ML20215A864
| ML20215A864 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/04/1987 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20215A805 | List: |
| References | |
| NUDOCS 8706170118 | |
| Download: ML20215A864 (28) | |
Text
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Surveillance Page No.
3.7 . CONTAINMENT SYSTEMS 4.7 152 A. Primary Containment A 152
- 8. Standby Gas Treatment System B 158 C. Secondary Containment C 159 1
3.8 RADI0 ACTIVE MATERIALS 4.8 177 A. Liquid Effluents A 177
- 8. Airborne Effluents B 179 C. (Deleted)
D. Environmental Monitoring Program D 182 E. Mechanical Vacuum Pump E 184 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 194 A. Auxiliary Electrical Equipment A 194 B. Operation with Inoperable Equipment B 195 3.10 CORE ALTERATIONS 4.10 202 A. Refueling Interlocks A 202 B. Core Monitoring B 202 C. Spent Fuel Pool Water Level C 203 3.11 REACTOR FUEL ASSEMBLY 4.11 205A A. Average Planar Linear Heat Generation Rate (APLHGR) A 205A B. Linear Heat Generation Rate (LHGR) B 205A-1 C. Minimum Critical Power Ratio (MCPR) C 2058 3.12' FIRE' PROTECTION '4.12-206 A. Fire' Detection Instrumentation A 206 B. Fire Suppression Water System B 206 ,
C. Spray and/or Sprinkler Systems C 206c D. Halon System D 206d E. Fire Hose Stations E 206e F. Penetration Fire Barrier F 206e i G. Dry Chemical Systems G 206e-1 H. Yard Hydrants and Exterior Hose Houses H 206e-1 4.0 MISCELLANEQUS RADI0 ACTIVE MATERIALS SOURCES 206k 1
4.1 Sealed Source Contamination 206k j 4.2 Surveillance Recuirements 206k i 4.3 Reports 2061 I 4.4 Records Retention 2061 f
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Amendment No. 11 i
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1 :. 0 DEFIN!TIONS (Cont'd) valv'e closure, are bypassed-when reactor pressure'is less than c 600 psig,.the-low pressure main steam line isolation valve ;
closure trip is bypassed, the reactor protection system is energized with IRM neutron monitoring system trips and control ;
rod withdrawai interlocks in service.
- 2. Run Mode - In this mode the reactor system pressure is at or ;
above 880 psig and the reactor protection system is energized with APRM protection and RBM interlocks in service.
1
- 3. Shutdown Mode - The reactor is in the shutdown mode when'the reactor mode switch is in the shutdown mode position and no i core alterations are being performed.
- a. Hot Shutdown means conditions as above with reactor coolant temperature greater than 212*F.
- b. Cold Shutdown means conditions as atove with reactor ,
coolant temperature equal to or less than 212'F. ;
- 4. Refuel Mode - The reactor is in the refuel mode when the mode j switch is in the refuel mode position. When the mode switch is !
in the refuel position, the' refueling interlocks are in service.
L. Design Power - Design power means a steady-state power level of 1998 thermal megawatts. l M. Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:
- 1. All manual containment isolation valves on lines connected.to the reactor coolant system or containment which are not required to be open dur.ing. accident. conditions:are" closed.
- 2. At least one door'in each airlock'is~' closed'and sealed.
- 3. All blind flanges and manways are closed.
- 4. All automatic primary containment isolation valves are operable or at least one containment isolation valve in each'line'having an inoperable valve shall be deactivated in the: isolated condition.
- 5. All containment isolation check. valves are operable or at least one containment valve in each line having an inoperable valve is secured in the isolated pos_ition.
N. Secondary Containment Integrity - Secondary containment integrity means that the reactor building is intact and the following conditions are met:
Amendment No. 3
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1.0 DEFINITIONS (Continued)
AA. Action'- Action shall be that part of a specification which prescribes remedial measures required under designated conditions.
BB. Member (s) of the Public' - Member (s) of the public shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the !
