ML20028H682
| ML20028H682 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/22/1991 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML19310E532 | List: |
| References | |
| NUDOCS 9101280049 | |
| Download: ML20028H682 (34) | |
Text
. _ _ _.. _ _ _. _ -.. _.. _ _ _ _ _ _ _ _
k LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.1 REACTOR PROTECTION. SYSTEM 4.1_ REACTOR PROTECTION SYSTEM Appikibili ty:
Acolicability:
Applies to the instrumentation Applies to the surveillance of and associated devices which the instrumentation and initiate a reactor scram, associated devices which initiate reactor scram.
Obiettive:
Qhjective:
To assure the operability of the reactor protection system.
To specify the type and frequency of surveillance to be applied to Soecification:
the protection instrumentation.
The setpoints, minimum number of Soetification:
trip systems, and minimum number-of instrument channels that must Instrumentation systems shall be be operable for each position of functionally tested and the reactor mode switch shall be chlibrated as indicated in Tables as given in Table 3.1.1.
The 4.1.1 and 4.1.2 respectively.=
system response times from the opening of the sensor contact up to and including the opening of the trip actuator contacts shall not exceed 50 milli-seconds.
4 Amendment No - 42, 129, 133 ~
26-9101200049 91o122 PCR ADOCK 05000293 P
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3.1 f E (Cont'd) 4.1 ILAS15 (Cont'd) been used in transient analyses A comparison of Tables 4.1.1 and cealing with coolant inventory 4.1.2 indicates that two decrease.
The results show that instrument channels have not been scram at this level adequately included in the latter Table, protects the fuel and the These are: mode switch in pressure barrier, because HCPR shutdown and manual scram. All remains well above the safety of the devices or sensors o
limit MCPn in all cases, and associated with these scram system pressure does not reach functions are simple on-off the safety valve settings.
The switches and, hence, calibration scram setting is approximately 25 during operation is not in, below the normal operating applicable, i.e., the switch is range and is thus adequate to either on or off.
avoid spurious scrams.
The sensitivity of LPRH detectors Jyrbine Stpp Valve Closun decreases with exposure to neutron flux at a slow and The turbine stop valve closure approximately constant rate.
scram anticipates the pressure.
This is compensated for in the neutron flux and heat flux APRH system by calibrating every increase that could result from three days using heat balance rapid closure of the turbine stop data and by calibrating valves. With a scram trip individual LPRH's every 1000 setting of 1 10 percent of va'ive effective full power hours using closure from full open, the TIP traverse data, resultant increase in surface heat flux is limited such that HCPR remains above the safety limit HCPR even during the worst case transient that assumes the turbine bypass is closed.
l Turbine CoatIgl Valve fast Ciontti 4
The turbine control valve fast closure ! cram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves.
The reactor protection system initiates a scram when fast 1
closure of the control valves is j
initiated by the acceleration relay.
This setting and the fact that control valve closure time is approximately twice as long as Amendment No. 42, 13) 40
4 3.1 QAi{$ (Cont'd) biased rod block and scram. The.RH's provide.dditional protection in the " Refuel" and "Startup/ Hot Standby" Sodes, thus, the IRH and APRH 16%
{
scram are required in the " Refuel" and "Startup lot Standby" modes.
In s
the power range the APRM system provides tae rei:Jired protection. Ref.
l Section 7.S.7 iSAR.
Thus, the IRH system is not required in the "Runa mode.
{
i The high reactor pressure, high drywell pressure, reactor low water level i
and scram discharge volume high Isvel scrams are required for Startup/ Hot Standby and Run modes of plant operation.- They are th to be operational for these modes of reactor operation,erefore, required
]
l The requirement-to-have--the scram functions, as indicated in Table 3.1.1, operable in the Refuel mode is to assure that shifting to the Refuel. mode during reactor power operation does not diminish the need for the reactor protection system.
I The turbine conoenser low vacuum sc ara is only required during power operation and must be bypassed to start up the unit.
Below 305 psig turbine first stage pressure (45% of rated), the scram signal due to turbine stop valve closure or fast closure of turbine control valves is i
bypassed because flux and pressure scram are adequate to protect the reactor.
If the scram signal due to turbine stop valve closure or fast closure of turbine control valves is bypassed at lower powers -less 1
l conservative MCPR and MAPLHGR operating limits may be appited as specified in the CORE OPERATING LIMITS REPORT.
The requirement that the IRH's be inserted in the tore when the APRH's l
read 2.5 indicated on the scale assures that there is proper overlap in l
the neutron monitoring systems and thus, that adequate coverage is l
provided for all ranges of reactor operation.
The provision of an AFRH scram at,(15% design power in the " Refuel" and "Startup/ Hot Standby" modes and the backup _IRH scram at 1120/125 of full scale assures that there is proper overlap in the neutron monitoring systems and, thus, that adequate coverage is provided for all ranges of reactor operation.
