ML19282C211
| ML19282C211 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/23/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19282C212 | List: |
| References | |
| NUDOCS 7903220229 | |
| Download: ML19282C211 (27) | |
Text
.
O e,8'"
UNITED STATES j*.
NUCLEAR REGULATORY COMM!sslON
[ Yt- '? (d[ $
WASHINGTo N, D. C. 2C555 a-e t.~..... e CCMONWEALTH EDISON COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC CCMPANY DOCKET NO. 50-254 00AD CITIES UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. DPR-29 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated November 20,1978, as supplemented December 15, 1978, and February 14, 1979, complies with the standards and requirements 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; 8.
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 (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 Commission's regulatiens; 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 Commission's regulations and all applicable requirements have been satisfied.
I 7 9 03220 AA,9
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraphs 3.B and 3.C of Facility License No.
OPR-29 are hereby amended to read as follows:
3.B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 50, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.C.
Restrictions Reactor power level shall be limited to maintain pressure margin to the safety valve set points during the worst case pressurization transient.
The magnitude of the power limitation, if any, and. the point in the cycle at which it shall be applied is specified in the Reload No. 4 licensing submittal for Quad Cities Unit No.1 (NEDO 24145).
Plant operation shall be limited to the operating plan described therein. Subsequent operation in the coastdown mode is permitted to 70% power.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION.
f h P-l Thomas Ar.,/Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Speci fica tions Date of Issaance:
February 23, 1979
l ATTACHMENT' TO LICENSE AMENDMENT f'O. 59 FACILITY OPERATING LICENSE NO. DPR-29 i
DOCKET NO. 50-254' Replace the following pages of the Appendix "A" Technical Specifications i
with the enclosed pages. The revised pages are identified by the cap-l tioned Amendment number and contain vertical lines indicating the area of change.
i Remove Insert i
1.1/2.1 -2 1.1/ 2.1 -2 1.1/2.1-4 1.1/2.1-4 l.1/2.1 -5 1.1/2.1-5 1.1/2.1 -8 1.1/ 2.1 -8 1.1/ 2.1 -9 1.1/2.1-9 1.2/2.2 1.2/2.2-1 3.1/4.1-11 3.1/4.1-11 3.2/4.2-5 3.2/4.2-5 3.2/4.2-Sa 3.2/4.2-6 3.2/4.2-6 3.2/4.2-7 3.2/4.2-7 3.2/4.2-8 3.2/4.2-8 3.2/4.2-11 3.2/4.2-11 3.2/4.2-12 3.2/4.2-12 3.2/4.2-14 3.2/4.2-14 3.3/4.3-3 3.3/4.3-3 3.3/4.3-4
- 3. 3/4. 3-4 3.3/4.3-9 3.3/4.3-9 3.3/4.3-10 3.3/4.3-10 3.5/4.5-10 3.5/4.5-10 3.5/4.5-11 3.5/4.5-11 3.5/4.5-12 3.5/4.5-12 3.5/4.5-14 3.5/4.5-14 3.5/4.5-15 3.5/4.5-15 Add page 3.2/4.2-Sa.
=
9
aun c:e - oe:::te - sas :u:
QUAD-CITIES DPR-29 D.
Reactor Water leel (Shutdown Condition) curve in Figure 2.1-2. at which point the actual peaking factor vale: snall be When:ver the reactor is in the shutdown condt-used.
tion with irradiated fuel in the reactor vessel, the water level shall not be less than that corre-LTPF = 3.06 (7 x 7 fuel assembiies) sponding to 12 inches above the top of the 3.03 (8 x 8 fuel assemblies) active fuel
- when it is seated in the 3.00 (8x8R fuel assemblies) l core.
_ Mode)
When the reactor mode switch is in the Refuel or Startup Hot 5:andby posi-tion. the. APRM scram shall be set at less than or equal to 15Fc of rated neutron flux.
3.
IRM Flux Scrarn Trip Se: ting The IRM flux scram set::n shall be set at less than or equal to 120/125 of full scale.
4.
When the reactor mode sw::ch is in the s:artup or run position, :he reactor shall not be operated in the na: ural circula.
tion flow mode.
B.
APRM Rod Block Setting The APRM rod block setting shall be as shown in Figure 2.1-1 and shall be:
S s (.65W + 43)(LTPF/TPF)
The definitions used above for the APRM scram trip apply.
C. Reactor low water level scram setting shal1be 2 143 inches above th: top of the active fuel at normal operating conditions.
D. Reactor low water level ECCS initiation shall be 83 inches ( + 4 inches /-0 inch ) above the top of the active fuel at normal operating conditions.
E. Turbine stop valve scram shall be s 10% valve closure from full open.
F. Turbine control va!ve fast c!csure scram shall initiate upon ac: at:an of the fas: c:csar: scie-noid valves which trip th: tur:;ne control
- Top of active fuel is valves.
defined to be 360 inches above vessel zero (see G. Main steamline isolation valve c!:sure scram Bases 3.2).
shall be s 10% valve closure from fu 1 open.
H. Main steamline low-pr:ssure in::ia: ion of main st:amline isola: ion vah; cicsure shall be 2 850 psig.
1.1/ 2.1 -2
QUAD-CITIES DPR-29 1.1 SAFETY LIMIT LASES "D.: fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal opera.tional transient. E:cause fuel damage is not directly observab::. a st:p-back approach is us:d to establish a safety limit such that the minimum critical power ratio (MCPR)is no 1:ss than 1.07MCPR > 1.07 represents l a conservative margin relative to the conditions required to maintain fuel claddmg integrity.
l l
The fuel cladding is one of the physical barricts which separate radioactive materials from the environs. The i
integrity of this cladding barrier it related to its relative freedom from perforations or cracking. Although some corrosion or us:-related cracking may occur during the life of the c: adding, fusion product migration from this source is incrementally cumulative and continuously measurable. Fu:1 cladding perforations. however can result from thermal stresses which occur from reactor operation signi5cantly above design conditions and the protection system safety s:ttings. While fission product migration from claddirrg perforation isjust as measurable as that from use related cracking, the thermally caused cladding perforauons signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding detericration. Therefore, the fuel cladding safety limit is dermed with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0).
These conditions represent a signific:nt departure from the condition' intended by design for planned operation.
A.
Reactor Pressure > 800 psig and Core Flow > 10% of Rated Onset of transition boiling results in a decrease in heat transfer from the cladding and therefore elevated cladding temperature 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 boiling 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 power ratio (CPR). which is the ratio of 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 the core is the minimum critical power ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables (Figure 2.13).
The safety limit (MCPR of 1.07 has sufficient conservatism to assure that in the event of an abnormal l operational trar.sient initiated from the normal operating condition, more than 99.9% of the fuel rods in
' the core are expected to avoid boiling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the safety limit.1.07as derived from a detailed statistical analysis considering all of the l uncertainties in monitoring the core operating state, including uncertainty in the boiling transition correlation (see e.g. Reference I). 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 1.07would not produce boiling transition.
However, if boiling transition were to occur. cladding p:rforation would not be expected. Cladding temperatures would increase to approximately 1100
- F. which is below the perforation temperature of the cladding material.This has 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 time (30 minutes) without cladding perforation.
If reactor pressure should ever exceed 1400 psia during normal power operation (tne limit of j
applicability of the boiling transition correlation) it would be assumed that the fuel cladding int:grity safe:y limit has been violated.
I In addition to the boihng transition limit (MCPR).operatten is censtrained to a maximum LHGR: 17.5 f
kw/ft for 7 x 7 fuel and 13A kw/ft for 8 x 8 fuel.This constraint is established by Specincations 2.1.A.1 and 3.5.1. Specification 2.1.A.I established limiting total peaking fac: ors (LTPF) *hich constrain LHGR's to the maximum values at 100G power and estab:ished proc:dures for adjusting APRM scram 1.1/ 2.1-4
QUAD-CITIES DPR-29 5:: tings which maintain equivalent safety margins when the total peak factor (TPF) exc::ds the LTPF.
Specification 3.5J established the LHGR maximum wiich cannot be exc::ded under s't:ady power cperation.
B.
Core hermal Power 1.imir (Reactor Pressure <S00 psia)
At pressur:s below S00 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi.
