ML20237H442
| ML20237H442 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/30/1986 |
| From: | AFFILIATION NOT ASSIGNED |
| To: | |
| References | |
| LRP-B-022, LRP-B-22, NUDOCS 8708170216 | |
| Download: ML20237H442 (20) | |
Text
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1 9.3.4.2 Chemical Addition Svste
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9.3.4.2.1 Design Bases me YON Chemical. addition operpticus are requirec to alter the concentration 'of various chemicals in the reactor coolant and auxiliary systems. The chemical addition system is designed to add boric acid < cne reactor. coolant system for reactivity.
control, lithium hydroxide for pH control, and hydra:ine for ' oxygen control.
The' system also'provides boric acid for other plant components, and is sized to be able co' add sufficient boric acid to maintain the core 1* ak/k sub '
critical at any time during lif e.
1 A single beric acid six tank is provided as a basic source of concentrated boric acid solution for use throughout the nuclear portions of the unit.
j The quantity of boric acid retained in this tank in addition to that retained in the aclai:ed boric acid tanks of the radioactive liquid waste system provide more than stificient boric acid solution to increase the boren cmentration of the reactor coolant system to that required for cold'shutdova.
Two boric acid pumps are required to facilitate transfer of the concentrated boric acid solution frem the boric. acid =1x tank to upstrea= cf the' makeup filter, the spent fuel storage pools, the borated water storage tank, the core flooding makeup tank, and the neutrali:er tank. The two pumps are sized so that when bein are operating, a complete charge of concentrated boric acid solution from the boric acid six cank may be injected into the reactor. coolant
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system in approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Heaters are provided in the beric acid six cank and the boric acid pumps and piping system are heat traced to ensure that
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the boric acid solution contained herein is always =aintained in the dissolved.
state.
The volume of the lithium hydroxide =ix tank is' based on maintaining a suf-ficient quantity of lithium hydroxide solution available fer addition to the reactor coolant system so that a concentration between 3 and 6 ppm can be j
maintained in the primary coolaat while letting down at the tximu=' race.
The capacity of the lithium hydroxide pump is also based on this criterien.
The core flooding makeup tank and pu=p provide beric acid makeup.of a prede-l ter:1;.M boren concentration in the core flooding canks. The syse d is de-signed to deliver makeup in the event cf a check valve leak at the core flood-ins reacter coolant interface.
i The reclaimed boric acid tank provides storage for the reclaimed beric acid from the reactor coolant evaporator.
If. the beric acid radioactivity is acceptable, it is reused in the beric acid mix tank; otherwise, it is sent to the Unit 2 reactor coolant evaporator; i
The volume of the caustic mix tank is based on providing adequata caustic to the neutralize tanks, the =1sceJlaneous vaste tank and the sodium hydr xide tank.
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k The resin addition tank as shown on F1;ure 11.2.1 is sized to per=it gravity i
replace =ent of resin in any of the dc=ineralizers of the =akeup and purifi-I f
cation, radioactive liquid waste, or spent fuel purification syste=s.
The routing of resin to the proper pair of de=ineralizers is assured by the =anuai installation of a flexible connecticn between the resin tank outlet and the I
resin fill line to the decineralizers being serviced.
The hydrogen supply syste= consists of a hydrogen manifold which =aintains adequate hydrogen partial pressure in the gas space of the =akeup tank to limit the dissolved oxygen content in the reactor coolant to an acceptable level during nor=al operation. Redundant isolation devices will be provided in the hydrogen charging line, with safety grade control for valves and position indication.
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Two nitrogen =anifolds are provided.
One supplies fresh nitrogen blanket 4
gas to the =akeup tank, the low pressure vent header portion of the radie l
active waste gas disposal syste= and provide =1scellaneous nitrogen source l
throughout the nuclear sy:ce;:s. The second nitrogen =anifold provides high pressure nitrogen to the core flooding tanks.
9.3.4.2.2 syste= Description I
l The che 1 cal addition =yste= is ent' prised of a nu=ber of individual syste=s (each hav1=g their =ajor equip =ent components and piping systc= with valves
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and controls) as required to perfor= various functions relate _ to the operatien of the pri=ary syste=, the spent fuel pools, and the radioactive liquid and I
gas waste disposal syste=1.
The chemical addition syste= is shown in Figure 1
9.3-7.
