ML20249C710
| ML20249C710 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/26/1998 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20249C709 | List: |
| References | |
| NUDOCS 9807010073 | |
| Download: ML20249C710 (17) | |
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.o Attachment B 1
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3 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3 REACTIVITY CONTROL (CONT) 4.3 REACTIVITY CONTROL (Cont)
B. ' Control Rods (Cont) 8.
Control Rod _s (Cont)
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- 2. The control rod drive housing
- b. When the rod is fully l
support system shall be in place withdrawn the first time during reactor power operation and subsequent to each when the reactor coolant system is refueling outage or after pressurized above atmospheric mciatenance, observe that pressure with fuelin the reactor the drive does not go to the vessel, unless all control rods are overtravel position, fully inserted and Specification 3.3.A.1 is met.
2.
The control rod drive housing support system shall be inspected after reassembly and the results of the inspection recorded.
3.
- a. No control rods shall be moved
- 3. Prior to control rod withdrawal for I
when the reactor is below 20%
startup or insertion to reduce l
rated power, except to shutdown power below 20% of the the reactor, unless the Rod operability of the Rod Worth Worth Minimizer (RWM) is Minimizer (RWM) shall be operable. A maximum of two verified by:
rods may be moved below 20%
design power when the RWM is
- a. verifying the correctness of the control rod withdrawal inoperable if all other rods sequence input to the RWM except those which cannot be moved with control rod drive computer.
pressure are fully inserted,
- b. Control rod patterns and the
- b. performing the RWM sequence of withdrawal or computer diagnostic test.
insertion shall be established such that:
- 1) when the reactor is critical
- c. verifying the annunciation of I
and below 20% design power the selection errors of at the maximum worth of any least one out-of-sequence insequence control rod which control rod in each distinct is not electrically disarmed is RWM group.
less than 0.010 delta k.
- d. verifying the rod block
- 2) and when the reactor is above function of an out-of-20% design power the sequence control rod which is maximum worth of any control withdrawn no more than three L
rod, including allowance for a notches.
single operator error, is less than 0.020 delta k.
Revision l
Amendment No. 39 3/4.3-3 L _
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I o
1 i
BASES:
l 3/4.7 CONTAINMENT SYSTEMS (Cont) l A.
Primary Containment (Cont) i 1
l The maximum permissible bulk suppression pool temperature of 120 F is acceptable since a complete accident blowdown can be accommodated without exceeding the bulk I
suppression pool temperature limit of 17&F immediately after blowdown. This 170 F LOCA blowdown limit is not a limit for the heatup of the suppression pool after the vesselis depressurized. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to
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avoid the regime of potentially high pressure suppression chamber loadings. Current Technical Specification limits on suppression pool temperature ensure bulk pool
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temperature remains within an acceptable range to condense steam discharged to l
the suppression pool during a LOCA or SRV actuation.
j in addition to the limits on temperature of the supprer.nion chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include: (1) use of all available 1
means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other re'icf valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requking the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external l
visual examination lollowing any event whero potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discoruinuities in the vicinity of the relief valve l
discharge since these are expected to be the points of highest stress.
I If a loss-of-coolant accident were to occur when the reactor water ter 3erature is below approximately 330 F, the containment pressure will nr. 'xe 4 me 62 psig l
code permissible pressure, even if no condensation were to u
.ir. The maximum allowable pool temperature, whenever the reactor is above 212*F, shall be governed by this specification. Thus, specifying water volume-temperature requirements applicable for reactor-water temperature above 212 F provides additional margin above that available at 330 F.
Revision Amendment No. 39 53,83, "3 B3/4.7-2 7
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BASES:
3/4.7 CONTAINMENT SYSTEMS (Cont)
A.
Primary Containment (Cont) capability of the structure over its service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operationalleak rate. The allowable operationalleak rate is derived by multiplying the maximum allowable leak rate or the allowable test leak rate by 0.75 thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.
The primary containment leakage rate testing is based on the guidelines in
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Regulatory Guide 1.163 dated September 1995, NEl 94-01 Revision 0 dated July 25, 1995, and ANSI /ANS 56.8-1994. Specific acceptance crituria for as-found and as-left leakage rates, as well as methods of defining the leakage rates, are contained in the primary containment leakage rate testing program.
The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. The leak rate test frequency is in accordance with 10CFR50 App. J, Option B and Regulatory Guide 1.163 dated September 1995.
Type A, Type B, and Type C tests will be performed using the technical methods and
.!achniques specified in ANSI /ANS 56.8 - 1994 or other alternative testing methods approved by the NRC.
A note is included in Surveillance 4.7.A.2.a stating that definition 1.U is not applicable. The 25% allowable extension of surveillance intervals is already included in the primary containment leakage rate testing program; therefore, an additional 25% is not allowed.
The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends.
