ML20094J159
| ML20094J159 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/09/1984 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20094J139 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.1, TASK-TM NUDOCS 8408140245 | |
| Download: ML20094J159 (13) | |
Text
9 Proposed Amendments to Technical Specifications - Torus Temperature Monitoring and TMI Action Plan Requirements A.
Narrative:
1.
Torus Temperature Monitoring i
Presently, there are two indicators (ranges: 50 to 150*F) used to display suppression pool temperature on Panel C-7.
These indicators (TI-5047 and TI-5048) each receive their input signal f rom single RTD's, each being ' located 90* apart on the outer wall of the Torus.
This indication of suppression pool temperature is not considered a true presentation of the pool " bulk" or T-quencher " local" temperature. The proposed system modification will provide an accurate indication of this temperature and, therefore, of the pool's ability to quench steam.
The modification incorporates the new requirement mandated by the NRC
)
in NUREG-0661.and Reg. Guide 1.97 for providing a reasonable measure of suppression pool " bulk" and " local" temperature over the range of 30 to 230*F.
The new system is completely redundant meeting single failure criteria.
It derives its signal from 13 redundant, 3-wire RTD's.
Four sensors / channel are used to indicate " local" T-quencher temperatures, and along with the remaining nine sensors / channel are used for determining a bulk temperature of the pool.
All these sensors are located in thermal wells on the reactor side of the Torus.
The temperature signals are transmitted back to the Control Room to Panels C179 and C180 where processing is done in order to display, record and alarm the temperature of the suppression pool. The indicators are located on the Main Control Board, C-903, and the recorders and alarms are located on the PAM panel, C170 and C171.
This system design will satisfy both NUREG-0661 and Reg. Guide 1.97, Type A variable which require normal and post accident monitoring of the suppression pool.
7 -
2.
NUREG-0737 Technical Specifications NUREG-0737, " Clarification of TMI Action Plan Requirements,"
identifies those items for which Technical Specifications are required, and Generic Letter No. 83-36 provides guidance on the scope of Technical Specifications which would be found acceptable by the NRC. This submittal addresses the following instrumentation:
(a) Noble Gas Effluent Monitors
-(b) Sampling and Analysis of Plant Effluents (c) Containment High - Range Radiation Monitoring (d) Containment Pressure Monitoring (e)
Containment Water Level Monitoring-8408140245 84CM309 PDR ADOCK 05000293 b
P ppg m.
B.
Reason for Change 1.
Torus Temperature Monitoring NUREG-0661 and Reg. Guide 1.97 require that a "af ficient number and distribution of pool sensors be provided to measure bulk and local pool temperatures for the range of 30*F to 230*F.
The presentation is to be indicated, recorded and alarmed in the Main Control Room and will be used to assist the operator to make the appropriate decisions in mitigating the consequences of an accident. This indication will also be used to confirm operation of numerous ECCS safety systems and to verify Tech. Spec. limitations.
2.
TMI Action Plan Requirements After the incident at Three Mile Island-2, the NRC requested the installation of instrumentation to aid plant personnel in a post-accident situation.
In Generic Letter 83-36 dated November 1, 1983, guidance was provided for formulating appropriate technical specifications.
In accordance with that request this proposed amendment is submitted to address those instruments associated with NUREG-0737 items II.F.1.1, II.F.1.2, II.F.1.3, II.F.1.4 and II.F.1.5.
Item II.F.1.6 is not includeo because the plant modification has not been completed.
An amendment will be submitted after completion of the plant modification pertaining to II.F.1.6.
Item II.B.3, Post-Accident Sampling, is also not included in this submittal for the same reason, and will be handled in like manner.
Item II.B.1, RCS Vents, and item III.D.3.4, Control Room Habitability Considerations, are not included because Pilgrim does not have an isolation condenser and because reviews have indicated that toxic detectors are unnecessary.
The Limiting Condition for Operation for the various post-accident instrumentation is provided by the addition of the following instrumentation to existing Table 3.2.F:
l Torus Water Level (Wide Range)
Containment Pressure, High Range Containment Pressure, Low Range Containment High Radiation Reactor Building Vent Monitor Main Stack /ent Monitor Turbine Building Vent Monitor
' 'L
a The test and calibration f requencies for surveilling this 1
instrumentation is incorporated into existing Table 4.2.F.
l The surveillance and LCO requirements are either based on those recommended by Generic Letter No. 83-36, or are those which currently exist for Tables 3.2.F and 4.2.F e
As part of this change, Note (7) is added which states:
"With less than the minimum number of operable instrument channels, restore the inoperable channels to operable status within-7 days or prepare and submit a special report to the Regional Director of Inspection and Enforcement within 14 days of the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the channels to operable status."
