ML20097H746
| ML20097H746 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 09/06/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Boston Edison Co |
| Shared Package | |
| ML20097H749 | List: |
| References | |
| DPR-35-A-079 NUDOCS 8409200478 | |
| Download: ML20097H746 (14) | |
Text
_____ _______ ________ _ _ _
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[d NUCLEAR REGULATORY COMMISSION UNITED STATES ti j
WASHINGTON, D. C. 20555
%...../
BOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION l
AMENDMENT TO FACILITY OPERATING LICENSE
{
Amendment No. 79 License No. DPR-35 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Boston Edison Company (the licensee) dated June 26, 1984 complies with the standards and l
requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate' in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to.he common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPP.-35 is hereby amended to read as follows:
e40920o47s s40906 PDR ADOCK 05000 e
r f
f l B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 79, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
'~, :'j: '
,L,'
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: September 6,'1984
ATTACHMENT TO LICENSE AMENDMENT N0. 79 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the following pages of the Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 30 30 31 31 32 32 33 33 34 34 35 35 36 36 37 37 38 38 39 39 54 54 4
i i
t 4
(
Si TAllLE 4.1. I REACTOR PROTECTION SYSTEM (SCHAtl) INSTRllHENTATION FilNCTIONAL TESTS k
HINitillli FilNCTIONAL TEST FREQllENCIES FOR SAFETY INSTR. ANL) CONTROL CIRCUITS
- s
[
Group (2)
Functional Test Minimum Frequency (3) 5
, ( Hode Switch in Shutdown A
Place Mode Switch in Shutdown Each Refueling Outage e
llanual Scram A
Trip Channel and Alarm Every 3 Honths RPS Channel Test Switch (5)
A Trip Channel and Alarm Each Refueling Outage IkN liigh Flux C
Trip Channel and Alarm (4)
Once Per Week During Refueling and Before Each Startup Inoperative C
Trip Channel and Alarm Once Per Week During Refueling and Before Each Startup APRM High Flux B
Trip output Relays (4)
Once/ Week (7)
Inoperative B
Trip Output Relays (4)
Once/ Week Downscale 11 Trip Output Relays (4)
Once/ Week Flow Biase 11 Calibrate Flow Ilias Signal Once/flonth (I) liigh Flux (15%)
11 Trip output Relays (4)
Once Per Week During Refueling and Before Each Startup liigh Reactor Pressure A
Trip Channel and Alarm (1) liigh I)rywell Pressure A
Trip Channel and Alarm (1)
Reactor I,ow Water Level (6)
A Trip Channel and Alarm (1) liigh Water Level in Scram Discharge Tanks D
Trip Channel and Alarm Every 3 Honths l
Turbine Condenser Low Vacuum A
Trip Chasisiel and Alarm (1) flain Steam Line liigh Radiation 11 Trip Channel and Alarm (4)
Once/ Week Hain Steam Line Isolation Valve Closure
'A*
Trip Chanael and Alarm (1)
Tnibine Control Valve Fast Closure A
Trip Channel and Alarm (1)
Turbine First Stage Pressure Permi.ssive A
Trip Channel and Alarm Every 3 Honths Turhine Stop Valve Closure A
Trip Channel and Alarm (1)
Reactor Pressure Permissive A
Trip Channel and Alarm Every 3 Honths
- E$
NOTES FOR TABLE 4.1.1 1.
Initially 3once per month until exposure (H as defined on Figure 4.1.1) is 2.0 x 10 ; thereafter, according to Figure 4.1.1 with an interval not less than one month nor more than three months.
The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of PNPS.
t i
i 2.
A description of the four groups;is included in the Bases of this Speci-fication.
a 3.
Functional tests are not required when the' systems are not required to be operable or are tripped.
If tests are missed, they shall be performed prior to returning the systemn to an operable status.
4.
This instrumentation' is exempted from the instrument channel test defini-tion.
This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.
5.
Test RPS channel after maintenance.
6.
The water level in the reactor vessel will be perturbed and the corres-ponding level indicator changes-will be monitored.
This perturbation i
test will be performed every month after completion of the monthly func-l tional test program.
7.
This APRM testing will be performed once per week when in the run mode.
If the reactor is out of the.run mode for more than one week, the testing will be performed as soon as practicable after returning to the run mode.
I i
i Amendment ?!o. 79 31
k TAllLE 4.1.2 I
i REACTOR PROTECTION SYSTEN (SCRAH INSTRilNENT CALIBRATION 2
HINIHtlH CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUNENT CllANNELS a
Instrument Channel Group (1)
Calibration Test (5)-
Hinimum Frequency (2) 2
?
