ML20153B325
| ML20153B325 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/17/1988 |
| From: | Norrholm L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20153B329 | List: |
| References | |
| NUDOCS 8807130039 | |
| Download: ML20153B325 (24) | |
Text
s
/
'o UNITED STATES
~g NUCLEAR REGULATORY COMMISSION o
h WASHINGTON, D. C. 20555
\\...../
COMMONWEALTH EDIS0N COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET N0. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 104 License No. DPR-30 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated March 28, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health 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 will not be inimical to the common defense and. security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have baen satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions (as indicated in the attachment to this license amendment).
Further-more, paragraph 3.C. of Facility Operating License No. DPR-29 is deleted in its entirety, and paragraph 3.B is hereby amended to read as follows:
kgRo713co39 830617 p
ADOCK 05000265 PNu
. a B.
Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No.104, are hereby incorporated in the license. The licensee shall operate the facility in accordance with 'he Technical Specifications.
C.
(Deleted) 3.
This license amendment is effective as of the date of its issuance.
' FOR THE NUCLEAR ULATORY COMMISSION f'
l' I
g Leif orrholm, Acting Directo P Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects
Attachment:
Changes to the Technical
',pecifications Date of Issuance: June 17,1988 l
l i
O ATTACHMENT TO LICENSE AMENDMENT NO. 104 FACILITY OPERATING LICENSE N0. OPR-30 DOCKET N0. 50-265 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 1.1/2.1-1 1.1/2.1-1 1.1/2.1-4 1.1/2.1-4 1.1/2.1-7 1.1/2.1-7 Figure 2.1-3 Figure 2.1-3 3.2/4.2-14 3.2/4.2-14 3.2/4.2-14a 3.2/4.2-14a 3.3/4.3-5 3.3/4.3-5 3.5/4.5-5 3.5/4.5-5 3.5/4.5-10 3.5/4.5-10 3.5/4.5-12 3.5/4.5-12 3.5/4.5-14 3.5/4.5-14 l
3.5/4.5-14a 3.5/4.5-14a 3.5/4.5-14b 3.5/4.5-14b Figure 3.5-1 (Sheets 1 thru 6)
Figure 3.5-1 (Sheets 1 thru 5)
Figure 3.5-2 Figure 3.5-2 3.6/4.6-Sa 3.6/4.6-Sa 3.6/4.6-13a 3.6/4.6-13a
QUAD CITIES OPR-30 1.1/2.1 FUEL CLA00!NG TNTEGRITV
$AF(TY LIMIT LIMITING SAFETY $YSTEM SETTING Applicability:
Applicability:
The safety limits established to preserve The limiting safety system settings apply the fuel cladding integrity apply to to trip settings of the instruments and those variables which monitor the fuel devicJs which ark provided to prevent the thermal behavior, fuel cladding integrity safety limits from being exceeded.
Objective:
Ob.'ective:
The objective of the safety limits is to
-The objective of the limiting safety sys-establish limits below which the integ-tem settings is to refine the level of rity of the fuel cladding ts preserved.
the process variables st which automatic protective action is initiated to pre-vent the fuel cladding integrity safety limits from being exceeded.
SPECIFICATIONS A.
Reactor Pressure > 800 psig and Core A.
Neutron Flux Trip $ettings Flow > 10% of Rated The existence of a minimum critical The limiting safety systen trip set-power ratio (MCPR) less than 1.04 tings shall be as specified below:
shall constitute violation of the fuel cladding integrity safety limit.
1.
APRM Flux scram Trip Setting (Run Mode) 8.
Core Thermal Power Limit (Reactor Pressure 1 800 psig)
When the reactor mode switch is in the Run position, the APRM When the reactor pressure is 1 800 flux scram setting shall be as psig or core flow is less than 10% of shown in Figure 2.1.1 and shall rated, the core therrul power shall be:
not exceed 25% of rated thermal power.
5 1 (.5BWo + 62)
C.