-site to service equipment or to make deliverles. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the site.
1 CC. Site Boundary' - The site boundary is shown in Figure 1.6-1 in the FSAR.
DD. Radwaste Treatment System i
- 1. Gaseous Radwaste Treatment System - The gaseous radwaste treatment system is that system identified in Figure 4.8-2.
- 2. Liquid Radwaste Treatment System - The liquid radwaste treatment system is that system identified in Figure 4.8-1.
EE. Automatic Primary Containment Isolation Valves - Are primary ,
containment isolation valves which receive an automatic primary containment group isolation signal.
See FSAR Figure 1.6-1 Amendment No. Sb l
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3.2 -In addition to reactor protection instrumentation which initiates a l reactor scram, protective instrumentation has been provided which l initiates action to mitigate the consequences of accidents which are ]
beyond the operator's ability to control, or terminates operator errors l before they result in serious consequences. This set of specifications j provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The. objectives of the i Specifications are (i) to assure the effectiveness of the protective )
instrumentation when required by~ preserving its capability to tolerate a (
single failure of any component of such systems even during periods when j portions of such systems are out of service for maintenance, and (11) to i j
prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect ,
on safety. The set points of other instrumentation, where only the high ]
or low end of the setting has a direct bearing on safety, are chosen at 1 a level away from the normal operating range to prevent inadvertent )
actuation of the safety system involved and exposure to abnormal '
situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
l The low water level instrumentation' set to trip at 129.'5" above the top of the active fuel closes a'll' isolation valves'except'those in Groups 1, 4 and 5. This trip setting is adequate to prevent core uncovery in the l )
case of a break in the largest line assuming a 60 second valve closing I time. Required closing times are less than this.
The low low reactor water level instrumentation is set to trip when i reactor water level is 78.5" above the top of the active fuel (-49" on I the instrument). This trip closes Main Steam Line Isolation I
Amendment No. 68 I
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i 3.2 CA_SE_S (Cont'd)
Valves, Main Steam Drain Valves, Recirc Sample Valves (Group 1)
I activates _the CSCS subsystems, starts.the emergency diesel generators and trips the recirculation pumps. This trip' setting. level was chosen to be high enough to prevent spurious actuation but low enough to. I initiate CSCS operation and primary system isolation so that no fuel f damage'will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large. breaks up to the complete circumferential break of a 28-inch recirculation line- j and with the trip setting given above,.CSCS initiation and primary <
system isolation are initiated.in time to meet the above criteria. l i The high drywell pressure instrumentation is a' diverse signal 'to the water level instrumentation and in addition to initiating CSCS, it j causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low low water level instrumentation; thus.the.results given above are applicable here also. The low low water level ;
instrumentation initiates protection.for the full spectrum of loss-of-coolant accidents and causes isolation of Group l' isolation 'i valves.
Venturis are provided in.the main steam lines as a means of measuring.
steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. -The primary function of the.
instrumentation.is to. detect a break in the main steam-line. For the worst case accident, main steam line break outside the drywell, a trip. I setting of 1407. of rated steam flow in conjunction with the flow
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limiters and m&in steam line valve closure, limits the' mass inventory '
loss such that fuel is not uncovered, fuel temperatures remain approximately 1000*F and release of radioactivity to the environs is well below 10 CFR 100 guidelines. 1
- Temperature ' monitoring: instrumentation 11sprovided in-the main -steam line. tunnel.and the turbine basement'to' detect leaks in'these areas, i Trips are provided on'this' instrumentation ~and when" exceeded,-cause 1 closure of isolation valves. The setting of 170*F for the main steam 'l l line tunnel detector is low enough to detect leaks of~the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. !
For large breaks, the high steam flow-instrumentation is'a backup'to the temperature instrumentation.
High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop acci-
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Amendment No. 69 L
LIMITING. CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 . CONTAINMENT SYSTEMS l
Applicability: Applicability:
Applies to the operating status of the -Applies to the primary and secondary containment integrity.