IRM liigh Flux C
Comparison to APRH on Controlled Note (4) y Shutdowns 11 alibration Once/ Operating Cycle hPRH High Flux Output Signal B
lleat Balance Once every 3 Days Flow Bias Signal B
Internal Power and Flow Test Each Refueling Outage LPRH Signal B
TIP System Traverse Every 1000 Effective Full Power llours High Reactor Pressure A
Standard Pressure Source Every 3 Honths liigh Drywell Pressure A,.
Standard Pressure Source Every 3 Honths Reactor Low Water Level A
Prgssure Standard Every 3 Honths liigh Water Level in Scram Discharge Tanks D
Note (7)
Note (7)
Turbine Condenser Low Vacuum A
Standard Vacuum Source Every 3 Honths Main Steam Line Isolation Valve Closure A
Note (6)
Note (6)
Hain Steam Line High Radiation B
Standard Current Source (3)
Every 3 Honths Turbine First Stage Pressure Permissive A
Standard Pressure Source Every 6 Honths Turbine Control Valve Fast Closure A
Standard Pressure Source Every 3 Honths Turbine Stop Valve Closure A
Note (6)
Note (6)
Reactor Pressure Permissive A
Standard Pressure Source Every 6 Honths M
NOTES FOR TABLE 4.1.2 1.
A description of four groups is included in the bases of this Specifi-cation.
2.
Calibration tests are not required when the systems are not required to be operable or are tripped.
3.
The current source provides an instrument channel alignment.
Calibration using a radiation source shall be' made each refueling outage.
5.
Maximum frequency required is once per week.
5.
Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle.
6.
Physical inspection and actuation of these position switches will be performed during the refueling outages.
7.
Calibration of these devices will be performed during refueling outages.
1 1
l
~
l e
Amendment flo. 79 33
.A
. BASES:
BASES:
3.1 The reactor protection system 4.1 A.
The minimum functional test-automatically initiates a reac-ing frequency used in this tor scram to:
specification is based on a reliability analysis using 1.
Preserve the integrity of the concepts developed the fuel cladding.
in reference (6).
This con-cept was specifically adapt-2.
Preserve the integrity of ed to the one out of two the reactor coolant system.
taken twice. logic of the reactor protection system.
3.
Minimize the energy which The analysis shows that the must be absorbed following sensors are primarily respon-a loss of. coolant accident sible for the reliability and prevents criticality.
of the reactor protection system.
This analysis makes This specification provides the use of " unsafe failure" rate limiting conditions for opera-experience at conventional tion necessary to preserve the and nuclear power plants in ability of the system to tole-a reliability model for the 4
rate single failures and still system.
An " unsafe failure" perform its intended function is defined as one which ne-i even during periods when in-gates chanael operability strument channels may be out and which, due to its na-o'f service because of mainte-ture, is revealed only when j
nance.
When necessary, one.
the channel is functionally channel may be made inoperable tested or attempts to re-for brief intervals to conduct spond to a real signal.
required functional tests and -
Failures such as blown calibrations.
fuses, ruptured bourdon cubes, faulted amplifiers, The reactor protection system faulted cables, etc. which is of the dual channel type.
cesult in " upscale" or Ref. Section 7.2 FSAR.
The "downscale" readings on the system is made up of two in-reactor instrumentation are j
dependent trip systems, each
" safe" and will be easily having two subchannels of recognized by the operators tripping devices.
Each sub-during operation because channel has an input from at they are revealed by an least one instrument channel alarm or a scram.
which monitors a critical parameter.
The channels listed in Tables 4.1.1 and 4.1.2 are The outputs of the subchannels divided into four groups are combined in a 1 out of 2 for functional testing.
logic; i.e., an input signal These are:
on either one or both of the subchannels will cause a trip A.
On-Off sensors that system trip.
The outputs of provide a scram trip the trip systems are arranged function.
so that a trip on both systems is required B.
Analog devices enupled with bi-stable trips that provide a scram I""CEi "*
Amendment Mo. 79 34
3.1 BASES (Cont'd) 4.1 BASES (Cont'd) to provide.a reactor scram.
This C.
Devices which only serve a system meets the intent of useful function during some EEI - 279 for Nuclear Power Plant restricted mode of operation, Protection Systems.
The system such as startup or shutdown, has r, reliability greater than or for which the only prac-that of a 2 out of 3 system and tical test is one that can somewhat less than that of a 1 be performed at shutdown.
l of 2 system.
j D.
