ML20073J902
| ML20073J902 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 09/30/1994 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20073J900 | List: |
| References | |
| NUDOCS 9410070183 | |
| Download: ML20073J902 (28) | |
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SURVEILLANCE REQUIREMENTS (Continued) 1)
All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
Tubes in those areas where experience has indicated potential 2) problems, and A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall 3) be performed on each selected tube.
If any selected tube does t
not,'ermit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall i
be selected and subjected to a tube inspection.
In-eddi44on-to-the--3%-sampleaMJ* tubes-wi44-be inspec-tede
-c.
l The tubes selected as the second and third samples (if required by C. 'd s Table 4.4-2) during each inservice inspection may be subjected to a s
l partial tube inspection provided:
The tubes selected for these samples include the tubes from 1) i those areas of the tube sheet array where tubes with imperfections were prev;ously found, and
(
2)
The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
t Inspection Results Category Less than 5% of the total tubes inspected are C-1 degraded tubes and none of the inspected tubes are defective.
One or more tubes, but not more than 1% of the C-2 total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
More than 10% of the total tubes inspected are C-3 degraded tubes or more than 1% of the inspected tubes are defective.
In all inspections, previously degraded tubes must exhibit Note:
significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
AmendmenE No. 59 (Unit 1)
McGUIRE - UNITS 1 and 2 3/4 4-12 Amendment No. co(Unit 2) 2-5
st >[
e 6
d [Eo Y O
-REACTOR COOLANT SYSTEM
$<N L
E t
Qt C)ep 9-(>.
3(b SURVEILLANCE REQUIREMENTS (Continued)
/
N 4.4.S.3 steam generator tubes shall be performed at the following i
The first inservice inspection 'shall be performed afteri6 Effective a.
Full Power Months but within 24 calendar months of initial Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections following service under AVT conditions, not including the preservice inspection, all inspection results falling into the C-1 category or if two result in consecutive inspections demonstrate that previously observed degra-dation has not continued and no additional degradation has occurred the inspection interval may be extended to a maximum of once per 40 months; b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fa in Category C-3, the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency
(
i shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a; the interval may then be extended to a maximum of once per 40 months; and r
Additional, unscheduled inservice inspections shall be performed on c.
l each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
i
1)
Reactor-to-secondary tubes leaks (not including leaks originating 1
{
from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,
(
2)
A seismic occurrence greater than the Operating Basis Earthquake, 3)
A loss-of coolant accident requiring actuation of the Engineered t
Safety Features, and 4)
A main steam line or feedwater line br,eak.
i I
l l
t 2-6 McGUIRE - UNITS 1 and 2 3/4 4-13
[
(
SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this specification:
1)
Imperfection means an_ ext > tion to the dimensions, finish or contour of a tubehr--sleeve-- rom that required by f abrication j
drawings or specit1 cat 16'ns.
Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if l
detectable, may be considered as imperfections; 2)
Degradation !wans a service-induced cracking, wastage, wear or genera _cnnca Q n occurring on either inside or outside of a i
tube -op-s4eeve-; )
3 3)
Degraded Tube means a tub c--s4eevehontaining i perfect.jp s 3
greater than or equal to 20%'of7he nominal tub sl eeve-11 thickness caused by degradation; 4)
% degradation means the percentage of the tube {r-s4eeve-all l
thickness affected or removed by degradation; 5)
Defect 4n ns an imperfection of._su severity that it exceeds
(
th((Ceepah-himi t.
A tubeksiteD ontaining a defect' is defe'ct1 vet > - Q LQ Q
\\
6) imit means the imperfection depth at or beyond which thT t0be 7 r :lce;/^= hall be removed from service by pluggin -ee -
. ' epa #ed--by-g-
nd is equal to 40% of the nominal tube siceve WaT I thickness. -Th4re-definition does not apply to tha-i
-aeea--of-the-tubesheet--region-below-the F* di-stance-provided-the.
- tube i s not degraded (i. c., ac indicat-ions-of cracking-)-wi-tMe-44u f
- d+ stance.
If a tube i: cl ceved-due--40-degradation-in-
-the F* distance, then -any-defects in-the-tube-below-the--sleevel
-w-i4A-remain-in--serv 4ce ei thout--repair.