l Amendment No. 133 40c
PNPS YABLE 3.2.C-1 INSTRUENTATION THAT INITIATES ROO BLOCJK Minimum Operable Channels Required Trio Function Der tric function Operational Conditions Notes l APRM Upscale (Flow 4
Run (1)
Blased)
APRM Upscale 4
Startup/ Refuel (1) l APRM Inoperative 4
Run/Startup/ Refuel (1)
APRM Downsca'e 4
Run (I)
J l
Rod Block Monitor 2
Run, with Ilmiting cont 01 rod (2) r (Power Dependent) pattern, and reactor power > LPSP (S)
(
3 f
Rod Block Monitor 2
Run, with limiting control rod (2)
Inoperative pattern, and reactor power > LPSP (5)
Rod Block Monitor 2
Run, with limiting control rod (2)
Downscale pattern, and reactor power > LPSP (S) l f
IRM Downscale 6
Startup/ Refuel, except trip is (1)
(
bypassed when IRM is on its icwest range IRM Detector not in 6
Startup/ Refuel, trip is bypassed (1)
Startup Position when mode switch is placed in run IKM Upscale 6
Startup/ Refuel (1)
IRM Inoperative 6
Startup/ Refuel (1)
SRM Detector not in 3
Startup/ Refuel, except trip is by-(1)
Startup Position passed when SRM m unt rate is 1 100 counts /second or IRMs on Range 3 or above (4)
SRM Downstale 3
Startup/ Refuel, except trip is by-CD L
passed when IRMs on Range 3 or above (4) 54 Amendment 15, 21, 42, 55, 72, 79. 710, 129 L.
a _.
~
t PNPS TABLE 3.2 C-1 (Con't) 4 Mip' mum Operable Channels Required j _
Trio Function Der trio function Operational Conditions Notes SRM Upscale 3
Startup/ Refuel, except '. rip is by-(1) passed when the IRM range switches are on Range 8 or above (4)
SRM Inoperative 3
Startup/ Refuel, except trip is by-(1) passed when the IRM range switches I
are on Range 8 or above (4)
Ii Scram Discharge 2
Run/Startup/ Refuel (3)
I Instrument Volunt Mater Level - High l
Scram Discharge 1
Run/Startup/ Refuel (3) 2 Instrument Volume-Scram
]
Trip Bypassed I
i Recirculation Flow
-2 Run (1) l Converter - Upscale
- 1
,j Recirculation Flow 2
Run
-(1) 8 Converter - Inoperative Recirculation Flow 2
Run (1)
Converter - Comparator Mismatch
(
Amendment-No. 15, 27. 42, 65, 77, 79, 170, 129 548
l l
l Notes for Table 3.2.C-1 1.
With the number of operable channels:
.?.. One less than retaired by the minimum operable channels per trip function requirement, restore an inoperable channel to operable status within 7 days or place an inoperable channel in the tripped condition within the next hour.
b.
Two or more 1er.s than required by the minimum operable channels per
'l l
trip function requirement, place at least one inoperable channel in
]
the tripped ce dition within one hour.
2.
a.
With one RBH Channel inoperable:
1 (1) restore the inoperable RBH channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; otherwise place one rod block monitor channel in the tripped conaition within the next hour, and; i
(2) prior to control rod withdrawal, perform an instrument function i
test of the operable RBH channel, i
b.
Hith both RBH channels inoperable, place at least one inoperable rod l
block monitor channel in the tripped condition within one hour.
3.
If the number of operable channels is less than. required by the minimum operable channels per trip function requirement,-place'the inoperable channel in the tripped condition within one hour.
l 4.
SRH operability requirements during core alterations are given in Technical Specification 3.10.
5.
RBH operability is required in the run-mode in the presence of a limiting rod pattern with reactor power greater than the RBH low power setpoint (LPSP). A limiting rod pattern exists when:
HCPR < 'i.40_for reactor power 2 90%
HCPR < l.70 for reactor power < 90%
The allowable value for the LPSP is 1 29% of rated core thermal power.
Amenament No. 15, 27, 42. 55, 170 55
~. _. _ _. -.. _
PNPS TABLE 3.2.C
- CQfTROL ROD BLOCK INSTRUMENTATION SETPOINTS Trio Funtij.2D Trio Setooint APRH Upscale (1) (2)
APRM Inoperative Not Applicable-APRH Downscale 1 2.5 Indicated on: Scale Rod Block Honitor (Power Dependent)
(1) (3)
Rod Block Monitor Inoperative Not Applicable Rod Block Honttor Downscale (1)-(3)
TRH Downscale 1 5/125 of full Scale IRH Detector not in Startup Position Not Applicable IRH Upscale 1 108/125 of Full Scale IRH Inoperative Not Applicable SRM Detector not in Startup Position Not Applicable-SRM Downscale 1 3 counts /second-SRM Upscale 1 105 counts /second SRM Inoperative Not Applicable Scram Discharge Instrument Volume 1 IB gallons Hater Level - High Scram Discharge Instrument Volume -
Not. Applicable Scram Trip Bypassed Recirculation Flow Converter - Upscale 1 120/125 of Full Scale Rccirculation Flow Converter -
Not Applicable Inoperative Recirculation Flow Converter -
1 10% Flow Deviation for > 80%
Comparator Hismatch Rated Power,'and I 15% Flow Deviation for 1 80%=
1 Rated Power (1) The trip level setting shall be-as specified-in the CORE OPERATING LIMITS REPORT.-
(2) When the reactor mode switch is in the refuel or-startup posi.tions, the APRM rod block trip setpoint shall _be less than or equal to 13% of rated thermal power, but_always less than the APRM flux scram trip setting.
(3) The RBH bypass time delay (td2) shall be < 2.0 seconds.
Amendment No. 42, 170, 129, 133 55a m_...
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4 3.2 BASES (Cont'd)
The control rod block functions are provided to prevent excessive control rod withdrawal.
The trip logic.for this function is 1 out of l
n:
e.g., any trip on one of six APRH's, two RHB's, eight IRH'r, or four CAM's will result in a rod block.