At low powers and flows this pressure differentialis maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevatiot head the core pressure drop at low powers and flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 10'lb/hr bundle flow, bundle pressure drop is nearly independent oibundle power and has a value of 3.5 psi.Thus the bundle flow with a 4.56. psi driving head will be greater than 28 x 102 lb/hr. Full sdale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicare that th: fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of rated thermal power, the peak powered bund!: would have to be operating at 3.86 times the average powered bundle in order to achi:ve th_is bundle power. Thus, a core thermal power limit of 25% for reactor pressures below 300 psia is conservative.
C.
Power Transient During transient operation the heat flux (thermal power.to-water) would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel, which is 8 to 9 seconds. Also, the limiting safety system scram settings are at values which will not allow the reactor to be operated above the safety limit during normal operation or during other plar.t operating situations which have been analyzed in detail.
In addition, control rod scrams are such that for normal operating transients, the neutron flux transient is terminat:d before a significant increase in surface heat flux occurs. Scram times of each control rod are checked each refueling outage, and at least every 32 weeks,50% are checked to assure adequate inseration times. Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to less than the scram setting within 1.5 seconds does not necessarily imply that fuel is damaged: however, for this specification, a safety limit violation will be assumed any time a ceutron flux scram setting is exceeded for loager than 1.5 seconds.
If the scram occurs such that the neutron flux dwell time above thelimiting safety system setting is less than 1.7 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected.These analyses show that even if the bypass system fails to operate, the design limit of MCPR = 1.07 is not exceeded. Thus, use of a 1.5-second limit l
provides additional margin.
De computer provided has a sequence annunciation program whi:h willindicate the sequence in which scrams occur, such as neutron tiux pressure, etc. This program also indicates when the scram setpoint.is c!:ared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be availab!:
for analyzing scrams: however, if the computer information should not be availab!: for any scram analysis, Specification 1.1.C.2 will be relied on to determin: if a safety limit has been violated.
During p.eriods when the reactor is shut down, consideration must also ~b: given to water 1: vel requirements due to the effect of decay heat. If reactor water level should drep b: low the top of the active fu:1 during this time, the ability to cool the cere is redue:d. This :ducnon in core-cooling capability could 1:ad to c!:vated cladding temperatures and cladding ; :rioranen. The core wi!! b: coe':d sufic:ently to prevent cladding melting should the wat:r level be redue:d to r o-thirds the core height. Establish-ment of the safety limit at 12 inches above the top of the fuel
- provides adequate margin. This level will he continuously monitored whenever the recirculation pumps are not operating.
- Top of active fuel is defined.to be 360 inches above vessel zero (see Bases 3.2).
1.1/ 2.1-5
QUAD-CITIES DPR-29 3
An in:rease in the APRM scram.-;n setting wou!d decrease the margi.. present before the fuel cladding integrity _ safety limit is reached. The APRM scram trip setting was cetermined by an analysis of margins required to provide a reasonable range fer maneuvering during operation.
i Reducing this operating margm would increase the Trequency of spurious scrams. which have an adverse effect on reactor safety because of the resulting thermai stresses. Thus, the APRM 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.
The scram trip setting must be adjusted to ensure that the !.HGR transient peak is not increased for any combination ofTPF and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.I. when the maximum total peaking factor is greater than the limiting total peaking factor.
I
For operation in the Startup mode while the reactor is at low pressure, the APRM scram setting of 15% of rated power provides adequate thermal margin between the setpoint and the safety limit. 25?o of rated. The margin is adequate to accommodate anticipated maneuvers associated with oower plant startup. Effects ofincreasing pressure at zero or low void content are minor, cold water ft, m sources available during startup is not much colder than that already in the system, temperature cofficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Of all possible sources of reactivity input. uniform contrc, rod withdrawal is the most probable cause of significant power rise. Because the flux distribuuun 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 $?o of rated power per minute, and the APRM system would be more than adequate to assure a scram I
before the power could exceed the safety limit. The 15fo APRM scram remains active until the mode switch is placed in the Run position. This switch occurs when reactor pressure is greater than 850 psig.
- 3. IRM Flux Scram Trip Setting The IRM system consists of eight chambers, four in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being one half a decade in size.
The IRM scram trip setting of 120 divisions is active in each range of the IRM. For example,if the instrument were on Range 1. the scram setting would be 120 divisions for that range; likewise,if the instrument were on Range S. the scram would be 120 divisions on that range. Thus, as the IRM is ranged itp to accommodate the increase in power level. the scram trip setting is also ranged up.
The most significant sources of reactivity change during the power increase are due to control rod withdrawal. In order to ensure that the IRM provides adequate prote: tion against the single rod
{
withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting i
the accident at various power levels. The most severe case involves an initial condition in which the I
reactor is just subcritical and the IRM system is not yet on scale.
Additional consavatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is hypassed. The results of this analysis show that the reactor :s scran med and peak power limited to ICc of rated power. thus matntainmg MCPR above LC7. Based on the above f
analysis, the IRM provides protection against local con::c! rod withdrawai errers and continous withdrawal of control rods in sequence and provides backup protection for tne APRM.
~
l 1.1/ 2.1-8 f
i
QUAD-CITIES DPR-29 i
E.
APRM Rod Block Trip Setting Reactor pow:: level may be varied by moving control rods or by varying the recir:ulation Sow rate. The APRM system provides a control rod block to prevent roc withdrawal beyond a g:ven point at constant j
recirculation flow rate to prot:ct against the condition of an MCPR less than 1.C7.This rod block trip j
setting. which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor 1
power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage. assuming a steady-state operation at the trip setting, over the entire i
I rectreulation flow range. The margin to the safety limit increases as the flow decreases for the specified j
trip setting versus flow relationship; therefore the worst-case MCPR which could occur during
[
steady-state operation is at 1089 of rated thermal power because of the APRM rod block trip setting.The l
actual power distribution in the core is established by specified control rod sequences ahd is monitored I
continuously by the incore LPRM system. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if th maximum total peaking factor exceeds the limiting total peaking factor, thus preserving the APRM rod block safety margin.
C.
Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained. The scram setpoint is based on normal operating temperature and 8
pressure conditions because the level instrumentation is d:nsity compensated.
t' D. Reactor Iew Low Water Level ECCS Initiation Trip Point 4
I l
The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss.of-coolant accident and to limit fuel cladding temperature to well l
below the cladding melting temperature to assure that core geometry remains intact and to limit any cladding metal-water reaction to less than Ire. To accomplish their intended function, the capacity of each emergency core cooling system component was established based on the reactor low water level scram setpoint. To lower the setpoint of the low water level scram would increase the capacity requirement for each of the ECCS components. Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will not be set lower because of ECCS capacity requirements.
The design of the ECCS components to meet the a' ove criteria was dependent on three previously set o
parameters: the maximum break size the low wat:r level scram setpoint, and the ECCS initiation serpoint. To lower the setpoint for initiation of the ECCS could lead to a loss of efr :tive core cooling.To e
raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.
i l
E.
Turbine Stop Valve Scram The turbine stop valve closure scram trip anticipates the pressure. neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 10cc of valve closure from full open, the resultant increase in surface heat eux is limited such that MCPR rerhains above 1.07 i
even during the worst-case transient that assumes the turbine bypass is closed.
4 F.
Turbine Control Valse Fast Closure Scram The turbine contrel valve fast closure scram is provided to anticipate th: rapid iner:ase in pressure and neutron '.ux r:sulting from fast closure of the turbine contrcl valves due to 2 Jead rej:: tion and
~
subsequent failure of the bypass. i.e. it prevents MCPR from benming less than 1.0 7 for this transiedt.
For the load rejection from 100"o power. the LHGR incr:ases to only 106.5"c ofits rat:d value, which results in only a small d::::ase in MCPR.
' 1.1/ 2.1 -9
QUAD-CITIES t
i DPR-29 I
t I
i 1.2/2.1 REACTOR COOLA::T SYSTEM l
k SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING Applicability:
Applicability:
Applies to limits on reactor coolant system Applies to trip settings of the instruments and pressure.
devices wh:ch are proviced to prev;nt the reactor system safety limits from being exceeded.