Diverse Analog Transmitter /
With the exception of the trip unit devices that provide Average Power Range Monitor alarms, trips or scram functions.
(APRM) channels, the Inter-mediate Range Monitor (IRM)
The sensors that make up group channels of the Main Steam (A) are specifically selected Isolation Valve closure from among the whole family of 4
and the Turbine Stop Valve industrial on-off sensors that closure, each subchannel has have earned an excellent reputa-one instrument channel.
When tation for reliable operation.
the minimum condition for During design, a goal of 0.99999.
operation on the number of probability of success (at the operable. instrument channe19 50% confide' ace level) was adop-per untripped protection trip ted to assure that a balanced and system is met or if it cannot adequate design is achieved.
The be met and the affected Pro' probability of success is primar-tection trip system is placed ily a function of the sensor fai-in a tripped condition, the lure rate and the test interval.
effectiveness of the protecti~on A three-month test interval was system'is preserved; i.e., the plannq,d for group (A) sensors.
system can tolerate a single This is in keeping with good failure and still perform its operating practices, and satis-intended function of scramming fies the design goal for the the reactor.
Three APRM in-logic configuration utilized in strument channels are provided the Reactor Protection System.
for each protection trip system.
To satisfy the long-term objec-APRM's #1 and #3 operate con-tive of maintaining an adequate tacts in one subchannel and level of safety throughout the APRM's #2 and #3 operate con-plant lifetime, a minimum goal of tacts in the other subchannel.
0.9999 at the 95% confidence APRM's #4, #5, and #6 are level is proposed.
With the arranged similarly in the (1 out of 2) X (2) logic, this other protection trip system.
requires that each sensor have Each protection trip system an availability of 0.993 at the has one more APRM than is 95% confidence level.
This level necessary to meet the minimum of availability may..be maintained number required per channel.
This by adjusting the test interval as allows the bypassing of one APRM a function of the observed fail-
,per protection trip ~ system for ure history (6).
maintenance, testing or calibra-tion. Additional IRM channels (6) Reliability of Engineered have also Safety Features as a Func-tion of Testing Frequency, I.M. Jacobs " Nuclear Safety,"
. Amendment ?!o. 79 Vol. 9, No. 4, July-Aug.
1968, pp. 310-312
o 3.1 ' BASES (Cont'd) 4.1 SASES (Cont'd) been provided to allow for by-To facilitate the i=plenentatica passing of one such channel.
of this technique, Figure !. 1.1 The bases for the scram setting is provided to indicate an a::ro-for the IPJ!, AFF2!, high reactor priate trend in test interval.
pressure, reactor low water level, The procedure is as follows:
MSIV, closure, generator load rejection, turbine stop valve 1.
Like sensors are peeled closure and loss of condenser into one group for the vacuum are discussed in Speci-purpose of data acquisitien, fication 2.1 and 2.2 2.
The factor M is the exp:sure hours and is equal to the Instrumentation (pressure number of sensors in a switches) for the drywell are gr up, n, times the clapsed provided to detect a loss of t::e T (M = nT).
coolant accident and initiate the core standbv cooling equip-3.
The act glated nu-ber cf A high d'rywell pressure ment.
1 unsafe fa$ utes is p:otteu scram is provided at the same as an ordinate aga:nst M setting as the core cooling as an asseissa on Figure syste=s (CSCS) initiation to
' I 1-
~c:inimize the energy which must be accomodated during a loss 4.
After a trend is established, of coclant accident and to the appropriate monthly test prevent return to criticality.
interval t satisfy the goal This instrumentation is a will be the test interval to backup to the reactor vessel the left of the plotted water level instrumentation.
points.
High radiation levels in the
- 5., A test. interval of one ::ntb main steam line tunnel above will be used initially un-that due to the normal nitro.
til a trend is established.
gen and oxygen radioactivity is an indication of leaking Group (3) devices utili:e an ana-fuel.
A scram is initiated log sensor followed by an ampli-whenever such radiation level fier and a bi-stable trip circuit.
exceeds seven times normal The sensor and aeplifier are ac-background.
The purpose of tive c =p nents and a failure is this scram is'to reduce the almost always accc panied by an source of such radiation to alarm and an indication of the extent necessary to prevent the source of trouble.
In the excessive turbine contamina-event of failure, repair or sub-tion.