The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met.
The minimum instrument channel requirements for the RBH may be reduced by one for not longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without significantly increasing the risk of an inadvertent control rod withdrawal.
Reactor power may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point, thereby possibly avciding an APRM Scram.
The rod block setpoint is automatically reduced wita recirculation flow to form the upper boundary of the PNPS powi r/ flow map.
The flow biased APRM rod block is not necessary to prolib't fuel damage and is not included in the hnalysis of anticipated transients.
The RBH rod block function provides local protection of the core, for a single rod withdrawal error from a limiting control rod pattern.
The RBH bypass time delay (td2) is the delay between the time the signal is normalized to the reference signal and the time the signal is passed to the trip logic. Control rod withdrawal is unrestricted during this interval.
The RBH bypass time delay is low enough to-assure that control rod movement is minimized during the time RBH trips are bypassed.
The IRH rod block function provides local as well as gross core protection.
The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level.
A downstale indication on an APRM, RBH or IRH is an indication the l
instrument has failed or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.
The downscale trips are as shown in Table 3.2.C-2.
Amendment No. 15, 42, 110, 129, 133 71 l
L l-
j LIMITf NG CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT 3.3.8 Control Rods 4.3.B Control-Ro.dt 4.
Control rods shall not be 4.
Prior to control rod withdrawn for startup or withdrawal for startup or refueling unless at least two during-refueling, verify that source range channels-have an at least two source range observed count rate equal to-channels.have an observed or greater than three counts count rate of at least three y
per second, counts per second.
5.
The RBH shall be operable as required in Table 3.2.C-1, or control rod withdrawal shall be blocked.
C.
Scram Insertion Times C.
Scram Insertion Times 1.
The average scram insertion time, based on the 1.
Following each refueling.
deenergization of the scram outage, or after-a reactor.
pilot valve solenoids as time-
-shutdown that is greater.thanL zero, of all operable control-120 days, each operable.
rods in the reactor power control rod shall be subjected operation condition shall be to scram time tests from the no greater than:
fully withdrawn position.
If
. testing is_not accomplished
% Inserted -Average Scram with the nuclear _ system from Fully Insertion pressure.above 950 psig, the Withdrawn-Times (set) measured scram insertion time shall be extrapolated to 10
.55 reactor pressures;above 950 _
30 1.275 psig using previously 50 2.00
. determined correlations.
90 3.50 Testing:of all operable:
control rods shall-be completed prior to exceeding 40*/. rated thermal power.
J i
Amendment No. 75, 68, 724 B3 l
3 I
3.3 and 4.3 BASIS:
4.
The Source Range Monitor (SRM) system performs no automatic safety system function; i.e., it has no scram function.
It does provide the operator with a visual indication of neutron level.
The consequences of reactivity accidents are functions of the initial neutron flux.
The requirement of at least 3 counts per second 1
assures that any tra sient, should it occur, begins at or above the
- 'l initial value of 10- of rated power used in the analyses of
-- {
transients from cold conditions. One operable SRM channel would be l
adequate to monitor i!:0 approach to critical using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.
l S.
-The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal.
Two channels are provided, of which one may be bypassed for_not more than 24 l
hours without significantly increasing the risk from a rod withdrawal error.
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 written sequences.
Amendment No.
90
~._
3.3 and 4.3 3,
BASES:-
During reactor operation with aslimiting control rod pattern,1the reactor in run and reactor power greater than 29% of rated power, RBM opercbility is _ required to prevent _ the withdrawal of a single control rod which could result.in HCPR being decreased-below the MCPR safety limit.
A limiting control rod pattern exists when-MCPR is below a threshold MCPR-which does not provide adequate. margin to
,,~~-
the safety limit HCD1 in the event a' rod is fully withdrawn - When RBM operability is required and:one RBM channel is inoperabl en
,,s instrument functional test of the operable RBM channel' to control rod withdrawal will provide adequate assuyan that improper'
~
withdrawal does not occur.
C.
Scram Insertion Times The control rod system.is designed-to bring the reactor suberitical at a.
rate fast enough to prevent fuel damage;Li.e.,=to prevent the MCPR-_from; becoming less_than the Safety Limit MCPR.
Analysis of the111miting power transient shows that the-negative reactivity rates-resulting.-from_the scram with the average. response of. all the drives as given in =the above Specification, provide the required protection, and-MCPR remains greater than the Safety Limit MCPR.
The scram times.for all-control rods will be' determined at the time ofi each refueling outage.
A representativersample of. control rods will be:
1 scram tested during each cycle as a periodic _ check against-deterioration i
3 of the control rod performance.
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1 4
J Amendment No. 42 gl-1-
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+
4.
n,
p' LIMITING CONDITIONS _ LOR OPERATION e ,. '
SURVEILLANCE REOUIREMENTS y
L C.
Minimum Critical Pp er' Ration MCPR C.
Minimum Critical Power Ration MCPR (Cont'd)../, / ~'
-(Cont'd)
,/
/
2.
The operating limit MCPR values b)
The average scram time to the as a function of the t are given 30% insertion position is in Table 3.3.1 of the Core determined as follows:-
Operating Limits Report where t is given by specification n
4.11.C.2.
