ML20126G155
| ML20126G155 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/22/1992 |
| From: | Larkins J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126G158 | List: |
| References | |
| NUDOCS 9301040090 | |
| Download: ML20126G155 (14) | |
Text
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.j' UNITED STATES
[
NUCLEAR REGULATORY COMMISSION W ASHING TON, D. C. 20666 e...+
j NEBRASKA PUBLIC POWER DISTRICT j
DOCKET NO. 50-298 COOPER NVCLEAR STATION j
AMENDMENT TO FACIllTY OPERATING LICENSE j
Amendment No. 156 License No. OPR-46 1.
The Nuclear Regulatory Commission (the Commission) has found that:
i A.
The application for amendment by Nebraska Public Power District '(the i
licensee) dated November. 15, 1991, complies with the standards and i
recuirements of the Atomic Energy Act.of 1954, as amended (the Act),
anc the Commission's rules and regulations set forth in 10 CFR-j' Chapter I; l
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;
{
C.
There is reasonable assurance:
(1 that the activities' authorized by this amendment can be conducted)without-endangering the health 1
and safety of the public, and-(ii) that such activities will-be-i conducted in compliance with the Commission's regulations;.
D.
The issuance of this license amendment will not be inimical 'to the l.
common defense and security or to the health and safety of-the public; and 1
l.
E.
The issuance of this amendment is in accordance with-10 CFR Part E!
of the Commission's regulations and.all applicable requirements have 1
been satisfied, 5
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- 9301040090 921222
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. 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of facility Operating License No. DPR-46 is hereby amended to read as follows:
2.
lechnical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.156, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is etfective as of its date of issuance.
FOR THE NUCLEAR RE60LATORY COMMISSIDN John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: December 22, 1992
4 e
ATTACHMENT TO LICENSE AMENDMENT NO 156 FACIllTY OPERATING'LIC,Mif NO. DP8-li DOCKET NO. 50-298 Replace the following ) ages of the Appendix A Technical Specifications with the enclosed pages.
Tie revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES 18 18 93 93 4
l 94 94 1-95_
95 i
96-96 97 97 1
100 100 101 101 i
101a 101a i
216b4 216b4 l
'l i-l' l
j1 4.
i I'-
i
2.1 naans
(cont'd)
An increase in the APM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.
Reducing this operating margin would increase the frequency of spurious scrans which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.
b.
APRM Flux Scram Trio lettina (Refuel er Etart & Eat standhv Modai Por operation in the startup mode while the reactor is at low pressure,' the APM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety Limit, 25 percent of rated. The margin is adequate to l
accomodate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at aero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedure backed up by the Rod Worth Minimiser.
Worth of l
individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.
Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APM system would be more than adequate to assure a scram before the power could exceed the Safety Limit. The 15 percent ApM scram remains active until the mode switch is placed in the RUN position.
This change can occur when reactor pressure is greater than Specification 2.1.A.6.
i
(
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l Amendment No. Mr44r4n,. 156 18 l
l
.e...e.
se. o.m L
LIMITINC CONDITION FOR OPERATION SURVEILIANCE REDUIREliENT 3.3 REACTIVITY CONTROL 4.3 REACTIVITY CONTROL Annlicabilitv!
Anolicability!
Applies to the surveillance Applies to the operational status of requirements to the control rod the control rod system.
system.
Obinetivat-Obiectivat To assure the ability of the control To verify _the ability of the control rod system to control reactivity.
rod systen to control reactivity.
Snecification!
It3t* " III"
- t I*" '*-
A.
Esaglivity Limitations A.
Raaetivitv Limitatiana I'
"****IVI'Y ***8 " * * ** I"*dI"3 I
1.
