ML19224D813
| ML19224D813 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/23/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19224D814 | List: |
| References | |
| NUDOCS 7907170076 | |
| Download: ML19224D813 (15) | |
Text
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4 UNITED STATES
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NUCLEAR REGULATORY COMMISSION j) j WASHINGTON, D. C. 20555
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NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 57 License No. DPR-46 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Nebraska Public Power District (the licensee) dated November 29, 1978 and January 30, 1979, comply with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is rea:nnable assurance (i) that the activities authorized by 51s amer.dment can be conducted without endangering the nealth and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment otill not be inimical to the comon defense and security or to the health and safety df 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.
353 003 072.t p 7907170 1
~
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility License No. DPR-46 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 57, are hereby incorporatej in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Thomas ppolito, Chief Operatioq Reactors Branch #3 Division of Operatinq Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 23, 1979 353 004
ATTACHMENT TO LICENSE AMENDMENT N0. 57 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 65 65 96 96 102 102 115 115 122 122 124 124 128 l?o 137c 1
137d la/d 137e 137e 137f 1379 137h 137i 167a 167a 171 171 353 005
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT 3.3.3.3 (cont'd) 4.3.3.3.5 (cont'd) e.
If Specifications 3.3.3.3a 1)
The correctness of the control through d cannot be met, the rod withdrawal sequence input r.o reactor shall not be started, the RWM computer shall be veri-or if the reactor is in the fled.
run or srartup odes at less than 20% 7ted power, it shall 2)
The RWM computer on line diag-be brought to a shutdown nostic test shall be sucess-condition immediately.
fully performed.
f.
The sequence restraints imposed 3)
Proper annunciation of the se-on the control rods may be re-lection error of at least one moved by the use of the individual out-of-sequence control rod in rod position bypass switches for each fully inserted group shall scram testing only those rods be verified.
which are fully withdrawn in the 100% to 50% rod density range.
4)
The rod bloc' functiot of the RWM shall be verified by with-drawing the first rod as an out-of-sequence control rod no more than to the block point.
c.
When required, the presence of a second licensed operator or other qualified employee to verify the following of the correct rod program shall be verilied.
4.
Control rods shall not be with-4.
Prior to control rod withdrawal drawn for startup unless at least for startup, verify that at l
two source range channels have an least two source range channels observed count rate equal to or have an observed count rate of greater than three counts per at least three counts per second, second.
5.
During operation with limiting 5.
When a limiting control cod control rod patterns, as deter-pattern exists an instrument mined by the designated quali-functional test of the REM shall fied personnel, either:
be performed prior to withdrawal of the designated rod (s).
a.
Both R3M channels shall be operable:
or b.
Control rod withdrawal shall be blocked:
or c.
The operating power level shall be limited so that the MCPR will remain above 1.07 assuming a single error that results in complete withdrawal of any single operable control rod.
Amendment No. 57 353 007
3.3 and 4.3 BASES ( c on t.' d) flux. The requirements of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8% of rated power used in the analyses of transients cold conditions.
One operable SRM channel would be adequate to :onitor the approach to criticality using homogeneous patterns of scattered control rod with-drawal. A minimum of two operable SRM's are provided as an added conservatism.
5.
The Rod 31ock Monitor (REM) is designed to automatically prevent fuel da= age in the event of erroneous rod withdrawal from locations of high power density during high power level operation. Two channels are pro-vided, and one of these may be bypassed from the console for =aintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel da= age.
This system backs up the operator who withdraws control rods according to written se-quences. The specified restrictions with one channel out of service conservatively aosure chat fuel da= age will not occur due to rod with-drawal erro s when this condition exists.
A li=iting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit (i. e., MCPR = 1. 07 or LEGR = 18. 5kW/ f t). l During use of such patterns, it is judged that testing of the R3M system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur.
It is the responsi-bility of the Reactor Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Cther personnel qualified to rerform this function may be designated by the station superintendent.
C.
Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel da= age; i.e.,
to prevent the MCPR from becoming less than 1.07.
The limiting pcwer transient is that resulting l
from a turbine stop valve closure with failure of the turbine bypass system.
Analysis of this transient shows that the negative reactivity rates resulting from the scram (FSAR Figure III.6.15) with the average response of all the drives ss given in the above specification, provide tae required protection, and MCPR remains greater than 1.07 l
On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate caterial (probably construction debris) plugging i internal control rod drive filter.