E Ng tj save 1-1 n
E Ni 1-1 Where: an n - number of surveillance tests performed to date in-the cycle.
l Ni - number of active control rods-l measured in the-ith surveillance test.
l ti - average scram time to the 30%
a insertion position of all rods measured in the ith I
surveillance test.
c)
The adjusted-analysis mean scram time (tg) is calculated as follows:
tB p + 1.65 o
n ENj 1-1 Where:
p - mean of the distribution for average scram insertion time-to the 30% position. 0.945 sec.
Ni - total number of active-control rod =
o - standard deviation of the distribution for average scram' insertion time to the 30% position 0.064 sec.
Amendment No. 54, 133 205c
6.9.A Routine ReDorts (Cont 10 Nil).
3.
Qg_gspational Ernosure Tabulation A tabulation of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g. reactor operations and surveillance inservice inspection, routine maintenance, special maintenance (including a description), waste processing, and refueling shall be submitted on u
an annual basis.
This tabulation supplements the requirements of 20.407 of 10 CFR 20.
The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or flim badge measurements.
Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.
4.
Core.0peratinL!imits Reoort a) Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle, b) The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (the approved version at the time the reload analyses are performed shall be identiflad in the CORE OPERATING LIMITS REPORT),
l in NE00-21696, " Loss of Coolant Analysis Report for Pilgrim Nuclear Power Station," dated August 1977, (the approved version at the time the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT), and in NEDC-31312-P, " ARTS Improvement Program Analyses for Pilgrim Nuclear Power Station", dated September 1987, (the approved version at the time the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT).
c) The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) are met, d) The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
6.9.B Deleted i
l l
Amendment No. 88, 122, 133 217a
i
.i Attachment C Technical Specification Markun for ARTS 1morovement Procram 1
i j
'/I? i f
LIMl?tN M ONDITIONS FOR OPERAff0N SURVEILLANCE REOUTREMENTS l
3.1 REACTOR PPOTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Arolicability:
ADD 11cability:
Applies to the instrumentation Applies to the surveillance of and associated devices which the instrumentation and initiate a reactor scrar associated devices which initiate.
reactor scram.
Obiective:
Obiective:
To assure the operability of the reactor protection system.
To'soecify the type and. frequency of surveillance to be applied to Seecification:
the protection instrumentation.
The setpoints, minimum number of SDecification:
trip systems, and minimum number of instrument channels that must
--Instrumentation systems shall be be operable for each position of functionally tested and the reactor mode switch shall be calibrated as indicated in Tables as given in Table'3.1.1. The 4.1.1 and'4.1.2 respectively.
system response times from the C
opening of the sensor contact up B.-
Verify the maximum fraction o
\\
}
to and including the opening of limiting power density i ss the trip actuator contacts shall than or equ e
action of-not exceed 50 milli-seconds.
rated p r once wi hin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> aftey hermal power is greater
[
B.
The maximum fraction of lim -ting tbah or equal to 25% of rated f
j power density (HFLPD) sha be -
/ hermal power and every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
[
less than or equal to t thereafter.
fraction of rated =powe (FRP)
~
when thermal power is greater than or equal to 25 of rated 7
thermal power.
1.
If MFLPD i greater-than FRP, adjust t APRH high flux scram d rod block trip setpc s to the rela onships specified in the ORE OPERATING LIMITS RE RT within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
2.
he required actions and associated completion times of Specification 3.1.B.1, above cannot be met, reduce j
thermal power to less than 25% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Amendment No d2,129, 26 Ctv
. 4 M W's M w eg **d u te -w e M ww w an-mem a a g
4 e.
3.1
[LAiG (Cont'd) 4.1 BAlfj (Cont'd) been used in transient analyses A comparison of Tables 4.1.1 and dealing with coolant inventory-4.1.2 indicates that two I
decrease. The results show that instrument channels have not been scram at this level adequately included in the latter Table.
protects the fuel and the These are: mode switch in-pressure barrier, because MCPR shutdown and manual scram. All remains well above the safety of the devices or sensors limit MCPR in all cases, and associated with these scram system pressure dces not reach functions are simple on-off-
,1 the safety valve settings. The switches and, hence, calibration scram setting is approximately 25 during operation is not in, below the normal operating
- applicable, i.P., the switch is range and is thus adequate to either on or off.
avoid spurious scrams.
B.
/ The Marimum Fraction of limiti g Turbine Stoe Valve Closure fPowerDensity(HFLPD)shal e
I checked once per day t etermine The turbine stop valve closure if the APRM scram r ires scram anticipates the pressure, adjustment. This ill normally neutron flux and heat flux be done by chec ng the LPRH increase that could result from readings a small number of rapid closure of the turbine stop contr rods moved daily and j
valves. Hith a scram trip th the HFLPD is not expected to
[
5etting of 1 10 percent of valve e ange significantly and thus a closure from full open, the ally check of the HFLP0 is resultant increase in surface adequate.
heat flux is limited such that u
HCPR remains above the safety The sensitivity of LPRH detectors limit MCPR even during the worst decreases with exposure to case transient that assumes the neutron flux at a slow and-turbine bypass is closed, approximately constant rate.
This is compensated for in the Turbine Control Valve Fast Closure APRM. system by calibrating every three days using heat balance The turbine control valve fast data and by calibrating l
j closure scram anticipates the individual LPRH's every 1000 pressure, neutron flux, and heat effective full power hours using flux increase that could result TIP traverse data, from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves.