Reactivity margin core loading sufficient control rods shall be A sufficient number of control rods withdrawn following a refueling shall be operable so that the core outage when core alterations were l
could be made suberitical in the performed to demonstrate, with a most reactive condition during the margin of 0.38% Ak/k, that the core operating cycle with the strorgest can be made suberitical at any time control rod fully withdrawn and all in the subsequent fuel cycle with other operable control rods fully the analytically determined
- inserted, strongest operable control rod fully withdrawn and all other operable rods fully inserted.
Amendment No.. 20, 50,_156 93
-.m.__.
l LIMITINC CONDITION VDR OPERATION SU5VIII.iANCE REDUIREMENT
[
3.3.A (cont'd.)
4.3 (cont'd) 3 1
2.
Reactivity margin inoperable 2.
Reactivity margin inoperable control rods control rods
]
a.
Control rods which cannot be moved a.
Each partially or fully withdrawn l
with control rod drive pressure operable control rod shall be 1
shall be considered inoperable.
If exercised one notch at least once i
a partially or fully withdrawn each week, when operating above 301 control rod drive cannot be moved power. This test shall be performed with drive or scram pressure the at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when reactor shall be brought to a operating above 301 power in the j
shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> event power operation is continuing 1
unless investigation demonstrates with three or more inoperable i
that the cause of the failure is not control rods or in the event power i
due to a failed control rod drive operation is continuing with one j
mechanism collet housing, full or partially withdrawn rod whic cannot be moved and for.which b.
The control rod directional control control rod drive mechanism damage j
valve for inoperable control rods has not been ruled out.
The 1
shall be disarmed electrically, surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of l
c.
Control rods with scram times inoperable rods has been reduced to i
greater than those permitted by less than three and if it has been l
Specification 3.3.C.3 are demonstrated that control-rod drive inoperable, but if they can be mechanism collet housing failure is j
inserted with control rod drive not the cause of an immovable pressure they need not be disarmed control rod.
a electrically.
b.
Deleted.
l l
d.
Control rods with a failed " Full in" or *rull out" posi+. ion switch may be c.
Once per week, check the status of i
l considered operable if the actual the pressure and level alarms for j
rod position is known. These rods each accumulator.
l must be moved in sequence to their correct positions (full in on insertion or full out
-on l
withdrawal).
i e.
Control rods with inoperable i
accumulators or those whose position cannot be positively determined 4
shall be considered inoperable, n
f.
Inoperable control rods shall be positioned such that Specification 3.3.A.1 is set. In addition, during reactor power operation, inoperable y
control rods shall be separated by at least two control rod cells.
I If this Specification cannot be met i
the reactor shall not be started, or l
if at power, the reactor shall be brought to a shutdown condition i.
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
T Amendment No. 43r 156-
.g4
. _. - _, _. - _. - _ - - _.. -. ~ _ _ -.... _, - - -. _ _ _,
1_1MITIMO CONDITION FOR OF*mATION RLfRVEfttAMCI hamtthWWmr?
l L
3.3.5 (cont'd) 4.3.5 (cont'd) l l
c.
During each refueling outage observe that-any drive which has been uncoupled free and subsequently recoupled to its control rod does 2.
The control rod drive housing not 30 to the overtravel position, support system shall be in place during reactor power operati n er 2.
W control rod drive housing when the reactor coolant systen is support system shall be inspected pressurized above steospheric after reassembly and the results of pressure with fuel in the reactor the inspection recorded.
vessel. unless all control rods are fully inserted and specification 3.3.A.1 is set.
3.a. Deleted.
3a.
Deleted.
l-
-b. Deleted.
b.' - Prior to the start of control rod withdrawat ' towards criticality and
- c. Whenever the reactor -is in the prior to attaining lot rated power
-- l
-l startup or run modes below 101 rated during rod insertion while shutting power the Rod Vorth Minimiser shall down. the capability of the Rod be operable or a second licensed Worth Minimiser (RWM) to properly operator or other qualified employee fulfill its function-shall be shall verify that the operator at verified by the following checks:
l the reactor console is following the control rod program. Reactor startup shall-not be -initiated more fre-quently than once per calendar year with the RWM inoperable.