The =ajor equipment co=ponents of the chemical addition and sa=pling syste=s are liste.d with the perti::ent data in Table 9.3-6.
l Other che=ical addition equip =ent is provided in the turbine building as re-
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quired to maintain the quality of feedwater to the stea= generators as indi-i cated in Table 10.4-1.
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~he chemical addition system provides. gas manifolds, tanks.. pumps, piping.
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systems. and associated valves and contrelsLas required to furnish the' following functions:
Item Function
- 1. Boric Acid Mix Tank and Provides a, source of fresh, concentrated-Pumps boric acid solution for chemical shim-control. in the primary system, makeup to spent, fuel pool, and borated, vater to the borated water storage tank.~-
- 2. Lithium Hydroxide Mix Provides a" strong basic' solution for pH' Tank and Pump.
controlief primary coolant.'
- 3. Core Flooding Makeup Provides~ makeup'bdric acid to the core t
Tank and Pump flooding tanks.
- 4. Reclaimed Boric Acid Tank Provides storage forfreclaimed boric' acid free the reactor coolant evaporator.
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- 3. Caustic Mix Tank and Pump Provides Na0R for regenerating deborating resins or a strong acid for neutralizing-solutions in the'neutra11zer tank and mis-cellaneous vaste tank. 'Also used-for pro-viding makeup to the' sodium hydroxide g
storage tank.
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- 6. Hydrazine Drums and Pump Provides ~ a reducing agent for. control of exygen dissolved in reactor coolant.
- 7. Rasin Addition Tank Provides gravity fed chargers of'f:esh resin to dominera112ers in thel makeup and purification,. spent fuel purification, and radioactive liquid vaste systems.
The-resin is gravity fed.co the proper l
dominera11:ers by the manual installation of'a flexible" connection'.
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- 8. Hydrogen Manifold Provides hydrogen for control'of dis-solved' oxygen'in reactor coolant.-
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- 9. Nitrogen WM fold Provides. blankat gas for makeup-tank and:
vaste gas system and makeup gas to core.
flood" tanks.
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The. individual systa=s of the chemical addition system are operated' independently i
as' required co perform the variousL functions each is designed to provide.. The 1
only fully automatic functioning systems are the nitrogen: manifolds.. :The functions cf all other systems comprising the chemical: addition' system are initiated'by.
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operator action and nay. thereaf ter be subject to local'or remote manual ~ control.,
automatic control, or a. combination of chase.
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filter, the spent fuel s:crage pccis, the berated vater s:crage tank, the ccre
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ficediss makeup tank, e.=d the neutrali:er task.
I 9 3.'.2.3 safety Evalua:fo:
i 9 3.i.2 3.1 Reliabi i:7 considera:1 ens I
Tc.e chemical additic: syste is c: required :c fu:ctics during r.: e=ergency i
conditica. Redundant beric acid pu=ps and independent beric acid addi:ic:
-- e s e.-.,.-. v. de'. *. e a. d.
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syste inadequate. Scric acid 's available for berati:g froc the beric acid mix tack as well as frc the eclai=ed beric acid s:srage ta:k. To facili-tate the handling cf eccee : rated beric acid, flush ctnecticas are provided that alicv the lines cac fing 1: : be cleased when necessary.
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handled by this syste: can be aistained abcve its cryst,tilicatic te=pers-I ture by any c:e cf a: less: Ovc independent, full capacity her.:ing systems.
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Even if beric acid sciiiiricatic: iid =ake the systen incperable, the reac-tcr can still be shutdown safely.
In such a situatics, sufficient bercn 3
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v pumps, fic: the berated vater s:cre.ge tank.
9,3.L.2 3 2 7e:1ure Analysis 4
i A failure e. a'.ysis cf the chemical edditic' systes is give: ?: Te.ble 9.3-7 c de ctstrate that the syste: has sufficie:: redundancy c maintai: -he i
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Ne contai==ect vessel isciatic: is required of this cyste: since its beund-tries dc ne penetrate the ecstainment vessel.
9.3.k.2 3.L Leakage Ccesideratic:
This syste: delivers additives Oc -he st.en; fuel s:crar.e =.cci a:4 the takeu; and berated vater s:crage taxis.
2ackficv frce -he ta:Es c -he beric acid
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The functien of the hydrogen supply sys'.em is to provide anc saintain a blanket of nydrogen witnin the makeup tank.