Power Transient with a maximum setpoint of 120%
1.
The neutron flux shall not fogcoreflowequalto98x exceed the scram setting estab-10 lb/hr and greater.
11shed in specification 2.1A for longer than 1.5 seconds as ind1-where cated 'oy the process computer.
S setting in percent of rated 2.
When the process computer is out power j
of service, this saf ety 1*: alt shall be assumed to be exceeded WO s percent o' drive flow if the neutron flux exceeds the required to produce a rated core scram setting established by flow of 98 million 1b/hr. In Specification 2.1. A and a con-the event of operation with a l
trol rod scram does not occur.
maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
E12 51 (.58WD + 62 ( MFLPD ]
1 l
l 09258 1.1/2.1-1 Amendment No.
104 1
l 1
1
.I
Qua0 C!nts OPR-30 j
1.1 SAFETY LIMIT BASIS The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damage is not directly observable, a step-back approach is used to establish a safety limit such that the mintmm critical power ratto (MCPR) is no less than the fuel cladding integrity safety limit MCPR > the fuel cladding integrity safety limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical boundaries which separate radioactive materials from the environs. The integrity of the fuel cladding is related to its relative freedom from perforations or cracking.
Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumlative anJ j
continuously measur4ble. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operatton significantly above dtsign conditions and the protection system safety settings. While fission product migration from cladding perforation is just as 'neasurable as that f rom use-related cracking, the thermally caused cladding perforations signal a thre'shold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding safety limit is defined with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0).
These conditions represent a signiftcar.t departure from the condition intended by design for planned operation. Therefore, the fuel cladding integrity safety limit is established such that no calculated fuel damage shall result from an abnormal operational transtant. Basis of the vahles derived for this safety limit for each fuel type is documented in References 1 and 2.
A.
Reactor Pressure > 800 pstg and Core Flow > 107, of Rated Onset of transition boiling results in a decrease in heat transfer from the cladding and therefore elevated cladding temcerature and the possibility of cladding failure. However, the existence of critical power, or boiling transition is not a directly observable parameter in an operating reactor.
Therefore, the margin to bo11tng transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the critical po,ier ratio (CPR), which is the ratio for the bundle power which would produce onset of transition boiling divided by the actual bundle power.
The minimum value of this ratio for any bundle in tne core is the minimm critici1 power fatio (MCaR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables (Figure 2.1-3).
The MCPR fuel cladding integrity safety limit has sufficient conservatism to assure that in the event of an abnormal operational transient initiated from the normal operation condition, more than 99.97,of the fuel rods in the core are expected to avoid bolling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the safety limit, is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the l
core operating state, including uncertainty in the bolling transition I
correlation (see e.g., Rgference 1).
Because the boiling transition correlation is based on a large quantity of full-scale data, there is a very high confidence that operation of a fuel assembly at the condition of MCPR -
the fuel cladding integrity safety limit would not produce boiling transition.
However, if boiling transition were to occur, cladding perforation would not be expected. Cladding temperature would increase to approximately 1100*F, which is below the perforation temperature of the cladding material. This had been verified by tests in the General Electric Test Reactor (GETR), where similar fuel operated above the critical heat flux for a significant period of timo (30 minutes) without cladding perforatton.
l If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of the boiling transition correlation), it would be assumed that the fuel cladding integrity safety Itmit has been violated.
1 In addition to the bolling transition limit (MCPR) operatton is constrained to a maximm LHGR of 13.4 kw/f t for fuel types P8x8R and BP8x8R, and 14.4 kw/f t for fuel types Gt8x8E and Gt8x8E8. This constraint is established by l
Specification 3.5.J. to provide adequate safety margin to 17, plastic strain for abnormal operating transients initiated from high power conditions.
i l
Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from lower power conditions by adjusting the APRM flow-biased scram l
Setting by the ratto of FRP/MFLPO.