~
primary and secondary containment-systems.
Objective:
Objective:
To verify.the integrity of the primary To assure the integrity of the primary and secondary containment. ,
and secondary containment systems. j i'
Specification: Specification:
A. Primary Containment A. Primary Containment Suppression Pool Suppression Pool
{
- 1. At any time that the nuclear system 1. a. The suppression chamber water is pressurized above atmospheric level and temperature shall be pressure or work is being done which checked once per day.
has the potential to drain the vessel, the pressure suppression b. Whenever there is indication of pool water volume and temperature . relief valve. operation or shall be maintained within the . testing which adds heat to the following limits except as specified suppression pool,~the. pool in 3.7.A.2 and 3.7.A.3. temperature shall be continually monitored and also
- a. Minimum water volume - 84,000 ft* observed and logged every 5 minutes until the heat addition
- b. Maximum water volume - 94,000 ft' is terminated,
- c. Maximum suppression pool bulk c. Whenever there is Indication of
-temperature during normal relief-valve operation with the continuous power operation shall bulk temperature of the be 180*F, except as specified in suppression pool reaching 160*F 3.7.A.I.e. or.more and the primary coolant system pressure greater than
- d. Maximum suppression pool bulk 200 psig, an external visual temperature during RCIC, HPCI or examination of the suppression ADS operation shall be 190*F, chamber shall be conducted except as specified in 3.7.A l.e. before resuming power. operation,
- d. Whenever there is indication of relief valve operation with the local temperature of the suppression pool T-quencher reaching 200*F or:more, an external visual-examination of the suppression chamber shall be conducted before resuming power operation'.
Amendment,No. 152 1 l
r; 1 l
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
'3.7 CONTAINMENT SYSTEMS.(Con't) 4.7 CONTAINMENT SYSTEMS (Con't)
- e. In_ order to continue reactor e. A visual inspection of the power operation, the suppression suppression chamber interior, chamber pool bulk temperature including water line regions, must be reduced to <80*F within
~
shall be made at each major 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. refueling outage.
- f. If the suppression pool bulk f. The pressure differential temperature exceeds the limits between.the drywell and of Specification 3.7.A.l.d, suppression chamber shall be RCIC, HPCI or ADS testing shall recorded at least once each be' terminated and suppression shift when the differential pool cooling shall be initiated. pressure is required.
- g. If the suppression pool bulk g. Suppression chamber water temperature during reactor power level shall be recorded at j operation exceeds 110*F, the least once each shift when !
reactor shall be scrammed. the differential pressure is required.
- h. During reactor isolation conditions, the reactor pressure vessel shall be depressurized to l less than 200 psig at normal cool down rates if the pool bulk temperature reaches 120*F. !
- 1. Differential pressure between the drywell and suppression chamber shall be maintained at i equal to or greater than 1.17 j psid, except as specified in j ;
and k. j
- j. The differential pressure shall l be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l of placing the reactor in the run mode following a shutdown.
The differential pressure may be i 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 operability testing of the HPCI system, the relief valves, the RCIC system and the drywell-suppression chamber vacuum breakers.
Amendment No. IS3 l
o l
l LIMITING CONDITIONS FOR OPERATION. -SURVEILLANCE' REQUIREMENTS q J
G 3.7 CONTAINMENT; SYSTEMS-(Con't)- 4.7 CONTAINMENT-SYSTEMS (Cont'd)
~1. If the specifications-of Item 1, above, cannot be met,.and :
the ' differential. pressure- ;
- cannot be. restored.within:the subsequent (6) hour period...
an orderly shutdown shall be j initiated and the reactor j 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 aidowncomer submergence'of 3.00.and 3.25 l feet respectively. ;
1
.l n. The suppression chamber can i be drained if the conditions i as specified in Sections i 3.5.F.3 and 3.5.F.5 of this Technical. Specification are adhered to.