Four manifold storage bottles.
three spares and the pressure reducing valves are located in the yard shed j
adjacent to che, Auxiliary Building.
Each bottle contains 7 fc3 of hydrogen l
at 2'00 psi.
i The hydrogen line is routed from the yara shed to the roof of the auxiliary building, where it enters the auxiliary buf1 ding and becomes a seismic Category I line run inside of a seismic Category guard pipe. The guard pipe is i
capped and terminated at the makeup cank enclosure vall (see Fig.1.2-9).
l At no point along its routing is the hydtcgen line exposed to danger of fai*ure either of high energy piping or non-seismic Category I equipment. In the un-I likely event of a hydrogen line rupture, hydrogen vill be vented out the guard i
pipe vent.
If a break occurs within the makeup tank cubicle, the gas vill be I
dispersed by the Auxiliary Building EVAC system. A line break in the cubicle results in a 126 scfm flew race for appro -1,sately 8 minutes. HVAC dispersion netvichstanding, there is no safety related equipment in the makeup tank cubicle. Should the hydrogen regulating valve diaphragm fail, the hydrogen i
line vill be protected from overpressurization by either a relief valve which l
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vents near the yard shed or a rupture disc which relieves out the guard pipe I
vent.
There is no hydrogen piping routed in any other safety related areas of the plant.
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'A 9.3.4.2.4 Tests and Inspections Active and passive components of the chemical addition system will.be' examined periodically to determine the operating condition.' - Periodic visual inspections and preventive maintenance will be conducted according to sound maintenance practice.
1 9.3.4.2.5 Instrumentation ' Application'.
The instrumentation of the chemical addition system provides measurements which are used to indicate, record, alarm, ' interlock and control process variables such as level and flow as follows:
The beric acid mix tank temperature is measured, indicated locally a.
and a signal is transmitted that will actuate alarms in the control A signal is also providec for heater control to maintain room.
fluid temperature within a predetermined range.
b.
Ste boric acid mix tank level is measured, indicated locally and.
a signal is transmitted that will actuate alarms located in the control room.
An' interlock signal is also provided to de-energiz'e.
the heaters on low level.
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The following process variables are measured 'and locally indicated:
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1 1.
Hydrazine pump discharge pressure 2.
Hydrazine pump flow rate 1
3.
Lithium hydroxide pump discharge pressure 4.
Lithium hydroxide pump flow rate.
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1 5.
Lithium hydroxide mix tank level 6.
Boric acid pump discharge pressure 7.
Sodium hydroxide pump discharge pressure b
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5.2.6 PUMP FLYM1 EELS An analysis to determine the adequate margins of safety for operation of pump flywheels will be presented as discussed in 5.5.1.5.
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5.2.7 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS Principally, the reactor coolant system is located within the secondary shield-ing and is inaccessible during reactor operation. Any leakage in the.contain-I ment normally drains to the reactor builditg sump. Any reactor coolant leakage to the, containment atmosphere will be in the form of liquid and vapor. The liquid will drain to the reactor building sump and the vapor will be condensed l
in the reactor building air cooling units drip pans and will also reach the J
cump via a drain line from the coolers.
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All power operated valves containing reactor coolant, located within the reac-tor building, have two sets of stem packing, with a leakoff connection between tha packing sets which ultimately drain to the Reactor Coolant drain tank.
j For the reactor coolant pumps, any leakage past the backup seal drains to i
the reactor coolant drain tank. The reactor pressure vessel head gasket is also provided with a drain line to the teactor coolant drain tank.
I Reactor coolant pressure boundary leakage is detected by several techniques.
The most direct method is measurement of reactor coolant makeup requirements.
l Additional methods include reactor building sump mlume measurement, reactor
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building air cooling co21 condensate volume measurement, and released radio
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activity monitoring. Each one of the additional techniques has limitations.
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Sump or condensate volume measurements will be a measurement of total primary, t
sacondary, and auxiliary system leakage within the reactor building; activity monitoring is most sensitive when there is defective fuel cladding and the retetor has been at power for some period of Mme; condensate volume measure-ment will be af fected by ambient humidity conditions.
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In order to continue reactor operation, reactor coolant leakage must not ex-ceed the makeup capability of the system. Additionally, unidentified leakage cannot be at a rate high enough to cause rapid propagation of a crack in the i
reactor coolant pressure boundary.