--T he-B a bcock-&-W44eox-p ro cesfr-(-o e-meth od-)-e quivale ntr-to-thea metho d-dese nibed-4n--To p4ea4-Repo r-t--BAW4045(P-)= A-w444--be-used-a 7)
Unserviceable describes the condition of a tube er 0100'/^ if it l
leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c, above; 8)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and 2-7 McGUIRE - UNITS I and 2 3/4 4-14 Amendment No.107 (Unit 1)
Amendment No. 89 (Unit 2)
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 9)
Preservice Inspection means an inspection of the full length of each tube in each steam generator ;,erformed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection shall be performedTefter the N Mad hydrostatic tect and prier to-initial POWER OPERATION X6 using the equipment and techniques expected to be used during 0 subsequent inservice inspections.
% )-FA-D is t-ance-4s-the-d i sta n ee-in to-the-t u b e s heet-f-nom-t he-top-face-of-t he-t ub es hee t-or-the-top-o f-the-les t-ha rd rolh-whi chever-i+
iowe r-tfu rth e r-i n te-t he-ttbesh eet-)-that-has-b e en-cons e rva t-i ve4y-chosen-to-be 2-ineh m (T
-lb)-P-TUBE--is-a-tube-wi-th-degradat-ion-equal-to-or-greater-than-4GL, <
-below-the-F-d4+tance-and-not-degraded--Gi,4._,--no indications _of-OD
~
-eracking3-in-the-A-distence,-
DD d
b.
The steam generator shall be determined OPERABLE af ter completing th corresponding actions (plug-or-eepa%11 tubes exceeding the + epa 4+p l
limit and all tubes containing through wall cracks) required by Table 4.4-2.
k 4.4.5.5 Reports Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.
This Special Report shall include:
3)
Number and extent of tubes inspected, 2)
Location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes pl ugged.oe-eepa4ted--
l
---Th e-res u l ts-o (-i n s p e c t i on s-o f-F *-t ub e s-s hall-be-re p o r ted-to-th e.
-c.
-Comndss4nn in-report -prior-to-the-res-tart-of-the-unit-following-7
-the--i nspeui o n:-T h i s-repo rt-sha ll-i nclude+--
4-)--n;en t+f i ca L ion-o f4 *-tuber,v_and2
--;t).
L-oca t-i on a nd +iv e -o f-t,ho-deg rada ti on:-
2-8 McGUIRE - UNils 1 and ?
3/4 4-15 Amendment No. 107 (Unit 1)
Amendment No.
89 (Unit 2)
M e
TABLE 4.4-1 2
m a
c5 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION
$a N
Prearvice Inspection No Yes t
No. of Steam Generators per Unit Two Three Four Two Three Four First inservice Inspection oMer fit.e, Stea m Gene e-/or ffplaccrne #
All One Two Two
{$
Second & Subsequent inservice inspections One One One One l
l 2
3 a
n Table Notation:
)
o 1.
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operatin onditions in one er more steam generators may be found to be more severe than those in other steam eneg nder suc. ci tum-i stances the sample sequence shall be modified to inspect the most severe condi gg 7 g,,,,,, r-)
n
- 2. The other steam generator not inspected during the first inservice inspection *sha bbeJashdhy ent inspections should follow the instructions described in I above.
3.
Each of the other two stearn generators not inspected during the first inservice inspectionsThall be inspected during the second and third inspections. The fourth and subscquent inspections shall follow the instructions described in 1 above.
7
REAC_ TOR C00LAi4T SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERAT10N
- 3. 4. ft. 2 Reactor Coolant Sy5 tem 1eakage shall be 1imited to:
No PRESSURE Bout 1DARY LEAKAGE, a.
b.
1 gpm Ul1 IDENTIFIED LEAKAGE, 0.27 c.
gp gpm to'.nl primary-to-secondary leakage through all steam generators and gallons per day through any one steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of e.
2235 20 psig, and f.
1 gpm leakage at a Reactor Coolant System pressure of 2235 20 psig from any Reactor Coola' t System Pressure Isolation Valve specified n
in Table 3.4-1.
APPLICABILITY:
MODES 1, 2, 3', and 4.
ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hou:rs.
b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage f rom Reactor Coolant System Pressure Isolation. Valves, redu,ce the le.akage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the followi ng 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With any Reactor Coolant System Pressure Isolation Valve leakaoe c.
greater than the above limit, isolate the high pressure portior. of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY Within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
McGUIRE - UNIlS 1 and ?
3/4 4-19 2-10 4
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REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.
Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
(
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. -The BD! procccc (or method) equivalent to the 4nspect4en-metted-desc+ibed-+n-Topiea+-Repor-t S!"<!-20A 5(P) ^ wi be uced.
4aservicc iaspest4on-of-steam-gene-rator-sleevc i: c1:oweguined-t+-ensure-RGS
-int +g r4ty,--Beeau se-the-s4eeves-int-roduce-eha ng c : ia the-wa44-thtekness and--
dame-ter r-t-hey-eeduce the sendt-ivity-of--eddy-euerent-tes-t4-ng-therefore2,
-spec 4a4--inspest4on-methods-must-be-used.
A method b deser4 bed in Topical Repert BAW=2045(-P-)=A-with-suppor44ng-validat-ion-date-that-demonstrates-the-
-i-nspectab-i1ity ef the-sleeveand-under-ly4ng-tube.
^c required-by-NRG-f+r-
-Licensees-author 4 red-te uce this-repair-processdeGui-ee-commits-to validate-4he-adequacy ef anyy-system-that-+s-used-for-pee-iod4e-inservicc iaspecticn3 o f-tho e i new, and will e, valuate-and,-as deemed-appropniate-by--Duke 4ower
-Gompa ny ;--i mplement--testeg-methods-arr-bet 4e r-methodsa re-developed-a nd va44 dated-fee-eommere4C ucc.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Second_ary Coolant System _ l37 t
a (reactor-to-secondary leakage = -50dN c!1ons per day per steam generator).
3
\\
Cracks having a reattor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstan N oads imposed during normal operation and by postulated accide s.
Operathg-p" ants have demonstrated that reactor-to-secondary leakage of 0 gallons per day per steam generator can readily be detectedAf--r-adiation moniterc o f " t e am-
-fie m kw-4 h 4w m-Leakage in excess of this limit will require plant 2 12 Amendment No.107 (Unit 1)
McGUIRE - UNITS 1 and 2 B 3/4 4-3 Amendment No. 89 (Unit 2)
BASES 3/4.4.S STEAM GENERATORS (Continued)
OD shutdown and an unscheduled inspection, during which the leaking tubes will be C
located and plugged.
0%
g Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service, it Q s
L will be found during scheduled inservice steam generator tube examinations. fN
'IRepa4t will be required for all tubes with imperfections exceeding the-r-epair-limit of 40% of the tube nominal wall thickness. -Instal 4ed-sleeves-with-impee-te ction s-e xc e ed i ng-40%-okthe-sJ e e v e-noni naLwaLL-thickne s s-wi1L_be-plugged.
-Defec-tive-ste am-generato r-tub e s-ca n-be-r-e pai r-ed-by-the-i n sta Lla ti-on-of--sleeve s.
-whieh-span-the-area-of-degradat-ionWod-serve an a-replacement-pressure-boundary
-4eMhe-degraded-por-t4on-of the-tubeallowing tho t ube.to_ccmain in corvica_
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20%
of the original tube wall thickness.
Sor-t ubes-with-deg ra dation-b elow-the-A rH r t anca -and-not-degr-aded-wi-thin-the-E*-- di s t a nce, repain-_ts -not r
enquired-Lf_
-a-tube-4s rAeeved-due-te-degradation-in the c* distance A en-any-defects-in-
-the-ttbe-below-the-sl eev e-w i44-rema in-i n-se r+ ice-witho ut-r+pa4r.
8 i
i s
L i
Amendment No.107(Unit 1)
McGUIRE - UNITS 1 and 2 B 3/4 243 Amendment No.
89(Unit 2) l
REAER7R COOLANT SYSTEM BASES
.7~_
STEAM GENERATORS (Continued)
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to 10 CFR Sections 50. 72 and 50.73 prior to resumption of plant operation.
j Such cases will be considered by the Commission on a case-by case basis and may result in a requirement for analysis, laboratory exami-nations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm.
low to ensure early detection of additional leakage.This threshold value is sufficiently The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig..