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. Amendment.No. 154 t
E 4
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primary Containment'(Con't) 4.7.A Primary Containment (Con't)
Primary Containment Integrity Primary Containment Integrity l 2.a Primary containment integrity 2.a The primary containment integrity shall be maintained at all times shall be demonstrated by when the reactor is critical or performing Primary Containment when the reactor water temperature Leak Tests in accordance with 10 is above 212*F and fuel is in the CFR 50 Appendix J, as amended thru reactor vessel except while Sept. 22, 1980, with exemptions as performing "open vessel" physics approved by the NRC and exceptions test at power levels not to exceed as follows:
5 Mw(t).
(1) The main steam line isolation Primary containment integrity valves shall be tested at a means that the drywell and pressure 223 psig, and pressure suppression chamber are . normalized to a value intact and that all of the equivalent to 45 psig each following conditions are satisfied: operating cycle.
(1) All manual containment (2) Personnel air lock door. . seals isolation valves on lines shall be tested at a pressure connected to the reactor 210 psig each operating coolant system or containment cycle. Results shall be which are not required to be normalized to a value open during accident equivalent to 45 psig. i conditions are closed.
If the total leakage rates listed (2) At least one door in each below are exceeded, repairs and airlock is closed and sealed. retests.shall be performed to correct the conditions.
(3) All blind flanges and manways are closed. (1) ,All double-gasketed seals:
10% L. (x)
(4) All automatic primary . .
containment isolation valves (2) All testable penetrittons and and all instrument line flow isolation valves:
check valves are operable 60% L. (x) except as specified in '
i 3.7.A.2.b. (3) Any one penetration or isolation valve except main (5) All containment isolation steam line isolation valves:
check valves are operable or 5% L. (x) at least one containment ,
isolation valve in each line (4) Any one main ~ steam line j having an inoperable valve is isolation valve:
secured in the isolated 11.5 scf/hr @23 psig.
postion.
where x - 45 psig L .75 L.
L. - 1.0% by weight of the contained air @ 45 psig for 24 hrs.
155 Amendment No.
7.
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i LIMITING CONDI,TIONS FOR OFERATION SURVEILLANCE REQUIREMENTS l 1
3.7.A Primary Containment (Con't) 4.7.A Primary Containment (Con't) l Primary Containment Isolation Valves Primary Containment Isolation Valves i 2.b. In the event any Primary' 2.b.1 The primary containment Containment Isolation Valve that isolation valves surveillance receives an automatic isolation shall be performed as follows: 1 signal listed in Table 3.7-1 becomes inoperable, at least one a. At least once per operating ,
containment isolation valve in cycle the operable j each line having an inoperable isolation valves that are valve shall be deactivated in power operated and the isolated condition. (This automatically initiated requirement may be satisfied by shall be tested for deactivating the inoperable simulated automatic valve in the isolated initiation and closure condition. Deactivation means times.
to electrically or pneumatically disarm, or otherwise secure the b. At least once per quarter:
valve.)*
- 1. All normally open power operated isolation valves (except for the main steam line power operated isolation valves) shall be fully closed and reopened.
- 2. Trip the main steam isolation valves individually and verify closure time.
- c. At least twice per week the
. main. steam line-power operated. isolation valves shall be exercised by partial closure and subsequent reopening.
i
- d. At least once per operating )
cycle the operability of l the reactor coolant system j instrument line flow check )
valves shall be verified. l 1
2.b.2 Whenever a primary containment i
- Isolation valves closed to satisfy isolation valve, that receives these requirements may be reopened on an automatic isolation signal, an intermittent basis under ORC listed 'n Table 3.7-1 is approved administrative controls. Inoperable, the position of the ;
isolated valve in each line having an inoperable valve shall be recorded daily.
Amendment No. 155a
r 1+ I 1
.g LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A ~ Primary-Containment' (Con't) 4.7.A- Primary Containment (Con't) 2.c Continuous Leak Rate Monitor l Whent 'he primary containment is inerted, the containment shall be
- continuously monitored for gross leakage by review of the inerting system makeup requirements. This o
- monitoring system may be taken t
out of service for maintenance but shall be retur ed to service as soon as practicable.