Locating the actual point of reactor coolant system leakage can be most rapidly accomplished when the reactor is in the hot shutdown condition, ing personnel access inside the secondary shield.
thereby allow-
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Location of leaks can then ba determined by visual observation of escaping steam or water, or the presence of leaking coolant.
5.2.7.1 Detection Methods i
S:vstal methods are employed to detect reactor coolant leakage.
Continuous monitoring of the pressurizer and the makeup tank levels provides a measure cf leakage from the reactor coolant pressure boundary.
b2 dzrived from the difference in flows between the letdown and the makeup as A second estimate can data mined by flow monitors located at tor coolant seal injection and return flews.these points taking into account reac-the reactor building sump provides a third indicationIntegrated flow of water from of leakage to the reac-(
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tor building. The condensate level in the reactor building drip pans can also provide an indication of leakage inside the reactor building. Reactor building i
i activity monitors detect reactor coolant leaka6e to the containment by increased i
activity levels due to the release of radioactive isotopes from the leaking fluid.
5.2 7.2 Indication of Reactor Coolant Pressure Boundary Leakage All of the leakage detection systems utilized provide positive indications of leakage from the reactor coolant pressure boundary. The parameters monitored are as follows-i a.
Tank Levels
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Both the makeup tank and the pressurizer levels are displayed on recorders in the control room.
The reactor coolant drain tank level is indicated in the control room. After correction of these records for changes in reactor power level and reactor coolant pressure and temperature, the level traces prcvide an indication of leakage across the reactor coolant I
pressure boundary.
b.
Flow Indication i
Both the flow upstream of the makeup flow control valve and the flow down-stream of the letdown block orifice are indicated and recorded in the control room together with the difference in flow. Both the total seal injection flow and the four individual seal return flows are indicated in the control room I
c.
Sump Level The sump water level is continuously monitored. A "s.lgh" sump water lev el is annunciated at panel WDL 301A, located in the auxiliary building, and su=p pump operation is automatically initiated. The pumps continue to operate until a predetermined "lov" level is reached. By observing either the frequency of pump operation and knowing the sump capacity removed or the level trace, an estimate of the leakage rate can be made.
d.
Reactor Building Cooling Unit Condensate level Condensate level in the drip pans of the reactor building air cooling units provide an alarm signal from each' unit whenever there is a flov out of each unit drip pan of more than h GPM.
An alarm provides a means of corroborating other potential leak indications.
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Reactor Building Activity Monitors The reactor building air activity monitors provide readout in the control room with a high level alarm setpoint.
1 5.2-17 Am. 38 (2-20-76)
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5.2.7.3 Maximum Allovable Leakage
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l The maximum proposed total leakage rate for the reactor coolant pressure boun-I dary is specified by the Technical Specification, 3.1.6, as 30 gpm.
The reactor coolant makeup system is capable of providing 300 gpm of makeup under normal operating conditions. The maximum unidentified allowable leakage
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is 1 gpm.
5.2.7.h Unidentified Leakage i
fm anticipated leakage rate of 10 gpd has been assumed. After a B&W nuclear i
steam supply system similar to 'INI-2 provides operating experience, this leak-4 age rate vill be re-evaluated and adjusted if necessary.
3 The sensitivities of the various monitoring systems for detection of departure i
from the anticipated leakage rate are as follows:
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a.
Tank Levels l
The pressurizer level is monitored over the range of 0 to k00 inches of vater by measuring the differential pressure with instrumentation having
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i a sensitivity of 0.01% over the span and an overall accuracy of.0.707% -
l at reference conditions. The instrument is calibrated for the normal operating environment. However, it is assumed that a change of 1% of j
the range vould, as a minimum, be observable to the operator. At a rate
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l of 10 gpd, as much as 10 days could pass before a change in the pressuri-zer level of this magnitude vould be detectable. Any increase in leaka6e vould be detected in a much shorter period of time. As 1 gpm, the same level vould begin to be detectable in 1-1/2 hours.
i The reactor coolant drain tank is L.onitored over a range of 70 to 95 inches with an overall accuracy of f.1% full scale. Assuming 1% scale change, a 10 gpd leak would not be detected for 3 days. A 1 gpm leak would be detected in 1/2 hour.