- d. Deleted.
b
'Anendment No, M4r 156 95 q
2________
d
LIMITING CONDITION FOR OPERATION SURVEILIANCE REOUIREMEffr 3.3.B.3 (cont'd) 4.3.B.3.b (cont'd) l e.
If Specification 3.3.B.3.c 1)
The correctness of the Banked cannot be
- met, the reactor Position Withdrawal Sequence shall not be started, or if the input to the RWM computer reactor is in the run or start.
shall be verified.
I up modes at less than 10% rated power it shall be brought to a 2)
The RWM computer on line diag.
shutdown condition immediately, nostic test shall be success.
I fully performed.
4.
Control rods shall not be with.
drawn for startup unless at 3)
Proper annunciation of the least two source range channels selection error of at least have an observed count rate one out of sequence control equal to or greater than three rod in each fully inserted counts per second.
group shall be verified.
5.
Doeratdon with a Limiting 4)
The rod block function of the Control Rod _ Pattern (for Rod RWM shall be verified by with.
Withdrawal _ Error. R R),
drawing the first rod as an out of sequence control rod no a.
A Limiting Control Rod Pattern more than to the block point.
for RWE exists when either:
c.
When required, the presence of
- 1) Core thermal power is 2 30%
a second licensed operator or and < 90% of rated power other qualified employee to and the MCPR is less than verify the following of the 1.70, or correct rod program shall be verified.
- 2) Core thermal power is 2 90%
of rated power and the MCPR 4
Prior to control rod withdraw.
is less than 1.40.
al for startup, verify that at least two source range chan.
b.
During operation with a Limit.
nels have an observed count ing Control Rod Pattern for RWE rate of at least three counts either:
per second.
- 1) Both rod block monitor 5.
DoeratLon with_a Limitine (RBM) channels shall be Control Rod Pattern (for Rod operable, or Withdr.ayal Error. RWE)
- 2) With one RBM channel inop.
During operation when a Limit.
erable, control rod with.
ing Control Rod Pattern for drawal shall be blocked RWE exists and only one RBM within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or channel is operable, an in.
strument functional test of
- 3) Vith both RBM channels in.
the RBM channel shall be per.
operable, control rod with.
formed prior to withdrawal of drawal shall be blocked the control rod (s).
A Limit.
until operability of at ing Control Rod Pattern for least one channel is re.
RWE is defined by Specifica.
- stored, tion 3.3.B.S.
Amendment No. 117, 151 r 156 96 1
_ _... = _...
i o,
LIMIT 7NG CONDITION FOR OPERATION SURVEILIANCE REOUIREtifg 3.3 (cont'd)
- 4.' 3 ' (cont'd) l C.
Scram Insertion Times C.
Scram Insertion' Times j
1.
The average scram insertion time, l '. - ' After-- each refueling outage-all S
based on the deenergitation of the operable rods shall be scram time scram pilot valve solenoids as time
- tested -from the fully withdrawn sero, of'all operable control rods position with - the nuclear system i
in the reactor power operation con-
' pressure above 800 psig. - This test-
_ l dition shall be no greater than:
ing - shall be completed prior to exceeding 40%. power.
During 1 all-
% Mnserted From Ave. Scram Inser-scram time testing below 10% power, l
FuL1v Vithdrawn tion Times (see) the Rod Worth' Minimirar. shall be 5
0.375 operable or a second l' msed opera-20 0.90.
tor or--other-qualit. d employee 1
50 2.0 shall verify that the operator at.
l 90
.3,50 the reactor console'is following the control rod program.
)
i 2.
At 16 week intervals. 10% of' the
- operable control rod drives shall be scram timed above 800 psig. Whenev.
j
- er such scram-time measurements are.
t made, r.n evaluation shall be made to provide reasonable assurance that-j-
proper control rod drive performance 2.