Objectise:
Objective:
To establish a limit be ow w hich the integrity of the To d 5ne the level of the process variables at w hich reactor coolant system is not threatened due to an automatic protective action is initiated to prevent overpressure condition.
the safety limits from being exceeded.
SPECIFICATIONS e
A.
The reactor coolant system pressure shall not A.
Reactor coolant high pressure scram shall be exceed 1325 psig at any time when irradiated s1060 psig.
I fuelis present in the reactor vessel.
B.
Priman system safety valve nominal settings shall be as follows:
I valve at !!15 psig'"
2 valves at 124u psig 2 valves at 1250 psig 4 valves at 1260 psig
"' Target Rock combination safety / relief valve The a!!awable setpoint error for each valve shall be i1%
l r
i
~
t
{
l.2 / 2.2 - 1 i
QUAD-CITIES DPR-29 TABLE 3.1-4 NOTES FCR TASLES 3.11.3.12. AND 3.14 1.
There sha!! be two cperate tre systems er ere c;erable and cne tr;;ed syste-f:r ea:n h:nctett 2.
If the frst celumn cann:t be rnet for one of the tre systems, that tre system shall be trc:ed. If the frst celumn carrct be met fer both try systems, the a;;r:criate actons I:ste becw snail be taken:
A. Inrtiate inse::cn of cperab' r:ds and com:!ete estrica of all caerab!e rods a: n:n a h urs.
e
- 8. Red' e power level to IRM range and p!a:e mode swit:h n the Startu;/ Hot Standby ::s: ton witnn x
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C. Reduce turbre lead and cbse main steam!ce isc!aten valves withm 8 hcurs.
3.
An APRM wi!! be constered in:perable d there are fewer than 2 LPRM r;uts per level:t t ere are yss than 50% of the ncrrnal ecmplement of LPPMs to an A?RM.
4.
Permestie to twass, w:th centrei rod bicek fcr rea: tor protecten system reset n refuel and snut::en p0s: tons of the Tea:ter mode switch.
5.
Not recured to be cperable when prrnary contamment integrity is r:ct re;ured.
6.
Re design perm:ts closure of any one hne w:thcut a s::am being inttiated.
7.
Aut:matically bypassed when reactor pressure is*<1060 psig.
8.
The + Stich tre peint is the water 'avel as measured by tr:e estrumentation outste the shreud. The water levelinste the shroud will decrease as power is increased to 100% n c: par:sca to the ' vel e
cutste the shrod to a carrnum of 7 oches. Th:s is cue to the pressure d cp a::ss tne steam dryer.
Therefore, at 100% power, an edication of + 8. inch water level will actually be,I c:n r:sce the j
shroud 1 inch on the water level instrumentation is >500' above vessel zero.
(See Bases 3.2).
9.
Permeste to typass when frst stage turbre pressure e less than that which c:rres;o.:s to 45%
rated sLarn flow. (<400 ps0
- 10. Tres upon a:tuaten of the fast <bsure solened when tres the turbr.e centrol vaves.
1: The APRM downs: ale trg function is automatca!!y bypassed wnen - RM estrumentaten s c: era::e and not h gh.
- 12. Channei shared by the rea: tor prctecten ind contarment sclaten system.
O 3.1/4.1-11
QUAD-CITIES DPR-29 3.2 LIMITING CONDITIONS FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reacter scram. protective instrumentation has
~
been proviced which initiates action to mitigate the consequences of ac:idents which at beyond.the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specincations j
provides the limiting conditions of operation for the primary system isoladon function, initiation of the emergency core cooling system, control rod block, and standby gas treatment systems. The objectives of the specincations are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such sptems are i
f out of service for maintenance, and (2) to prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for briefintervals to conduct required functional tests and calibrations. Some of the settings on the instrumentation that initiates or controls core and containment cooling hav'e tolerances explicitly stated wher: the high and low values are both critical and may have a substantial effect on safety. It should be noted that the serpoints of other instrumentation. where only the high or low end of the setting has a direct bearing on safety. are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
~
Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of. coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation
~
of these va!ves is initiated by the protective instrumentation which senses the conditions for which isolation is required (this instrumentation is shown in Table 3.2 1 ). Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an ac:ident.
4 The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus the discussion given in the bases for Specincation 3.1 is applicable here.
The low-reactor water level instru=entation is set to trip at >8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel zero) af ter allowing.for the full power pressure drop across the steam dryer the low level trip is at 504 inches above vessel zero, or 144 inches above top of active fuel. Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs, however, present trip setpoints were used in the LOCA analysis (NEDO 24146). This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps (ref arence SAR, Section 7.7.2).
For a trip setting of 504 inches above vessel zero and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximu= break, the setting is, therefore, adequate.
The low-low reactor level instrumentation is set to trip when reactor water level is 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero, -390 is 84 inches above the top of active fuel).
This trip j
initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems starts the emergency diesel generator, and trips the recirculation pumps. This trip setting level was chosen to be high enough to prevent spurious operation but low enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding l
will occur and so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded. For the complete circumferential break of a l
2S-inch recirculation line and with the trip settin:; given above, ECCS initiation and primary system isolation are initiated and in time so meet the above criteria (reference SAR Secticas 6.2.7.1 and 14.2.4.2).
The instrumentation elso covers the full spectrum of breaks and meets the above criteria (reference SAR Sections 6.2.7.1).
3.2/4.2-3
QUIS-CITIES DPR-29 The high-drvwell pressure instrumentation is a backup to the water leve! instrumentation and. in addition initiating ECCS,it causes isolation of Group 2 isolation valves. For the breaks discussed above. this instrument tion will initiate ECCS operatton at about the same time as the low low water levelinstrumentation; thus the res given above are applicable here also. Group 2 isolation val' es include the d.well vent. purge. and su valves. High-drywell pressure activates only these valves because high dryw e!! pressure could occur as of non safety related causes such as not purging the dowell air during startup. Total sys:em isolation is not desirable for these conditions, and only the valves in Group 2 are requeed to close. The low low water level instrumentation initiates protection for the full spectrum ofloss-of coolant accidents and causes a trip of Group primary system isolanon valves.
4 e
3.2/4.2-5a
QUAD-CITIES DPR-29 Venturi tubes are provided in the main steamlinc. n a means of measuring steam now and also limutng the loss of mass inventory from the vessel during a steamline break accident. In addition to menitoring steam 6ow, instrumentation is provided which causes a trip of Group I tsolation,alves. The primary function of the instrumentation is to detect a break in the main steataline, t.ius only Group i valves are closed For the worst-case accident, main steamiine break outside the drpell. this trip setting cf 120"c of rated steam Bow,in conjunction with the flow limiters and main steamline valve closure, limits the mass inventory loss sect that fuel is not uncovered, fuel temperatures remain less than !$00' F, and release of radioactisity to the environs is well below l
10 CFR 100 guidelines (reference SAR Sections 14.2.3.9 and 14.2.3.10).
i Temperatur: monitoring instrumenta. ion is provided in the main steamline tunnel to detect leaks in this area.
l Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves. Its j
setting of 200
- F is low enough to detect leaks o f the order of 5 to 10 gpm; thus it is capable orcovenng the entire spectrum of breaks. For large breaks. it is a %kup to high-steam flow instrumentation discussed above, and for small breaks with the resulting small release v radioactivity, gives isolation before the guidelines of 10 CFR 100 are exceeded.
I
.High-radiation monitors in the main steamline tunnel have been provided to detect gross fuel failure. This instrumentation causes closure of Group i valves, the only valves required to close for this accident. With the established setting of 7 times normal background and main steam!iae isolation valve closure, 6ssion product release is 'imited so that 10 CFR 100 guidelines are not exceeded for this accident (reference SAR Section 12.2.1.7 ).
Pressure instramentation is provided which trips when main steamline pressure dropt below 850 psig. A trip of this instrumentation results in closure of Group 1 isobtion valves. In the Refuel and Startup/ Hot Standby modes this trip function is bypassed. This function is provided primarily to provide protection against a pressure regulator malfunction which would cate the control and/or bypass valve to open. With the trip set at 850 psig. inventory loss is limited so that fuelis not uncovered and peak claddin~ temperatures at: much less than 1500
- F; thus, there g
are no 6ssion products available for release other than those in the reactor water (reference SAR Section e
1 1.2.3 ).