This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.
a d & c. n S y d w w go,ti,4 & p u ge e h The ca7
.otal steam generator tube leakage limit of g gpm for all steam generators a the tube leakage will be limited to gnsure g at the dosage contribution from traction of 10 CFR Part 100 dose guideline valuerp edr=
the analysis of these accide its.The g gpm amanhan consistent with the assumptions use in The @ gpd leakage limit per steam generator ensures thi t stear generator tub? integrity is maintained in event of a main sten i line upture or under LOCA conditions.
as o.a.7 d rm yd UA %
f~ c> c "\\\\ $.5 RR Che<pieC I5 McGUIRE - UN1'IS I and 2
- mweeh B 3/4 4-4 Amen. _
Amendment 2-14 l
i
REACTOR COOLANT SYSTEM BASLS N
OPERATIONAL LEAKAGF (Continued)
PRL55URE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System ieakage or failure due to stress corrosion.
Main-taining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.
The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within~the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the. Steady State Limits.
The Surveillance Requirements provide adequate assurance thaf concentrations in excess of the limits will be detected in sufficient time to take corrective ACTION.
3/4.4.6 SPECIFIC ACTIVITY O.27 The limitations on the specific activity of the reactor coolant ensure that.the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will non exceed an appropriately small fraction of Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to secondary steam generator leakage rate of jjpegpm.
The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the McGuire site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-3, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
Amendment No. 66(Unit 1)
McGUlRE - UNITS 1 and 7 B 3/4 4-5 Amendment No. 47 (Uni t 2) 2-15
[
DESIGN FEATURES 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE
\\
S.4.1 The Reactor Coolant System is designed and shall be maintained:
i In accordance with the Code requirements specified in Section 5.2 of a.
the FSAR, with allowance for normal degradation pursuant to the I
applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and For a temperature of 650 F, except for the pressurizer which is c.
680 F.
l y LUME p
5.4.2 The total water and steam volume of the Reactor Coolant System is lh 040-+ 100 cubic feet at a nominal T f 525 F.
avg
- 5. 5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
(
5.6 FUEL STORAGE CRITICALITY
}
- 5. 6.1 The new and spent fuel storage racks are designed and shall be maintained with:
i l
a.
Ak equivalent to less than or equal to 0.95 when flooded with unb8f$ted water, which includes a conservative allowance for i
uncertainties as describe <1 in Section 9.1.2.3.1 of the FSAR, and b.
A nominal 21-inch center-to-center distance between fuel assemblies placed in the new fuel storage vault racks, hnd A nominal 10.4-inch and 9.125-inch center-to center distance between c.
fuel assemblies placed in Region 1 and Region 2 storage racks, respectively, in the spent fuel storage pool.
DRAINAGE
- 5. 6. 2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 745 ft. 7 in.
i CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a i
storage capacity limited to no more than 1463 fuel assemblies (286 spaces in Region 1 and 1177 spaces in Region 2) having an initial enrichment less than or equal to 4.0 weight percent U-235.
McGUIRE - UNITS 1 and 2 N
Amendment No.16 (Unit 2)
Amendment No. 35 (Unit 1) t
ADMJf{LSTRATIVE CONiR0p _ _ _ _ _ _ _ _ _ __ _ _ _ _ _
[0Rf_0PFRAT!!JGljMIM M ORI 2.
WCAP-10216-P-A, "RELAXA110ft Of CONSTANT AXIAL OFFSET CONTROL fQ SURVEllt.ANCE lECilNICAL SPEClfICA110N," June 1983 (W Proprietary).
(Methodology for Specifications 3.2.1 - Axial Flux Difference ' Relaxed Axial Of fset Control) and 3.2.2 - Heat llux Hot Channel Factor (W(Z) surveillance requirements for FQ Methodology.)
3.
WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUA110N MODEL USING BASH CODE," March 1987 ()4 Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
BAW-10168P, Rev.
4.
1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," September 1989 (B&W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel 'f actor.)
5.
DPC-NE-20llP, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
(
6.
DPC-NE-300lP, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," March 1991 (DPC Proprietary).
t (Methodology for Specification 3.1.1.3 - Moderator Temperature Coeffi-cient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
7.
DPC-NE-2010P, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," April 1984 (DPC Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature
~
Coefficient.)
Re\\lk i
8.
DPC-NE-3002A "FSAR Chapter 15 System Transient Analysis Methodology,"
August 1991.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
~
~
licGUIRE - UNITS I and 2 6-2 g Amendment No.
143 (Unit 1)
Amendment No.
125 (Unit 2)
(.
ADMINISTRATIVE CONTR01.5 COREOPERATit1G1IMITSJpFJt__
Rev \\ 2$"JJ ermal-llydraulic Transient Analysis Methodolog 9.
