ML20086R806
| ML20086R806 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/21/1995 |
| From: | IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT |
| To: | |
| Shared Package | |
| ML20086R805 | List: |
| References | |
| NUDOCS 9507310289 | |
| Download: ML20086R806 (11) | |
Text
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L.-.
RTS-285 Attachmint 2 to NG-95-2067 Page1of1.
PROPOSED CHANGE RTS-285 TO THE DUANE ARNOLD ENERGY CENTER'
- TECHNICAL SPECIFICATIONS The holders oflicense DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said license as indicated on the attached marked up pages. The List of Affected Pages is given below.
LIST OF AFFECTED PAGES 3.5-10 3.8-6 5.5-1 j
6.5-3 6.8-1 6.8-2 i
SUMMARY
OF CHANGES-The following list of proposed changes is in the order that the changes appear in the Technical Specifications.
j Page Description of Changes 3.5-10 Revise Surveillance Requirement to allow credit when an OPERABLE EDG has been demonstrated to be OPERABLE within the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, so that it is not required to be run again within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after finding the other EDO inoperable.
3.8-6 Revise Surveillance Requirement for one ESW pump or loop inoperable to delete the reference to Surveillance Requirement 4.5.G.1 and reiterate the requirement to verify all low pressure are cooling and containment cooling subsystems and the diesel generator associated with the OPERABLE ESW are also OPERABLE.
i 5.5-1 Reformat Section 5.5 to be consistent in content and format with NUREG 1433, Improved Standard TS.
6.5-3 Delete "and implementing procedures" from items i and j.
6.8-1 Replace " Procedures required by the Emergency Plan" with the word
" Deleted."
6.8-2 Replace " Procedures required by the plant Security Plan" with the word
" Deleted."
95073102e9 950721 DR ADOCK 05000331 PDR
DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS t
G.
M__inimum Low Pressure Cooline and G.
Minimum Low Pressure Coolino and Diesel Generator Availability Diesel Generator Availability 1.
During any period when one diesel 1.
-dh
-it is termined that -- ne s
generator is inoperable,
-diesel-generatee-le-- i nope rabler (
continued reactor operation is
/ th; r;;;ining-diesel venerster permissible only during the shal-1 b: trn treted to ;.-
succeeding seven days unless such
-OPERABLE-in-socordanee-with -
diesel generator is sooner made
-Specification -4. 0. A. 2. a. l... withitr-(
OPERABLE, provided that the the fis.L 24 hews -and ;very-
/
remaining diesel generator and subsequett-22 hepr;j.eceaf_4er r_, /
all low pressure core and In7 ddition, all low pressure core centainment cooling subsystems cooling and containment cooling supported by the OPERABLE diesel subsystems supported by the generator are OPERABLE.
If this
{OPERABLEdieselshallbeverified requirennt cannot he me*.,
an t( be OPERABLE.
orderly SHUTDOWN shall be v
initiated and the reactor shall 4mJ 3,n,r.+or kw r4 W*
gg
,u be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in 5 " "*,
E"f uI m e n f 4. S. A. 2. 4.l.JL pe rGeb "fo demces f rade COLD SHUTDOWN within the O^ U be following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 0F 4 Af5n.Iry.f 4,
, c,,,, m,-
2.
Any combination of inoperable gen e to 24 h ours, components in the core and
"" Ic 5 ' P
- d " "'"2 QMW I-k containment cooling systems shall PreAo".
24 b rs,med hvery not defeat the capability of the swbs u <., f 72 hou r s (r. ~ Al remaining OPERABLE components to fulfill the cooling functions.
- " "" g^l
- ' " Q cc' 3.
When irrodiated fuel is in the N q reactor vessel and the reactor is
^#
in the COLD SHUTDOWN Condition or REFUEL Modes
~
a.
If no work is being
)
performed which has the potential for draining the reactor vessel, both core spray and RHR systems may be inoperable; or b.
If work is being performed which has the potential for draining the reactor vessel, at least two of any combination of core spray and/or RHR (LPCI or shutdown cooling mode) pumps shall be OPERABLE (including the capability to inject water into the reactor vessel with suction from the suppression pool) except as RTS-285~
Amendment No. IS$ I97 07/9g
DAEC-2 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
-E.