The makeup tank level is monitored with the same degree or accuracy and sensitivity as the pressurizer although the range is 0 to 100 inches.
Assuming the same 1% of scale change as an observable indication of. change, a 10 gpd leak would not be detectable for approximately 3 days. However, the makeup flow and, subsequently, the makeup tank level are controlled by the pressurizer level which, in the worst case, would change by as much as 3 inches before the makeup flov (and tank level) change.
At 10 gpd, 10 days could elapse before the makeup tank lei vould start to change, and an additional 3 days before the change u.cank level was detectable.
At 1 gpm a change would be noticeable in approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The above assumes that the leakage is upstream of the letdown flow centrol element and downstream of the makeup control valve.
Any leakage elsewhere in the makeup and purification system would also change the maxeup tank level.
However, since such leakage would be outside of the reactor
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building, it could be detected and eliminated from further consideration.
l 5.2-18 Am. 36 (12-17-75)
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Flev Indication
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Both the letdown and makeup flows are continuously monitored with the same degree of accuracy and sensitivity. These correspond to O to 20%,
l 11.6% accuracy and 20 to 100% scale,11.53% accuracy, respectively.
The seal injection flow is monitored with an accuracy of 11% of 80 gpm (full scale). The individual seal return flovs are monitored with an accuracy of 12% of 3.5 gym (full scale).
The sensitivity of each conitor is 0.0707% for a range cf 0 to 160 gps.
Assu=ing a 1% change in scale necessary for operator detection, a minimum change of 1.6 gpm would be detectable. As noted previously, a change in makeup flov is subject to pressurizer level change and with a 10 spd l
leak, the flov might not initially change for a period of 10 days. ~'
Except for the pressurizer the above sensitivities are independent of variable magnitude at the time of the change; however, correction vill be required as indicated and for changes in pressure and te=perature l
fluctuations, during the reporting period, in the reacter coolant syste=.
c.
Su=p level Observation of the water level in the reactor building sump indicates the total leakage of reactor coolant, secondary coolant, 'and auxiliary
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cooling vater within the reactor building. The reactor building sump level is continuously monitored with readout located in the auxiliary building.
"Eigh" and " low" levels are annunciated and indicate a level change of 6 inches which corresponds to a volu=e of 105 gallons. The 4
sump pump starts at the "high" level and ceases operation at the " low" level. Changes in sump water level as indicated by the "high" and "lov" level alarms or frequency of pu=p operation vill lead to an estinated leakage rate. At the anticipated leakage rate of 10 gpd, approximately 10 daya vould elapse before the leak would be detected by this method.
In addition to the level alarms, the level of the su=p is monitored over its entire 6 foot depth. The minimum detectable level change is 1 inch corresponding to 17 5 gallons. At the anticipated normal estimated leakage rate, h2 hours vould elapse before the leakage vould be detected and quantified.
The above sensitivities in terms of sump level and equivalent volu=e are independent of the magnitude of the leak. Any increase in leakage abcve the anticipated rate. vould result in a smaller period of time necessary for leakage evaluation. However, the readings vould have to be corrected for leakage other than reactor coolant leakage into this sump. In any event, sump level readings can be used as an independent leakage moniter-ing method to be compared with other monitoring schemes.
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d.
Reactor Building Cooling Unit Ccndensate Level Another potential indication of leakage is the monitoring of condensation on the coils of the reactor building air cooling units.
Each of the five cooling units has a drip pan which collects condensate and any cooling vater which may leak through a defect in the cooling coil.
Each drip pan is monitored by means of a level switch, which is set to actuate at a level that indicates a flow of h GPM through the drip pan drain line.
Since the reactor building humidity is not controlled directly and can vary I
due to the humidity associated with the most recent purge as ven as changes in the temperature of the cooling vater, this method is not to be relied upon for quantitative'infor=ation. Rather it. serves as an independent method to corroborate other measurement techniques.
i e.
Reactor Building Air Activity Monitors Leakage from the reactor coolant pressure boundary may also be detected by the reactor building air activity menitor as long as activity is present in the reactor coolant. Potential sources of activity are fission pro-ducts escaping as a result of defects in the fuel cladding, activated corro-sion products and neutron interaction with tramp uranium.