Whenever a bolted double-gasketed penetration is broken and remade, the space betwean the gaskets is pressurized to determine that the seals are performing properly. It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of eccidents are to be minimized. The personnel air lock is tested at 10 psig, because the inboard door is not designed to shut in the
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opposite direction.
Primary Containment isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the vaives in each line would be sufficient to maintain the integrity of the pressure suppression system.
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Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.
' Revision Amendment No. 4 13,13S,1S7,172 B3/4.7-4 L
- - - -. - - - - - - - - -- _ _ o
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- 9 BASES:
3/4.7 CONTAINMENT SYSTEMS (Cont)
A.
Primary Containment (Cont) i Group 1 - process lines are isolated by reactor vessel low-low water level in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems. The valves in group 1 are also closed when process instrumentation detects excessive main steam line flow, low pressure, main steam space high temperature, or reactor vessel high water level.
Group 2 - isolation valves are closed by reactor vessel low water level or high drywell pressure. The group 2 isolation signal also " isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal.
Group 3 - isolation valves can only be opened when the reactor is at low pressure and the core standby cooling systems are not required. Also, since the reactor vessel could potentially be drained through these process lines, these valves are closed by low water level.
Group 4 and 5 - process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines.
The signals which initiate isolation of group 4 and 5 prccess lines are therefore indicative of a condition which would render them inoperable.
Group 6 - process lines are normally in use and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the intet to the cleanup system. Also, since the vessel could potentie!!y be drained through the cleanup system, a low level isolation is provided.
Group 7 - The HPCI veceum breaker line is designed to remain operable when the HPCI system is required. The signals which initiate isolation of the HPCI vacuum breaker line are indicative of a break inside containment and reactor pressure below that at which HPCI can operate.
The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
In satisfying this design intent an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valva performance, prior to exceeding the design closure times.
In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting
}
from the accident occur prior to closure of the main steam line isolation valves.
l Analyses indicate that fuel rod cladding perforations would be avoided for main sttam valve closure times, including instrument delay, as long as 10.5 seconds.
Revision i
Amendment No-143r467 B3/4.7-5 l
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,.a LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3 REACTIVITY CONTROL (CONT) 4.3 REACTIVITY CONTROL (Cont)
B. ' ontrol Rod.= (Cont)
B.
Control Rods (Cont)
C
- 2. The control rod drive housing
- b. When the rod is fully support system shall be in place withdrawn the first time during reactor power operation and subsequent to each when the reactor coolant system is refueling outage or after pressurized above atmospheric maintenance, observe that pressure with fuel in the reactor the drive does not go to the vessel, unless all control rods are ove,rtravel position.
fully inserted and Specification i
3.3.A.1 is met.
I 2.
The control rod drive housing support system shall be inspected after reassembly and the results of the inspection recorded.
3.
- a. No control rods shall be moved
- 3. Prior to control rod withdrawal for when the reactor is below 20%
startup or insertion to reduce rated power, except to shutdown power below 20% of the the reactor, unless the Rod operability of the Rod Worth Worth Minimizer (RWM) is Minimizer (RWM) shall be operable. A maximum of two verified by:
rods may be moved below 20%
design power when the RWM is
- a. verifying the correctness of the control rod withdrawal inoperable if all other rods except those which cannot be sequence input to the RWM moved with control rod drive computer.
pressure are fully inserted.
- b. Control rod patterns and the
- b. performing the RWM sequence of withdrawal or computer diagnostic test.
insertion shall be established such that:
- 1) when the reactor is critical
- c. verifying the annunciation of and below 10'/ design power the selection errors of at the maximum wo of any least one out-of-sequence insequence control ro which 20%
control rod in each distinct is not electrically disarmed is RWM group.
less than 0.010 delta k.
- d. verifying the rod block function
- 2) and when the reactor is above of an out-of-sequence control 20% design power the rod which is withdrawn no maximum worth of any control more than three notches.
rod, including allowance for a single operator error, is less than 0.020 delta k.
Revision Amendment No. 39 3/4.3-3 I
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.A i
BASES:
j 3/4.7 CONTAINMENT SYSTEMS (Cont)
Deleted
)
A.
Primary Containment (Cont) j Exper" enta! dat: !" dica!Oc ? Sat exceccive cten-' condenc!ng !Onde can be avoided i' tSO f
peak !cc ! temperature O! the preccure cuppreccion poc!! maintained be!cy' 2002c dur"'g j
eny per!cd Of re!!cf-va!ve Operat!Or w"5 conic conditienc at +50 diccharge cit ^
!ycic j
Sac been perfor-'ed te verify that the !cca! p00! temperature v'"! c! y be!Ow 2002F and the j
insert bu!' temperature w"! ct y bc!cw 1SO C for ;!! SPV 'rancientedThe maximum permissible
]
bulk suppression pool temperature of 120 F is acceptable since a complete accident 1
blowdown can be accommodated without exceeding the bulk suppression pool temperature l
limit of 170 F immediately after blowdown. This 170*F LOCA blowdown limit is not a limit Mr tha bantup of the suooression pool after the vesselis depressurized specifications have been placed on the envelope of reactor operating conditions so mat the reactor can be depressurized in a timely manner to avoid the regime of potentially high pressure suppression chamber loadings. Current Technical Specification limits on suppression pool l
temperature ensure bulk pool temperature remains within an acceptable range to l
condense steam discharged to tha suppression pool during a LOCA or SRV actuation.