Emercency Service Water System E.
Emeroency Service Water System 1.
Except as required in Specification 1.
Emergency Service Water System 3.8.E.2 below, both Emergency surveillance shall be as follows:
Service Water System loops shall be OPERABLE whenever irradiated fuel a.
Simulated auto-once/
is in the reactor vessel and matic actuation OPERATING CYCLE reactor coolant temperature is test.
greater than 212*F.
b.
Pump and r40 tor As specified in I
operated valve the IST Program i
- r..
Flow Rate Test i
Each Emergency After major pump Service Water maintenance and I
pump shall once per 3 months, deliver at except weekly east that flow during periods of 87-06 etermined from time the river 88-01 Figure 4.8.E-1 water temperature for the exceeds 80*F.
existing river water temperature.
2.
With one of the Emerr-cy Service 2.
With one Emergency Service Water Water System pumps ops System pump or loop inoperable, the inoperable, REACT 0k > n R OPERATION OPERABLE pump and loop sh
'be must be limited to seven days s
vu if_ted_to,he OP 92-02 unless OPERABILITY of that system hdditien,th: F quira c,t:--cf-is restored within this period.
Specificatien-- 4.5.0.; 2;11 bc met.
During such seven days all active I In 4 M f:o n, a los tres-e- c a r e 1
components of the other Emergency co o l r ~3 and coM -r a+ c""""3 Service Water System shall be sa ys+<-s m,J e6 dresc /
OPERABLE, provided the requirements 3cne:c<d-r :" P r""*' d l'7 ""
of Specification 3.5.G are met.
( o rce Ast s tw laer s%d h'
x v e r : f, < A 4e be o f c a A BL E.
3.
If the requirements of Q_
Specification 3.8.E cannot be met, be in at least HOT SHUTDOWN within the next 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 following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
XT s-2ss Amendment No.10,32,2 30,l'3,100,1:;7, 210 3.8-6
%W DAEC-1 5.5 SEENTA ND NEW EUEkSTOR, AGE Af
( 1.
The new fuel storage facility shall be such that t effective neutron niultiplication factor (k,,)
of the fuel,' dry i ess than 0.90 and
(
flooded-is less than 0.95.
These k, val are satisfied if the maximum infinite lattice multiplica,,i actor (k.)'of the individual s
t, fuel bundles is'ssl.31.
s s
of the fuel iDt 4 pent fuel storage pool spai be les's,(than 2.
The k,,,l to 0.95.
or equa Tji k value is satisfied if,the maximum,
)
exposure-depende, of tee individual feel bundles is s 1.31 and the initial unifo average enrichment ib n4.6 wtt U-235.
3.
Spent f el shall only be stored in the pent uel pool in a vertical orientation in approved storage racks'j Bases The basis fcr the k, limit is described in Reference 1 for the GE-designed new fuel storage racks.
Compliance with this specification is demonstrated by comparing the beginning-of-life, uncontrolled k, values for the fuel type of interest to the 1.31 limit.
For GE-supplied fuel, k values can be found in Reference 2.
The k, values found in Reference 2 represent the maximum, exposure-dependent lattice reactivity and can be conservatively applied to the new fuel limit.
Calculations have been performed (Reference 3) to determine the bounding reactivity limits for bundles of GE-designed fuel, when stored in the spent fuel storagt racks of an approved design.
These analyses were perfortred 4
conservatively assuming uniform average initial enrichments in a parametric
)
evaluation for fuel with enrichments up to 4.6 wt% U-235 initially.
The bounding limit of an infinite multiplication factor of 1.31 for fuel of 4.6 wtM enrichment (or less) was evaluated at the maximum k, over burnup and includes a conservative allowance for possible differences between the rack design calculations and the fuel vendor calculations.
References 1)
General Electric Standard Application for Reactor Fuel, NEDE-240ll-P-A.*
2)
General Electric Fuel Bundle Designs, NEDE-31152-P.*
3)
Licensing Report for Spent Fuel Storage Capacity Expansion, Duane Arnold Energy Center, Holtec Report HI-92889.