The reactor protection system initiates a scram when fast closure of the control valves is initiated by the acceleration relay. This setting and the fact that control valve closure time is approximately twice as long as k
ln, T
t 3.1 S115 (Cont'd) biased rod block and scram. The IRM's provide additional protection in the " Refuel" and "Startup/ Hot Standby" modes. Thus, the IRH and APRM 15%
scram are required in the " Refuel" and "Startup/ Hot Standby" modes.
In the power range the APRM system provides the-required protection. Ref.
Section 7.5.7 FSAR. Thus, the IRH system is not required in the "Run" mode.
The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume blgh' level-scrams are required for Startup/ Hot Standby and Run modes of plant operation.
They are, therefore, required to be operational for these modes of reactor operation.
The requirement to have the scram functions, as indicated in Table 3.1.1, operable in the Refuel mode is to assure that shifting to the Refuel mode during reactor power operation does not diminish the need for the reactor protection system.
The turbine condenser low vacuum scram is only required during power operation and must be bypassed to start up the unit. Below^305 psig turbine first stace pressure (45% of rated), the scram signal due to W d*M turbine stop valve closurejis bypassed because flux and pressure scram are OTT p 7dequatetoprotectthereactorg herequirementthattheIRH'sbeinsertedinthecorewhentheAPRH's read 2.5 indicated on the scale assures that there is proper overlap in the_ neutron monitoring systems and thus, that adequate coverage is provided for all ranges.of react 9r operation.
The provision of an APRH scram at 515% design power in the "Refu(l" and "Startup/ Hot Standby" modes and the bukup IRH scram at 1120/125 of full h
scale assures that there is proper nvorlap in the neutron monitoring
[
systems and, thus, that adequate icyorage 13 provided for all ranges of reactor operation, n
0-The scram trip setting must be adjusted to ensure that the LH M peak is not increased for any combination of maximumpactT6'n of limiting power density (HFLPD) 3a 4-u ret r 7,
rmal r.
The scram setting is adjusted in apertrance with the formula the CORE OPERATING LIMITS-1 l
REPORT wheMlle HFLPD is greater than the
. ction of rated power (FRP).
i In a jimtTar manner, the APRM rod block trip setting is: adjusted downward TLPD exceeds FSP, thus preservir,g the APRM rod block safety margin.
+w,-p 4 m m, A pr<> of h powu.r, Au mr.,m.h ruf A M fnof L.wG A o fQ NIh Amendment No 40c l
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} L PS P Pt4P5 1 ABLE 3.2.f.-1 h-I_NSTR..t__IMTNTAI]ON THAT INITI ATES CONTRot Roo BttXKS WNG Minimum Operable Channels Required per Trip Function Operational Conditionj u
Irip funsg gn
'I 4
Run (1) l APRM tfpscale (Flow Biased) 4 Startup/ Refuel (1)
APRM tipicale (Setdown) 4 Run/Startup/ Refuel (1)
AP9M Inoperative 4
Run (1)
APRM Downscale g
O 0
(2) l
) (g) ", _. _-srvi i y ; t, ; - e ? '-
2 t
Rod Block Monitor fH 3;.>e4 %
l
_.________,w, r-t-4 " -- St (2)
Rod Block Monitor Inoperative 2
( d -"
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- :: : ;: - s 3 'nv
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Startup/ Refuel, e= cept trip is bypassed (1) i when IRM is on its lowest range i
6 Startup/ Refuel, trip is bypassed (1)
[
I IRM Detector not in Startup Position when mode switch is placed in Run G
Startup/ Refuel (t) l IRH (fpscale IRM Inoperative 6
Startup/ Refuel (1)
SRM Detector not in Startup Position 3
Startup/ Refuel. e= cept trip is bypassed (1)
'*i when SRM count rate is 1 100 counts /
" [
second or IRMs on Range 3 or above (4) i'
}
SRM Dewnscale 3
Startup/ Refuel. esceot trip is bypassed (1)
[
when IRMs on Range 3 or above (4) l SRM t)pscale 3
Startup/Pefuel, e= cept trip is t*ypassed (t)
I when the IRM rasge switches are on Range 8 or above (4)
SRM Incperative 3
Startup/ Refuel. += cept trip is (1) bypassed when the IRM range switches are on Range 8 or above (4) 2 Sce am Discharge Instramnent Volume Water level - High 2
Run/Startup/ Refuel (3)
.. s.am Di n tiarrte Ins t reame nt Volmae - Scram T rip Hypasse.d I
fon/Startup/ Refuel.
(3)
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Egtes for-Table 3.2.C-1 1.
With the number of operable channels:
a.
One less than required by the minimum operable channel's per. trip function requirement, restore an inoperable channel.to operable status within 7 days or place an inoperable channel in the tripped condition within the next hour.
I b.
Two or more less than required by the minimum operable channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour.
_y
,,4
% A'W' 2.
The minimum operable channels for the rod block monitor trip function may
/
be reduced by one for maintenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period.
3.
If the number of operable channels is less than required by the minimum operable channels per trip function requirement, place the inoperable channel in che tripped condition within one hour.
4 SRM operability requirements during core alterations are given in Technical Specification 3.10.
C.