DPC-NE-3000, May 1989.
(Modeling used in the system thermal-hydraulic analyses) 10.
DPC-f1E-1004A, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P,"
November 1992.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident-analysis limits) of the safety analysis are met.
The CORE OPERATIl1G LIMITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
SPECIAL REPORTS
(
6.9.2 Special reports shall be submitted-to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
i l
~
McGUIRE - UNITS 1 and 2 gb Amendment flo.
p[g (Unit 1)
Amendment No.
16 (Unit 2) l
l TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TP1!P $ETP01ET_5 9
s E
[MiCTIONAL uni.
_ TRIP SETPOINT ALLOWABLE VALUE l
E
- 12. Steam Generator Water Level Low Low glo.fc7p noncta can - c.
2 9 7, g nar reto rc(n c spGn 2 o of peu
. sp e S l of.,,an -
~
a.
Unit 1 0% to 30%
0% to 30%
incre'*.- linearly increas inearly to 0.0% of span to ^
% of span
' om 30% to 100% RTP*
m 30% to 100% RTP*
236.8% of narrow 235.1% of narrow b.
Unit 2
~
range span range span E
- 13. Undervoltage - Reactor 277% of bus voltage 276% (5016 volts)
(5082 volts) with a
?
Coolant. Pumps O.7s. response time 256.4 Hz with a 255.9 Hz
- 14. Underfrequency - Reactor 0.2s response time Cooiant Pumps
- 15. Turbine Trip 2550 psig 2500 psig 2R Stop Valve EH
- 3 a.
Pressure Low EE RR b.
Turbine Stop Valve Closure 21% open 21% open g{
- 16. Safety Injection Input from ESF N.A.
N.A.
SS 22
==
?
- RTP - RATED THERMAL POWER 3
6 l
I hop _osed Revision to TechnicalSprcification Ta.hle12-1 The low-low steam generator water level reactor trip setpoint is modified from a variable serpoint that is proportional to nuclear power to a constant level setpoint of i
16.7% of narrow range span.
t TechnicalJustification l
These setpoints were chosen to maximize the plant operating region wSile still ensuring that reactor trip on low-low level would occur following a feedline break inside containment. The new low-low level setpoints are consistent with all reanalyzed licensing basis safety analyses. All of the reanalyzed transients that take credit for this trip function meet the applicable acceptance criteria.
horasedlexision to TechnicaLSpecification.Iahle13-4 a) The high-high steam generator water level setpoint is changed to 83.9% of narrow range span.
b) The low-low steam generator water level auxiliary feedwater actuation setpoint will be modified from a variable setpoint that is proportional to nuclear power to a constant level setpoint of 16.7% of narrow range span.
l Iechnicallusfificalion i
a) This setpoint was chosen to maximize the plant operating region while still ensuring that feedwater isolation on high-high level would occur prior to the actual water level in the generator reaching the flood point. The new high-high level setpoint is consistent with all reanalyzed licensing basis safety analyses. The results of the increase in feedwater flow analysis, which is the only FSAR Chapter 15 transient i
which relies on this trip function, show that all applicable acceptance criteria are met.
i b) Refer to the technical justification for the proposed revision to Technical Specification Table 2.2-1.
DoposedEerision to TechnicalSpecification 3/4.4.4 The Surveillance Requirements are changed to delete repair methods which are no longer applicable after the replacement of the steam generators. References to F*, and sleeving are deleted. Clarification is added to the surveillance requirements on 3-1 i
t
(
performing initialinspections after replacement of the steam generators and when they will be performed.
IechnicaljuMfication This proposed change to the Technical Specifications deletes repair criteria which will no longer be applicable after the replacement of the steam generators. References to F*
criteria and sleeving are removed because these methods of repair were approved specifically for use on the current steam generators. Clarification is also added to show that initial inspections will be performed after replacement of the steam generators and that the baseline inspection of the tubing will be performed after the manufacturer performs the hydrostatic test. These changes will not alter the way surveillance's are performed, and continue to meet the current intent of the requirements.
Eropmed Revision to Technical Specification 3/4.4.6.2 i
This proposed Technical Specification revision reduces the allowable primary-to-secondary system leakage to a total of 0.27 gpm through all steam generators and 135 gallons per day through any one generator.