- Latest NRC-approved revision.
Amendment No.
JJE. 195 5.5-1 j
RTS-z85~
07/cjg
1 3
e' INSERTy m.
p
~
5.5.1 Criticality N
5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:
'N
- a. Fuel assemblies having a maximum k. of 1.31 in the normal reactor core configuration at cold conditions and a maximum initial uniform average U-235 enrichment of 4.6 weight percent.
- b. k,, s 0.95 flooded with unborated water.
\\
5.5.1.2 De new fuel storage racks are designed and shall be maintained with:
)
- a. Fuel assemblies having a maximum k. of 1.31 in the normal reactor core configuration at cold conditions and the maximum initial uniform average U-235 enrichment is 4.6 weight percent.
3
\\
i
- b. k,y < 0.90 dry and s 0.95 flooded with unborated water.
j i
5.5.2 Capacity
/
1
'J 5.5.2.1 The spent fuel storage pool has been analyzed to allow storage of a maximum of 3152
(
f fuel assemblies in a venical orientation only.
T l
i 5.5.2.2 He new fuel storage vault is equipped with racks for storage of up to 110 fuel assemblies in a vertical orientation only.
^%
'"~
/
f I
6 l
t i
i l
i
I DAEC-1 f.,
Review of all Reportable Events, g.
Review of facility operations to detect potential safety hazards.
h.
Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Safety Committee.
1.
ReviewofthePlantSecurityPla,andi.,,le.atir.g;;b$ihb.
J.
Review of the Emergency Pla
,,,,,,1 ;._,, t i ng 7;;eiras.-
k.
Review of every unplanned release.of radioactivity to the environs for which a report to the NRC is required.
1.
Review of changes to the Offsite Dose Assessment Manual and changes'to the Process Control Program.
Review of the Fire Protection Program and implementing procedures.
m.
6.5.1.7 Authority The Operations Committee shall:
Recommend to the Plant Superintendent-Nuclear written approval or a.
disapproval of items considered under Specification 6.5.1.6 (a) through (d) above.
RTS-285~
Amendment No.198.198 6.5-3 O7/95
-.. ; ~ ',1^
DAEC-1 6.8 PLANT OPERATING PROCEDURES h
6.8.1 Written procedures involving nuclear safety, including applicable check-off lists and instructions, covering areas listed below shall be prepared, and approved as specified in Subsection 6.8.2.
All procedures shall be implemented and maintained.
l 1
1.
Normal startup, operation, and shutdown of systems and components l
of the facility.
l 2.
Refueling operation.
j l
3.
Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary system leaks, and abnormal reactivity changes.
l 4
Emergency and off-normal condition procedures.
5.
Preventive and corrective maintenance operations which could have an effect on the nuclear safety of the facility.
6.
Surveillance and testing requirements of equipment that could have an effect on the nuclear safety of the facility.
f 7.
recedures -equired by the-E cr;cncy al:n. Devle;fea/
6.8-1 Amendment No. 109 RTS - 2 85 07/95
a w
i DAEC-1 i
d.
e;ed.ur:-7;Uir;d 57 th: Ph:t h:OrityPhr.-dei *Y#d 9.
Operatiori of r aste s 10.
Fire Protection Program implementation.
11.
A prever.tive maintenance and periodic visual examination program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient to as low as practical levels. This program shall also include provisions for performance of periodic systems leak tests of each system once per OPERATING CYCLE.
12.
Program to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions, including training of personnel, procedures for
-i monitoring and provisions for maintenance of sampling and analysis equipment.
13.
Administrative procedures for shift overtime for Operations personnel to be consistent with the Commission's June 15, 1982 policy statement.
14.
OFFSITE DOSE ASSESSMENT MANUAL.
15.
PROCESS CONTROL PROGRAN.
4 16.
Quality Control Program for effluents.
6.8.2 Procedures described in 6.8.1 above, and changes thereto, shall be reviewed by the Operations Committee as indicated in Specification 6.5.1.6 and approved by the Plant Superintendent-Nuclear or designee prior to implementation, except 'as 'provided in 6.8.3 below.
