ML20065E132
| ML20065E132 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 09/19/1990 |
| From: | Hannon J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20065E133 | List: |
| References | |
| GL-88-01, GL-88-1, TAC-69009, NUDOCS 9010010318 | |
| Download: ML20065E132 (12) | |
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NUCLEAn REGULATORY COMMISSION n,.$
-E-WASHINGTON, D. C. 20065
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i I_0WA ELECTRIC LIGHT AND POWER COMPANY CENIRAL IDWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE 5
i DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER c'
I AMENDMENT TO FACILITY OPERATING LICENSE
' Amendment No. 169' pr C"#
License No. DPR-49 L
1.
The Nuclear Regulatory Commission (the Commission) has found that:
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-A.
' The application for amendment by Iowa Electric Light and Power
. Company, et al., date-luly 27, 1988,-as rer W June 29.-1990, j'
complies with the st ds and requirements vi the Atomic. Energy-Act of 1954, as ame'
.the Act), and the Commission's rules and
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regulations set for.
.a 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 Comission; C.
There is reasonable assurance (i; that the activities. authorized by this amendment can be conducted without endangering the health.
- i and safety of the public, and (ii) that such activities will be
- onducted in compli2nce with the Comission's regulations; p
D.
The issuance of this amendment will not be' inimical to the common defense and security or to the hiialth and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part c.
51 of the Connission's regulations and all applicable requirements have.been satisfied.
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~2.
'Accordingly the license is amended by changes to the Technical Specifi-cationsasIndicatedintheattachmenttothislicenseamendmentand paragraph 2.C.(2)ofFacilityOperatingLicenseNo.DPR-49:ishereby amended to read as follows:
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o (2) ~ _ Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.169. are hereby incorporated-in the license. The licensee shall operate the facility in_
accordance with the Technical Specifications.
.3.-
The license. amendment is effective as of the date of issuance and shall be implemented within 30 days of the date of. issuance.
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FOR THE NUCLEAR REGULATORY COMMISSION i
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-i John N. Hannon, Directo-Project Directorate III.3 Division of Reactor Projects - III,
.IV, V and Special Projects -
Office of Nuclear Reactor. Regulation Attachment-Changes to the Technii:a1 Specifications 1
Date of issuance: September 19, 1990
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ATTACHMENT TO LICENSE AMENDMENT NO.169:
FACILITY. 0PERATING LICENSE NO. DPR-49
-DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with
. the_ enclosed pages. The revised areas are indicated by marginal lines.
4 Pages l
l 3.2-20 3.2-45 3.6-5 3.6 \\.
L 3.6-9 3.6-25 3.6-26 L
3.6-36 3.6-37 L
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.y TABLE 3.2-E 65 j-INSTRUMENTATION THAT MONITORS DRYWELL LEAK DETECTION 2e 2
Minimum No.
No. of' Instrument-P of Operable
- Channels Provided instrument Instrument by Design
. Action Channels 1
Sep System (1) 6
-(3) 1 Air Sampling System (2) 6
-(3)
NOTES FOR TASLE 3.2-E (1) The. Sump Systen is comprised of the Equipment Drain Sump and Floor nrain Sump Sub-systems.
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The Equipment Drain Sep Sub-system consists of one Equipment Drain Sump Flow Integrator and g
two Equipment Drain Sump Flow Timers. The Floor Drain Sump Sub-system likewise eansists of-one Floor Drain Sump Flow Integrator and two Floor Drain Sump Flow Timers. The Sump Sub-system is: operable when 'any one of these six devices operable.
(2) The Air Sampling System provides a backup system to the Sump System.
Action for Table 3.2-E (3) See Specification 3.6.C.
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timer is set to annunciate before the values specified in Specification 3.6.C are exceeded. An: air sampling system is also provided, as a backup to the' sump system, to detect leakage inside the primary containment.
For each parameter monitored, as listed in Table 3.2 F. there are two (2) channels-of instrumentation. By comparing readings between the two (2).
channels, a near continuous surveillance of instrument perfcrmance is available. Any deviation in readings will initiate an early recalibra-e tion, thereby maintaining the quality of the instrument readings.