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i PNPS TABLE 3.2.C-2 CONTROL ROD BLOCK INSTRUw!NTATION SEtPOINTS TriD Function triD SetDoint 1
APRH Upscale (1) (2)
APRM Inoperative Not Applicable APRH Oownstale 1 2.5 Indicated on Scale few</D*fdd Rod Block Honitor U!n hed^r (1)(~5) l Rod Block Monitor inoperative Not Applicable
( Q U)
Rod Block Honitor Downstale
- F.ni25 vi ivii 5uie(
IRH Downscale 1 5/125 of Full Scale IRH Detector not in Startup Position Not Applicable-IRH Voscale 1 108/125 of Full Scale IRH Inoperative Not Applicable SRH Detector not in Startup Position Not Applicable 1
SRH Downstale 2 3 counts /second 5 counts /second-SRM Upscale 1 10 SRH Inoperative Not Applicable Scram D,aharge Instrument Volume 1 18 gallons Water Level - High Scram Discharge Instrument Volume -
Not Applicable Scram Trip Bypassed Recirculation Flow Converter - Upscale 1 120/125 of Full Scale Recirculation Flow Converter -
Not Applicable Inoperative-Recirculation Flow Converter -
1 101 Flow Deviation for > 80%
Comparator Hismatch Rated Power, and i 15% Flow Deviation for 1 80%
Rated Power (1) The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT, (2) When the reactor mode switch is in the refuel or startup positions, the APRH rod block trip setpoint shall'be less than or equal to 13% of rated thermal power, but always less than the APRM flux scram trip setting.
j G) A RBM %gr b% CW,)St.M ke. c 2.oM 55a Amendment No. 42, !!0. 129 x
v/o
3.2 8.Al[j(Coat'c)
The control rod block functions are provided to prevent excessive [
control rod withdrawal a tu t C da ; a t de:r;a ; to th; O fet q -li"'* W The trip logic for this function is I out of n:
e.g., any trip on one of sir APRM's, eight IRH's. or four SRH's will result in a rod block.
No gomg The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBH may be reduced by
'i fw ma h m e. te:t %, w c61;Lietien. TM: tim; period : ca L J y21 of t'e ope nt'ag-tir: ' --o month ed do;; not4significantly increah"-l T
I N
the risk of pr;;inting an inadvertent control rog withdrawal.
j
. @W ITartor power T m i may-be verted vy m ving control roos or oy varyir Ithe recirculation flow rate. The APRM system provides a control r alock to present rod withdrawal ie:or.d a given point at constan ecirculation flow rate, and th% do protect against the con on of a M,yp 4CPR less than the Safety Limit MCPR.
This rod block setg oint, which is automatically varied with recirculation loop flow r te, prevents an increase in the reactor power level to excessive y es due to control rod withdrawal. The flow variable trip setting ovides substantial margin from fuel damage, assuming a steady-stA e operation at the trip setting, over the entire recirculation floy dange. The margin to the r
safety limit increases as the flow decreues for the specified trip setting versus flow relationship; t e, the worst case HCPR which
- ould occur during steady-stat er ion is at 107% of rated thermal ower because of the APRH of block setting. The actual power istribution in the cp is established by specified control rod equences and is pofatored continuously cy the in-core LPRM system. _As ith the APRH < ram trip setting, the APRH rod block trip setting is i justed d ward if the maximum fraction of limiting-power density a ceed / e fraction of rated power,_thus preserving the APRM rod block s
afet>f margin, In the startup and refuel modes, the APRM rod bloc k....
(g (tion 3 utdown halnw th2 a o"" f ' u uram trip. -
The RBH rod block function provides local protection of the core, for a single rod withdrawal error from a limiting control rod pattern.
- yJc 0 &The IRH rod block function provides local as well as gross core protection.
The scaling arrangement is such that trip setting is less i
than a factor of 10 above the indicated level.
4TA0rn) t I
A downstale indication on an APRHfor IRM is an indication the instrument has failed or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.
The downscale trips are as shown in Table 3.2.C-2.
t Amendment No. Ib, 42, 110, 129, 71
$7
i insert A to Pace 71
.. -:/,
Reactor power tl,WsY may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod tjock to prevent rod withdrawal beyond a given point, thereby possibly avoiding an APRH scram. The rod block setpoint is automatically reduced with recirculation flow to form the upper boundary of the PNPS power / flow map. The flow-biased APRM rod block is not necessary to prohibit fuel damage and W n6 tic ~KiTIT&3 6 the
~ '9 analysis of anticipated transients.
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LiHITING CONDITION FOR OPERATION
_ SURVEILLANCE REOUIREMENT 3.3.B Control Rodi 4.3.8 Control Rods 4.
Control rods shall not be 4.
Prior to control rod withdrawn for startup or withdrawal for startup or refueling unless at least two during refueling, verify that source range channels have an at least two source range observed count rate equal to channels have an observed or greater than three counts count rate of at least three per second.
counts per second.
% Rom S 4 k opam W M.
s eith r e h T When a limiting control od 9 W @aaeraticacontrol rod M t M #
pattern exists, an rument 5.
ji't r"~ N de functional test the RBH
^ A h $ h ;termind by a in;;r,-eith:ri shall b d prior to wit wal of e destgrsted j
+ R t' onu channe!: tha!' W p g W r (s) and daily thereafter.
opr.aM e - ers pW, J
A-Gentrel red withdswat p_ n u_ %..., w. _
n i
. Thi operating power level shall be limited so that ugTE the HCPR will remain above the Safety Limit MPCR assuming,J etiqiTe e thayrtrsuits in plet withdrawal of a e
C.
Scram Insertion Times C.
Scram Insertion Times 1.
The average scram insertion 1.