6.8.3 Temporary minor changes to procedures described in 6.8.1 above which do not change the. intent of the original procedure may be made with the concurrence of two members of the plant management staff, at least one of whom shall hold a senior operator license.
Such changes shall be documented and promptly reviewed by the Operations Committee and by the Plant Superintendent-Nuclear or designee. Subsequent incorporation, if necessary, as a permanent change, shall be in accord with 6.8.2 above.
6.8-2 n
nt No.199,126,128,153,157, g_ g g 5-p7/pg
-.. =
a-RTS-285 to NG-95-2067 Page 1 of 2 SAFETY ASSESSMENT INTRODUCTION By letter dated July 21,1995, IES Utilities has proposed revisions to the Duane Arnold Energy Center (DAEC) Technical Specifications (TS) to provide administrative improvements. These changes include reformatting Section 5.5 on Spent and New Fuel Storage and Sections 6.5 and 6.8 to remove the requirement for Operations Committee review of procedures in support of the Emergency and Gecurity Plans. The current Surveillance Requirement 4.8.E.2 inappropriately requires demonstration of Emergency Diesel Generator (EDG) OPERABILITY when one Emergency Service Water (ESW) pump or loop is inoperable. The current Surveillance Requirement,4.5.G.1, requires demonstration of EDG OPERABILITY within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after having found the other inoperable. The proposed revision would allow credit for a previous demonstration of EDG OPERABILITY ifit occurred within the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ASSESSMENT The proposed revisions will provide administrative enhancements to the TS and the process for certain procedure revisions. No changes will be made to the existing limits on spent or new fuel storage. The previously approved analytical limits on fuel enrichment, design and quantity of fuel assembly storage will be incorporated. The proposed revisions are consistent with the Improved Standard TS, NUREG 1433.
Elimination of the requirement to review certain procedures will allow the Operations Committee to concentrate on other issues more pertinent to its ftmetion. The procedures implementing the Security and Emergency Plans will still be maintained and any changes will be reviewed by appropriate members ofIES staff. This revision is consistent with the guidance provided in NRC GL 93-07.
The changes to Surveillance Requirements correct an inappropriate conditional surveillance and improve another. The revised requirements will still serve to assure OPERABILITY of the affected systems. The current conditional surveillance for ESW requires demonstration of EDG OPERABILITY. The purpose of any conditional surveillance is to prove that whatever condition or event degraded one division of equipment is not common to the other. The link between the ESW conditional surveillance and the EDG conditional surveillance is erroneous. A condition which makes one division of ESW inoperable would not typically be suspected to make the opposite division EDG inoperable. Therefore this conditional Surveillance may be eliminated with no adverse impact. In both cases, for inoperable components, a review is performed of the degradation to determine the likelihood of a similar situation existing in i
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1 1
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RTS-285 NG-95-2067 Page 2 0f 2 the opposite division. The EDG conditional Surveillance requires that when one EDG becomes inoperable, the other must be tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This test is unnecessary when the OPERABLE EDG hasjust been tested within the preceding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The surveillance schedule would then begin with thejust performed test and continue every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter.
l 1
I I
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7 RTS-285 Attachmmt 4 to NG-95-2067 Page1of1 ENVIRONMENTAL CONSIDERATION 10 CFR Section 5: 22(c)(9) identifies certain licensing and regulatory actions which are eligible for categorical exclusion from the requirement to perform an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and, (3) result in an increase in individual or cumulative occupational radiation exposure. IES Utilities Inc. has reviewed this request and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR Section 51.22(b),
no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows:
i Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9) for the following reasons:
1.
As demonstrated in Attachment I to this letter, the proposed Amendment does not involve a significant hazards consideration.
2.
The proposed changes are administrative; no physical changes are made to the plant. The proposed changes do not alter any plant parameters, revise any safety limit setpoints or provide any new release pathways. Thus, there will be no change in the types or increase in the amounts of any effluents that may be released offsite.
3.
The proposed changes are administrative; no physical changes are made to the plant. The proposed changes do not alter any plant parameters, revise any safety limit setpoints or provide any new release pathways. Thus, there will be no increase in either individual or cumulative occupational radiation exposure.
.