which will also significantly reduce the stresses in.the
.(
Additional surveillance and limitations.on coolant
~
fuel clad.
and off-gas activity will assure that operation does not continue
~
with grossly failed fuel.
References:
" Technical Evaluation Adequacy of La Crosse Boiling Water 1.
Reactor Emergency Core Cooling System", Report SS-942, Gulf United Nuclear Corporation, May 31, 1972.
" Review of Densification Effects in La Crosse Boiling Water 2.
Reactor", Report SS-1085, Gulf United Nucle.ar Corporation, May 15, 1973.
3.
NRC Safety Evaluation Report, Letter, Reid to Madgett, dated August 12, 1976.
"ECCS Analysis for Type II and Type III Fuels for the 4.
La Crosse Boiling Water Reactor", Exxon N':: lear Company, Inc., XN-NF-77-7, March 1977.
i
" Transient Analysis for LACBWR Reload Fuel", Response to 5.
Inc., Report 81A0025, Question 4, Nuclear Energy, Services, February 18, 1977.
Amendment No. AT, M,,26,26 I
---,=<.,,-ww--,,
.-e--
,,,p-,~n
,,,---g w,,,--e
,-.m
,,,,----,_m.e,-w-,e
- - - -w-.-,w--
,-,, - - - - - - - ~ ~
,,-,-.n--
w--
-aw-,
e,
New
- 3211 -
f, POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 References - (Continued) 6.
" Description of Exxon Type III Nuclear Fuel for Batch 1 Reload in the LACBWR", Dairyland Power Cooperative, LAC-3929, May 17, 1976.
7.
Exxon Nuclear Co. Letter, J. A. White to C. W. Angle,
Subject:
MAPLHGR Limits for Type I (Allis-Chalmers) Fuel,-
dated. June 22, 1977.
j, 8.
DPC Letter, LAC-6846, Linder to Ziemann, dated April 1,^1980.
9.
DPC Letter, LAC-7572, Linder to Crutchfield, dated June 1,1981.
l (Next page is page 33)
Amendment No. 36, E6 I