Following each. refueling time, based on the outage, or after a reactor-doenergization of the scram shutdown that is greater than pilot valve solenoids as time 120 days, each operable zero, of all operable control control rod shall be subjected rods in the reactor power to scram time tests from the operation condition shall be fully withdrawn position._ If no greater than:
testing is not accomplished with the nuclear system
% Inserted Average Scram pressure above 950 psig, the-(
From Fully Insertion measured ~ scram insertion time Hithdrawn Times (set) shall be extrapolated to reactor pressures above 950 10
.55 psig using previously 30 1.275 determined correlations.
50 2.00 Testing-of all operable 90 3.50 control rods shall be completed prior to exceeding 40% rated thermal power.-
-4 M
Revision 129 AmendmentNo.15.68,(([
83 j
- .h.
"N
1 3.'3 and k.3 I
PASES:
h.
TheSourceRe.ngeMonitor(EPR)systemperforms i
no automatic safety system function; i.e., it l
has no scram function. It does provide the oper-ator with a visual indication of neutron icvel, a
The consequences of reactivity accidents are l
i l
functions of the initial neutron flux. The l
requirement of at least 3 counts per second I
assures that any transient, should it occur beginsatorabovetheinitialvalueof10bof rated power used in the analyses of transients from cold conditions.
One operable SRM channel vould be adequate to monitor the approach.to criticality using homogeneous patterns of scat-tered control rod withdrawal. A minimus of two operable SPR's are provided as an e.dded con-servatism.
5.
The Rod Block Monitor (P3M) is designed to auto..
matically prevent fuel damage in the event of erroneous rod withdrawal ^ :: 1;=ine of hi p rer den:it; during high p=r le < e'. c o a... %
- 'N so..s '...- c pr cid:d, 2nd = cf :h:: - ny "
~
- b: t m s s;d f..
- '.m wmsie_ im W.s.mucc j -.ni/c tcsting.f Tripping of one of Ethe channels vill block erroneous rod withdrawal soon enough to prevent fuel damage. - This system backs up the operator who withdraws control rods according to written sequences. The spccifie '.inri; a
+hne,dth r e%,,,
- a
- e. ft;; c;;;; 7.
tiis1/ nas.
that fu 1 m ge wil' n:t ea-"*
A"- t: ::d withA'2 ;l m.a a as. :h h ;nditi:n -
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w 3ASES During reactor operation with certain 11mied 1
- py
- T)"
control rod pattens, the withdrawal of sig-nated single control rod could result in one or k
nors fual rods with MCP1's less than the saf try 1.imit MCF1. During use of such patteru, it is y
judged that testing of the 13M system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not It is the responsibility of the Reactor
\\
occur.
Eosinear to identify these limiting pactarns and the designatad rods either when-the pactarus are initially ascab11shed or as they develop due-to the occurrence of inoperabia control nMa in
( othat than limiting patterns.
C.
Seram T.nsartien Times ne control rod system is designed to bring the tsac-ter suberitical at a rata f ast enough to prevent' fuel damags; i.e., to prevent the MCPR from becoming less than the Safety 1.imit MC?R. Analysis of the limiting power transient shows that the negative reactivity-ratas resulting from the scram with the average responsa of all the drives as given in the abova Specification, provide the recuired protection, and MCPR r dna graatar than the Saf ety 7 d #t MCP1.~
na scram timas.for all control rods will be datar-mined at the cima'of each refueling outage. 1A re' prasantativa sample of control rods vill be scram ~ tastad duri=g each cycle as a periodic check against decarioration -
of the control rod: performanca.
A=end=ent No.-
91
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hfWl?!NG CONDITIONS FOR OPERAVIOS SURVEILLANCE Rf0VIRMEN75 C.
Minimum Critical Power Ratio MCPR C.
Minimum Critical Power Ratio MCPR (Cont'd)
(Cont'd) for core flows other than rated b) The average scram time to the MCPR limits shall be the lik s 301 insertion position is identified above times K here Kr is determined as follows:
as shown in Figure 3.
of the Core n
Operating Limits Re rt,
[ Nj tj j
/
t,y,.
i.I
+
As an altern ethod providing I
n equivelept-th fma hydraulic E Nj protection at flows other than i=1 ratd,thecalculatedMCPRmaybe d fir'ed by Kr. where Kr is as shown in ; Where: cn n = number of $Urveillance igure 3.3.1 of the Core Operating tests performed to date in the cycle.
Limits Report.
L Ni. number of active control rods 2.
The operating limit HCPR values as measured in the'ith surveillance a function of the t are given in test.
Table 3.3.1 of the Core Operating ti average scram time to the 30%
Limits Report where t is given by insertion position of all rods specification 4.ll.C.2.
measured in the lith surveillance test.
c)
The adjusted analysis-mean scram time (t follows:g) O calculated as
~
~
hj tg p + 1.65 o
n ENj
_i.1 _
Where:
-p
. raean of the -distribution for average scram insertion time to the 30% position 0.945 sec.
Ni total number of active control rod o.-standard deviation of the distribution for average scram insertion time to the 301 position. 0.064 sec.
-Amendment No.
59, 205c 1
L C[D z
6.9.A Routine ReDorts (Con?inued) 3.
Occueational Ereosure Tabulation A tabulation of the number of station, utility and other personnel-(including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work antt job functions, e.g. reactor operations and surveillance inservice inspection, routine maintenance, special maintenance (including a description), waste processing, and refueling shall be submitted on an annual basis.
This tabulation supplements the requirements of i
20.407 of 10 CFR 20.. The dose assignment to various duty functions may be estimtes based on poctet dosimeter, TLD, or film badge measurements.