Discharge of excessive stitution can start ic=ed i a te ly amounts of radioactivity to An "as-ts" failure is one that the site environs is prevented
" sticks" mid-scale and is not by the air ejector off gas Capable of going either up or de-a monitors which cause an isola-in response to an out-of-11=:ts tion of the main condenser input.
This type of failure fer off-gas line, analog devices is a rare occur-rence and is detectable by an A reactor mode switch is pro-peratcr who observes that one vided which actuates or by-passes the various scram func-sienc1 does not track rSe other tions appropriate to the par-three. For purpose of analysis.
ticular plant operating status.
Ref. Section 7.2.3.7 FSAR.
Na 3+
3.1 BASES (Cont'd) 4.1 BASES (Cont'd)
The 2RM system and APRm (157,)
it is assumed that this rare scrar provide protection failure will be detected within ataitst excessive power levels two hours.
and short reactor periods in the startup and intermediate power The bi-stable trip circuit which ranges.
is a part of the Group (B) de-vices can sustain unsafe failures The control rod drive scram sys-.
which are revealed only on test.
j tem is designed so that all of Therefore, it is necessary to the water which is discharged test them periodically.
frem the reactor by a scram
~
can be accc=odated in the dis-A study was conducted of the in-charge piping.
strumentation channels included The two scram discharge volumes in the Group (B) devices to calculate their " unsafe" failure accorcedste in excess of 39 gal-rates.
The analog devices (sen-lens of water each and are at the low points of the scrs= discharge sors and amplifiers) are predict-ed to have an unsafe of less than 20 X 10 {ailure rate piping. No credit was taken for
~
failure /
these volu=es in the design of the hour.
The bi-stable trip cir-discharge piping as concerns the -
cuits are predicted to have an a= cunt of water which =ust be accorcodated during a scram.
unsafefailurerateoflessthan
~
2 X 10 failures / hour.
Con-During normal operation the scram sidering the two hour monitoring interval for the ana~ log devices discharge volume system is er.pty; as assumed above, and a weekly however, should it fill with' water, test interval for the bi-n.able the water discharged to the piping trip pircuits., the design reli-could not be accorcodated,which ability goal ~of 0.99999 is at-would result in slow scram times tained with ample margin.
or partial control rod insertion.
To preclude this occurrence, redun-The bi-stable device; are moni-dant and diverse level detection tored during plant operation to devices in the scram discharge record their failure history and instru=ent volumes have been pro-establish a test interval using vided which will alar = when water the curve of Figure 4.1.1.
There level reaches 4.5 gallons, ini-are numerous identical bi-stable tiate a control rod block at 18 devices used throurhout the gallons, and scram the reactor plant's instrumentation system.
when the water level reaches 39 Therefore, significant data on gallons. As indic::te? above, the failure rates for the bi-there is sufficient volume in stable devices should be accumu-the piping to accommodate the lated rapidly.
scram without impairment of the scram times or amount of The frequency of calibration of insertion of the control rods.
the APRM Flow Biasing Network This function shuts the reac-has been established as each tor down while sufficient volume remains to acco==odate the discharged water'and pre-cludes the situation in which a scram would be requested but not be able
. Amendment No. 79
3.1 BASES (Cont'd) 4.1 BASES (Cant'd) to perform its function adequately, refueling outage.
The flow bias-ing network is functiocally test-A source range monitor (SRM) sys-ed at least once per month and, tem is also provided to supply in addition, cross calibration additional neutron le. vel informa-checks of the flow input to the tion during start-up but has no flow biasing network can be scram functions.
Ref. Section made during the funct:enal test 7.5.4 FSAR.
The APRM's cover the by direct ceter reading.
There
" Refuel" and "Startup/ Hot Standby" are several instruments which modes with the APRM 15% scram, and must be calibrated and it will the power range with the flow bia-take several days to perfor= the sed rod block and scram.
The IRM's calibration of the entire network.
provide additional protection in While the calibration is be:ng the " Refuel" and "Startup/ Hot Performed, a zero flow signal Scandby" modes.
Thus, the IRM and will be sent to half of the APCM's resulting in a half scrs=
APRM 15% scram are required in the and rod block condition.
- Thus,
" Refuel" and "Startup/ Hot Standby" modes.
In the power range the if the calibration were perform-APRM system provides the required ed during operation, flux shap-protection.
Ref. Section 7.5.7 ing would not be possible.
Based FSAR.
Thus, the IRM system is not on experience of other generating required in the "Run" mode.
stations, drift of instrc=ents, such as those in the Flow Bias-The high reactor pressure, high ing Network, is not significant drywell pressure, reactor low.
and therefore, to evoid spurious water level and scram discharge scra=s, a cali ration frequency volume high level scrams are f each refueling outage is established.
required for Startup/ Hot Stand-by and Run modes of plant operation.