The design of the present control rod d.ve (Model CRD31443) is grossly i= proved by the relocation of the filter to a location out of the scram drive path; i.e.,
it can no longer interf ere with scram performance, even if completely blocked.
353 008 Amendment No. 57
-102-
LIMITINC CO'OITIONS FOR OPERATIO.
SI'RV E_
.E REC" IRE"E'C 3.5 A (cont'd.)
4.5.A (cont'd.)
2.
From and after the date that one of 2.
I.* hen it is determined that one core the core spray subsystems is =ade or spray subsystem is inoperable, the found to be inoperable for any *eason, operable core spray subsystem, the continued reactor operation is per-LPCI subsystem and the diesel gener-missible during the succeeding seven ators shall be de=enstrated to be days provided that during such seven operable innediately. The operable days all active components of the core spray subsyste= shall be demon-oth(r core spray subsystem and active strated to be operable daily there-components of the LPCI subsystem and after.
the diesel generators are operable.
3 '.
Both LPCI Subsystems shall be opera-3.
LPCI Subsystem Testing shall be as ble:
follows:
i (l)' prior to reactor startup from a Item Freauency Celd Condition, except as specified in 3.5.E 7, or a.
Si=ulated Auto-Once/ Operating matic Actuation Cycle (2) when there is irradiated Test fuel in the vessel and when the reactor vessel pressure b.
Pu=p Operability Once/ month is greater than atmospheric pressure, except as specified c.
,4 t r Operated 0~nce/ month in 3.5. A. 4 and 3.5. A. 5 below.
valve operability d.
Pu=p Flow Rate Once/3 months Each RER pump shall deliver at least 8400 gpm but no more than 8800 gmp against a reactor vessel pressure at atmospheric conditions and a total system flow of at least 16,000 gpm if two RHR pumps arc in-jecting into the same recirculation loop.
e.
Recirculation Pu=p discharge valves shall be tested each refueling outage to verify full open to full closed in 20 < t < 26 seconds.
4 From and after the date that one of 4.
When it is deterr'4ed that one of the the RER (LPCI) pumps is made or found RHR (LPCI) pumps is inoperable at a to be inoperable for any reason, con-time when it is required to be operating tinued reactor operation is per=issi-the remaining active cceponents of the ble anlj during the succeeding thirty LPCI Subsystems, the containment cool-days provided that during such thirty ing subsysten, bc:b core spray system lays :5.e remaining active components and the diesel generators shall be i
of :he LPCI Subsyste= and all active demonstrated to be operab]e innediately components of both core spray sub-and the operable LPC pumps daily s7stens and the diesel generators are I thereafter.
- er2ble.
I Amendment No. 57 115 009
LIMITING CONDITICNS FCR OPER.CION SURVEILLANCE RECUIREMENT 3.5.F (cont'd) 4.5.F (cont'd) h.
A special flange, capable of sealing a leaking control red housing, is available for i==ediate use.
1.
The control rod hcusing is blanked folicwing the re= oval of the cen-trol rod drive.
j.
No work is being perfor:ed in the vessel while the housing is open.
6.
During a refueling cutage, refueling operation =ay entinue with one core spray systen the L2CI sys tem in-operable for a period of thirty days.
'7.
The LPCI System is required to be operable while perferning training startups at at=ospheric pressure at power levels less than II of rated ther=al power with the exception that the RRR systen may be aligned in the shutdown cooling mode rather than the LPCI mode.
G.
Maintenance of Filled Dischar2e Pice G.
Maintenance of yilled Dischar2e ? ice The follcwing surveillance recuire=ents
'ahenever core spray subsystems. LPCI shall be adhered to, to assure that the subsystem, EPCI, or RCIC are required discharge piping of the core spray
.o be operable, the discharge pipin3 subsystems, LPCI subsystem, EPCI and frcs the pu=p discharge of these sys-RCIc are filled:
te=s to the last bicek valve shall be filled.
L.
'ahenever the Core Spray, LPCI., E?c t or RCIC syste=s are =ade ope..
l.,
the discharge piping shall be ven:ed from the high point of the sys te= and water ficw observed initially and on a
=cnthly basis.
2.