2.d Drywell-Surfaces l
The interior surf. aces of the drywell and torus above the water line shall be visually inspected every refueling outage for ,
evidence of deterioration, i 1 I
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s Amendment No. 155b
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primary Containment 4.7.A Primary Containment 5.b. Within the 24-hour period subsequent to placing the reactor in the Run mode following a shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 4% by volume and maintained in this condition.
De-inerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.
- 6. If the specifications of 3.7.A.1 l thru 3.7.A.5 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 i
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BASES:
' 3.7.A & 4.7.A Primary Containment l'i l
-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 100 in the event of a break in the primary system i piping. Thus, containment integrity is specified whenever the potential for- l violation of the primary reactor system integrity exists. Concern about such i a violation exists whenever the reactor is critical and above atmospheric pressure. An exception was made to this requirement during initial core ,
loading and while the low power test program was being conducted and ready- !
access to the reactor vessel was required. There was no pressure on the system at this time, thus greatly reducing the char.ces of a pipe break.
Should this type of testing be necessary in the future, the reactor may be taken critical; however, restrictive operating procedures would be in effect again-to minimize the probability of an accident. Procedures and the Rod Horth Minimizer would limit control worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an !
excursion did. occur, the secondary containment and standby gas treatment- I system, which shall be operational 'during this time, offer a sufficient I r- barrier to keep off-site doses well below 10 CFR 100 limits.
The pressure suppression pool water provides the heat sink for the reactor j primary system energy release following a postulated rupture of the system. 1 The pressure suppression chamber water voluu must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig. Since all of the gases in the drywell ace purged into the pressure suppression chamber air space during a loss-of-c%lant 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 i that the total volume of reactor coolant to be condensed is discharged to the suppression ch&mber and that tN J.,well volume is purged to the suppression chamber.
Using"the minimum or maximum water volumes given-in the" specification, j containment pressure during the design basis accident is approximately 45 psig i which is below the maximum of 62 psig. Maximum water volume of 94,000 ft' !
results in a downcomer submergency of 4'-0" and the minimum volume of 84,000 '
ft results 8
in a submergence approximately 12-inches less. Mark I )
Containment Long Term Program Quarter Scale Test Facility-(QSTF) testing at a 1 downcomer submergency of 3.25 feet and 1.17 psi wetwell to drywell pressure-differential shows a significant suppression chamber load reduction and Long Term Program analysis and modifications are based on the above submergence and differential pressure.
Should it be necessary to drain the suppression chamber, provision will be ,
made to maintain those requirements as described in Section 3.5.F BASES of this Technical Specification.
Amendment No. 165 -
r I
BASES:
3.7.A & 4.7.A Primary Containment Experimental data indicates that excessive steam condensing loads can be ;
avoided if 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. Analysis has been performed to verify that the local pool temperature will stay below 200*F and the bulk temperature will stay below 160*F for all SRV transients. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high pressure suppression chamber loadings.
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 sticks open. This action would include:
(1) use of all available means to close the valve, (2) initiate suppression i pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressuriza 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 ds-ly is sufficiant to establish any temperature trends. By requiring 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 visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are ,
expected to be the points of highest stress. j 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 i 62psig 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 reactor-water temperature above 212*F provides additional margin above that available at 330*F.
Amendment No. I ti6
3 BA " ;
3.7.A & 4.7.A Primary Containment l
Primary Containment Testing The primary containment pre-operational test pressures are based upon the [
calculated primary containment pressure response in the event of a loss-of-coolant accident. The calculated peak drywell pressure is about 45 psig which would rapidly reduce to 27 psig following the pipe break. Following the pipe break, the suppression chamber pressure rises to 27 psig, equalizes with drywell pressure and therefore rapidly decays with the drywell pressure decay.
The design pressure of the drywell and supression chamber is 56 psig. The design leak rate is 0.5%/ day at a pressure of 56 psig. Based on the calculated containment pressure response discussed above, the primary containment j pre-operational test pressures were chosen. Also, based on the primary j containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit ,
rather than the individual components separately.
The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.25%/ day at 45 psig. .
Calculations made by the AEC staff with this leak rate and a standby gas j treatment system filter efficiency of 95% for halogens and assuming the fission i product release fractions stated in TID 14844, show that the maximum total whole body passing cloud dose is about 13 REM and the maximum total thyroid dose is ,
about 110 REM at the site boundary over an exposure duration of two hours. The {
resultant doses that would occur for the duration of the accident at the low l population zone distance of 4.3 miles are about 3 REM total whole body and 70 REM total thyroid. Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident.
These doses are also based on the assumption of no holdup in the secondary cor..ainment resulting in a direct release of fission products from the primary containment through the filters and stack to the environs. Therefore, the sm cified primary containment' leak rate and filter efficiency are conservative ano provide margin between expected off-site dose and 10 CFR 100 guidelines. 1 The maximum allowable test leak rate is 1.0%/ day at a pressure of 45 psig. This value for the test condition was derived from the maximum allowable accident leak rate of 1.25%/ day when corrected for the effects of containment environment under accident and test conditions. In the accident case, the containment atmosphere initially would be composed of steam and hot air whereas under test conditions the test medium would be air at ambient conditions. Considering the differences in mixture composition and temperatures, the appropriate correction factor applied was 0.8 as determined from the guide on containment testing.
Establishing the test limit of 1.0%/ day provides an adequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness 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 operational leak rate. The allowable operational leak 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.
Amendment No. 167 ,
l
1 i
' BASES: ,
3.7.A & 4.7.A Primary Containment, !
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 10 CFR 50 App. J as amended through Sept. 22, 1980.
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 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, I it is possible that-leakage into other parts of the facility could occur. 1 Such leakage paths that may affect significantly the consequences'of accidents are to'be minimized. The personnel air lock.ls tested at 10 psig, because the inboard door is not designed to shut in the opposite direction.
Primary Conta.inment Isolation Valves ,
l Double isolation. valves are provided 1 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.
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, high rauiation,, low pressure, main steam space high !
temperature, or reactor vessel high water level. "
i Group 2 - isolation valves'are closed by reactor vessel: low water level.or i high drywell pressure. The group 2 isolation signal also " isolates" the 1 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. j 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 process lines are therefore indicative of a condition which would render them inoperable. :
Amendment.No. 168 J
{ l BASES:
3.7.A & 4.7.A Primary Containment 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 i
. containment and reactor vessel isolation control system have been selected in -
consideration of the design intent to prevent core uncovering f_ollowing pipe q breaks outside the primary containment and the need to contain released ;J 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 ac ure that the, doses that may result from a steam line break do j 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 li v isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including-instrumer' delay, as long as 10.5 seconds. l l
These valves are highly reliable, have iow service requirements and most are j normally ' clos ~ed. The initiating sensors and' associated trip channels are.also checked to demonstrate'the capability for' automatic isolation. The test-interval 'of once per operating c )
failure probability of 1.1 x 10 ycle for-a automatic that initiation line will not results-in isolate. a-More frequent-testing for valve operability results in a greater assurance.that the valve will be operable when needed.
The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.
The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment. A program for ,
periodic testing and examination of the excess flow check valves is in place. j Primary Containment Paintin_g ,
The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage, i approximately every 18 months, assures the paint is intact. Experience with l this type of paint at fossil fueled generating stations indicates that the inspection Interval is adequate. !
Amendment No. 169 i
)
E 1 1
I B_ASES:
3.7.A & 4.7.A Primary Containment Vacuum Relief I
The purpose of the vacuum relief valves is to equalize the pressure between !
the drywell and suppression chamber and reactor building so that the I structural integrity of the containment is maintained. The vacuum relief system from the pressure suppression chamber to reactor building consists of two 100% vacuum reliet breakers (2 parallel sets of 2 valves in series). < '
Operation of_either system will maintain the pressure differential less than 2 psig; the external design pressure. One valve may be out of service for repairs for a period of seven days. If repairs cannot be completed within seven days, the reactor coolant system is brought to a condition where vacuum i relief is no longer required. -
The capacity of the 10 drywell vacuum relief valves is sized to limit the {
pressure differential between the suppression chamber and drywell during ]
post-accident drywell cooling to the design limit of 2 psig. They are sized j on the basis of the Bodega Bay pressure suppression system tests. The ASME ]
Boiler and Pressure Vessel Code,Section III, Subsection B, for this vessel allows a 5 psig vacuum; therefore, with two vacuum relief valves secured in the closed position and eight operable valves, containment integrity is not impaired.
Reactor operation is permissible if the bypass area between the primary containment drywell and suppression chamber does not exceed an allowable area. The allowable bypass area is based upon analysis considering primary system break area, suppression chamber effectiveness, and containment design pressure. Ana.lyset show that the maximum allowable bypass area is 0.2 ft',
which is equivalent to all vacuum breakers open 3/32". (See letters from Boston Edison to the Directorate of Licensing, dated May 15, 1973 and October 22, 1974)
Reactor ~ operation is~not' permitted if differential pressure decay rate is demonstrated to' exceed 25% of allowable, thus providing-a-margin of safety for the primary containment in the event of a small break in the primary system.
Each drywell suppression chamber vacuum breaker is equipped with three switches. One switch provides full open indication only. Another switch provides closed indication and an alarm on Panel C-7 should any vacuum breaker come off its closed seat by greater than 3/32". The third switch provides a separate and redundant alarm on Panel 905 should any vacuum breaker come off J its closed seat by greater than 3/32". The two alarms above are those l referred to in Section 3.7.A.4.a(3) and 3.7.A.4.d.
]
The water in the suppression chamber is used only for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a daily check of the temperature and volume is 3dequate to assure that adequate heat. ,
removal capability is present.
l l
Amendment No. 170 l
I
BASES:
3_.7.A & 4.7.A Primary Containment j 1
l Inerting The relatively small containment volume inherent in the GE-BWR pressure suppression containment and the large amount of zirconium in the core are such that the occurrence of a very limited (a percent or so) reaction of the zirconium and steam during a loss-of-coolant accident could lead to the ;
liberation of hydrogen combined with an air atmosphere to result in a j flammable concentration in the containment. If a sufficient amount of !
hydrogen is generated and oxygen is available in stoichiometric quantities, the subsequent ignition of the hydrogen in rapid recombination rate could lead to failure of the containment to maintain a low leakage integrity. The 4% ]
oxygen concentration minimizes the possibility of hydrogen combustion following a loss-of-coolant. ;
)
The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the <
loss-of-coolant accident upon which the specified oxygen concentration limit j is based. Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration.
The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the containment but air could not leak in to increase oxygen concentration. Once the'c'ontainment is' filled with nitrogen to the~ required concentration, no monitoring'of oxygen concentration is necessary. However, at least twice a week the' oxygen concentration'will be determined as added assurance. Mark I Containment Long Term Program testing showed that maintaining a drywell to ;
wetwell pressure differential to keep the suppression chamber downcomer legs !
clear of water significantly reduced suppression chamber post LOCA )
hydrodynamic loads. A pressure of 1.17 psid is required to sufficiently clear j the water legs of the downcomers without bubbling nitrogen into the suppression chamber at the 3.00 ft. downtomer submergence which corresponds to approximately 84,000 ft. of water. Maximum downcomer submergence is 3.25 ft. I at operating suppression chamber water level. The above pressure differential j and submergence number are used in the Pilgrim I Plant Unique Analysis. )
1 l
l l
Amendment No. 171 l
L ..