The RCIC and the HPCI high flow and temperature instrum:ntation are provided to detect a break in their respective pipinj;. Tripping of this instrumentation results in actuation of the RCIC or of HPCI isolation valves.
Tripping logic for this function is the same as that for the main steamline isolation valves, thus all sensors are required to be operable or in a tripped condition to meet the single-failure criteria.The trip settings of 200
- F and 300% of design flow and valve closure time are such that core uncovery is prevented and fission product release is within limits.
The instrumentation which initiates ECCS action is arranged in a one.ou -of two taken twice logic c reuit. Unlike the reactor scram circuits, however, there is one trip system associated with each function rather thar the =o trip systems in the reactor prct:ction system. The single-failure criteria are met by virtue of the fact that redundant core cooling functions are provided, e.g., sprays and automatic blowdown and high-pressure coolant injection. The specification requires that if a trip system becomes inoperable, the system which it activates is declared inoperable.
j For example,irthe trip system for core spray A becomes inoperable, core spray.iis de6ted inoperable and the t
out-of service specifications of Speci6 cation 3.5 govern. This speci6 cation preserves the effectiveness of the system with respect to the single failure criteria even during periods when maintenance or testing is being performed.
l The control rod block functions are provided to prevent excessive control rod withdrawalso that MCPR does not I
approach I.07 The' trip logic for this function is one out of n; e.g., any trip on one of the six APRM's. eight IRM's.
I four SRM's. vill result in a red block. The minimum instrument channel requirements assur: sufici:nt l
instrumentatiop to assure that the single. failure critetta are met. The minimum instrument channel requir==:nts for the RBM may be reduced bv one fer a shcrt pened cf time to 41:cw for maintenanc:.~ testing, or calibration.
i I
This tirne period as only-3cc of the operating time in a month and do:s not significantly increas: the risk of l
preventing an inadvertent control rod withdraw.t!
~
e 12 /.t.2 -o
QUAD-CmF.S DPR-29 Th: APRM red block function is Sow biased and pre.:nts a significant reduenon in MCPR. espe:: ally during cperation at reduced now. The APRM provides gross core protection,i.e.. limits the gross cere c:ntrol rods in the normal withdrawal s:quen::. The trips are set so that MCPR is main:iined gre:ter than 1.C7, t
e Th: APRM rod block function, which is set at 12"e of rated power. is functional in the Refuel and St.:rtup/ Hot Stanr. w 9 ~
.-erc' rod b!ack provides the same type of protection in the Refuel and Startap/ Hot Standby rnoces as th: APR.M hissed rod block does in the Run mode,i.e.,it prevents MCPR from cc:reasing below 1.07during control rod withdrawals and prevents control rod withdrawal before a scram ts rea:ned.
The RBM rod block function provides local protection of the core,i.e., the preventior, of transition boiling in a local region of the core for a single rod withdrawal error from a limiting control rod pattern.The trip point is now biased. The worst case single control rod withdrawal error has been analyzed. and the results show that with the specified trip settings, rod withdrawalis blocked before the MCPR reaches 1.07,thus allowing adequate margin l
(Reference 1),
B:!ow 30% power, the worst case withdrawal of a singla control rod results in a MCPR gr:ater than 1.07 without rod block action. Thus it is not required below this power level.
The IRM block function provides local as wc!! as gross core protection. The scaling arrang:ident is su:h that the trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-cas accident results in rod block action before MCPR approaches 1.07.
A dowascale indication on an APRM or IRM is an indication the instrument has failed or is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion, and the control rod motion is thus pr: vented.The downscale trips are set at 3/125 of full scale.
The SRM rod block with s 100 CPSand the detector not fully inserted assures that the SRM's at: not withdrawn frorn the core prior to commencing rod withdrawal for startup. The scram discharge volume high water level rod block provides annunciation for operator action. The alarm setpoint has been sele:ted to provide adequat: time to allow determination of the cause oflevelincrease and corrective action prior to automatic scram initiation.
For effective emergency core cooling for small pipe breaks, the HPCI system must function. sinc: reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPC1 does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimiz: spurious operation.The trip setings given in the specification are adequate to assure the above criteria are rnet (referen:e SAR Section 6.2.6.3 ).
The specification preserves the effectiveness of the system during periods of maintenanc:. testing. or calibration and also minimizes the risk of inadvertent operation, i.e., only one instrument channel out of service.
' Two air ejector off-gas monitors are provided and, when their trip point is reached, cause an isolation of the air ejector off gas line. Isolation is initiated when bcth instruments reach their high trip point or one has an upscale trip and the other a downscale trip. There is a 15 minute delay before the air ejector off-gas isolation valve is closed.
This delay is accounted for by the 30-minute holdup time of the off gas before it is released to the chimney.
Both instruments are required for trip. but the instruments are so designed that any instrument failure gives a downscale trip. The trip settings of the instruments are set so that the chimney release rate limit given in Specification 3.8.A.2 is not exceeded.
Four radiation monitors are provided in the reactor building ventilation ducts which ini.tiate isolation of the r: actor building and' operation of the standby gas tr:atment system. The monitors at: located in the reactor building ventilation duct. The trip logic is a one-out-of-two for each set.and each set can in nat: a t ip independent cf ": other set. Any ups:2!: trip will cause tne desired acnon. Trip settings of 2 -R t-for nonitors in the v:n ;!ation duct are based upon initiating normal ventilation isolation and standby gas treatm:nt 5. stem operation so that the ventilation :ta:i release rat: limit given m Specincanon 3.S.A.3 is not ete::ded.Tw o ra::auen monitors at: provided on the refue!:ng door which initiat: isolation of the reactor tiuildmg and cperation of the standby gas treatmeni systems. Th: trip logic is one out-of-two. Trip settings of 100 mR/hr for the ~.onitors on th:
r: fueling floor are bas:d upon initiating normal ventilation isolation and standby gas treatment s: stem operation 3.2 /.t.2 -7
QUAD-CITIES DPR-29 so that none of the activity released during the refueling accident leaves the reactar building sia the normal ventilation stack but that all the activity is processed by the standby gas t. atment system.
The instrumentatien which is provided to monitor the postaccident condition is listed t-Tat:e 3.2-4 The instrumentation listed and the limiting conditions for operation on these systems ensure adec. ate rc.onttoring of the containment following a loss-of-coolant accident. Information from this instrurnentation wil proude the operator with a detailed knowledge of the conditions resulting from the accident; based on this inferrnation he can make logical decisions regarding postaccident recovery.
The specifications allow for postaccident instrumentation to be out of service for a period of 7 days. This eeriod is based on the fact that several diverse instruments are available for guiding the operator should an accident occur.
on the low probability of an instrument being out of service and an accident occurring in the 7 day period, and on engineering judgment.
The normal supply of air for the control room ventilation system come< from outside the service building. In the event of an accident this source of air may be required to be shut down to prevent high doses of radiation in the ccattol room. Rather than provide this isolation function on a radiation monitor installed in the intake air duct.
signals which indicate an accident. i.e.. high drywell pressure, low water level. main steamline high flow, or high radiation in the reactor building ventilation duct. will cause isolation of the intake air to the control roo n. The above trip signals result in immediate isolation of the control room ventilation systern and thus minimize any radiation dose.
References 1.
GE Topical Report NEDO-24145, "Ganeral Electric Boiling Water Reactor Reload No. 4 Licensing Submittal for Quad-Cities Nuclear Power Station (Unit 1) ", Section 6.3.3.2, September, 1978.
G e
O e
G 3.2 / 4.2 -8
QUAD-CITIES DPR-29 TABLE 3.21 INSTRUMENTATION THAT INITIATES FRIMARY CONTAINMENT 150LATiON FUN 0i!ONS Kni =n Meter of Operable or Triped in Mnest 2
chsanetsm lastruments Trip Level Setting 1:*arr )
4 Rea:ter len waterD
>l44 inches above A
top of active fuel *
~ 4 Rea:ter icw low water
>84 inches above A
l top of active fuel
- I 4
High drywell pressure:
s2ps:gt A
S 5'
s120% of rated steam w.