The creation of this proposed amendment also postulated to us
)
that some of the Notes f rom Table 3.2.F could be made clearer.
We therefore propose the following changes to clarify the meaning without altering it.
Currently, Note (1) states:
"From and af ter the date that one of these parameters is reduced to one indication, continued operation is permissible during the succeeding thirty days unless such instrumentation is sooner made operable."
This would be changed to:
"With less than the minimum number of instrument channels, restore'the inoperable channel (s) within 30 days."
Currently, Note (2) states:
"From and after the date that one of these parameters is not indicated in the control room, continued operation is permissible during the succeeding.seven days unless such instrumentation is sooner made operable."
This will be changed to:
"With the instrument channel (s) providing no indication to the control room, restore the indication to the control room within seven days."
Note (5) is altered by adding the word " indicators" af ter " parameter" in its first line. This is done because parameters are not the true subject of a limiting condition of operation, the indicators of parameters are.
The Table associated with instruments for monitoring safety / relief and safety valves is altered by removing the asterisk f rom the title t
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" Secondary Tail Pipe Temperature Thermocouple" and placing it next to the four dual element thermocouples which monitor safety / relief valve (SRV) tail pipe temperatures. The asterisk references Note (6), which is concerned with an additional-restriction concerning (SRV) tail pipe temperature monitoring thermocouples. The SRV restriction does not apply to the safety valves.
This change more clearly identifies which instruments are subject to Note (6) and is proposed for clarity.
Currently, Note (6) states:
"At a minimum, the above listed (SRV) tail pipe temperature, one of the dual thermocouples, will be operable for each valve when the valves are required to be operable.
If a thermocouple becomes. inoperable, it shall be returned to an operable condition within 31 days or the reactor shall be placed in a shutdown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."
Note (6) will be changed to state:
"At a minimum, for thermocouples providing (SRV) tail pipe temperature, one of the dual thermocouples will be operable for each SRV when the valves are required to be operable.
If a thermocouple becomes inoperable, it shall be returned to an operable condition within 31 days or the reactor shall be placed in a shutdown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."
Note (6) describes the action to be taken when one of the dual thermocouples is inoperable.
However, the current chart labels the thermocouples with a "8" designation, which is only one of the dual elements. To correct this contradiction, we propose to remove the "B" from the dual element (SRV) tail pipe thermocouple designation from the chart.
This change is proposed to clarify a confusing footnote and to emphasize that only those sensing devices for SRV valves are subject to Note (6).
Currently, Table 3.2.F names the parameter being indicated as
" Instrument." The proposed change substitutes " Parameter" for
" Instrument" because it seems more accurate a description and provides a more straight forward clarification to the Table. This a pro forma change.
-C.
Safety Considerations These changes do not present an unreviewed safety question as defined in
.10CFR50.59. They have been reviewed and approved by the Operations Review Committee and reviewed by the Nuclear Safety Review and Audit Committee.
D.
Sionificant Hazards Consideration The NRC has provided guidance concerning the application of standards for determining whether license amendments involve signifirint hazards considerations by providing certain examples (48FR14870). The Torus a:
i
. Temperature Monitoring System is an example of an amendment which is considered not likely to involve a significant' hazards consideration, and provides, "(ii) a change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications:
for example, a more stringent surveillance requirement,"
which is provided by the addition of. thirteen (13) sensors to ensure a reasonable measure of suppression pool " bulk" and " local" temperature which will provide a more accurate indication of pool temperature and the pool's ability to quench steam.
Proposed Technical Specifications of this nature add restrictions to address plant design changes which conform to NRC requirements.
The TMI Action Plan requirements are examples of amendments which are considered not likely to involve a significant hazards consideration, and provides, "(ii) a change that constitutes an additional limitation, restriction..or control not presently included in the technical specifications:
for example, a more stringent surveillance requirement of additional instrumentation provided to assist plant personnel in a post-accident situation.
Proposed Technical Specifications of this nature 5
add restrictions to address plant design changes which conform to NRC requirements.