Small exposures totalling less than 207,of the individual total dose need not be accounted for.
In the aggregate, at least 80Y of the total whole body dose received from external-sources shall be assigned to specific major work functions.
4.
Core Ooeratina Limits Reoort a) Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle, b) The analytical methods used to determine.the core operating limits shall be those previously reviewed and approved by the NRC in NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," (the approved version at the time the reload analyses are erformed shall be identified in the CORE OPERATING LIMITS REPO )
.s4)in NE00-21696, " Loss of Coolant Analysis Report for Pilgrim aclear Power Station," dated At. gust 1977, (the approved version at the time the reload analyses are performed shall-be identified in the CORE OPERATING LIMITS REPORT)g out. A us CE -3l7(24,' A/2TJ c) The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits-such as shutdown margin, and transient and accident analysis limits) are met, d) The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto. shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
6.9.8 Deleted m
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l ATTACIIMENT D GE Report. NEDC-31312-P (Orn-Copy. Proprietary) 1 B
E 1
General Eiectric Company AFFIDAVIT I, Janice S, Charnley, being duly sworn, depose and state follows:
m I am Manager, Fuel I.icensing, General Electric Company, and have been delegated the func-1.
tion of reviewing the information described in paragraph 2 which is sought to be withheld fron public disclosure and have been authorized to apply for its withholding.
The information sought to be withheld is contained in ARTSimprovemmt Prognun Analysirfo, 2.
Pilgn'm NuclearPowerStation (PNPS), NEDC 31312 P, September,1987.
3.
In designating material u proprietary, General Electric utilizes the definition of proprietary ir formation and trade secrets set forth in the American 1.aw Institute's Restatement of Torts, Section 757. 'Ihis definition pioMes:
"A trade secret may consist of ar y formula, pattern, device or compilation ofinfor-mation which is used in one'a b'4smess and which gives him r.n opportunity to obtain an advantage over competitors who do not know or use it... A substantial element of secrecy must exist, so that, except by the ese ofimproper means, there would be difficulty in acquiring information.... Some factors to be considered in determining whether given information is one's trade secret are: (1) the extent to which the infonnation is known outside of his business; (2) the extent to which it is known by employees and others involved in his business; (3 the extent of mea-sures taken by him to guard the secrecy of the information;)(4) the value of the in-formation to him and to his competitors; (5) the amount of effort or money ex-pended by him in developing the information; (6) the ease or difficulty with which the information could be properly acquired or duplicated by others?
4.
Som,e examples of categories of information which fit into the definition of proprietary infor mation are:
information that disclosed a process, method or apparatus where prevention of its use a
by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies; b
Information consisting of supporting data and analyses, including test data, relative to a process, method or apparatus, the application of which provide a competitive economic advantage, e.g., by optimization or improved marketability; (n'ormation which, if used by a competitor, would reduce his expenditure of resources c.
or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality or Ucensing of a similar product; d.
Information which reveals cost or price information, production capacities, budget level:
or commercial strategies of General Electric, its customers or suppliers; Information which reveals aspects of past, present or future General Electric cus-e.
tomer-funded development plans and programs of potential commercial value to General Electric; 1
p
....c f.
Information which discloses patentable subject matter for which it may be desirable to obtain patent protectio ';
g.
Information whien Ocneral Electric must treat as proprietary according to agreements with other pa; ties.
5.
Initial approval of proprietary treatment of a document is typically made by the Subsection manager of the originating component, who is most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within the Company is limited on n 'nced to know" basis and such documents are cicarly identified as proprietary.
6.
The procedure for approval of external release of such documentation typically requires review by the Subsection Manager, Project manager, Princi3al Scientist or other equivalent authority, by the Subsection Manager of the cognizant Mar ceting function (or delegate) and by the legal Operation for technical content, competitive effect and determination of the ac-curacy of the proprietary designation in accordance with the standards enumerated above.
Disclosures outside General Electric are genen.lly limited to regulatory bodies, customers and potential customers and their agents, suppliers and licensee: and then only with appro.
priate protection by applicable regulatory provisions or proprietary greements.
7.
The documentation mentioned in para graph 2 above has been evaluated in accordance with the above criteria and procedures and has been found to contain information which is propri-etary and which is customarily held in confidence by General Electric.
)
8.
The information to the best of my knowledge and belief has consistently been held in confi.
dence by the General Electric Company, no public dir r te has been made, and it is not available in publ"ources. All disclosures to third iini have been made pursuant to regulatory providons or proprietary agreements wha : covide for maintenance of the Infor-mation in confidence.
9.
Public disclosure of the information sought to be withheld is likely to cause substantial harm to the competitive position of the General Electric Company and deprive or reduce the availability of profit making opportunities because it would provide other parties, including competitors, with valuablelnformation.
w 2
l
r
~
t STATE OF CALIFORNIA
)
COUNTY OF SANTA CLARA ) 88 Janice S. Charnley, twing duly sworn, deposes and says:
That she has read the foregoing affidavit and the matters stated therein are true and correct to the best of b's knowledge, information, and belief.
Executed at San Jose, California, thi.
day of h 1990.
0 rA W
ce S. CiarnTUy eneral Electric Company Subscribed and sworn befoat ;a this day of hp,k(1990.
3 OFFICI AL SEAL
=
NOTA C'
- ANIA b
18SM-l J
swn cum coum gs
- ~~
~ #7 " " *'" #" 8-) $,
Notary Public California 4
Santa Clara County 3-
-