B 16 High Ecw ma' steam!ce r
i 16 H:gh, temperature man s200
- F S
steamice tunnel 4
High radiatcn man s7 normal rated power B
steamline tunnefU ba:xground 2.50 ps:g B
4 Low main steam pres:ure(8) 5 4
High Row RC C steam!ine s300% cf ' rated steam ta C
16 RC C turtee area high
$200*F C
tem 0erature 4
Hth Row HPCI steamline s300% of rated steam hw D
16 HPCI area h:gh temperature
's200 ' F D
Netes 1.
Whenever prenary contam<nent etegnty is recuired, there shall be two caeratie or tncped systems for ea:9 fuc.:tv excect for low-pfessufe inain steamtme which Only need be avadatie a the Run ecsition 2.
Action:If the 6rst CCllima CaMot be met for One of the trip systems, that tng system shall be tric;ed.
l
~
A. inrtate an creerty shutdown and have the reactor a cold shutsc n conciten a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the erst cewmn cannot be met for tnth tne mtems. the a:orconate actons hstH tee. sha:t be timen.
l l
B. inrtate an oram icad reacton and have rea:!cr e mt stnce, mtam a nours e
C. cw.e issaton vanes a R:!c miem.
l D. cbst iscuten vatves a MPCI setsystem.
t
^
3.
Nead itet he c0tf atte when primar) Dn'ammefit stegrity is M:t required.
4 De f5Clatert triD sigSjl ts DfCassed when the mode seit:n is a SeNel Of t!ar!.:P:t Shgt :n1 1*
11 mstrumeetaton also isclates iflt contret rocm ve*ti;atcri !vrem 6 Ns sral a :.: a.t: mat.:a4 cieses tae m*:~anca%a:nm :. : :=a te we s: at On.a;ves I
- Tcp of active fuel is defined as 360" above vessel cero for all water levels used in the LOCA analysis (see Bases 3.2).
i.
t 3.2 /.t.2 - 1 1
QUAD-CITIES DPR-29 TABLE 3.2 2 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING STITEMS Winimum hmter of Opernis or inpped lastrument channed!3 ins Functen Tng Lani satteg Remam 4
Rea:ter icw ! w
>84 inches (+ 4 inches /
1.
In cenjun: ten wth k s.rea: tor pressure water level 20 inch) above top of c.tiates are spray and LPCl.
active fuel *
- 2. In conjun: en with higM crywe0 ;tessure 123.second tirne de!ay and icw. press =e c:'e cocirg hterbck innates auto ti:wdown.
- 3. k.tates HPCI ard RC:C.
4 Ir.:iates starting cf desel generat:rs.
t83 HO.drywell s2 ;sig
- 1. Idates core s; ray, LPCl, HPC1. and pressure,m
$373, m
2.
In conjun: ton with bw icw water level, f
123-se: nd t:me delay, and icw. pressure l
c:re cooling interock rutates auto bcwd:wm.
- 3. Inastes startirg of diesel generaters.
4.
Intates iso!ation of control room ventilation.
- 1. Pe.~nissive fer cpenrg core spray and LPCI 2
Reactor tw 200 psigsps350 psig j
pressure a: meson vanes.
- 2. In conjuncton w;th tw Icw reae:ct water level n:tates core s; ray and LPCI.
Contantnent spray Prevents r. advertent operaten of c0nta ment sterlock
- ray durrg a
- :ident c:nditens.
2m 2/3 core he:ght 2:2/3 cere he@t 4m
, contar. ment 0.5 ;sigsps1.5 ps:g high pressure 2
icer auto sl:0 seccc:s In cenrr,cten witn 1:w ion rea:ter water bcwdown level. hign-dry *eil pressure. and icw. pressure c:re c:ciirg eteric:k ' fiates auto bicw.
n C:hn.
l 4
L:w. press:e ::re 75 ;s gs:s103 :s g Oefers 'FR a:::aten ;er.deg c:n'irmaten of
- 'eg ;um; drs.
ca.; essure ::re ::::rg system c:eratcn.
.cr.arge pressure 2
Ur:eNetage On.
N/A
- Mates : 2 t: g :f Ottel gererat:rs.
I e ergency tuses 2.
f t-~ss:ve f:t starn g ECCS ;;m:2 3.
E!?:ves n:nesser.t al ::::s 'r:m tuses.
- Tep of active fuel is defined as 360" above vessel :ero for all water levels used in the LOCA analysis.
3.2 / -l.:- 12
QUAD-CITIES DPR-29 TABLE 3.2 3 1-i INSTRU!.iENTAil:N THAT INiilATES RCD ELOOK M.1%rn k!.:er cf Operat!t tr frt:pe! tr.nrument a
Ct.tates pt Trip Systed" tiztviirner:t frl; Lart! Letteg 2
AFFJ.1 u;s:a:e (.aca biasfD 50.EMW + 432 2
AF"JA c:s: ate (Rebel and Startup/H:t s!2/125 fu'! s:ae Stancy tr: ode)
- 2 AFM d:wns:aWh 23/125 full s:a'e 1
Rod ht:k rnon: tor u;s: ale ('l:w b:asy" so.65M + 42 (2) i 1
Rod t!::k momter d:*ns: ale'4 23/125 fut! s:23 3
1RM dcwns:a'e * '8' 23/125 full s:a'e 3
IFJ.! c;s:a'e8' s102/125 fd s:a's ZM SEM dete:ter cet h Startes pes: ten'4 22 feit heb c:re :ar.ter.
line 3
IRM dete:ter n:t h Sta.tp pcstton;8' 22 feet beca ccre :er.ter.
h 2'u s>
SFJAupscah s105 ::unts/se:
ZH SRM downs:aie*
210? ::ar.ts/se:
1 H:;h water level h s: tarn dts:natze volut.e s25 ga!!c s etes 1.
For tre 5:artuoMct Stutt, a*3 R.n :: sit.:ns :t the f ea:tof fr. cst st!ceter saitch. there s'a't te tao c;t atie o: tie:ts tre syste-s f:t en:* fe:te, ti:tst the !%M fod biotis. IRM c 1:a'i a*d 6P"'I C'#1:a'e me,$ act Bt C;t'ah6e in tit Run g: Sit on. ANf.t (seet:a'e SW vCMJ t tfca D asedi Ii'.I gr$;a't. and R.5M cmwa's nets n:t te citiat:0 in tre Strtap/htt star.dby fncet if the tvst column canitet et n t! tu cat et t*e tas tri sfstt-s te.rs C: : ton fr.tv es st hr t! t0* Cart pf0Vided that (Jf'*{ !. at !@t !St C0t 300 Stitt915 funCice.a%y feif t$ ifwet.att',46 !3at !*t'ta!!tf. af t*d$ CO*f.f ci tal!1 vgt' tPai I (Jyt tre system Sha3 he tit 0;t3 If tre titt COVn CanfCl t4 f."tt tCf DCth t!iD Sy5 tit 1. the system $ SM3!! be ll "St?
I., W 5 tit f tactar fectcutaton bo9 fce m Streett Trip level $ttfsg is an perttSt of f alt $ PCatt C$11 E All.
3.
EM C:wf*L:3't may be b) a:14$ mhts it rl On I'lIC*tSt IaR(t 4.~
na f.neton s typasses abt. tit cetet f att s 2100 CPS.
L Dra s' trf faaf SRM rputs f".45 te tr: asses.
L Tha :/M ten: ton inar te ty:assee m (ne hfei IRM farges Outes 8. 9. and 10) arts tr.e IRM ucsca'e rod tic:a rs c;itat'e 7.
pct fMuirPd 13 tt Opt'at't wis4 Ser4"5mg ice Scatf Sny1lCS 181!1 at at.Tc1*fitflC 9't11#t duff { 3r af:sf it?.steg 31 pCatt 'tyt31:ct 10 ta'ft1 $ M At.
L D3 FV f.*;!ca CGC' wit phti t.ht fia:!ar fRSCf tes :$ is s3 !*lt f t?Lelce $tartgp/**:t ty:ty g:Sition I.