Based on this guidance, it has been determined that the amendment request involves no significant hazards consideration. Under the NRC's regulations in 10CFR50.92, this means that operation of the Pilgrim Nuclear Power Station in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or dif ferent kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
D.
Schedule of' Change These amendments will be ef fective upon receipt of approval by the NRC.
E.
Application Fee Pursuant to 10CFR 170.21, an application fee of $150.00 is submitted with this amendment request.
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C I-u
' TABLE 3.2.F'
. SURVEILLANCE INSTRUMENTATION
.Minimua #'of Operable. Instrument Type. Indication Channels
' Instrument #
Parameter and Range Notes.
l 2
640-29A & B Reactor Water Level Indicator 0-60" (1) (2) (3) 2
'640-25A & B Reactor. Pressure Indicator 0-1200, psig (1) (2) (3) 2 TRU-9044 Drywell Pressure Recorder 0-80 psia (1) (2) (3)
.TRU-9045 2
TRU-9044 Drywell Temperature Recoraer, Indicator (1) (2) (3)
TI-9019 0-400*F 2
TRU-9045 Suppression Chamber Air Recorder, Indicator (1) (2) (3)
TI-9018 Temperature 0-400*F 2
LR-5038 Suppression Chamber Water Level Recorder 0-32" (1)
(,, 2 ) (3)
LR-5049
~
1 NA Control Rod Position 28 Volt Indicating )
Lights
)
)
. (1)- (2) (3) (4) 1 NA-
. Neutron Monitoring SRM, IRM, LPRM
)
0 to 100% power
)
P TI-5021-01A Suppression Chamber Water Dual Indicator /
TRU-5021-OlA Temperature Multipoint Recorder (4) (7) (2) (3) e 30-230*F (Bulk / Local) 2 1
TI-5022-OlB Suppression Chamber Water Dual Indicator /
q TRU-5022-OlB Temperature Multipoint Recorder (4) (7) (2) (3) 30-230*F (Bulk / Local) 1 PI-5021 Drywell/ Torus Diff. Pressure Indicator
.25 4>3.0 psig (1) (2) (3) (4)
I fPI-5067A Drywell Pressure Indicator
.25-1> 3.0 psigi, 1PI-5067B Torus Pressure Indicator -1.0-* +2.0 (l) (2) (3) (4) ps19.
Amendment No.
58
TABLE 3.2.F (Cont'd)
SURVEILLANCE INSTRUMENTATION Minimum # of Operable Instrument Type Indication
..l' Channels Instrument #
Parameter and Range Notes 1/ Valve a) Primary Safety / Relief Valve Position-a) Acoustic monitor (5) or (5) b) Thermocouple-b) Backup 1/ Valve a) Primary Safety Valve Position Indicator a) Acoustic. monitor--
(5) or (5) b) Thermocouple b) Backup 1/ Valve See Note (6)
Tail Pipe Temperature Thermocouple (6)
Indication
( LI 1001-604A Torus Water Level Indicator /Multipoint (4) (7) (2)-(3)-
, LR 1001-604A (Hide Range)
Recorder 0-300"H O 2
2 LI 1001-604B Torus Hater Level Indicator /Multipoint (4) (7) (2) (3) ulR 1001-604B (Hide Range)
Recorder 0-300"H O 2
f*PR 1001-600A PI 1001-600A Containment Pressure, Indicator /Multipoint
-(4) (7) (2) (3)
(High Range)
Recorder 0-225 psig 2
PI i001-6008 Containment Pressure, Indicator /Multipoint (4) (7)-(2) (3)
( PR 1001-600B (High Range)
Recorder 0-225 psig 1"FI 1001-601A Containment Pressure, Indicator /Multipoint (4) (7) (2) (3) j PR 1001-600A (Low Range)
Recorder -5 to 5 psig 2
PI 1001-601B Containment. Pressure.
Indicator /Multipoint (4) (7) (2) (3)
PR 1001-600B (Low Range)
Recorder -5 to 5 psig 5
'RIT 1001-606A Containment High Radiation Monitor /Multipoint 1
g RIT 1001-606B (Drywell)
Recorder (4) (7)
RR 1001-606A 1 to 1x10' R/hr
, RR 1001-606B
- RIT 1001-607A Containment High Radiation Monitor /Multipoint I
RIT 1001-6078 (Torus)
Recorder (4) (7)
RR 1001-606A 1-to 1x10' R/hr s RR 1001-606B Amendment No.
58a
' TABLE 3.2.F'(Cont'd)
SURVEILLANCE INSTRUMENTATION
-Minimum # of:
~ Operable Instrument Type Indication
~
[_,
Channels Instrument #-
Parameter!
and Range Notes 1
RI 1001-607 Reactor Bu'11 ding Vent Indicator /Multipoint (4) (7)
RR 1001-608 Recorder 10 to 10*'R/hr il
'RI'1001-608 Main Stack. Vent Indicator /Multipoint (4)'.(7)
RR 1001-608 Recorder 10-' to 10* R/hr 1
RI'1001-610 Turbine Building Vent Indicator /Multipoint (4) (7)
RR 1001-608 Recorder 10 to 10*R/hr s
Amendment'No.
58b
.n:
- a n --
n
'1 am r- -
n
i Notes for Table 3.2.F-(1) With less than the minimum number of instrument chahnels, restore the
. inoperable channel (s) within 30 days.
(2) With the instrument channel (s) providing no-indication to the control room, restore the indication to the control room within seven days.
~(3) If the requirements of notes (1) or (2) cannot be met, an orderly shutdown i
shall be initiated and the reactor shall be in the Cold Shutdown Condition with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
s (4) These surveillance instruments are considered to be redundant to each other.
(5) At a minimum, the primary or back-up* pirameter indicators shall be l
operable for each valve when the valves are required to be operable.
With both primary and backup
- Instrument channels inoperaole either return one (1) channel to operable status within 31 days or be in a. shutdown mode within'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The following instruments are associated with the safety / relief and safety valves:
5 Primary Secondary Valve Acoustic Monitor Tail Pipe Temperature Thermocouple 203-3A ZT-203-3A TE6271'
- 203-38 ZT-203-3B TE6272 203-3C ZT-203-3C TE6273 203-3D ZT-203-3D TE6276 203-4A ZT-203-4A TE6274-8 j
203-48 ZT-203-4B TE6275-8
- See Note-(6)
(6) At a minimum, for thermocouples providing SRV tall pipe temperature, one of the dual thermocouples will be operable for each SRV when the valves are required to be operable.
If a thermocouple becomes inoperable, it shall-be returned to an operable condition within 31 days or the reactor shall be placed in a shutdown raode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(7) With less than the minimum number of operable = instrument channels, restore the inoperable channels to operable status within 7 days or prepare and submit a special report to the Regional Director of Inspection and Enforcement within 14 days of the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the chaanels to operable status.
1 l
C Amendment No.
59 E
PNPS TABLE 4.2.F.(Cont.)
MINIMUM TEST AND CALIBRATION FREQUENCY FOR SURVEILLANCE INSTRUMENTATION Instrument Channel Calibration Freauency Instrument Check
[
l
- 13) Torus Water Level (Wide Range)
Each refueling outage Once every 30 days l
I 14)' Containment Pressure Each refueling outage Once every 30 days l
l
- 15) Containment High Radiation Each refueling outage' Once every 30 days I
- 16) Reactor Building Vent Radiation Monitor Each refueling outage Once every 30 days.
l l
- 17) Main Stack Vent Radiation Monitor Each refueling outage Once every 30 days l
I
- 18) Turbine Building Vent Radiation Monitor Each refueling outage Once every 30 days I
Amendment No.
66A
~
~
LIMITINGCONDITIONSFORdPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Applicability:
Applicability:
Applies to the operating status of the primary Ahplies to the primary and secundary and secondary containment systems.
co,ntainment integrity.
Objective:
Objective:
To assure the integrity of the primary and To verify the integrity of the primary srcondary containment systems, and secondary containment.
Soecificaticn:
_ Specification:
A.
At any time that the nuclear system is 1.
a.
The suppres.sion chamber water pressurized above atmospheric pressure level and temperature shall or work is being done which has the be checked once per day, potential to drain the vessel, the pressure suppression pool water volume b.' Whenever there is indication and temperature shall be maintained of relief valve operation or within the following limits except as testing which adds heat to the specified in 3.7.A.2 and 3.7.A.3.
suppression pool, the pool 3
t P"
a.
Minimum water volume - 84,000 ft t
al n t red an also b.
Maximum water volume - 94,000 #t observed and logged every 5 minutes until the heat addition c.