}
B__ASES :
'3.7.A'& 4.7.A Primary Containment (Cont'd)
Post LOCA Atmosphere Dilution In order'to ensure that the containment atmosphere remains inerted, i.e. the oxygen-hydrogen mixture below the flamr.able limit, the capability to inject i '
nitrogen into the containment after a LOCA is provided. A minimum of 1500 gallons of liquid N2 in the storage tank assures that a three-day supply of ,
Na for post-LOCA containment inerting is available. Since the inerting.
makeup system is continually functioning, no periodic testing of the system is-required.
The Post-LOCA Containment Atmospheric Dilution (CAD) System is designed to I meet the requirements of AEC Regulatory Guides'1.3, 1.7 and 1.29, ASME.Section i III, Class 2 (except for code stamping) and seismic Class I as defined in the PNPS FSAR. ]
In summary, the limiting criteria are: l
- 1. Maintain hydrogen concentration in the containment during post-LOCA conditions to less than 4%.
- 2. Limit the buildup in the containment pressure due to nitrogen addition I to less than 28 psig.
- 3. To limit the offsite dose due to containment venting (for pressure control) to less than 300 Rem to the thyroid.
By maintaining at least a 3-day supply of N2 on site there wi11 be sufficient time after the occurrence of a LOCA for obtaining additional nitrogen supply from local commercial sources.) The system design j contains sufficient redundancy to ensure its reliability. Thus, it is
-sufficient to test the. operability of the whole system.once-per operating cycle. The Hz analyzers'will provide' redundancy for the drywell i.e., there are two H2 analyzers for the Unit. By permitting reactor operation for 7 )
days with one of the two Hz analyzers inoperable, redundancy of analyzing capability will be maintained while not imposing an immediate interruption in plant operation. Monthly testing of the analyzers using H 2 will be adequate to ensure the system's readiness because of the design. Since the analyzers are normally not in operation there will be little deterioration due to use. ,
In order to determine H2 concentration, the analyzers must be warmed up 6 l hours prior to putting into service. This time frame is acceptable for accident conditions because a 4% H level will not be reached in the drywell 2 l until 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> following the accident. Due to nitrogen addition, the pressure l in the containment after a LOCA will increase with time. Under the worst expected conditions the containment pressure will reach 28 psig in approximately 45 days. If and when that pressure is reached, venting from the 1 containment shall be manually initiated per the; requirements of 10CFR50.44. 1 The venting path will be .'hrough the Standby Gas Treatment system in order to )
minimize the off site dose.
(1) As listed in Pilgrim Nuclear Power Station Procedure No. 5.4.6 " Post Accident Venting".
Amendment No. 171a (the next pg. is 172) .
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BASES:
-l 3.7.C - Secondary Containment The secondary containment is Jesigned to minimize any ground level release of l radioactive materials which might result from a serious accident. The reactor i building provides secondary containment during reactor operation, when the l drywell is sealed and in service; the reactor building provides primary !
containment when the reactor is shutdown and the drywell is open, as during l refueling. Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required'as well as during refueling.
Initiating reactor building isolation and operation of the standby gas I '
treatment system to maintain at least a 1/4 inch of water negative pressure within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leak tightness of the reactor building .
and performance of the standby gas treatment system. Functionally testing the I initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Performing these tests prior to refueling will demonstrate secondary containment capability prior to the time tne primary j containment is opened for refueling. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.
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l Amendment No. 175 (next pg. is 177) l L . , . . .
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TABLE 6.9-1 Area Reference Submittal Date I
- a. Secondary Containment 4.7.C.I.c Upon completion of each Leak Rate Testing (1) test (2) .
- b. In-service Inspection 4.6.G. Five years after .]
Evaluation commercial operation
- c. (Deleted)
- d. Gross Gaseous Release 4.8.B. Ten days after the 0.05 C1/sec for 48 release occurs Hours NOTES: 1. Each integrated leak rate test of the secondary containment reauired by 4.7.C.I.c. shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency ventilation flow rate. The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.
- 2. The report shall be submitted approximately 90 days after completion of each test. Test periods shall be based on the l commercial service date as the starting point.
Amendment No. 225