D3 0; 5 ty;asted pSt* !*t sRV 'l fdly est'!t$
i 3.2/4.2-14
QUAD-CITIES DPR-29
- 3. The control red drive housing support
- 3. The correctness of the control rod sys.em shall be in place during reactor withdrawal sequence input to the power operation and when the reactor RWM comput:r shall b: veri 5:d after coolant system is pressurized above loading the sequence.
atmospheric pressure with fuel in the Prior to the start of control rod with-reactor vessel, unless all control rods 6wal towards criticality, the capabil-are fully inserted and Specification ity of the rod worth minimizer to 3'3 p'operly fulfill its function shall be r
a.
Control rod withdrawal sequences veri 5ed by the following checks:
shall be established so that max-a.
The R%.M computer online d.iag-imum reactivity that could be nostic test shall be successfully a'dded by dropout of any incre-performed.
ment of any one control blade would not make the core more b.
Proper annunciation of the selec-than 0.013 ak supercritical.
tion error of one out of-sequence c ntr i r d shall be veri 5ed.
b.
Whenever the reactor is in the Startup/ Hot Standby or Run c.
The rod block function of the l
mode below 20% rated thermal RWM shall be verified by with-power, the rod worth minimiz:r drawing the Srst rod as an out-shall be operable. A second opera-of sequenc: control rod no more tor or qualified technical person than to the block point.
may be used as a substitute for an
?
inoperable rod worth minimizer which fails after withdrawal of at least 12 control rods to the fully withdrawn position. The rod worth minimizer may also be bypassed for low power physics testing to demonstrate the shut-down margin requirements of j
' Specification 3.3.A if a nuclear engineer is present and verifies the step-by step rod movements of the test procedure.
l 4.
Control rods sha!! not be withdrawn 4.
Prior to control rod withdrawal for j
for startup or refueling unless at least startup or during refueling. verify that
{
two source range channels have an at least two source range channels observed count rate equal to or greater have an observed count rate of at least than thre counts per second and these tnre: counts pt: second.
SRMYare fully inserted.
5.
During operation wi'h limiting con-5.
When a limiting control rod pattern trol rod patterns. as determined by the exists. an :nstrument functional tes't of nuclear engine:r. either-the RBM shall be cerformed prior to withdra;x iand d uy therea:j:
i t.
cesign ted r d(s) a.
both RBM channels shall be ter_
t operabli.
~
b.
control rod withdrawal shall be blocked: or 3.3 / 4.3-3
l QUAD-CITIES DPR-29 e
i c-the operatinc power level shall be limited so that the MCPR will remain above 1.07 assuming a sin-gle error that results in complete withdrawal of any single operable control rod.
C.
Scram Insertion Times C.
Scram Insertion Times 1.
The average scram insertion time, ba.
- 1. After refueling outage and prior to sed on the deenergization of the scram operation above 30re power, with re-pilot valve solenoids at time zero, of all actor pressure above 800 psig, all con-operable control rods in the reactor trol rods shall be subject to scram-time power op: ration condition shall be no measurements from the fully with-greater than:
drawn position. The scram times shall be measured ~ without reliance on the
-Average Scram control rod drive pumps.
% Inserted From insertion Fully Withdrawn Times (sec) 5 0.375 20 0.900 g
50 2.00 90 3.50 The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two by two array shall oc no greater than:
% inserted From Average Scram Fully Withdrawn Times (sec) 5 0.398 20 0.954 50 2.12 90 3.80 Ii 2.
The maximum scram insertion time 2.
Following a controlled shutdown of f
for 90% insertion of any operable con-the reactor, but not mere frequently trol rods sh'a!! not exceed 7 seconds.
than 16 w::is ner less frecuentiv than 32-we:k int::vals. 50cc ci the c'ontrol 3.
If Soeci5 cation 3 3.C.1 car. net be met.
r d drives in ea:h cuadrant of th:
^
(
the res: tor shall not be made super-re et : cor shall be' measured for the l
crttical if op::atinc. the reactor shall be shut down imr.:diat:!y upon deter-scram times specided in Spe:i5 cation 3.3.C. All control rod drtves shall have mination that averace scram time is I
d&ie.s experienced scram test measurements each year. Wheneser all of the contro!
4 If Speci5 cation 3.3.C.2 cannot be met, rod drive scram times have been m:a-the defcient control rod shall be con-sured. an evaluation sha:1 be made to t
3.3 / 4.3 -4
QUAD-CITIES DPR-29
- b. an end-of-cycle de!ayed neutron fractm of 0 005, a beginning-of life Doppler reactivity feedback.
c.
- d. the rod scram insertion rate shown in Speci5 cation 3.2.C.
the maximum possible rod drop velocity of 3.1I fps.
c.
f.
the design accident and scram reactivity shape function, and
- g. the mod:rator temperature at which criticality occurs.
In most cases the worth ofinsequence rods or rod segments will be substantially less can 0.013 ak.
Further, the addition of 0.013 ak worth of reactivity, as a r:sul: of a red drop and in conjunction with the actual values of the other important ac:ident analysis parameters describd above, would most likely result in a peak fu:1 enthalpy substantially less than 250 cal /g design lir.it. However, the 0.013 ak limit is applied in order to allow room for futute reload changes and es.se of verification l
without repetitive technical specification changes.
Should a control drop accident result in a peak fuel energy content of 250 cal /g, fewer than 660 (7 x
- 7) fuel rods are conservative!y estimated to perforate. This would result in an offsite dose well below i
the guideline value of 10 CFR 100. For 8 x 8 fuel, fewer than 850 rods at: conservatively estimated to perforate, with nearly the same consequences as for the 7 x 7 fu:1 case because of the rod power i
differences.
The rod worth rninimizer provides automatic supervision to assure that out of sequene: control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences (reference SAR Section 7.9). It serves as a backup to procedural control of control rod worth. In the event that the rod worth minimizer is out of service when required, a licensed operator i
or other qualified technical employee can manually fulfill the control rod pattern conformance function of the rod worth minimizer. In this case, the normal procedural controls are backed up by independ:nt procedural controls to assure conformance.
4.
The source range monitor (SRM) system performs no automatic safety system function, te.,it has no scram function. It does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and efficient reactor startup at low neutron levels. The consequences of l
reactivity accidents are functions of the initial neutron flux. The requirement cf at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 104 of rated j
power used in the analyses of transients from cold conditions. One operab!: SRM channel would be adequate to monitor the apptcach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's is provided as an added conservatism.
- 5. The rod block monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the consol for maintenance and/or..
t: sting. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator. who withdraws control rods according to a written i
i sequence. The specified restrictions with one channel out of servie conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists. During reactor j
operation with certain limiting control rod patt:rns, the withdrawal of a d:signat:d single controlf f
rod could r:sult in ene of more fuel rods with MCPR's less than 1.07.During use of such patterns.
l it is judged that t: sting of the RBM system to assure its operability prior to withdrawal of such rods l
will assure that irnproper withdrawal does not c::ur. It is the responsibility cf the Nu:!::.r Engineer i
to identify these lirniting patterns and tne designated rods eitner when in: patterns are initially.
establish:d or as they develop due to the occurr:ne: ofinop:rabi: control rods in other th.n limiting patterns.
I i
3.3s.t3-9 f
e
QUAD-CITIES DPR-29 C.
Scram Insertion Times The control rod syst:m is analyzed to bring the reactor subcritisal at a rate fast enough to prevent fuel damage, i.e., to pr: vent the MCPR from becoming less than 1.07.The limitmg power transient is that resulting from a turbine stop valve closure with failure of the turbine bypass syst:m. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the avera;c response of all the drives as given in the above specification provide the required protection. and MCPR remains i
greater than 1.07.Referene: I shows the control rod scram reactivity used in analyzing the transients.
l
}
Reference I should not be confused with the total control rod worth 18"o Ak, as listed in some amendments to th: SAR.The ISto ak value represents the amount of r: activity available for withdrawal in the cold clean core, whereas the control rod worths shown in Reference I represent the amount of Jeactivity available for insertion (scram)in the hot operating core. The minimum amount ofreactivity to be inserted during a scram is controlled by permitting no more than 10Co of the operab!: rods to have long scram times. In the ana!ytical treatment of the transients. 390 milliseconds are allowed betwe:n a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and l
conservative when empared to the typically observed time delay of about 270 mittiseconds. Approx.
imately 70 milliseconds after neutron flux reaches the trip point, the pilot scram valve solenoid j
d: energizes. Approximately 200 milliseconds later, control rod motion begins. The time to deenergize the pilot valve scram sole 6oids is measured during the calibration tests required by Specification 4.1.The 200 milliseconds are included in the allowable scram insertion times speci5ed in Specification 3.3.C.
The scram times for all coritrol rods will be determined at the time of each refueling outage. A representative sample of control rods will be scram tested at increasing intervals following a shutdown.
Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times following initial plant operation at power are expected. The test schedule at increasing time intervals provides reasonable assurance of detection of slow drives before system deterioration beyond the limits i
of Specification 3.3.C. The program was developed on the basis of the statistical approach outlined below and judgment.
The probability that the mean 90?o insertion time of a sample of 25 control rod drives will not exceed 0.25 seconds of the mean of all drives is 0.99 at a risk of 0.01. If the mean time exceeds this range or the mean 90"e insertion time is greater than 3.5 seconds, an additional sample of drives will be measured to verify the mean performance.
Since the differences between the expected observed mean insertion time and the limit of Specification 3.3.C greatly exceed the expected range, this sampling technique gives assurance that the limits of Specification 3.3.C will not be exceeded. As further assurance that the limits of Specification 3.3.C will not be exceeded, til operable dr.
will be scram tested to determine compliance to Speci6 cation 3.3.C if the enlarged sample of 50 control rods exceeds 4.25 seconds. The 0.75 second margin to the limit is greater than the maxtmum expected deviation from the mean and therefore gises assurance that the mean will not exceed the limit of Speci6 cation 3.3.C. In addition. 50"e of the control rods will be checked every 16 weeks to verify the performance and for correlation with the sampling program.
i The history of drive performance accumulated to date indicates that the 90"e insertion tir es of new and overhauled drives approximate a normal distribution about the mean which tends to b:come skewed toward longer s: ram tirnes at operating time is accumulated. The pronability of a dr:ve not ex::: ding the m:an 90"e insertion tim ev 0.75 seconds is greater than 0 999 for a normal distribution. The m:ssur:m:nt of the scram performance of the drives surrounding a drive ex::: ding the :xpected range of scram performance will detect local variations and also provide a3surance : hat Ic:al scram time limits at: not exc:eded. Continued monitorinc of other drives exceeding-the exp::ted range of scram tim:s a
provides surveillance-of possible anomalous perfcrman::.
l Th: numeri:al values assigned to the predicted scram performance are based en the analysis of the Dr:sd:n 2 startup data and of data from other BWR's such as Nin: Mil: Point and Oyst:r Cre:k.
I i
3.3 /.'.3 -10
QUAD-CITIES DPR-29 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reacter 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 />. Surveil-lance and corresponding action shall continue until r:a: tor operation is within the prescribed limits.
LH G R.,,
LHGR,1 -(aP/P),,,(L/L )r
17.5 kW/ft, 7 x 7 fuel assemblies 13.4 kW/ft,8 x 8R fuel assemblies 8x!
=
( a P/ P ),,,
- maximum power spiking penalty
.035 initial core fuel
=
.029 reload I, 7 x 7 fuel
=
.022 reload, S x 8 fuel
=
.MR reload I mixed otide
=.UUO reload 8 x 8R peluel assemblies
=
L
= total core length 7
= 12 feet L
Axial distance from bottom of core
=
K.
Minimum Critical Power Ratio (MCPR)
K.
M nimum Critical Power Ratio (MCPR)
During steady-state operation MCPR shall be The MCPR shall be d:termined daily during greater than or equal to steady-state power operation above 259, of rated thermal power-
- 1. 23 (7 x 7 fuel) 1,29 (S x 8 fuel) er and(riow. If at any time during 8 x 8 BLTA) at rated pow.32 1
operation it is determined by normal surveil-lance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minut:s to restore operation to within the pre-scribed 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 b: 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 />. Surveil-lance and corresponding action shall continue until rea: tor operation is within the prescribed
!;.:ts. For core i'cws o,:her than rat:d. tnese nemir.al values of MCPR 5: a:1 be increased by a factor of k. where k is as shown in Figure r
r 3.5-2.
3.5 /.t.5 - 10
QUAD-CITIES DPR-29 i
3.5 IlMITING CONDITION FOR OPERATION BASES A.
Core Spray and LPCI Mode of the RHR Sprem This specification assures that.idequate emergency cooling capabi!:ty is availsh!: whenever irtsdi ted fuel is in the reactor vessel.
s Based on the lesa-or coolant analyt test sc.thod s deser tied in General electete Tootest neport Kno-2:56 6 a-d the s pe c t f te snelysis in m.Do 24146
- Loa n-o r-coo t ar t Annys ts sonort ror I
Deesden unita 2 a and oued-c a t tes t/nats 1. 2 Nuclear Power l
StatLons, SeptemDer 3.976 core coolinpyste s provide suf f tetent cooling to the core to dise tpate the energv i
anacciated intth the loss-of-coolant a ce nd e nt, to 1:. mat casc alava fuel claeding temperature to less than 22:Co, to assure that r
core geometry renaans antact, to limat cladding metal-water easetion to less than 1%. and to is.nst the calculated lo:a1 matal-water resetton to less than 17%.
The limiting conditions of o;:rstien in Specifications 3.5.A.I through 3.5.A.6 spe:ify the cornbinations of cperab'e subsys: cms'io ass:.re the svailability of the minimum cooling systems ceted above. No single failure of ECCS equipment c::urring during a loss-of-coolant accident under these limiting conditions of operation will result in inadequate cooling of the rea: tor core.
Core spray distribution has been shown, in full scale tests of systems similar in design to that of Quad Citi:s I and 2. to exceed the minimum requirements by at least 257. In addition. coolin; effectiveness has been demonurated at leu than half t!: rateil flow in umulated fuel auctnbhes wuh hester rods to duplicate the dreay heat charat ternte of irra.h.ned fuel. l'he aunlent.malpn n additionally conservauve in that no credit n taken for spray coolitig of the reactor core before the internal pressure has fallen to 90 psig.
The LPCI mode of the RHR system is designed to provide emergenev cooling to the core by flooding in the event ora loss-of-coolant accident. This system funcnons in combination with the core spray system -
to prevent excessive fuel cladding temperature. The LPCI mode of the RHR system in combination with the core spray subsystem provides adequate cooling for break areas of approximately 0.2 ft: up te and in:!udir:g 4.1S ft', the latter being the douNe ended recirculation line break with the equahzer hne between the recirculation loopseloeedwithout assistance from the high pressure emerg:ncy core cooling subsystems-The allowable repair times 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 Reference 1. ll sing the results deve! aped in this reference, the repair period a found to be lew than half the test intersal. This assumes that the cere spray subsystemt and 1.lTl wmutute J one out of two setem. howeser the corubined eilect of the two systems to hmit euenne daJJmg tempetatuie must aho be conuJered. Ibe test mtersal vecmeJ m Specification 4.5 was 3 montns lhetetore, an allowable repair pers J wh.ch ma.nta ns the basic ran considering single failures should be lea than 30 days and tnn spe itication ts ws:hm inis oenod. For multiple failures, a shorter interval a specineJ. to unprove the awarance that the remaming wuems will function, a daily test is ca!!ed tot. Although it is rewgn: zed tnat the :nformat;on g;ven in Reference i provid:s a quantitaave method to esumate allowable repair ames, the laa of cpe:aang Jata to sup:crt th: ana!yti:al appreach presen:s comp!ct: acceptan:e of :h:s m:tho: at th;s t:me. Ther: fore, tne times s:a::4 in th: sp:cifi: items were estachsned w:th du: rep. J to judg=:n:.
Sh:uld on: core spisy su: system 6::ome inoperao!e. the remaining :::e sp ay subsystem and :h: entire LFCI mode of the RHR spte= are avat!aD;e snouid in: need for cor: cochng anse. To assure that tn:
r:maining cor: spray, the LPC! m:Je of tne RH R system- :.d tne d.nel genera c:s are ava.:_:.c. ;.ey are demonstrat:J to be'eperao.e immediate!). This J m:nstranon :n:!udes a manual ininauen of the pumps and assoc:ated valves and d:esel g:nerators SaseJ :n f uegr".:n'.s citn: rehabhty cf tne remsem;
- sp
- :ms. i.e., the cer spray anJ LPCI. a 7 day repair p:::04 was ob m:4.