Maximum suppression pool bulk tempera-is tenninated.
ture during normal contfnuous power operation shall be $ 80 F, except as c.
Wher.9ver there is indication of specified in 3.7.A.1.e.
relit.f valve operation with the bulk temperaturegfthesuppressionpool d.
Maxumum suppression pool bulk tempera-reaching 160 F or more and the primary tureduringRCIC,gPCIorADSopera-coolant system pressure greater than tion shall be 190 F, except as 200 psig, an external visual examina-specified in 3.7.A.1.e.
tion of the suppression chamber shall be conducted before resuming power e.
In order to continue reactor power operation.
j operation, the suppression chamber d.
Whenever there is indication of poolbufktemperaturemustbereduced to < 80 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
relief valve operation with the local temperatureofthesuppgessionpool
~
f.
If the suppression pool bulk tempera-T-quencher reaching 200 F or more, ture exceeds the limits of Specifica.
an external visual examination of the tion 3.7.A.1.d, RCIC, HPCI or ADS suppression chamber shall be conducted testing shall be terminated and before resuming power operation.
suppression pool cooling shall be initiated.
e.
A visual inspection of the suppresion chamber interior, including water g.
If the suppression pool bulk tempera-line regions, shall be made at each ture during reactor power operation major refueling outage.
g exceeds 110 F, the reactor shall be scrammed.
152 Amendment No.
a LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS (Cont'd) 4.7 CONTAINMENT SYSTEMS (Cont'd)
- h. During reactor isolation
- f. The pressure differential l
conditions, the reactor pressure between the drywell and vessel shall be depressurized suppression chamber shall be to less than 200 psig at normal recorded at least once each cooldownratesifthegoolbulk shift when the differential temperature reaches 120 F.
pressure is required.
- i. Differential pressure between the
- g. Suppression chamber water l
drywell and suppression chamber level shall be recorded at shall be maintained at equal to or least once each shift when greater than 1.17 psid, except as the differential pressure specified in j and k.
is required.
- j. The differential pressure shall be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of placing the reactor in the run mode following a shutdown. The differential pressure may be reduced to less than 1.17 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.
- k. The differential pressure may be reduced to less than 1.17 psid for a maximum of four (4) hours for maintenance activities on the
)
differential pressure control system and during required oper-ability testing of the HPCI system, the relief valves, the RCIC system and the drywell-suppression chamber vacuum breakers.
- 1. If the specifications of Item 1, above, cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty-four(24) hours.
- m. Suppression chamber water level shall be maintained between -6 to -3 inches on torus level instrument which corresponds to a downcomer submergence of 3.00 and 3.25 feet respectively.
(
Amendment No.
152a
)
BASES:,
3 7.A & 4.7.A Primary Containment The integrity of the primary containment and operation 'of the core standby cooling system in combination limit the off-site doses to values less than those suggested in 10 CFR 1.n0 in the event of a break in the primary system piping. Thus, contain-ment integrity is specified whenever the potential for violation of the primary r: actor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this r:quirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, thus greatly reducing the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel damage.
In building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below 10 CFR 100 limits.
The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heac released during primary system blowdown from 1035 psig. Since all of the gases in the drywell are purged into the pressure supression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in the specification, containment pressureduringthedesignbasisaccidentisapproximajely45psigwhichisbelowthe maximum of 62 psig. Maximumwatervolumeof94,000fj results in a downcomer sub-mergency of 4'-0" and the minimum volume of 84,000 ft results in a submergence
(
approximately 12-inches less. Mark I Containment Long Term Program Quarter Scale Test Facility-(QATF) testing at a downcomer submergency of 3.25 feet and 1.17 psi wetwell to dry well pressure differential shows a significant suppression chamber load reduction and Long Term Program analysis and modifications are based on the above submergence andA P.
l Should it be necessary to drain the suppression chamber, provision wi D be made to l
maintain those requirements as described in Section 3.5.F BASES of this Technical Specification.
(
Experimental data indicates that excessive steam condensing loads can be avoided if U the peak local temperature of the pressure suppression pool is maintained below 200 F during any period of relief-valve operation with sonic conditions at the discharge exit. Analysig has been performed to verify that the local pool tgmperature will stay below 200 F and the bulk pool temperature will stay below 160 F for all SRV tran.
)
sients. Specifications have been placed on the envelope of reactor operating con-ditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high pressure suppression chamber loadings.
c Amendment No.
166