Leakage of reactor coolant into the reactor building vill release noble gases, iodines i
and particulate into the atmosphere. Each of these three releases is sub-ject to quantitative measurement by an atmospheric monitor. Unlike the volumetric means of detection, all of the radiological measurement sensiti-vities are dependent upon the existing background at the time the measure-g ment is taken; as the background increases, sensitivities decrease. Unlike
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the volumetric methods, leakage from sources other than the reactor coolant system into the reactor building are not expected to significantly affect the atmospheric activity.
I Each of the means of radiological measurement has been analyzed to detemine l
its sensitivity under the vorst conditions. In all cases a 10 gpd leak into the reactor building was assumed to exist continuously for a period of time sufficient to establish a peak background. At that time a step increase to 1 gpm leakage was assumed and the time period to achieve a 100% change in the gpm/ min was calculated. The 100% level change was chosen as the change required to 1.rovide a positive visual indication of operating personnel that an increase in atmospheric activity had occurred.
To account for uncertainties in particulate behavior as well as plateout in the monitor sample lines, it was assumed that only 1 part in 10,000 of particulate reached the detector. For iodine it was assumed that 1 part in 200 would reach the detector.
the above activity releases, particulate monitoring was the most sensitive regardless of whether corrosion products alone or fission products correspond-ing to 0.1% defective fuel cladding existed in the coolant. The n(xt sensi-tive means of detection vould be iodine monitoring followed by noble gas detec-tion. The results of this analysis are contained in Table 5.2-13. " Sensitivity of Beactor Building Atmospheric Monitor to Reactor Coolant Leakage into the Reactor Building. " Reactor coolant activities are taken from Tables 11.1 L and 11.1-6, contained in Chapter 11.
In order to determine the crack size required for a leakage rate of 1 gym, the following crack form was assu=ed:
5.2-20 Am. 38 (2-20-76)
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x v = Displacement, in.
r = distance from crack tip C = 1/2 crack length, in.
According to reference 1, the displacement (V) in the Y-direction can be found to be:
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7 From reference 2 the velocity (U) out of the crack may be shown to be:
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P2 = containment pressure (15 psi),
I o = density, 1bm/ft3, f = friction factor (assumed 0.06),
f T = thickness of shell or pipe considered, in.,
j d = equiv. diameter of crack, in.,
K = shock factor for expansion and contraction (1.5),
2 g = gravitational acceleration (32.2 ft/s ),
I For the 28-in. ID cold leg piping:
l ID = 28.625 in.,
T = 2.25 in. (excluding cladding thickness),
p = h6.4 lbm/ft3.@ 557F, y = 30.h x 103 O
psi 6.570F (106. Grade C),
G = 11.5 x 106 p31, l
i y = 0.3 Assuming a crack length of 1-1/2 in. (C = 3/h in.)
(
V = 0.00192 in.
flow area ( A) = (b) (C) (V/2) = 2CV = 0.00296 in.2 L
h(2CV)
(
d = y,A
= 2V = 0.0039h in.
=
p, g
1 (2200-15) (2) (32.2) (1kh)
U=
(k6ek).(0.06) (2.25)
- 1*5- = 110 5 ft/s 0.0039h flow rate in gal / min = 1.0 gpm.
f Therefore, a crack length of 1-1/2 inches in the cold leg piping may be expected to leak at the rate of 1 gpts.
Similarly, for the reactor vessel 4
i a crack length of 1-5/8 inches was determined.
Based on the above and the analysis described below, the ratios of throu6n-vall crack length for the proposed leak limit to the critical through-vall I
f crack are-1 1
l r
~5.2-22 1
[
l
y.
p i
t j
J
?
I' 1.
Irradiated' reactor vessel shell q
ratio =
= 0.'15 2.
Primary piping cold leg j
l ratio =
'+ 0.17 Consider a through-vall crack in the irradiated reactor, ver.sel shell. '
1
.. Because of the large diameter 'of the shell and' vall' thickness, the mathe-'
matical model considered most reasonable is for a through-vall crack'in a-flat plate. ' The equation for this, ciack.taken from reference.1 is:
j K = o /id j
L
'l i
where K = stress intensity factor, C = 1/2 crack length, 1
l 0 = membrane stress.
- j l
When K reaches a critical value called Ky, the fracture toughness of.the.
material,'the crack length (20) is the cr tical crack length.