They are, there-Group (C) devices are active only fore, required to be opera-during a given portion of the tional for these modes of operational cycle.
For example, reactor operation.
the IRM is active during startup and inactive during full power The requirements to have the operation.
Thus, the,culy test that is meaningful is the one scram functions,as indicated in Table 3.1.1, operable in Perfor=ed just prior to shut-down the Refuel mode is to assure or startup; i.e.,
the tests that that shifting to the Refuel are perforned just prior ta use of the instrument.
mode during reactor power operation does not diminish Creup (D) devices, while similar the need for the reactor in descrip:icn to these in Group protection system.
(3), are different in use be:ause they (:be anale; :7anse:::cr/::::
The turbine condenser low uni: devices) previde 21:r s.
vacuum scram is only required trips or scram func:icns.
during power operation and An availability analysis is de: ailed must be bypassed to start up in NEDO-21617 (4/77).
the unit.
Below 305 psig turbine first stage pressure Surveillance frequencies far :he S T.
(45% of rated), the scram systa= instrumen:stien is dc:a:1ed in Amendnen: Nu=ber 65.
'F.
con-Amendment "o.
79 currence w::h :his sureeillance pre-
l J.1 SASIS (Cont'd) 4.1 BASES (Coct'd) signal due to turbine stop
.gra= is con:ained in the Safe:y valve closure is bypassed Evaluation Report and its associated because flux and pressure Technical Evaluation Repor: (TER-C-scram are adequate to pro-5506-66) da:ed 11/10/62.
tect the reactor.
Calibration frequency of the The requirement that the instrument channel is divided.
IRM's be inserted in the core into two groups. These are as when the APRM's read 2.5 in-follows:
dicated on the scale assures that there is proper overlap 1.
Passive type indicating in the neutron monitoring devices that can be compared systems and thus that ade-with like units on a contin-quate coverage is provided uous basis.
for all ranges of reactor operation.
2.
Vacuum tube or semiconductor devices and detectors that the The provision of an APRM drift or lose sensitivity.
scram at615% design power in the " Refuel" and " Start-Experience with passive type in-up/ Hot Standby" modes and struments in generating stations the backup IRM scram at and substations indicates that 6 120/125 of full scale as-specified calibrations are ade-sures that there is proper quate.
For those devices which overlap in the neutron employ amplifiers, etc., drift monitoring systems, and, specifications call for drift thus, that adequate cover- -
to be less than 0.4%/ month; age is provided for all i.e., in the period of a month ranges of reactor opera-a dcift of.4% would occur and tion.
thus providing for adequate mar-gin.
For the APRM system, drift of.. electronic apparatus is not the only consideration in deter-mining a calibration frequency.
Change in power distribution and loss of chamber sensitivity dic-tate a calibration every seven days.
Calibration on this fre-quency 4ssures plant operation at or Felow thermal limits.
A comparison of Tables 4.1.1 and 4.1.2 indicates that two instru-ment channels have not been in-cluded in the latter Table.
These are:
mode. switch in shut-down and manual scram.
All of i
the devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during opera-tion is not applicable, i.e.,
the switch is either on or off.
Amendment No. 79 l
39
k I'lil'S j'.
TAllt.E :).2.C g
INSTHi#lENTATION TilAT INITI ATES ROI) Ill.0CKS
?,
.j, flinimum # of P
Operalile ins t e iunent Cliannel s l'er Tr ip_ Syn.t ems (1)
Instrument Trip I.evel Setting
~
2 APlut lipscale (Flow (0.58 W + 50%)
FHP ' (2) lii a seil)
,HFl.PD 2
Al'Rt1 llownscale 2.5 indicated on scale 1 (7)
Hoel Illock tionitor (0.65 W + 42%)' FRP ' (2)
(FIou lliaseil)
HFl.PD,
I (7)
Roit litock tionitor 5/125 of full scale llownscale 3
'IRtl llownscale (3) 5/125 of full scale 3
IHil Detector not in (8)
Sta r tup l'os i t ion 3
IHit tipscale
$108/125 of full scale 2 (5)
Silli lletect or not in (4)
Startup Position 5
2 (5) (6)
Slui tipscalc
< 10 counts /sec.
I (per tank) (9)
Scram llischarge Volume 518 gallons Wa t e r 1.cve 1 -Ili gli Y: