ML20148M764
| ML20148M764 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 10/30/1978 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148M769 | List: |
| References | |
| NUDOCS 7811220140 | |
| Download: ML20148M764 (18) | |
Text
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b UNITED STATES
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4 NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20655 f
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,o NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO PROVISIONAL OPERATING LICENSE a
Amendment No. 36 License No. DPR-22 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The submittal by Northern States Power Company (the licensee) dated November 5,1976 as supplemented April 15 and August 29, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; i
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Corrinission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment, and 1
paragraph 3.B of Facility Operating License No. DPR-22 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A and B as revised through \\mendment No. 36, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
7811220140
2 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Thomas A ppolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: October 30, 1978
ATTACHMENT TO LICENSE AMENDMENT NO. 36 FACILITY OPERATING LICENSE NO. OPR-22 DOCKET NO. 50-263 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Pages 139 140 147A 156 157 157A 158 158A 158B 159 160 161 161A 165 167
1.0 SURVEILLANCE REQUIREMENTS 3.0 LIMITING (X)NDITIONS FOR OPERATION 4
I.7 COffrAIIMEfff SYSTEMS 3.7 CONTAI194ENT SYMDG 4
Applicability:
Applicability:
Applies to the operating status of the primary Applies to the primary and secondary and secondary containment systems.
containment integrity.
Objective:
Objective:
To assure the integrity of the primary and To verify the integrity of the primary and secondary containment systems.
Specification:
Specification,:
A.
Primary Contairrnent.
A.
1.
Suppression Pool Volume and Temperature 1.
Suppression Pool Volume and Temperature At any tilne that the reactor water tanp-erature exceeds 212 F or work is being done which has the potential to drain the vessel, except as permitted by specification 3.5.G.4, the following requirements shall be met:
a.
Water temperature during normal opera-a.
The suppression chamber water temperature shall tion shall be g90 F.
be checked once per day.
b.
Water temperature during test operation b.
Whenever there is indication of relief valve which adds heat to the suppression pool operation which adds heat to the suppression shall befr100 F and shall not be 290 F pool, the pool temperature shall be continually for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
monitored and also observed and logged every c.
If the suppression chamber water tempera-5 minutes until the heat. addition is terminated.
ture is?110 F, the reactor shall be c.
A visual inspection of the suppression chamber scranuned immediately.
Power operation interior including water line regions and the shall not be resumed until the pool temp-interior painted surfaces above the water line e ra ture is s90 F.
shall be made at each refueling outage.
3.7/4.7 139 Amendment No. 36
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS d.
During reactor isolation conditions d.
Whenever there is indication of relief the reactor pressure vessel shall be valve operation with a suppression pool depressurized to4200 psig at normal temperature;t160 F and the primary coolant cooldown rates if the suppression system pressure > 20,0 psig, an extended pool temperature exceeds 120 F.
visual examination of the suppression e.
The suppression chamber water volume chamber shall be conducted before resuming shall be2: 68,000 and:E77,970 cubic power operation, feet.
e.
The suppression chamber water volume shall f.
Two channels of torus water level instru-be checked once per day.
mentation shall be operable.
From and f.
The suppression chamber water volume indi-af ter the date that one channel is made cators shall be calibrated semimnnually.
or found to be inoperable for any reason, reactor operation is permissible only during the succeeding 30 days unless such channel is sooner made operable.
If both channels are made or found to be inoperable for any reason, reactor opera-2.
PrLaary Containment Integrity tion is permissible only during the The primary containment integrity shall be succeeding six hours unless at least one demonstrated as follows:
channel is sooner made operable.
I
- a. Integrated Prbmary Containment Leak Test 2.
Primary Containment Integrity PrLaary containment integrity, as defined (IPCLT)
(1) An integrated leak rate test shall be in Section 1, shall be maintained at all times when the reactor is critical or when perfoomed prior to initial unit opera-the reactor water temperature is above tion at an initial test pressure (Pt) 212 F and fuel is in the reactor vessel of 41 psig.
except while performing low power physics (2) subsequent leak rate tests shall be tests at abmospheric pressure during or performed without prelbainary leak de-tection surveys or leak repairs after refueling at power levels not to exceed 5 Mw(t).
immediately prior to or during the test, at an initial pressure of approximately 41 psig.
(3) Leak repairs, if necessary to peonit integrated leak rate testing, shall be preceded by local leak rate measurements where possible. The leak rate differ-140 l
3.7/4.7 Amendment No. 36
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS d.