2' For the reactor vessel the-conditions are:
l Material SA-533-GR B-CL1 Temperature 570F oy (570F) =/k2.2 kai l
o = 29 0 ksi (from contract design ' reports).
bC si on. Mrom reference 3)
=
The KIC value is actually for SA-533-GR.B-CL1' plate but SA-508-GR B-CL1 is an equivalent forging specification. Since..no facture toughness in.
'i formation is available on SA508CL2 material', the use of K for SA533CL1 has been justified byl comparing Charpy, dropveight,' and Nh data for both-materials. Rearranging this equation and. substituting K I#
- IC 0)',$,g5n,_
C=
=
5c
-w(29)2 2
Therefore, the critical crack length = 2C = 10 9":
]
Consider s through-vall crack '
h ld les' of the reactor coolant piping.
(
.The equation proposed by Hahn(ig~t e'cofor high t'oughness materials is applicable. 'i 51 I _
This equation is:
j.5 2 d
f c
=oM o = flow stress of the material = 1.0Lo + 10 ksi, Y
membrane loop stress, o =
h M = stress magnification factor = (1 + 1.61 g) /2 C2 1 I
yield stress, o =
C = 1/2 critical crack length, i
R = average radius of pipe, 1
t = pipe wall thickness, j
t I
For'the reactor coolant piping, the conditions are:
l 1
Material SA-106-GR C Temperature 570F, oy (570F) = 30.h25 ksi
)
ID = 28-5/8" max.
t = 2-1/h" min.
l i
R = 15-7/16" h"#
= 30.h25 (assumed as vorst case) y Rearranging the above equation and making appropriate substitutions:
(
M = "--
h 2
2
[1.Oka + 10h 1 + 1.61 C Y
=l Rt o
(
y
/
C = h.35" Therefore, the critical crack length = 2C = 8.7".
< Reference h stipulates the flow stress (o ) equation is valid if:
(K / yN C
L 5 Y '1di"5 t
C K 1 C
y 1 30.h25 15(h.35)l l/2 = 1h2
.('
5 2-2h
l l
i 1
(
=
a s ess intens W fa m (ksi en.) as MenMed h vhere KC reference L.
K data for %10M C mauM are not avaHam. hever, C
reference 5 represents pipe rupture tests on SA-106-CR B material, some of which could be classified as SA-106-GR C.
The lowest K value for the SA-g 106-GR B material that could be classified as SA-106-GR 'C is 168 ksi /in.
Therefore, the equation is valid.
Based upon the above analysis the critical parameter would be a crack in the I
cold leg piping of approximately 9 spm. In establishing' a maximum unidenti-fied leakage, the following criteria are considered:
1.
The magnitude of the leakage should be well below the leakage associated with a crach of critical size.
i 2.
The magnitude should be well vithin the capability of the normal makeup j
system.
)
3.
The magnitude should be sufficiently large to allow for ease of detec-
)
tion within a reasonable period of time.
4.
Offsite releases should be within 10 CFR 20 limits.
Accordingly, a 1 gpm leak was selected as the maximum allovable unidentified leakage rate. This value is well' below the leakage associated with a crack
{
of critical size. It can be detected within a reasonable period or time as
]
discussed previously. It is believed that continued operation at this level 1
for some period of time to allow for corrective action vill not jeopardize plant safety nor vill external releases exceed 10 CFR 20 limits. Details concerning continued operation at this level are discussed extensively in Technical Specification 3.1.6.
5.2.7.5 Maximum Allevable Total Leakage The normal makeup capacity is 300 gpm. As stated previously, the maximum allow-able total leakage rate is 30 gpm. The makeup capacity is thus greater by a factor of ten. The containment water removal system is capable of removing 200 gpm via the reactor building sump pump.
5.2.7.6 Leakage Source Identification The plant design has provisions for identifying the source of reactor coolant leakage or at least for systematically eliminating candidates from considera-tion, during normal operation.
A review of potential causes of reduction in inventory of reactor coolant indi-cates that fluid losses could be postulated at the following areas with the means of identification and location stated, a.
Leskage from reuctor coolant system valve stems:
All primary system motor operated valves are equipped with leakoffs. All leakoffs within the secon-l dary shield vall which are considered to be inaccessible are fitted with drains to a collecting point and monitored.
a ML s
'I' 1
l i
b.