IechnicalJustificahon The primary-to-secondary leak rate has a major impact on the results of the offsite dose calculation for the locked rotor, single uncontrolled rod withdrawal, and rod ejection events. This leakage is the mechanism for the release of the fission products. The taller tubes in the feedring steam generators potentially result in a longer period of tube bundle uncovery during the transient; this necessitated the recalculation of the offsite doses. In order to ensure that the doses do not exceed the acceptance criterion, the allowable leak rate must be reduced.
Eropmedlerision to Technical Sprcification Table 17-3 The steam line safety valve lift settings for banks 4 and 5 are changed to 1210 and 1220 psig, respectively.
Technica1 Justification A preliminary analysis showed the peak secondary system pressure for the turbine trip event to exceed the acceptance criterion of 110% of design pressure. The design basis of these safety valves is specifically to prevent such an overpressurization. Primarily due to the increased lift setting tolerance of 3% assumed in the analysis, the final two 3-2
banks of safety valves were not fully open at the point in the analysis when the peak system pressure occurred. Reducing the lift settings for these two safety banks results in acceptable turbine trip transient results.
Proplosed RevisiendolechnicalSpecification 5.4.2 The volume of the Reactor Coolant System changes from 12,040 100 cubic feet to 13050 100 cubic feet.
Iechnicallusfification The mass and energy release for postulated loss of coolant accidents inside contairunent is analyzed to ensure that the peak containment pressure limit is not exceeded. Since the Reactor Coolant System volume is greater, the total mass released into containment will be greater. In addition, during the depressurization of the RCS, the steam generators actually function as heat sources. Since the feedring steam generator full power liquid mass is greater than that of the Model D steam generators, the total energy available for removal by the RCS is increased. Both of these effects have the potential to yield more severe mass and energy release results. This event and all other reanalysis which was required for the replacement steam generators has been done assuming the new reactor coolant system volume. The results of these analyses show that the applicable acceptance criteria continue to met.
PIpposed_ Changes to TechnicalSpecification 6.9.1.9 The references to DPC-NE-3000 and DPC-NE-3002 have been updated to reflect the use of the most current approved revision to the topical reports for the replacement steam generators.
Technica1 Justification This revision, which reflects the use of the most current revision to the above topical reports, is administrative in nature.
3-3
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No Significant Hazards Analysis McGuire Nuclear Station The following analysis, required by 50.91, concludes that the proposed amendment will not involve significant hazards considerations as defined by 10 CFR 50.92.
10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)
Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)
Involve a significant reduction in a margin of safety.
Operation of McGuire Nuclear Station in accordance with the proposed changes to the Technical Specifications will not involve a significant increase in the probability or consequences of an accident previously evaluated. The low-low steam generator water level reactor trip setpoint, the high-high steam generator water level setpoint for turbine trip and feedwater isolation, and the low-low steam generator water level setpoint for auxiliary feedwater initiation are changing to support operation with the replacement steam generators. These setpoints were chosen both to optimize plant operation, and ensure that all t
applicable acceptance criteria are met for licensing basis safety analysis. These i
setpoints do not contribute to the initiation of any accident evaluated in the McGuire FSAR and have no adverse impact on system operation, therefore it can be concluded that these changes will not significantly increase the probability or consequences of an accident evaluated in the FSAR.
The reduction in the primary to secondary leakage rate for McGuire will not -
increase the probability of an accident evaluated in the FSAR. This lower limit will require corrective action more quickly than is currently required in the event that there is a steam generator tube leak. This change will not significantly affect the consequences of an accident previously evaluated. The allowable leakage is being lowered because this leakage has a major impact on the results
)
of the offsite dose calculation for the locked rotor, single uncontrolled rod i
withdrawal, and rod ejection events. The taller tube bundle in the replacement steam generators will potentially result in a longer period of tube bundle uncovery during the above transients. The revised allowable leakages of 0.27 4-1
gpm through all steam generators and 135 gallons per day through any one generator ensure that the dose analysis results are within the applicable fraction 10 CFR 100 limits.
The increase in Reactor Coolant System volume due to the replacement steam generators will not increase the probability or consequences of an accident previously evaluated. The increase in volume has no effect on the probability of occurrence of any accident evaluated in the FSAR. The mass and energy release due to postulated loss of coolant accidents inside containment has been analyzed to ensure that the peak containment pressure limit is not exceeded. All Chapter 15 reanalysis which was required due to the replacement steam generators assumed the new Reactor Ccolant System volume. Since the results of these analyses show the applicable acceptance criteria continue to be met, it can be concluded that the consequences of an accident previously evaluated are not significantly increased due to this change.
The changes in the steam line safety valve lift settings to 1210 and 1220 psig respectively ensure that the peak secondary system pressure for the limiting ANS Condition II event, turbine trip, does not exceed the acceptance criterion of 110% design pressure. The design basis of these valves is to prevent such an overpressurization. Since reducing these lift serpoints results in acceptable turbine trip transient results by ensuring that the valves perform their design basis function, it can be concluded that the probability or consequences of an accident previously evaluated is not significantly increased.
Operation of McGuire Nuclear Station in accordance with the proposed changes to the Technical Specification will not create the possibility of a new or different accident from any accident previously evaluated. The proposed changes to revise the low-low steam generator water level reactor trip setpoint, high-high steam generator water level setpoint for turbine trip and feedwater isolation, and low-low steam generator water level setpoint for auxiliary feedwater initiation ensure that the appropriate acceptance criteria for FSAR Chapter 15 transients which rely on these functions are met for operation with the replacement steam generators. The proposed change to lower primary to secondary leakage for operation with the replacement steam generators will require that corrective action be taken more quickly in the event that steam generator tube leakage is experienced during operation. As discussed in the technical justification, this will cause the dose results for transients affected by tube bundle uncovery to be within acceptable limits. The proposed change to Table 3.7-3 to reduce steam line safety valve lift settings allows the valves to perform their design basis function of ensuring that the peak secondary system pressure of 110% design is not exceeded for the turbine trip event, which is the limiting ANS Condition II event. The increase in Reactor Coolant System volume is taken into account in the analysis of the mass ard energy release due to a postulated loss of coolant inside containment and Chapter 15 events which have been reanalyzed due to 4-2
l replacement of the steam generators. As discussed above, the proposed changes will not introduce the possibility of a new or different accident from any previously evaluated; they will ensure that transients that take credit for these functions and dose analyses meet applicable acceptance criteria for operation with the replacement steam generators.
Operation of McCJre Nuclear Station in accordance with the proposed changes to the Technical Specifications will not involve a significant reduction in a j
margin of safety. The proposed changes are being made to ensure that transients l
that rely on low-low steam generator water level reactor trip setpoint, high-high steam generator water level setpoint for turbine trip and feedwater isolation, and low-low steam generator water level setpoint for auxiliary feedwater actuation meet applicable acceptance criteria. The reduction in allowable primary to secondary leak rate will ensure that transients with dose analyses which are affected by the replacement steam generators meet the current acceptable limits.
The reduction in the steam line safety valve lift settings will ensure that the design basis of these valves is met. The proposed change in the Reactor Coolant System volume will not involve a significant reduction in a margin of safety.
The increased volume affects the mass and energy release due to a postulated j
loss of coolant accident inside containment and the other Chapter 15 events which were reanalyzed due to replacement of the steam generators. These events have been analyzed and the results are within current acceptable limits.
As discussed above, the acceptance criteria for FSAR transients which are affected by these proposed changes continue to be met, therefore there is no significant reduction in the margin of safety.
Changes to the steam generator surveillance requirements will simply delete inspection requirements and repair methods which are no longer applicable after installation of the replacement steam generators. The only exception to this is Surveillance Requirement 4.4.5.4.a.9. This requirement is modified to clarify that the manufacturer will perform the hydrostatic test for the replacement steam generators. This change will not affect the probability or consequences of an accident previously evaluated, the purpose of the preservice inspection is to establish the baseline condition of the tubing. The baseline condition of the tubing in the replacement steam generators will be established prior to installation. The possibility of a new or different accident from any previously evaluated will not be created. No new accident initiation mechanisms will be introduced by this change, and the intent of the requirement, to establish the l
baseline condition of the tubing, will be met. Since the baseline condition of the tubing will be obtained for use in the monitoring of tubing degradation, as is currently required by the surveillance requirement, there will not be a significant reduction in the margin of safety.
The changes to Technical Specification 6.9.1.9 are administrative in nature.
These changes are being made to reflect the most recent revisions of DPC-NE-4-3
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3002 and DPC-NE-3000, which include changes associated with the replacement steam generators. These topical reports revisions will be reviewed and approved for use regarding Catawba and McGuire Nuclear Stations. Since these changes are administrative in nature, no significant hazards considerations are involved.
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