One position alarm circuit can be inoperable b.
When the position of any dryvell-providing that the redundant position alarm suppression chamber vacuum breaker valve c ircuit is operable. Both position slarm is indicated to be net-fully closed at a circuits may be inoperable for a period not time when such closure is required, the to exceed seven days provided that all vacuum dryvell to suppression chamber differential breakers are operable.
pressure decay shall be demonstrated to be less than that shown on Figure 3.7.1 immediately and following any evidence of subsequent operation of the inoperable valve until the inoperable valve is restored to a normal condition.
c.
When both position alarm circuits are made or found to be inoperable, the contral panel indicator light status shall be recorded daily to detect changes in the vacuum breaker position.
5.
Oxygen Concentration 5.
Oxygen concentration l
a.
The primary containment atmosphere shall Whenever inerting is required, the primary be reduced to less than 57. oxygen with containment oxygen concentration shall be nitrogen gas whenever the reactor is in measured and recorded on a weekly basis.
the run mode, except as specified in 3.7.A.5.b.
b.
Within the 24-hour period subsequent to placing the reactor in the run mode following shutdown, the containment atmosphere oxygen concentration shall be reduced to less than SI by weight, and maintained in this condition. Deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to leaving the run mode for a
reactor shutdown.
f3.7/4.7 Amendment No. 36 147A
'me Bases:
3.7 A. Primary Containment The integrity of the primary containment and operation of the emergency core cooling system in combination, limit the off-site doses to values less than 10 CFR 100 guideline valties in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists.
Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure.
An exception is made to this requirement 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 which will greatly reduce the chances of a pipe break.
The reactor may be taken critical dtiring this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring.
k Procedures and the Rod Worth Minimizer would limir incremental control worth to less than 1.37. ok. A drop of a 1.37. Ak increment of a rod does not result in any fuel damage.
In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offers a sufficient barrier to keep off-site doses well within 10 CFR 100 guide line values.
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 heat released during primary system blow-down from 1000 psig.
Since all of the gases in the drywell are purged into the pressure suppression 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 maximum allowable primary containment 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. Reference Section 5.2 3 FSAR.
Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approximately 41 psig which is below the allowable pressure of 62 psig.
The nominal downcomer submergence for the Monticello wetwell desigr)1{1s 4 feet which is in conform-ance with most of the Bodega tests. The majority of Bodega tests l were run with a submerged (1) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.
3.7 BASES 4
g s Continuedt I
length of four feet, which resulted in complete condensation. Thus with respect to downconer submergence, this specification is adequate.
The maximum temperature at the end of blowdown tested during the Humboldt Bay and Bodega Bay (2) 0 tests was 170 F and this is conservatively taken to be the limit.for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170 F.
Ilmperimental data indicate that excessive steam cond< nsing loads can be avoided if the peak temperature of the suppression pool is maintained below 160*F daring any period of relief valve operation with sonic conditions at the discharge exit.
Speci fications have been placed on the envelope of reactor operating conditions so that the reactor can he depressurized in a timely manner to avoid the regiate of potentially high 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 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 separated from that of the stuck-open relief valve to assure mixing and unifonnity of energy insertion to the pool.
7er an initial maaimum suppression chamber water temperature of 90 F and assuming the normal com-pleak nt of containr.cnt cooling piunps (2 LlY I renps and 2 containment ed inc service wt ter pgs).
containment pressure is not required to renin,ain adequ:ite net posit,1ve soct.ic,a head (NPSil) for the core spray, LpCI and HPCI pumps. However, during an atproximately one-day period strrting a few a
hours after a loss-of-coolant, accident, should one Rl!R loop be inoperable and should the containneht pressure be reduced to atmospheric pressure through any means, adequate Upsil would not be.available.
Since an extreme}y degraded conriition must. exist, the period of vulnerability to this event is re-stricted by Specification 3.7. A.I.b by limiting the suppression pool initial temperature and the period of operation with one inoperable Rl[R loop, llJ Robbins, C.
H., " Tests of Full Scale 1/18 Segment of the Humboldt Bay Pressure 4
Suppression Containment," CEAP-3596, November 17, 1960.
(2) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.
l3.7 BASES 157 Amendment No. 36
Bases Continued:
1 1
If a loss of coolant accident were to occur when the reactor water temperature is below 330*F, the containment pressure will not exceed the 62 psig design pressure, even if no condensation were to The maximum allowable pool temperature, whenever the reactor is above 212*F, shall be governed occur.
by this specification.
Thus, specifying water volume-ten.perature requirements applicable for reactor-water temperatures above 212*F provides additional margin above that available at 330*F.
The large amount of water that must be added or removed to cause a significant change in the suppression chamber water inventory is not likely to go un-noticed. With a daily check of water volume, there is an extremely low probability that a loss of coolant accident will occur sbaultaneously with water volume 4
being outside of the specified range. Two indicators provide redundant readings for comparison (with no automatic action initiation). The provisions allowing one or both indicators out of service are consistent with the need for a redundant indicator and the frequency for checking the volume, respectively.
In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed 4
which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The maintenance of a suppression chamber water level corresponding to a downcomer submergence range of 4.54 to S.62 feet will assure the integrity of the suppression
~
chamber when subjected to post-LOCA suppression pool hydrodynamic forces.
d 4
?
T l
3.7 BASES gg7g i
Amendment No. 36
Bases Continued:
1 The purpose of the vacuum relief valves is to equalize the pressure betueen the'drywell and suppression chamber and between the suppression chamber and reactor building during loss of coolant accident so that structural integrity of the containment is maintained.
The vacuum relief system between the pressure suppression chamber and reactor building consist of two 100% vacuum relief breakers (2 parallel sets of 2 valves in series). Operation of either system will maintain the pressure dif ferential less than 1 psig. The external design pressure is 2 psig.
One valve may be out of service for repairs for a period of seven days.
This per'.od is based on the low probability that system redundancy would be required during this time.
If.rpairs cannot be completed within seven days, the teactor coolant system is brought to a condition where vacuum relief is no longer required.
The capacity of the ten (10) drywell vacuum relief valves is sized to limit the pressure differential between the suppression chamber and drywell during post-accident drywell cooling operations to less than the design limit of 2 psi.
The relief valves are sized on the basis of the Bodega Bay pressure suppression system tests.
Since they are in series with the reactor building to suppression chamber vacuum relief valves pressure drop across these valves must be included in the evaluation of drywell negative pressures, even though there does not appear to be a mechanism for causing negative pressures in excess of the 2 psi design pressure. With eight of the ten valves in service, the dif ferential pressure across the valves for maximum flow conditions would increase. With this additional pressure drop the total differential pressure would still be less than the 2 psi design valve.
Containment integrity would therefore not be impaired.
In addition to the above considerations, postulated leakage through the vacuum breaker to the suppression chamber air space could result in a partial bypass of pressure suppression in the event of a LOCA or a small or intermediate steam leak. This ef fect could potentially result in exceeding containment design pressure. As a result of the leakage potential, the containment response has been analyzed for a number of postulated conditions.
It was found that the maximum allowable bypass area for any l
postulated break size was equivalent to a six-inch diameter openinR.1 This hypass corresponds to a I Report on Torus to Drywell Vacuum Breaker Tests and Modifications for Monticello Nuclear Generating Plant, dated March 12, 1973, submitted to Mr. D. J. Skovholt, AEC-DL, from Mr. L. O. Mayer, NSP l3.7 BASES 158 Amen &nent No. 36
l Bases Continued:
l One inch opening of any one valve or 0.1 inch opening for all ten valves, measur ed at the 1ottom of the disc with the top of the disc at the seat.
The position indication systen is designed to detect closure within 1/8 inch at the bottom of the disc.
each refueling outage and following any sigificant maintenance on the vacuum breaker valves, At positive seating of the vacuum breakers will be verified by leak test.
The leak test is conservatively designed to demonstrate that leakage is less than that equivalent to leakage through a one-inch orifice which is about 3% of the maximum allowable. This test is planned to establish a baseline for valve performance at the start of each operating cycle and to ensure that vacuum breakers are maintained as nearly as possible to their design condition. This test is not planned to serve as a limiting condition for operation.
During reactor operation, an exercise test of the vacuum breakers will be conducted monthly. This test will verify that disc travel is unobstructed and will provide verification that the valves are closing fully through the position indication system.
If one or more of the vacuum breakers do not seat fully as determined from the indicating system, a leak test will be conducted to verify that leakage is within the maximum allowable. Since the extreme lower limit of switch detection capability is approximately 1/16", the planned test is designed to strike a balance between the detection switch capability to verify closure and t he maximum allowable leak rate.
A special test was performed to establish the basis for this limiting condition. During the first refueling outage all ten vacuum breakers were shimmed 1/16" open at the bottom of the disc.
The bypass area associated with the shimming corresponded to 63% of the maximum allowable.1 The results of this test are shown in Figure 3.7.1.
When a drywell-suppression chamber vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights at the remote test panels are designed to function as follows:
Full Closed 2 Green - On 2 Red
- Off Intermediate Position 2 Green - Off 2 Red
- Off Full Open 2 Green - Off 2 Red
- On The remote test panel consists of a push bucton to actuate the air cylinder for testing, two red lights.
l 3.7 BASES Amendment No. 36 158A
l Bases Continued:
and two green lights for each of the ten valves. There are four independent lirit switches on eac' valve.
The two switches controlling the green lights are adjuste1 to provide an indication of disc opening of less than 1/8" at the bettom of the disc.
These switches are also used to activate the valve position alarm circuits. The two switches controlling the red lights are adjusted to provide injication of the disc very near the full open position.
The control room alarm circuits are redundant and fail safe.
This assures that no simple failure will defeat alarming to the control room when a valve is open beyond allowable and when power to the switches falls. The alarm is needed to alert the operator that action must be taken to correct a malfunction or to investigate possible changes in valve position status, or both.
If the alarm cannot be cleared due to the inability to establish indication of closure of one or more valves, additional testing is required.
The alarm system allows the operator to make this evaluation on a timely basis. The frequency of the testing of the alarms is the same as that required for the position indication system.
Operability of a vacuum breaker valve and the four associated indicating light circuits shall be established by cycling the valve.
The sequence of the indicating lights will be observed to be that previously described.
If both green light circuits are inoperable, the valve shall be considered inoperable and a pressure test is required immediately and upon indication of subsequent operation.
If both red light circuits are inoperable, the valve shall be considered inoperable, however, no pressure test is required if positive closure indication is present.
The 5% oxygen concentration minimizes the possibility of hydrogen combustion following a loss of coola ' accident. Significant quantities of hydrogen could be generated if the core cooling systems failed to sufficiently cool the core. The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is more probable than the occurrence of the loss of coolant accident upon which the specified oxygen concentration limit 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 prbnary 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 prim ary contaignent is noomally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could i
leak out of the containment but air could not leak in to increase oxygen concentration. Once the con-tainment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is
)
necessa ry.
Ilowever, at least once a week the oxygen concentration will be determined as added assurance.
j 3.7. BASES 158B Amendment No. 36
-=-__-_
.-__---=--_-_=-- _-- - - _ __
- _ = =
= ~ _ - _
B:toca Continued:
B.
Standby Gas Treatment System and C. Secondary Containment The secondary containment is designed to minimize any ground level release of radioac:.ive materials which might result from a serious accident.
operation, when the drywell is sealed and in service;'ilte reactor building provides secondary containment during reacto the reactor is shutdown and the drywell is open, as during refueling.the reactor building provides primary containment an integral part of the complete containment system, secondary containmentBecause the secondary containment is is required at all times that primary containment is required except, however, for initial fuel loading prior to initial power testing.
lhe standby gas treatment system is designed to filter and exhaust the reactor building atmosphere to the chimney during secondary containment isolation conditions, with a minimum release of radioactive materials from the reactor building to the environs. One standby gas treatment system circuit is designed to auto-matica11y start upon containment isolation and to maintain the reactor building pressure at the design negative pressure so that all leakage should be in-leakage.
Should one circuit fail to start, the redundant alternate standby gas treatment circuit is designed to start automatically. Each of the two circuits has 1007. capacity.
Only one of the two standby gas treatment system circuits is needed to cleanup the reactor building atmosphere upon containment isolation.
If one system is found to be inoperable, there is no inmedia te threat to the containment system performance. The re fo re, reactor operation or refueling operation may continue while repairs are being made.
If neither circuit is operable, the plant is placed in a condition that does not require a standby gas treatment system.
3.7 BASES 159 Amendment No. 36
-_ =_
~
i Bases Continued:
I While only a small amount of particulates are released from the primary containment as a result
~
of the loss of coolant accident, high-efficiency particulate filters before and after the charcoal l
filters are specif*ed to minimize potential particulate release to the environment and to. prevent clogging of the charcoal adsorbers. The charcoal adsorbers are installed to reduce the potential-release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1% bypass leakage for ";e charcoal adsorbers using halogenated hydro-carbon and a HEPA filter efficiency _ of at least 99% removal of DOP particulates. Labora tory carbon sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. Operation of the standby gas treatment circuits significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.
If the performance requirements are met as specified, the calculated doses would be less than the guidelines stated in 10 CFR'100 for the accidents analyzed.
D.
Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary 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. Details of the isolation valves are discussed in Sections 5.2 and 7.2 of the FSAR.
4 i
]
T i
Amendment No. 36 h
~
Bases:
4.7 A. Primary Containment Tiye 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 weekly check of the temperature and volume is adequate to assure that adequate heat removal capability is present.
For additional margin, these will be checked once per day.
The interiors of the drywell and suppression chamber are painted to prevent rusting.
The inspec-tion of the paint during each major refueling outage, approximately once per year, assures the paint is intact and is not deteriorating. Experience with this type of paint indicates that the inapection interval is adequate.
Because of the large volume and thermal capacity of the suppression pool, the volume ar.d temperature normally changes very slowly and monitoring these parameters daily is sufficient 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. partit ular 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.
Visual inspection of the suppression chamber including water line regions each refueling outage is adequate to detect any changes in the suppression chamber structures.
The primary containment preoperational test pressures are based upon the calculated prinary containment pressure response in the event of a loss of coolant accident. The peak drywell pressure would be ebout 41 psig, which would rapidly reduce to 25 psig within 10 seconds follow-ing the pipe break.
Following the pipe break, the suppression chamber pressure rises to 25 psig within 10 seconds, equalizes with drywell pressure and thereafter' rapidly decays with the dry-well pressure decay. See Section 5.2.3 FSAR.
The design pressure of the drywell and absorption chamber is 56 psig. See Section 5.2.3 FSAR.
The design leak rate is 0.5%/ day at a pressure e. 56 prig.
As indicated above, the pressure response of the drywell and suppression chamber following an accident would be the same af ter about 10 seconds. Based on the calculated cantainment pressure response discussc i above, the primary containment,preoperational test presr.ures were chosen. Also, t e e'l on the pritmry containmen pressure response and the fact that the dryi,:cIl and suppression charnber ibnction as a unit, the primary containment will be tested as a unit rather than the individual compo-nents separately.
4.7 BASES 161 Amendment No. 36
1 4
Bases Continued:
1 The design basis loss of coolant accident was evaluated at the primary containment maximum allowable accident leak rat e of 1.51 day at 41 psig. The analysis showed that with this leak 4
4.7 BASES 161A Amendment No. 36
Bases Continued:
l B.
Standby Gas Treatment System, and C.
Secondary Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain the design negative pressure within the secondary containment provides an adequate test of the reactor building isolation valves and the standby gas treatment system.
Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system operational capability.
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Standby gas treatment system inplace testing procedures will be established utilizing applicable sections of ANSI N510-1975 standard as a procedural guideline only.
Redundant heaters in the standby gas treatment system room prevent moisture buildup on the adsorbent.
If painting, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use.
Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52 Revision 1 (June 1976). The charcoal adsorber efficiency test procedures will allow for the removal of one representative sample cartridge. The sample will be at least two inches in diameter and a length equal to the thickness of the bed.
If the lodine removal efficiency test results are unacceptable, all adsorbent in the system will be replaced. liigh efficiency particulate filters are installed before and after the charcoal filters to prevent clogging of the carbon adsorbers and to minimize potential release of particulates to the environment. An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by inplace testing with DOP as the testing medium. Any HEPA filters found defective will be replaced with filters qualified pursuant to regulatory guide position C.3.d of Regulatory Guide 1.52 Revision 1 (June 1976). Once per operating cycle demonstration of HEPA filter pressure drop, operability of inlet heaters at rated power, automatic initiation of each standby gas treatment system circuit, and leakage tests after maintenance or testing which could affect leakage, is necessary to assure system performance capability.
h.7 BASES 165 Amendment No. 36
_. _ _ _. -_j
Bases Continued:
The containment is penetrated by a large number of small diameter instrument lines.. A program for the periolic testing (see Specification h.7.D) and exar.ination of the valves in these lines has been developed and a report covering this program was submitted to the AEC on July 27, 1973.
The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.
4 4.7 BASES 167 Amendment No. 36
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