The pressure swi:ches which =cnitor the L2CI, c0re spray, EPCI and RCIC lines to ensure they are full shall be func:1cnally tes ted and calibra:ed every :hree conths.
Amendment No. 57
-122-
3.5.A SAS c.
Core Sorav and L?CI Subsystems This specification assures that adequate e=ergency cooling capability is available whenever irradiated fuel is in the reactor vessel.
Based on the loss-of-coolant analysis included in General Electric Tooical Repor: NEE 0-10329 and the sensitivity studies given in Supple =ent 1 thereto and subsection 6.5 of the FSAR and in accordance with the AIC's "Interi=
Acceptance Cri:eria for I=ergency Core Cooling Systems" publianed on June 19, 1971, any of the following cooling syste=s provides suf ficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, to linit calculated fuel clad te=perature to less than 2300 F, to assure that core geometry re=ains intact, and to l' 4 e clad metal-water reaction to less than 17.; the two core spray subnyste=s; or either of the two core spray sub-syste=s and three RER pu=ps operacing in the LFCI code with operable L?CI injection valves.
The liz1 ting conditions of operation in Specifications 3.5.A.1 through 3.5. A.6 specify the combinations of operable subsystems to assure the availabill:7 of the mini =us cooling syste=s noted above. During reactor shutdown when the residual heat re=cval syste= is realigned fm L?CI to the shutdown cooling code, the L?CI Systa= is considered operable.
Core spray distributica has been shcwn, in full-scale tests of syste=s simil.r in design to that of Cooper Nuclear Station, to exceed the =ini=us require-
=ents by at least 25%. In addition, cooling ef fectiveness has been de=en-strated at less than half the rated flow in si=ulated fuel asse=blies with heater rods to duplicate the decay heat characteristics of irradiated fuel.
Tae accident analysis is additionally conservative in that no credit is taken for spray coolant entering the reactor before the internal pressure has f allen to 113 psig.
The LPCI subsysta= is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolanc accident. This systen functions in combination with the core spray syste= to prevent excessive fuel clad te=perature. The L?CI subsyste= and the core spray subsyste= provide ade-quate cooling for break areas of approxi=ately 0.2 square feet up to and including the double-ended recirculation line break without assistance fre=
the high pressure e=ergency core cooling subsystems.
The allcwable repair ti=es are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in ref erence (1). Using the results developed in this ref erence, the repair period is found to be 1/2 the test interval.
This assu=es tha t the (1) Jacot I.M., " Guidelines for Determining Saf e Tes t Intervals and Repair Ti=es f or Engineered Saf eguards", General Electric Co. A.P.I.D., April, 1969 (APID 5736).
i
~~
Amendment No. 57
_t 3_
35,a 011
3.5 3ASES (cont'd)
=ent is available at all ti=es.
It is during refueling cutages that =ajor
=aintenance is perfor=ed and during such ti=e that all low pressure core cooling syste=s may be cut of service. Specification 3.5.F.4 provides that should this occur, no work will be perfor=ed on the primary syste= which could lead to drainiag the vessel. This work would include work on certain control rod drive cc=ponents and recirculation syste=.
Thus, the specifica-tion precludes the events which could require core cooling.
Specifica ica 3.5.F.5 recognizes that, concurrent with control rod drive =aintenance during the refueling outage, it =ay be necessary to drain the suppression chamber for =aintenance or for the inspection required by Specification 4.7.A.2.h.
In this case, if excessive control red housing leakage occurred, three levels of protection against loss of core cooling would exist. First, a special flange vculd be used to stop the leak.
Second, sufficient inven-tory of water is =aintained to provide, under worsa case leak conditions, approxi=ataly 60 sinutes of care ccoling while atte= pts to secure the leak are =ade.
This inventory includes water in the reactor well, spent fuel pool, and condensate storage tank.
If a lcak should occur, =anually operated valves in the condensate transfer system can be opened to supply either the core spray systen or the spent fuel pool. Third, surficient inventory of water is -MntMned to per=it the water which has drained from the vessel to fill the torus to a level above the core spray and LPCI suc.icn strainers. These syste=a could then recycle the water to the vessel.