The average of the scrao insertion times for the three fastest control
_ is_being maintained.
l rods of all groups of four control.
rods in a two by+two array shall be no greater than:
I
% Inserted From Ave.' Scram Inser-l Fully Withdrawn tion Times (see) i' 5
0.398 20 0.954 50 2.120 i
90 3.71-(
l 4
4 1:
4
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Amendment No. 16, 32. 80.'156 97-d s
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' t and 4. 3 BASES (cont'd.)
cannot be moved with drive pressure.
If the rod is fully inserted and then disarmed electrically, it is in a safe position of maximum contribution to shutdown reactivity. If it is disarmed electrically in a non fully inserted position, that position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A.l.
This assures that the core can be shutdown at all times with the remaining control rods assuming the strongest operable control rod does not insert.
An allowable pattern for control rods valved out of service, which shall meet this Specification, will be determined and made available to the operator.
If damage within the control rod drive mechanism and in particular, cracks in drive internal housings, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out.
Circumferen-tial cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of
- c. racking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the af fected rods. Limiting the period of operation with a potential-ly severed collet - housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collet housings.
B.
Control Rod 1.
Control rod drop accidents as discussed in the USAR can lead to significant core damage.
If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated.
The overtravel position feature provides a positive check as only uncoupled drives may reach this position.
Neutron instrumentation response to rod movement provides a verification that the rod is following its drive. Absence of such response to drive movement could indicate an uncoupled condition.
Rod position indication is required for proper function of the Rod Worth Minimizer (RWM).
4 l
Amendment No, 44J 156 100 7
a 3.3.B and 4.3.B EASES (cont'd.)
l 2.
The control rod housing support restricts the outvard movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The design basis is given in Section 111 8.2 of the USAR and the safety evaluation is given in Section 111-8.4 This support
- .s not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing. Additionally, the support is not required if-all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain suberitical even in the event of complete ejection of the strongest control rod.
3.
The Rod Worth Minimizer (RW) restrict withdrawals and insertions of control l
rods to prespecified sequences. These sequences are established such that the drop of any in sequence control rod or control rod segment (i.e., one or more notches) would not cause the reactor to sustain a power excursion resulting in a peak fuel enthalpy in excess of 280 cal./gm.
An enthalpy of 280 cal./
which rapid fuel dispersal could occur (i.e.gm. is well below the level at 425 cal./gm.).
Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Ref. Subsections 111 6.6 and I
XIV 6.2 of the USAR and Reference 1.
In performing the function described above, the RW is not required to impose any restrictions at core power levels in excess of 10% of rated. Material in the cited references shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than 10%
regardless of the rod pattern.
This is true for all normal and abnormai patterns including those which maximize the individual control rod worth.
At power levels below 10% of rated, abnormal control rod patterns could produce i
rod worths high enough to be of concern relative to the 280 calories per gram rod drop limit. In this range the RW constrains the control rod sequences and i
patterns to those which involve only acceptable rod worths.
The RW provides automatic supervision to assure that out of sequence control j
rods will not be withdrawn or inserted; i.e.,
it limits operator deviations from planned withdrawal sequences. It serves as a backup to procedural control on control rod sequences, which limits the maximum reactivity worth of control rods.
In the event that the RW is out of service, when required, a second j
l licensed operator or other qualified technical plant employee can manually fulfill the control rod pattern conformance functions of this system.
I The function of the RW makes it unnecessary to specify a license limit on rod i
i worth to preclude unacceptaU e consequences in the event of a centrol rod drop.
At low powers, below 10%, this system forces adherence to acceptable rod patterns. Above 10% of rated power, no constraint on rod pattern is required l
to assure that rod drop accident consequences are acceptable.
Control rod pattern constraints above 10% of rated power are imposed by power distribution requirements as defined in Section 3.3.B.5 of-these Technical Specifications.
Power level for automatic cutout of the RW function is sensed by feedwater and j
steam flow.
Functional testing of the RW prior to the start of control rod withdrawal at startup, and prior to attaining 10% rated thermal power during rod insertion l
while shutting down, will ensure reliable operation and minimize the probability of the rod drop accident.
The Reduced Notch Worth Procedure for control rod withdrawal allows CNS to take l
advantage of the Banked Position Withdrawal Sequence (BPWS) (Ref. 4). The BPWS has the advantage of having been proven statistically to have such low individual control rod worths that the possibility of a control rod drop accident (CRDA), which exceeds the 280 cal /gm peak fuel enthalpy limit, is precluded (Ref. 1).
l Amendment No. Elr 156
-101-
j,-.
4 b.
The Reduced Notch Worth Procedure.. is. programmed into - the RWM. In the pre..
checkerboard pattern (1001 to 50% control rod density).--the RWM will enforce the Reduced Notch Worth Procedure, 3
i j
4 The Source Range. Monitor (SRM) system performs no automatic safety system function; i.e., it has no scram function.
It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents
(
are functions - of the initial neutron flux. - The requirements of at.least 3 counts per second assures thatigny transient, should it occur, begins at or above the initial-value of 10 1 of rated power used in the analyses - of transients cold conditions.
One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of-scattered control rod withdrawal.
A minimum of two operable SRM s are provided as an added conservatism, i
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1 T
1 T
I 4
Amendment No. 44 h 156-
-lola-4 m
~
?
LIMITING CONDITIONS FOR OPERATION-SURVEILIANCE REOUIREMENTS i
-3.22 SPECIAL TESTS / EXCEPTIONS (CONT'D)
- -4.22 SPECIAL TESTS / EXCEPTIONS (CONT'D)
T 2.
Deleted.
3.
RHR System The RHR system may be aligned in the shutdown cooling mode with at least one shutdown cooling mode loop OPERABLE while performing the Shutdown Margin Demonstration.
4 Containment Systems Primary containment is not required while performing the j
Shutdown Margin Demonstration
' when. reactor water tempera.
j.
ture is equal to or-less than l
212'F.
l B.
Trainine Startun -
B.
Trainine Startun j.
1.
LPCI System The reactor vessel'.shall be verified The LPCI System is-required to be.
t be ungenurized 'and the themal-operable with the exception that the RHR system may be - aligned in : the Power verified to be less than 1% of rated thirmal power at least once l
shutdown cooling mode while perform.
ing training startups at atmospheric.:
Per hour during training.startups.
pressure at power levels less than j.
3% of rated thermal power.
a a
I i,
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4 i
r J.
i,.
1' Amendment No. 97, 152, 156 216b2.-
0
)
3.32 6 4.23 BASES A.
Shutdown Margin Demonstration Performance of shutdown margin demonstrations requires additional h
restrictions in order to ensure that_ criticality does not occur.
Addi-tional surveillance requirements ensure that shutdown margin requirements and individual rod worths do not exceed values assumed in the safety analysis. Since power levels attained during the demonstration are kept below the level of significant heat addition, the residual heat removal system can remain aligned in the shutdown cooling mode.
B.
Trainine Startuo Specification 3.22.B provides for the performance of training startups without realigning the residual heat removal systeW from the shutdown cooling mode to the LPCI mode. Power levels during training startups are kept below the level of significant heat addition.
This exception is made in order to minimize contaminated water discharge to the radioactive waste disposal system.
C.
Physics Tests An exception is made to primary containment integrity during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required.
There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break.
The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect'again to minimize the probability of an accident occurring. Procedures and the rod worth mini-mizer would limit control worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an, excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below 10CFR100 limits.
D.
Startuo Test Procram Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the initial startup and testing phase nf operatica. Without this access the startup and test program could be restricted and delayed.
The recirculation flow exception permits reactor criticality under no-flow conditions and is required to perform certain startup and-physics tests while at low thermal power levels.
Amendment No. 43, 156
-216b4