On July 26,1984 the NRC published their final rule on Anticipated Transients Without Scram ( ATWS), (10 CFR 5 50.62). This rule requires all BWR's to make certain plant moat'icutons to mitigate the consequences.
of the unlikely occurrence of a failJre-to scram during an anticipated operational transient.
The bases for these modifications are described in NEDE-31096-P-A, " Anticipated Transients Without Scram; Response to NRC ATWS Rule,10 CFR 50.62," December 125. The Standby Liquid Control
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System (SLCS) was modified for two-pump operation ta provide the minimum required flowrate and boron. concentrati :1 required by the ATWS rule (see section 3.4 Bases). The existing ATWS Recirculation Pump Trip (RPT) was modified from a one-out-of-two-once logic to trip each recirc pump to a two-out-of-two-once logic to trip both recirc. pumps ("Monticello
design). This logic will also initiate the Alternate Rod Insertion (ARI) system, which actuates' solenoid valves that bleed the air off the scram air header, causing ~ the control rods to insert. The instrument setpoints are chosen such-that the normal reactor protection system (RPS) scram setpoints for reactor high pressure or low water level will be exceeded before.the ATWS RPT/ARI setpoints are reached. Because ATWS is considered a very low probability event and is. outside the normal design basis for the DAEC, the surveillance frequencies and.LC0 requirements are less v 'ngent than. for safety-related instrumentation.
, A End-of-Cycle (E0C) recirculation pump = trip was added to the plant to j
improve the operating margin to fuel thermal limits, in particular Minimum Critical'PowerRatio(MCPR).
The E0C-RPT trips the recirc. pumps to lessen the severity of the power increases caused by either a closure of turbine 3.2-45 Amendment No. A41,169
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'3 DAEC-1 1.IMITING CONDITION FOR OPERATION' SURVEILLANCE REQUIREMENT:
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Coolant Leakaae-C.
Coolant Leakaoe J
1.
Any time irradiated fuel is in the-1.
Reactor coolant system leaka9e reactor vessel and reactor coolant temperature is at,ove 212*F, sha~1 be checked by the sump reactor coolant leakage into the system and recorded at least once
,4 primary' containment shall be every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, limited to:
2.
The air sampling system shall be a.
5 gpm unidentified leakage.
checked and recorded at least once b.
2 gpm increase in unidentified every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
leakage within a 24 hr. period.
c.
25 gpm total leakage.
2.
The sump system shall be. operable any time irradiated fuel is in the
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vessel and reactor coolant temperature is above 212*F.
From l
and after the date that the sump a
system is made or found to be inoperable for any reason, continued reactor operation is l
permissible during the succeeding 1
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the system is made I
operable sooner, provided the air-l i
ampling system is operable.
3.
If the conditions in 1 or 2 cannot 1
be met, an orderly shutdown shall l'
be initiated and the reactor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I D.
Safety and Relief Valves D.
Safety and Relief Valves 1.
At least one safety valve and 3 1 '
1.
During' reactor power operating relief valves shall be checked or-conditions and prior to reactor replaced with bench checked valves 4
i startup-from a Cold Condition, or once per operating cycle. All l
whenever reactor coolant pressure valves will be tested every two is greater than atmospheric and cycles.
l temperature greater than 212*F, The setpoint of the safety valves-both safety. valves and the safety shall be as specified-in j
modes of all relief valves shall Specification 2.2.
l be operable,.except as specified in 3.6.D.2.
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Amendment No. JO,169 3.6-5
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' 4.l DAEC-15 LIMITING CONDITION FOR OPERATION-SURVEILLANCE REOUIREMENT.
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Following 1-pump uperation, s
the discharge.';alye of the lower speed pump may not be opened unless the speed of thet f aster ~ pump is: less than 50%
of its rated speed.
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. Structural Integrity G.
Structural Integrity The structural integrity of 1.
In-service' inspection of ASM'E.
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the pressure boundaries shall Code Class 1, Class 11. and be maintained at the level Class Ill components shall be required by the original performed in accordance with 4
acceptance standard throughout Section XI of the'ASME Boiler-i the life of the plant.
and Pressure Vessel Code and i
applicable Addenda as-required by 10 CFR 50, Section-50.55a(g), - except where.
Specif.ic written relief has been granted by the NRC pursuant to-10 CFR 50, Section-50.55a(g)(6)(i).
a.
The; second 10 year interval for the inservice inspection program cescribed above comenced on November 1, '1985.
2.
In-service testing of ASME Code Class I, Class II.and Class 111 pumps and valves shall be performed in accordance.with :
Section XI of the ASME Boiler and, Pressure Vessel Code and applicable Addenda-as required by 10 CFR 50, Section 50.55a(g), except where t.
specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section o
l 50.55a(g)(6)(1),
a.
The second 10-year interval for-
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the. inservice testing program-described above commenced or.
p February 1, 1985, a
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DAEC-1 s
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' LIMITING CONDITION FOR OPERATION
' SURVEILLANCE REQUIREMENT -
'3.
The' inservice inspection program for pipi.ng;' identified!
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in NRC Generic Letter'88 ;
shall: be performed.in' l
accordance with ' the-staff.
l positions!on' schedule, methods m
and personnel, and sample:
expansion included in this
- t generic letter.
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AmendmentL No. 32,169 3.6-9 c
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establishment of allowable' unidentified leakage greater than that given.it
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f 3.6.C:on the basis ofa the data presently available would' be premature w
because of uncertainties associated with the data. For leakagetof the order:
.g of 5 gpm,- as specified in 3.0.C. the experimental and analyticalJdata
_ suggest a'. reasonable mergin of safety that such' leakage' magnitude would: not realt from a crack approaching the critical size ior rapid Dropagation.
Leakage less than the magnitude specified can be detected reasonably in ~ a matter of a:few hours utilizing the availab'e leakage detection schemes, and:
.i if the origin cannat - be. determined in a _ reasonably short ' time - the! plant should be shut-down to allow further investigation and orrective-action.
Identified and unidentified leakage are defined,in the DAEC_ Updated FSAR, Section 5.2.5.2.2.
Total leakage is defined as the sum of identified and unidentified leakage.
L ihe capacity of the drywell floor sump ptnps is 50 gpm and the capacity of:
the drywell: equipment sump pumps is also 50 gpm.
Removal of 25 gpm from either of these sun 9s can be accomplished with margin.
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DAEC surveillance procedures require both ' identified and unidentified j
l-leakage to be determined at approximately 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> intervals..Should leakage-L
_ exceed the allowed limits,. control room alarms actuated by the equipment
-1 drain sump and floor drain sump pump timers are provided.to indicate this condition, thus, continuous leakage detection capability is provided by design.
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The requirement that an increase in unidentified leakage shall not exceed L
small leaks in a reasonat.y short time such that corrective action can be 2 gpm in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is based on maintaining the ability to detect I
initiated.
However, during reactor startup and ascension to normal operating pressure, leakage should be closely monitored until normal operating pressure is achieved and a " baseline" leakage rate can be established to which any leakage 1,1 crease can be compared.
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4 Amendment No. - /J,169 3.6-25 L
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.The primary containment ~ atmosphere radioactivity detector provides a sen'sitive and rapid indication of increased nuclear system leakage. The primary containment environmer:
quously sampled from one of three f
' locations which are chosen to both a representative gas mixture and i
an indication of the location of the leakage.
i The sample air undergoes three separate processes in which the radioactive noble gas, halogen, and particulate contents are determined. This system is thus a-three channel monitoring system.
The processed air is returned. to the drywell.
The primary containment atmosphere radioactivity detector serves as a k
sensitive, reliable backup to the other metnods of leak detection.- It is-l anticipated that the particulate detector will be the primary indication of leak 3ge, with the halogen and noble gas detectors serving as inoication of -
the primary containment environment if primary containment venting is required.
These detectors in conjunction with an isotopic analysis can b6 used to indicate whether the detected leak is from a steam or water system.
This system is not capable of accurately quantifying coolant leakage rates.
. Because the Air Sampling system is not capable of determining leak. rate, 'it is considered a backup system to the sump system, and no LCO is associated:
1 with it.
It is intended to be a compensatory measur'e used when the sump system is inoperable,
. Amendment No. /J,169 3.6-26 ax
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DAEC-1 The first 10-year. interval for inservice inspections' in accordance with the n
ASME Boiler and Pressure Vess'el Code,Section XI commenced on February 1, e
1975.
This interval was extended for 9 months because of a 9 month ~ cutage for; replacement of. recirculation system inlet nozzle safe-ends in 1978-79.
Therefore, the. first 10-year interval ended on October 31,.1985.
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J The-second '10-year interval for inservice inspections conrnenced on' November 1,1985 and is. scheduled to end on October. 31, 1995.
The second 10-year inspection program addresses the requirements of the ASME Code,Section XI, 1980 Edition with Addenda through Winter 1981,- subject to the limitations.
and modifications as stated in 10 CFR 50.55a.
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' Visual inspections for leaks will be made periodically on critical systems.
The inspection ' program specified encompasses the major areas of the vessel and piping systems within'the drywell.
The inspection period is based'on
- the observed rate of growth of defects from fatigue studies sponsered by'the :
NRCuand is delineated by Section XI of the ASME -Code. These studies - sho'w that it. requires. thousands of stress cycles at stresses teyond those u
expected to occur in a ' reactor system to.prepagate a crack'.
The test -
frequency established is at intervals such that in comparison to study L
L results, only a small' number of stress cycles, at values below limits will occur. On this basis, it is considered that the test frequencies are l
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. adequate.
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Th'e type of inspection planned for each component depends on location, L
accessibility, and type of expected defect.
Direct visual examination 's
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Amendment No. 7J,169 3.6-36 w
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proposed >wherever possible,since it is ~ f ast and reliable. Surface
. inspections are: planned 'where practical, and where added. sensitivity is required.- ' Ultrasonic testing or radiography shall be-usedlwhere ' defects can j
occur in concealed surf aces. Section 5.2.4 of the Updated FSAR provides J
dethils of the inservice _ inspection program.
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Starting with the Cycle 9D0 Refueling Outage, an ' augmented inspection program was implemented to address concerns relating to Intergranular Stress Corrosion Cracking (IGSCC) in reactor coolant pipir.g made of. austenitic stainless steel. _ The augmented inspection program conforms to.the NRC A
staff's positions set forth in Generic Letter 88-01 and NUREG-0313, Revision-2 for inspection schedule, inspection methods and personnel, and inspection k
sample expansion..
The first 10-year interval for inservice _ testing of-pumps -and valves inj accordance with the ASME Code,Section XI commenced on _ February 1,.1975 and ended on January 31, 1985.. The second 10-year inservice testing interval conenenced on February 1,1985 and is scheduled to end on January 31,1995.
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The second'10-year testing program addresses the requirements of the ASME Code,Section XI,1980 Edition with Addenda through Winter 1981, subject to theElimitations and modifications of 10 CFR 50.55a. Section 3.9.6 of the Updated FSAR describes the inservice testing program.
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'3.6. H & = 4.6.H BASES:
Shock Suppressors (Snubbers)
Snubbers are designedc to prevent unrestrained pipe motion under ' dynamic.
ib, loads as might occur-during an earthquake or other severe transient, while accommodating normal themal me+1on during system startup and shutdown. The j
consequence of an inoperable snut,ber is an increase in the probability of
' damage to piping as a result of a seismic or other event _ initiating dynamic' g
'_ loads or, in the case of a frozen snubber, exceedibg allowable stress limits n
M ouring-system thermal transients.
It is therefore required that all snubbers reqJired to protect the primary coolant system' or any other safety 4
system or component be operable during reactor operation.
Amendment No. JM,l'69 3.6-37 1
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