09568 1.1/2.1-4 Ainendment No. 104
QUAD CETIES OPR-30 2.1 LIMITING SAFETY SYSTEM SETTING BASES The abnorm &l operational transients applicable to operation of the units have been analyzed throughout the spectrum of planned operating conditions in accordance with Regulatory Guide 1.49.
In addition, 2511 MHt is the licensed maximum steady-state power level of the units.
This maximum steady-state power level will never knowingly be exceeded.
Conservatism incorporated into the transient analysis is documented in References 1 and 2.
Transient analyses are initiated at the conditions given in these References.
The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by technical specifications.
The effects of scram worth, scram delay time, and rod insertion rate, all conservatively aoplied, are of greatest significance in the early portion of the negative reactivity insertion.
The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion. By the time the rods are 60% inserted, approximately 4 dollars of negative reactivity have been inserted, which strongly turns the transient and accomplishes the desired effect.
The times for 507,and 90%
insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.
The MCPR operating limit is, however, adjusted to account for the statistical variation of measured scram times as discussed in Reference 2 and the bases of Specification 3.5.K.
I Steady-state operation without forced recirculation will not be permitted except during startup testing.
The analysis to support operation at varicus power and flow relationships has considered operation with either one or two recirculation pumps.
The bases for indiviriual trip settings are discussed in the following paragraphs.
l For analyses of the thermal consequences of the transients, the MCPR's stated in l
Paragraph 3.5.K as the limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transients.
A.
Neutron Flux Trip Settings l
i 1.
APRM Flux Scram Trip Setting (Run Mode)
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state I
conditions, reads in percent of rated thermal power.
Because fission chambers provide the basis input signals, the APRM system responds directly to average neutron flux. During transients the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to j
the time constant of the fuel.
0926B 1.1/2.1-7 Amendment No.104 1
DPR. 30 1
140 APRM BACXt'P SCALM
- ~'- ~*-.~.~.-
-.-.-...~.-..e.~.-.....
120 ~. ~. -. - - ~ * ~
o' s'
APRM RCD BLOCK LINE (0.54WD
- 30)
(100,87 100,108)**
100 s'
e APRM SCRAM LINE (0.54WD + 62
,e m
Eu a
Qll W
80,
mo N
8 v
- NAft'RAL
[
u CIRCt1ATION j
/
NOMINAL, CONSTANT KENON LING 100/100 POVIR/TLOW LINE o
60m
/
p.,
g j
g Operating Region Supported s
r By N.E.D.0. - 26167 and y
M.E.D.C. 22192 0
h
- 0perating on Stagle Loop or Na,tural Circulation is 60,
Li/4ted Per Tech. Specs.
o 3.6.M.3 and 2.1.A.4 t>
201 PL'MF SPEED LINE
- 0peration at greater than rated core flow is supported by NEDC-31449,
20 RATED CONDITIONS POVIA 2311 MWth CORI FLOW 98 M1bs/RR 1
1 I
0 O
20 60 60 to 100 12C W CORE FLOW RATI (1 07 RATED) 7 TICL'R1 2.1-3 (SCHEMATIC)
Amendment No. 104 APRM FLOW IIAS SCRAM RELATICMSNIP TO NORMAL CPERATING CON 01710NS x.
,.,,,-,,,m.,
,y, _ -,,. _,.. _,. -. _, _... - -..., -.. _.. -,, _. - -....,, _, _. - -,. _. - _,,
QUAD-CI?!ES OPR-30 VABLE 3.3-3 l
INSTRUMENTATION THat INITIATES R00 BLC K MinissJm Nunber of Operable or Tripped Instrument Channels per Trio system 111 Instrument Trio Level Settina 2
APRM upscale (flow bias)I73 1(0.58WD + 50)
.FRP (2)
MFLPD 2
APRM upscale (Refuel and 112/125 full scale Startup/ Hot Standby mode) 2 APRM downscale(7) 13/125 full scale 1
Rod nitor upscale (flow 10.65WD + 43( 1(1 1 1
Rod block monitor downscaleI73 13/125 full scale 3
IRM downscale(3) (8) 13/125 full scale 3
IRM upscale (8) 1108/125 full scale 2(5)
$RM detector (
t in Startuo 12 feet below core centerline position IRMdetectorggtinstartup 12 feet below core centerline 3
position I J 5
2(5) (6)
$RM upscale 110 enyngsfs,e 2(5)
SRM downscale I93 110 counts /sec 2
1 (per bank) High water level in scram i 25 gallons (per bank) discharge volume (50V) 1 50V high water level scram NA trip bypassed 10111 1.
For the Startup/ Hot Standby and Run positions of the reactor mode selector switch there shall be two operable or tripped trip systems for each function except the SRM rod blocks. IRM upscale and IRN downscale need not be operable in th6 Run position. APRM downscale. APRM upscale (flow biased).
and RBM downscale need not be operable ir. the Startup/ Hot Standby mode. The RSM upscale need not be operable at less than 30% rated thermal power. One channel may be bypassed above 30% rated thermal power provided that a limiting control rod pattern does not exist. For systems with more than one channel per trip system. If the first colune cannot be met for one of the two trip systems. this condition may exist for up to 7 days provided that during that time the operable system is functionally tested innediately and daily thereafter; if this condition lasts longer than 7 days the system shall be tripped. If the first colune cannot be met for both trip systems, the systeres shall be tripped.
(
09258 3.2/4.2-14 Amenenent No. 104 l
l I
{
QUAO-C' TIES OPR-30
)
2.
W is the percent of drive flow required to produce a rated core flow of 98 million Ib/hr. Trip level D
setting is in percent of rated power (2511 *t).
3.
IRM downscale may be bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is 1100 cos.
5.
One of the four SRM inputs may be bypassed.
6.
This SRM function may be bypassed in the high IRM ranges (ranges 8, 9, and 10) when the IRM upscale rod block is operable.
7.
Not required to be operable when performing low power physics tests at atmospheric pressure during or after refueling at power levels not to ex,ceed 5 W t.
8.
This IRM function occurs when the reactor mode switch is in the Refuel or Startup/ Hot Standby position.
9.
This trip is bypassed when the SRM is fully inserted.
lb/hr and greater.
l
- 10. The trip level setting shall be a maximum of 108% for core flow equal to 98 x 100 i
t t
i 1
3.2/4.2-144 0154H/00682 knendment No.
104 t
OVAD-CIT 8ES DPR-30 sidered inoperable, fully provide reasonable assurance inserted into the core, and that proper control rod drive electrically disarmed.
performance is being maintained. The results of 5.
If the overall average of the measurements performed on the 20% insertion scram time data control rod drives shall be generated to date in the current submitted in the annual cycle exceeds 0.68 seconds, the l
operating report to the NRC.
MCPR operating limit must be modified as required by 5.
The cycle cumulative mean scram
$pecification 3.5.K.
time for 20% insertion will be determined innedtately f ollowing the testing required in Specifications 4.3.C.1 and 4.3.C.2 and the MCPR operating limit adjusted, if necessary, as required by Specification 2.5.K.
D.
Control Rod Accumulators At all reactor operating pressures, a
- Once a shift, check the status of the rod accumulator may be inoperable pressure and level alanns for each provided that no other control rod in accumulator.
the nine-rod square array around that rod has:
1.
An inoperable accumJ14 tor, 2.
A directional control valve electrically disarmed while in a nonfully inserted position, or 3.
A scram insertion greater than maximum permissible insertion time.
If a control rod with an inoperable accumulator is inserted full-in and its directional control valves are electrically disarmed. it shall not be considered to have an inoperable accumulator, and the rod block asso-c14ted with that inoperable accumu-lator may be bypassed.
E.
Reactivity Anomalies F.
Reactivity Anonulies The reactivity equivalent of the dif-During the startup test progrsm and ference between the actual criticai startups following refueling outages, rod configuration and the expected the critical rod configurations will configuration during power operation be compared to the expected configur-shall not exceed 11 0 k.
If this ations at selected operating condi-limit is exceeded, the reactor shall tions. These comparisons will be be shutdown untti the cause has been used as base data for reactivity determined and corrective actions monitoring during subsequent power have been taken. In accordance with operation throughout the fuel cycle.
Specification 6.6 the NRC shall be At specific power operating condi-notified of this reportable occur-tions, the critical rod configuration rence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
will be compared to the configuration expected based upon approprtately l
corrected past data. This comparison will be made at least every equiva-1ent full power month.
F.
Economic Generation Control System F.
Economic Generstion Control system Operation of the unit with the eco-Prior to entering EGC and once per nomic generation control system with shift while operating in EGC, the EGC automatic flow control shall bt per-operating paraagters will be reviewed missible only in the range of 65% to for acceptability.
100% of rateJ core flow, with reactor power above 20%.
09258 3.3/4.3-5 Amendment No.104 l
QUAD-C3 TIES DpR-30 provided that during such 7 days operable immediately. The RCIC all active components of the system shall be demonstrated to automatic pressure relief be operable daily thereafter.
+
rubsystems, the core spray Daily demonstration of the subsystems, LPCI mode of the RHR automatic pressure relief system, and the RCIC system are subsystem operability is not
- operable, required provided that two feedwater pumps are operating at levels above 300 MWeg and one 3.
If the recjairements of feedwater pump is operating as specification 3.5.C cannot be normally required with one met, an orderly shutdown shall additional feedwater pump be initiated, and the reactor operable at power levels less pressure shall be reduced to 90 than 300 MWe.
pois within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Automatic pressure Relief Subsystams D.
Automatic Pressure Relief Subsystems.
Surveillance of the automatic 1.
The automatic pressure relief pressure relief subsystem shall be rubsystam shall be operable performed as follows:
whenever the reactor pressure is greater than 90 psis, irradiated 1.
The following surveillance shall fuel is in the reactor vessel be caeriod out on a six-month and prior to reactor startup surveillance interval:
from a cold condition.
e.
With the reactor at pressure each relief valve shall be 2.
Fromandafterthedatethattwoj manually opened. Relief of the five relief valves of the valve opening shall be automatic pressure relief verified by a compensating subsystem are made or found to l
turbine bypass valve or be inoperable when the reactor control valve closure.
is pressurized above 90 pois l
with irradiated fuel in the reactor vessel, reactor 2.
A logic system functional test operation is permissible only shall be performed each during the succeeding 7 days refueling outage.
unless repairs are made and provided that during such time the KPCI subsystem is operable.
3.
A simulated automatic initiation wnich opens all pilot valves shall be performed each 3.
If the requirements of Specifi-refueling outage.
cation 3.5.D cannot be met, an orderly shutdown shall be initi-ated and the reactor pressure 4.
When it is determined that two shall be reduced to 90 psis relief valves of the automatic within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
pressure relief subsystem are inoperable, the KPCI shall be demonstrated to be operable immediately.
0925B 3.5/4.5-5 Amendment No.104 s
e
QUAD-Cf7tES OPR-30 within the prescribed limits 1
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. Maximum allowable LHGR is 13.4 kw/ft, for fuel types P8xCR and BP8x8R. For fuel types GE8x8E and GE8x8E8 the maximum allowable LHGR is 14.4 kw/ft.
K.
Minimum Critical Power Ratio (MCPR)
K.
Minimum Critical Power Ratio (MCPR)
During steady-state operation at The M'CPR shall be determined daily durinq rated core flow. MCPR shall be steady-state power operation above 25% of greater than or equal :o:
' rated thermal power.
1.30 for T yg i 0.68 sec A
1.35 for TAVE 1 0.86 sec 0.278 TAVE + l'III for 0.68 see i TAVE i.86 sec where Tayg =
nuan 20% scrun insertion time for all surveillance data from specification 4.3.C which has been generated in the current cycle.
For core flows other than rated.
these noninal values of MCPR shall be increased by a factor of kg wnere kr is as shown in Figure 3.5.2.
If I
any time during operation it is determined by normal surveillance that the limiting value for MCPR is l
l being exceeded action shall be inittated within 15 minutes to restore operation to within the prescribed limits. If the steady-state MCPR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to the cold shutdown condition within l
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and j
corresponding action shall continue I
until reactor operation is within the l
prescribed limits.
l Amendment No.104 09258 3.5/4.5-10 I
l.
QUAD-CITIES DPR-30
\\
Based en the fact that when one lo p of the containment co311ng mode of the RHR system becomes inoperable, only one system remains, which is tested daily, a 7-day repair period was specified.
C.
High-Pressure Coolant Injection The high-pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.
The HPCI meets this requirement without the use of offsite electrical powar. For the pipe breaks for which the HPCI is intended to function, the core never uncovers and is continuously cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3).
The repair times for the limiting conditions of operation were set considering the use of the HPCI as part of ths isolation cooling system.
D.
Automatic Pressure Relief The relief valves of the automatic pressure reitef subsystems are a backup to the HPCI subsystem. They enable the core spray subsystem and l
LPCI mode of the RHR system to provide protection against the small pipe break in the event of HPCI failure by depressurizing the reactor vessel rapidly enough to actuate the core spray subsystems and LPCI l
mode of the RHR system. The core spray subsystem and/or the LPCI mode of the RHR system provide sufficient flow of coolant to limit fuel cladding temperatures to less than 2200'F to assure that core geometry remains intact, to limit the core widt clad metal-water reaction to less than 1%, and to limit the calculated local metal-water reaction to less than 17%.
Analyses have shown that only four of the five valves in tne automatic depressurization system are required to operate. Loss of one of the reitef valves does not significantly affect the pressure relieving capability, therefore continued operation is acceptable. Loss of two relief valves significantly reduces the pressure relief capability of the ADS: thus, a 7 day repair period is specified with the HPCI available, and a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> repair period with the HPCI unavailable.
E.
RCIC The RCIC system is providjd to supply continuous makeup water to the reactor core when the reactor is isolated from the turbine and when the feedwater system is not available. Under these conditions the pumping capacity of the RCIC system is sufficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts. The system may also be manually initiated at any time.
The HPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable.
Therefore, the specification calls for an operability check of the HPCI system should the RCIC system be found to be inoperable.
F.
Emergency Cooling Availability The purpose of Specification 3.5.F is to assure a minte m of core cooling equipment is available at all times.
If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only twc RHR pumps would be available.
Likewise, if two RHR pumps were out of service and two RHR service water pumps on the opposite side were also out of service no containment cooling would be available. It is during the refueling outages that major maintenance is performed and during such time that all low-pressure core coC ing systems may be e t of service. This specification provides that should this occur, no work will be performed on the primary system which eculd lead to draining the vessel. This work would include werk on certain control rod drive components and recirculation systems. Thus, the specification precludes the events which could require core cooling. Specification 3.9 m st also be consulted to determine other requirements for the diesel generators.
Quad-Cities Units 1 and 2 share certain process systems such as the makeup dominera112ers and the radwaste system and also some safety systems such as the staldby gas treatment system, batteries, andAmenenent No.104 09258 3.5/4.5-12
QUAD-Cl?!ES OPR-30 shown on Figure 3.5-1 as limits because conformance calculations have not been performed to justify operation at LHGR's in excess of those shown.
J.
Local LHGR This specification assures that the maximum linear heat-generation rate in any rod is 1(ss than the design linear heat-generation rate even if fuel pellet densification is postulated. The power spike penalty is discussed in Reference 2 and assumes a linearly increasing variation in axial gaps between core bottom and top and assures with 951 confidence that no more than one fuel rod exceeds the design LHGR due to power spiking. No penalty is required in Specification 3.5.L because it has been accounted for in the reload transient analyses by increasing the calculated peak LHGR by 2.2%.
K.
Hint'num Critical Power Ratio (MCPR)
The steady state values for MCPR specified in this specification were selected to provide margin to accontnodate transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself. These values also assure that operation will be such that the initial condition assumed for the LOCA analysis plus two percent for uncertainty is satisfied. For any of the special set of transients or disturbances caused by single operator error or single equipment malfunction. it is required that design analyses initialized at this steady-state operating Itmit yield a MCPR of not less than that specified in Specification 1.1.A at any time during the transient assuming instrument trip settings given in Specification 2.1.
For analysis of the thermal consequences of these transients the value of MCPR stated in this specification for the ilmiting condition of operation bounds the initial value of MCPR assumed to exist prior to the initiation of the transients. This initial condition, which is used in the transient analyses, w111 preclude violation of the fuel cladding integrity safety limit.
Assumptions and methods used in calculating the required steady state MCPR limit for each reload cycle are documented in References 2 and 4.
l The results apply with increased conservatism while operating with MCPR's greater than specified.
The most limiting transients with respect to MCPR are generally:
a)
Rod withdrawal error b)
Load rejection or turbine trip without bypass c)
Loss of feedwater heater Several factors influence which of these transients results in the largest reduction in critical power ratio such as the specific fuel loading, exposure, and fuel type. The currant cycle's reload licensing analyses specifies the limiting transients for a given exposure increment for each fuel. type. The values specified as the Limiting Condition of Operation are conservatively chosen to bound the most restrictive over the entire cycle for each fuel type.
The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the 00YN computer code for analyzing rapid pressur1Zation events. Generic statistical analyses were performed for plant groupings of similar design which considered the statistical variation in several parameters (initial power level. CR0 scram insertion time, and model uncertainty). These analyses (which are described further in Reference
[
- 4) produced generic Statistical Adjustment Factors which have been applied to plant and cycle specific ODYN results to yield operating limits which provide a 95% probability with 95% confidence r%at the limiting pressurization event will not cause MCPR to fall be.ow the fuel cladding integrity safety limit.
i l
09258 3.5/4.5-14 Amendment No.104
QUAD-CIT 8ES OPR-30 As a result of this 95/95 approach, the average 20% insertion scram time must be monitored to assure compliance with the assumed statistical distribution.
If the mean value on a cycle cumulative (running average) basis were to exceed a 5% significance level compared to the distribution assumed in the OCYN statistical analyses, the MCPR limit must be increased linearly (as a function of the mean 20% scram time) to a more conservative value which reflects an NRC determined uncertainty penalty of 4.4%.
This penalty is applied to the plant specific 00YN results (i.e. without statistical adjustment) for the limiting single failure pressurization event occurring at the limiting point in the cycle.
It is not applied in full until the mean of all current cycle 20% scram times reaches the 0.90 secs value of Specification 3.3.C.I.
In practice, however, the requirements of 3.3.C.1 would most likely be reached (i.e. Individual data set average
>.90 secs) and the regtrired actions taken (3.3.C.2) well before the running average ex:eeds 0.90 secs.
The 5% significance level is defined in Reference 4 as:
n TB - p + 1.65 (Nj/ { Nj)l/2 0 1-1 where:
Hean value for statistical scram time distribution to 20%
p inserted standard deviation of above distribution o
number of rods tested at BOC (All operable rods)
N1 n
total number of operable rods tested in the current cycle I Nj 1-1 The value for 1'8 used in Specification 3.5.k is 0.68 secs which is j
conservative for the following reasons:
a)
For simplicity in formulating and implementing the LCO, a n
conservative value for I Ng of 70a (i.e. 4x177) was used.
1-1 This repre:ents one full core data set at BOC plus 6 half core data sets. At the maximum frequency allowed by Specification 4.3.C.2 (16 week intervals) this is equivalent to 24 operating months.
That is, a cycle length was assumed which is longer than any past or contemplated refueling interval and the number of rods tested was maximized in order to simplify and conservatively reduce the criteria for the scram time at which MCPR penalization is necessary, b)
The values of p and a were also chosen conservatively based on the dropout of the position 39 RPIS switch, sirice pos. 38.4 is the precise point at which 20% insertion is reached. As a result Specification 3.5.k initiates the linear MCPR penalty at a slightly lower value'f aye.
This also produces the full 4.4%
penalty at 0.86 secs which would occur sooner than the required value of 0.90 secs. 09268 3.5/4.5-14a Amendment No. 104
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"Quad Cities Nuclear Power Otati % Unit..) 1 & 2 1'sGER/GT'CR-Lf.yA Loss of s
l CoolantAccidentAnalysis"MIK-313W(.*
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3 I
T 6(
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2.
Generic Reload Fuel Applicat hn." NEDE;-24011 4-A*r
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'I. p. Jacobs and P. W. Marriott, GE Topical Report A ED',;i36, 'ff,uidelines 3.
for,0stermining Safe Test Intervals and Repair Tl.ned fbe Br4 neered 1
Safeguards," April, 1969.
s 4.
Qualification of the One-Dimen'sional Core Tra.nsien's. Model for Boiling Water i
Reactors" Gen?ral Electric Co. Licensing Topical Report NEDO 2415/ Vols. I IJ and II and NEDA,-24154 Vol. III as supplemented by letter dated Seitember 5, 1980 from R.H. Buchholz (GE) to P. S. Check (NRC).
t Approved revisibt at time of phnt bjyrtion.
'N
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QUAD CITIES DPR-30 3.
Prior to Single Loop Operation for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the following l
restrictions are required:
a.
The MCPR Safety Limit shall be increased by 0.01.
(T.S. 1.1A);
b.
The MCPR Operating Limit shall be increased by 0.01 (T.S.*
3.5.K);
I c.
The flow biased APRM Scram and a
Rod Block Setpoints shall be reduced by 3.5% to read as follows:
T.S. 2.1.A.1; S 1 58WD + 58.5 T.S. 2.1.A.1;*
S s (.58WD + 58.5) FRP/MFLPD T.S. 2.1.B; S s.58WD t 46.3 T.S. 2.1.B;*
S s (.58WD + 46.5) FRP/MFLPD T.S. 3.2.C (Table 3.2-3);*
APRM upscale 1 (.08WD + 46.5)
FRP/MFLPD In the event that NFLPI, ex.ceeds FRP.
d.
The flow biased FBM Rod Block l
l setpoints shall *>e reduced by 4.0% to read as follows:
T.S. 3.7.C (Table 3.2-3); RBM Upsca16 1 65WD + 39 l
s.
The svetion valve in the idle l
loop sball be closed and electrically isolated except when the idle loop !A being prepared for return to service.
l 0925B 3.6/4.6-5a Amendment No. 104
QUAD-CITIES OPR-30 The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 15%.
Below 80%
power, the loop select logic would not be expected to function at a speed differential of 20%.
This specification provides a margin of 5% in pump speed differential before a problem could arise.
If the reactor is operating on one pump, the loop select logic trips that pump before making the loop selection.
Analyses have been performed which support indefinite single loop operation provided the appropriate restrictions are implemented within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The MCPR l Safety Limit has been increased by 0.01 to account for core flow and TIP reading uncertainties which are used in the statistical analysis of the safety limit.
The MCPR Operating Limit has also been increased by 0.01 to maintain the same margin to the safety limit as during Dual Loop operation.
The flow biased scram and rod block setpoints are reduced to account for uncertainties associated with backflow through the idle jet pumps when the operating recirculation pump is above 20 - 40% of rated speed.
This assures that the flow biased trips and blocks occur at conservative neutron flux levels for a given core flow.
The closure of the suction valve in the idle loop prevents the loss of LPCI flow through the idle recirculation pump into the downcomer.
l l
l l
l 09268 3.6/4.6-13a Amendment No.104