Si. ice the systes cannot be pressurired during refueling, the potential need for core flooding only exists and the specified cc=bination of the core spray or the LFCI systen can provide this. *his specification also provides fo.-
the highly unlikely case that both diesel generators are fcund to be inoper-able. The reduction of rated pcwer to 25% will provide a very stable operating condition. The allowable repair time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will provide an opportunit7 to repair the diesel and thereby prevent the necessity of taking the plant dcwn through the less stable shutdcun ccndition. If the necessary repairs cannot be =ade in the allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the plant will be shutdown in an orderly fashion. This will be acco=plished while the two of f-site sources of power required by Specification 3.9. A.1 are available.
Specification 3.5.E.7 provides for the perfor=ance of traintag startups withcut realigning the residual heat removal syste= from the shutdeva cooling
= ode to the LFCI = ode. Power levels during training startups are kept below the level of significant heat addition.
G.
Maintenance of Filled Discharge Pine If the discharge piping of the core spray, LPCI subsyste=, EPCI, and RCIC are not filled, a water ha==er can develop in this piping when the pu=p and/or pu=ps are s tarted If a uater ha==er were to occur at the ttne at which the syste= were required, the syste= would s'.ll perfor= its design functicus.
Ecwever, to =inimite damage to the discharge piping and to ensure added nar-gin in the operation of these systems, this Technical Specifica ica requires the discharge lines to be filled whenever the syste= is in an operable condi-tien.
H.
Engineered Safeguards Cc= cart =ents Cooling The unit eccler in each pu=p cc=part=ent is capable of providirg adequate 7en-tilatica flow and cooling. Engineering analyses indicate that the te=pera ture rise in saf aquards cc=part=ents without adequate ventilation ficw or cooling is such that continued operation of the safeguards equipment er associated auxiliary equipment cannot be assured.
4 Q
[
Amendment No. 57
-128-1L a
Table 3.6.1 SAFETY REIATED HYDPAULIC SHCCK SUPPRESSCRS (STUBBERS)
Snubber No.
Location Elevation AS-S-110 Torus Area 890'9" l
AS-S-lll forus Area 891'9" A3-S-112 Torus Area 891'9" Ab-9-ll3 Torus Area 893' 3S-S-1 B&R Torus 870'8" SS-S-15 Torus Area 893'4" CS-S-1 S.E. Quad 918' CS-5-2 S.E. Quad 929' g
CS-S-3 S.E. Quad 946'3" CS-S-10 Rx Bldg, 931' 946'3" CS-S-11 ax Bldg, 931' 946'3" i
HP-S-4 S.W. Quad 872'7" HP-S-11 S.W. Quad 869'11" HP-S-15 S.W. Quad 8/4'11" MS-S-1 S.W. Quad 864' MS-S-2 S.W. Quad 868'5" MS-S-3 S.W. Quad 880'4" MS-S-4 S.W. Quad 873'5" MS-S-7(2)
S.W. Quad 874'11" MS-S-8 Tor'is Area 885'2" l
MS-S-10 Torus Area 399'11" i
MS-S-Il Torus Area 897' MS-S-12 Torus Area 888' MS-S-13 S. RER Hx Rm 904'10" MS-5-14 S. RHR Ex Rm 923' MS-S-15 S. RRR Ex Rm 934' MS-S-16 Torus Area 885' MS-S-17 N. RER Hx Rs 904'10" MS-5-18 N. RHR Ex ?m 905'6" MS-S-19 N. RHR Ex Rm 921' MS-S-20 N. RHR Hx Rm 934' g
MS-S-23 Torus Area 898' MS-5-24 Torus Area 898' MS-S-25 N.E. Quad 877'6" MS-S-26 N.E. Quad 879'6" Amendment No. 57
-137c-
Table 3.6.1 SAFETY REIATED HYDRAULIC SHOCR St'PPRESSORS (SNU3BERS) (Cont'd)
Snubber No.
Location Elevation RCC-S-3 Rx Bldg, 931' 945'11" RCC-S-4 Rx Bld g, 9 31 '
943'6" RCC-S-20 Rx Bldg, 931' 953'3" RCC-S-21 Rx Bldg, 931' 953'3" RCC-S-22 Rx Bldg, 931' 953'3" RF-S-1 N.E. Quad 898'6" RF-3-2 Torus Area 896' RF-S-3 S.W. Quad 870' RF-S-4 Torus Area 894'6" RF-S-5 Torus Area 897'10" RF-S-6 Torus Area 891' RH-S-20 Rx Bldg, 903' 912'6" RH-S-21 Rx Bldg, 903' 911' RH-S-22 Torus Area 895'9" FE-S-23 Torus Area 892' RH-S-24 Torvs Area 897' RH-S-25 N. RHR Ex Rm 927' RH-S-26 N. Rha Hx am 929' RH-S-29 Rx Bldg, 903' 904'6" RH-5-30(2)
Torus Area 898'6" RH-S-32 Torus Area 894'7" RH-S-33D Iorus 892'3" RH-S-34 Rx Bldg, 903' 919'6" RH-S-35 S. RHR Hx Rm 912' RH-S-36 S. ?2R Hx Rm 914'3" RH-S-37 S. RHR Hx R=
916'4" RH-S-38 S. RER Ex Rm 930' RH-S-39 S. RHR Hx Rm 927'6" RH-S-40 S. RHR Hx Rm 915'6" RH-S-41 S.W. Quad 8'3' RH-S-42 S.W. Quad 874' RH-S-43 Torus Area 897' RH-S-44 S.W. Quad 884'6" RH-S-45 S.W. Quad 884' RH -S-48 N.W. Quad 884'6" RH-S-49 N.W. Quad 885' RH-S-51 N. RER Ex Rm 914'3" RH-S-52 N. RHR Ex Rm 915' RH-S-34 N.W. Quad 873'1" RH-S-55 N.W. Quad 874' P.H-S-36 N. RHR Hx Rm 927'6" RR-S-57 N. RHR Ex Rm 927'6" RH-S-58 N. RHR Hx Rm 921'11" RH-S-59 Torus Area 896' RH-S-65 S.W. Quad 887'2" RH-S-66 Rx Bldg, 903' 907'4" l
-137d-Amendment No. 57
Table 3.6.1 SAFETY RELATED HYDRAULIC SHOCK SUPPRESSCRS (SNU3BERS) (Cont'd)
Snubber No.
Locaticn Elevation RH-S-76(2)
Torus Area 898' RH-S-77 Torus Area 890'11" RH-S-78(2)
Torus Area 897' l
RH-S-80 N.W. Quad 889' RH-S-98 N.W. Quad 891' I
SbE-kE-23 A Intake Str.
904'3" SkE-WH-23B Intake Str.
904'3" SkH-bE-23C Intake S tr.
904'3" SWH-bE-23D Intake Str.
904'3" Amendment No. 57 353 015
-137e-
LIMITING CONDITIONS'FOR OPERATION I SURVEILLANCE REOUIREMENT 3.7 (cont'd) 4.7 (cont'd)
E.
Drywell-Sucoression Chamber E.
D rywell-Suc p r<
sion Chatber Differential Pressure Differential Pressure 1.
Dif ferential pressure between the 1.
The pressure differential drywell and suppression cha:ber between the drywell and shall be =aintained at equal to suppressicn cha ber shall or greater than 1.47 psid except be recorded at least once as specified in a, b, and c belew.
cach shift.
a.
This differential shall be established within 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> after placing the mode switch in run.
b.
This dif ferential =ay be de-creased to less than 1.47 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to placing mode switch in refuel or shut-down.
c.
This differential may be decreased to less than 1.47 psid for a maximum of four (4) hours during required operability testing of the IIPCI system pu=p, the RCIC sys tem pump, and the drywell-pressute suppression chamber vacuum breakers.
2.
If the differential pressure of specification 3.7.E.1 cannot be maintained, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initi-ated and the reactor shall be in Hot Standby in six (6) hours and in a Cold Shutdown condition within the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
53 014
-167a-V Amendment No. 57
TABLE 3.7.2 TESTABLE PENETRATICNS WITH DOU3LE O-RING SEALS PEN. NO DESCRIPTION X-1A Drywell equipment hatch X-13 Drywell equipment hatch X-4 Drywell head access hatch X-6 CRD removal hatch X-35A TIP "D" Penetration X-35B TIP "A" Euaetration X-35C TIP "C" Penetration X-35D TIP "B" Pener. ration X-35E TIP N3 Purge Connection X-200A Suppression chanber access hatch X-2003 Suppression chamber access hatch Drywell head Stabilizer Asse=bly Inspection Ports (3) l 353017 Amendment No. 57
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