1 In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or l
sticks open. This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be j
separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
l Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By rc
. ring the suppression pool e
temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage l
was encountered. Particular attention should be focused on structural discont!nnities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress, if a loss-of-coolant accident were to occur when the reactor water temperature is below approximately 330 F, the containment pressure will not exceed the 62 psig code permissible pressure, even if no condensation were to occur. The maximum allowable pool temperature, whenever the reactor is above 212 F, shall be governed by this specification. Thus, specifying water volume-temperature requirements applicable for i
Revision Amendment No. 39,53,83,113 B3/4.7-2 I
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..s reactor water temperature above 212 F provides additional margin above that available at 330 F..
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(Continued)
. Revision
. Amendment No. 39,52,82,143 B3/4.7-2 i
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l BASES:
3/4.7 CONTAINMENT SYSTEMS (Cont)
A.
Primary Containment (Cont) capability of the structure over its service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable l
operationalleak rate. The allowable operationalleak rate is derived by multiplying I
the maximum allowable leak rate or the allowable test leak rate by 0.75 thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.
B.
The primary conta!nmen+ teak rate tect fmquency !c baced on maintaining cdequate
> assurance that the !cak rate rsmain:"/ithin the cpecificatier 90 !cak rate tect Ddeta frequency !c in accordance "dth 10 CFR50,.^pp. J.
The primary containment leakage rate testing is based on the guidelines in Regulatory Guide 1.163 dated September 1995, NEl 94-01 Revision 0 dated July 25, 1995, and ANSI /ANS 56.8-1994. Specific acceptance criteria for as-found and as-left leakage rates, as well as methods of defining the leakage rates, are contained in the primary containment leakage rate testing program.
The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. The leak rate test frequency is in accordance with 10CFR50 App. J, Option B and Regulatory Guide 1.163 dated September 1995.
/
Insen Type A, Type B and Type C tests will be performed using the technical methods and techniques specified in ANSI /ANS 56.8 - 1994 or other alternative testing methods l
approved by the NRC.
A note is included in Surveillance 4.7.A.2.a stating that definition 1.U is not applicable. The 25% allowable extension of surveillance intervals is already included in the primary containment leakage rate testing program; therefore, an additional l
25% is not allowed.
I I
The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends.
j Whenever a bolted double-gasketed penetration is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents ere to be minimized. The personnel air lock is tested at 10 psig, because the inboard door is not designed to shut in the opposite direction.
i Revision Amendment No.
- 13,136,167,172 B3/4.7-4 l
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Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.
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Revision Amendment No. 113,139,1S7,172 B3/4.7-4 l
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delete high I
radiation I
3/4.7 CONTAINMENT SYSTEMS (Cont)
A.
Primary Containment (Cont)
Group 1 - process lines are isolated by reactor vessel low-low water level i order to i
allow for removal of decay heat subsequent to a scram, yet isolate in time r proper operation of the core standby cooling systems. The valves in group 1 are a so closed when process instrumentation detects excessive main steam line flow, high radiatioth-low pressure, main steam space high temperature, or reactor vessel high 1
water level.
l Group 2 - isolation valves are closed by reactor vessel low water level or high drywell i
pressure. The group 2 isolation signal also " isolates" the reactor building and starts the standby gas treatment system. It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal.
Group 3 -isolation valves can only be opened when the reactor is at low pressure and the core standby cooling systems are not required. Also, since the reactor vessel could potentially be drained through these process lines, these valves are f
closed by low water level.
Group 4 and 5 - process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines.
The signals which initiate isolation of group 4 and 5 process lines are therefore indicative of a condition which would render them inoperable.
Group 6 - process lines are normally in use and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.
Group 7 - The HPCI vacuum breaker line is designed to remain operable when the HPCI system is required. The signals which initiate isolation of the HPCI vacuum breaker line are indicative of a break inside containment and reactor pressure below that at which HPCI can operate.
The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
In satisfying this design intent an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance, prior to exceeding the design closure times.
In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation Revision l
Amendment Nc.
- 13,167 83/4.7-5 l
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8 a
resulting from the accident occur prior to closure of the main steam line isolation valves.
Analyses indicate that fuel rod cladding perforations would be avoided for main stea'm valve closure times, including instrument delay, as long as 10.5 seconds.
(Continued)
Revision l
Amendmer.t Nc.
- 13,167 B3/4.7-5 f