3.5/4.5-11
QUAD-CITIES DPR-29 Should the low of one RilR pump occur. a nearly full compicment of co.e and ennsam: e-a emhng equipment is av.nlable ~Ihree RilR pumps n coopmction witn ene core spray sunsystem w all rie: form tne core conting function. Recauw of the av.nlabihty of the majority of the core coohng equip-em. wh rh will be demc nstrated in be operabic, a 30 day repair period is patitied if the LPCI made of :ne R}iR system is not available. at le.ast two RHR pumps must he availanle to fultill the containtnent coohng fuisetion. The 7 day repair period is set on this basis.
B.
RHR Senice Water He containment cooling mode of the RHR system is provMed to remose heat energy from the containment in the event of a loss-of-coolant accident. For the flow speciSed. the containment Inn; term pressure is limited to less than 8 psig and is therefore more than ample to provide the required heat removal capability (reference SAR Section 5.2.3.2).
The containment cooling mode of the RHR system ennsists of two loops. esch centsining two RHR
. service water pumps. one hest exchanger, two RHR pumps, and the associated valves, piping. electrical equipment. and instrumentstion. Either set of equipment is espable cf performing the con:smme st cooling function. Loss of one RilR service water pump does not seriously jeopirdire the cont.unment cooling capability, an any one of the remanung three pumps (an sainfy the omiing requiremena Since there is some redundancy left. a 30 day repair period is adequate I.oss of one loop of the sontauunent cooling mode of the RHR sprem leaves one reinaining system to perform the contamment cooling function. The operable system is demonstrated to be operable each day u nen the shoye condition occurs.
Based on the fact that when one loop of the containment cooling mode of the RHR system be nmes inoperable, only one system remains, which is tested daily a 7-day repsir period was spect6ed.
C.
High.Prmure 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.
He HPCI meets this requirement without the use of offsite electrical power. For the pipe breaks for which the HPCI is intended to function, the core never uncovers and is continuously cooled, thus no cit.fding 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 llPCI as part of the isolation cooling system.
D Automatic Pressure Relief ne relief valves of the automatie pressure relief subsystem are a bsekup to the HPCI subsystem. They enable the core spray subetem or LPCI mode of the RilR system to provide protestien against the sms!!
pipe break in the event orilPCI failure by Jerrenuriting the reactor vewel rapidly enough to actuate the core spray subsystems or LPCI mode of the RilR sptem. The cere spray subsystem and tne LPCI mode of the RIIR sprem provide sutheient 11o* of uvlant to hnut feel claddmg tem [ieraturesaless than 22000F, to assure that core geometry remains intact, to li: nit the core wide cl-ad metal-water reaction to less than 1%,and to limit the calculated local metal-water reaction to less than 17%.
Loss of 1 of the relief valves affects the pressure relieving capability and, therefore, a 7 day repair period is soecified.
Loss of more than one relief valve significantly reduces the pressQre relief capability, thus a 24-hour repair period is specified based on the HPCI system availability during this period.
I E.
RCIC Re RCIC system n prosided to supply contmuous makeup water to the reacter core wnen the re:ctor is isolated from the turmne and whea the feat-ahr miem n nat.na !aNe. Under these cond;ned, tne.
pernping CJpJc1ty of the RClt' *) stem h sutlicient td manitaul(ne wats: lesel aOnse the gere %;t"; oat :n*,-
other water system in operanon if the w ater lesel m tne reactor vend Je,.reases to the RC!C m...:. n I: vel the system.u.tomaticaH3 starb. The aptem may alu ce mana !b nauated at any time 3.5/4.5-12
QUAlbCI'lli N DPR-29 H.
Condensate Pump Room Flood Protection See Specifiestion 3.5.H.
L Anrage PI:nar LHGR This specification assures that the pesk c!sdding temperature following de postulated d: sign basis loss of. coolant ac:ident will not esceed the 2200 F limit specnied in the 10 CFR 50 Appendix K 0
I considermg the postulated effcets of fuel pellet densification.
The peak cladding temperature following a postuisted lo;s of. cool. int accident is primanly a functiun of i
the average heat. generation rate of all the rods of.: fuel assembly at any axial location and is only f
secondstily dependent on the rod.to. rod power dntribution within.:n awembit Since expected local variations in power Jntribunon wahm a fuel awemi tv atresi ihe calculaint peak stadJm; temperature by less than110 F relative to the peat tem;wrature for a ispital suel deuen. the knui on she.ner.ge l
, plar' r L51GR is sullictent to awure that calculated temperatures are below the hnut. 'I he masunum average plansr LHG R's shown m Ficure 3.5 1 are based on calculations employing the modeh desenhed in Reference 2.
J.
Local LHGR This specifiestion assures that the maximum linear heat. generation rate in any nxiis less than the design linear heat. generation rate even if fuel nellet denuticanon is pmtulated The power spite penalty specified is based on that presented in Reference 3 and awumes a knearly inereeing varianon in.nial gaps between core bottom and top and apures with a 95" confidente that no rnore inan ene fuel rod exceeds the design linear heat.generatton rate due to power spilm; An irradi.inon growth factor of 0.25% was used as the basis for determining.1/F in accordance with Reference,4 and 5.
K.
Minimum Critical Power Ratio (MCPR)
The steady state values for MCPR spec 1Ged in this spenficannn were selected to proude margin to a::ommo.
date transients and uncertsmties m monnoring the core operatuig state as well as urn;ertstnties a the critical power correlation itself. These values also assure that operation wJLbc such that the intital conditu,n anumed for the LOCA analysts. an MCPR of 1.l8.issainhed. For any of the special set of tranuents or dnturbanees caused by single operator error or ung!c equipment mallunction. it is requued that design analyses imitshied at this steady. state operating limit yield a MCPR of not less than that speellied m Srcettication 1.1.A at - ny time during the trantent. assurntng instrument trip settings given in Speafication 2.1. For analyus of t!.e thermal consequences of these tranuents, the hmitmg value of MCPR stated in this spenficanon is co.:
servatively assumed to exist pner to the initiation uf the trannents. The results apply with meressed con.
servatism while operating with MCPR's greater than speettied.
The most limiting trartuents wtth respect to MCPR are generally; a) Rod withdrawal error b) Turbine tnp without bypass c) Loss of feedwater heater Several factors influence which nf these transaents results in tne largest reductien in entica!;ower rano such as the specific fuellosdmg. esposure. and fuel type. The current cycles re'usd h.ensq submitta! speed :s the licruttng trans:er.ts for s g..cn exposure mcrement tur ca.n fues type. The ducs speen:ed as :.c Lan;t:.ng Condiuon of Operatwn are conservatacly chosen as tne m:st restnetne over the :nt.re :yc!e for esen tue!
type.
s 9
3.5/4.5-14
l QUAD-CITIES DPR-29 For core flow rates less than rated. the steady state MCPR is increased hv the formula given in the speeiti:ation. This assures that the MCPR will be maintained greater than inat spec:ned in Spe:theation 1.1.A esen in the event that the motor generator set speed controller causes the scoop tute positioner for the fluid coupler to move to the maximum speed position.
References 1.
I. M. Jacobs and P. W. Marritt. GE Topical Report APED.5736. 'Guidatines for Determin ng S.:fe Test Intervals and Repair Tisnes for l'ngneered Safeguards / April 19(N 2.
" Loss-of-Coolant-Accident Analysis Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nticlear Po'mr Stations, " IEDO-24146, Septe::.ber 1978.
3.
GE Topical Report NEDM 10735.
- Fuel Densirication Effects on General Electric Boiling Water Resetor Fre!,* Section 3.2.1, Supplemen: 6. August 1973.
4.
J. A. Hinds. GE. Letter to V. A. Moore. USAEC.
- Plant Evaluation with GE GEGAP.!!!.* December 12.
1973.
~
5 USAEC Report. Supplement I to the Technical Report on Denstrication of General Electric Reactor Fuels.'
December 14. 1973.
I, l
i l
i I
1 i
f
~
l I
3.5/4.5-15