Reactor coolant pump seals: A reactor coolant pump mechanical seal leak j
could be identified by one or a combination of the following indications:
j I
1.
High seal water temperature at outlet of reactor coolant pump seal re-turn cooler.
i 2.
High flow in the seal leakage measuring device leading from each reactor coolant pump to the reactor coolant drain tank.
3.
Increased temperature of reactor coolant pump seal water returning to seal return coolers.
c.
Pressurizer relief valves: Reactor coolant inventory reduction as a j
result of seat leakage through a pressurizer relief valve may be iden-j tified by the following indication:
q 1.
Increased level in the reactor coolant drain tank. This will provide an indication of the magnitude of the leak, j
l d.
Saf ety injection and decay heat removal lines connected to the reactor coolant system: Back-leakage through a safety injection or decay heat i
removal line could be icentified as the point of reactor coolant leak-i age by the following:
i 1.
Level change in core flooding tank (s). This will provide an indica-tion of the magnitude of the leak.
l 1
2.
Contact temperature and/or radiation level readings on safety injection
(
1 lines located outside the reactor building.
3.
Increase in radioactivity concentration in safety injection fluid l
as determined by sample analysis.
4.
Increase in pressure in* safety injection and decay heat removal sys-tem lines located outside containment, Steam generator primary-to-secondary boundary: Steam generator primary-e.
to-secondary leakage can be identified as the cause of a reduction in reactor coolant inventory by one or a combination of the following:
1.
Sampling steam generator secondary side for radioactivity and/or boron.
2.
High activity as monitored in the condenser off-gas vent line.
3.
Indication of radioactivity in the condensate polishers.
f.
Reactor coolant system drain lines: Leakage could be identified as possibly originating from flow through one or more reactor coolant drain valves by observing reactor coolant bleed hold-up tank level or the miscellaneous waste hold-up. tank level.
g.
Makeup and purification system: A reduction in' reactor coolant inventory could be caused by leakage from portions of this system, most of which are
(
l l
5 5.2-26 Am. 36 (12-17-75) l.-
(
6 j
located outside the reactor building. Confirmation of leakage originating from this scurce can te obtained through observation of sump levels, sump pump operation, valve steam leakage, pump leakoff connections, and visual inspection of portions of the piping.
.s h.
Reactor coolant letdown cooler relief valve: The letdown coolers relief valves lead to the reactor building sump from which -drainage is pumped to j
the miscellaneous vaste stcrage tank. Relief valve leakage could be con-j firmed by alternately isolating one letdown cooler or the other and moni-l toring the reactor building sump collection rate and/or frequency of sump
{
pump operation.
)
5.2.7.7 Sensitivity and Operability Tests The various monitoring systems which are used for detection of unidentified leakage are given in 5.2.7.h.
By assuring the sensitivity and operi tility of the individual ecmponents of the system, the sensitivity and opera 1111ty of the complete leakage detection system are assured.
5.2.7.7.1 sensitivity l
\\
The output of a sensing device is the input of an indicating or tripping out-l l
put device. The sensitivity of an instrument vill be verified by the following method. The normal output of the sensor vill be replaced by a test source which is used as an input to the ultimate output device. The test source vill be var-
, [
ied to produce the minimum perceptible change in the output device. This devia-l tion is the sensitivity of the system.
L To verify the sensitivity, the instrument and its output device vill be taken out of service.
4 In general, sensitivity is not expected to deviate from the originally supplied and installed value, however, an overall sensitivity test for all the instru-ments in the leakage detection system will be performed.
5.2.7.7.2 operability The operability of an instrument and its final output device vill be verified by the following method. The instrument sensor vill normally be exposed to a range of the measured parameter. In most cases, the system in which the instrument is located can be temporarily rearranged to produce an abnormal condition at the sensor. location, without dama6 ng any part of the system. By progressively 1
changing the system until the various trip setpoints are reached, and observing i
l that the proper actions occur at the se1 points, the instrument operability can l
be checked. For an instrument sensor which cannot be varied by rearranging the system in which it is located, the operability can be tested by the use of. a built-in check source to vary the parameter measured. The only instruments l
vhich must be operability tested in this manner are the reactor building air activity monitors which have the required built-in sources.
The operability of remotely controlled valves vill be verified during the opera-l bility test for instruments. The operability of pumps will also be verified during the operability test for instruments.
5.2-27