ML20092K813
| ML20092K813 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 09/19/1995 |
| From: | WISCONSIN PUBLIC SERVICE CORP. |
| To: | |
| Shared Package | |
| ML20092K800 | List: |
| References | |
| NUDOCS 9509260155 | |
| Download: ML20092K813 (31) | |
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APPEICA8ILITY Applies to the limiting combination of thermal power, Reactor Coolant System pressure and coolant temperature during operation.
osascitvg To maintain the integrity of the fuel cladding.
SPECIFICATI0lf The combination of rated power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure TS 2.1-1. The safety limit is exceeded if the point defined by the combination of Reactor Coolant System average temperature and power level is at any time above the
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Proposed Amendment No. 130 TS 2.1-1 09/19/95 9509260155 950919 PDR ADOCK 05000305 P PDR
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's' l-l BASIS - Safety Limits. Reactor Core (TS 2.1)
To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) 7.nd at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure.
DNB is not, however, an observable parameter during reactor operation.
Therefore, the observable parameters of rated power, reactor coolant temperature and pressure have been related to DNB through the W-3 & "L" Grid DNB correlations. The "L" Grid DNB correlation has been developed to predict the DNB flux and the location of the DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio (DNBR), defined as the ratio of the '
heat flux that would cause DNB at a particular core location to the local heat l flux, is indicative of the margin to DNB. The minimum value of the DNBR, during {
steady 9 state operation, normal operational transients, and anticipated transients ,
is limited to 1.30. This minimum DNBR corresponds to a 95% probability at a 95% l confidence level that DNB will not ocg)ur and is chosen as an appropriate margin to DNB for all operating conditions.'
The curves of Figure TS 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation) represent the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which either the DNB ratio is equal to 1.3 or the average enthalpy at the exit of the core is equal to the saturation value. At low pressures or high temperatures the average enthalpy at the exit of the core reaches saturation before the DNB ratio reaches 1.3 and thus, this limit is conservative with respect to maintaining clad integrity. The area where clad integrity is assured is below these lines.
The curves are based on the following nuclear. hot channel factors:
l Flg . 1.55 F"n - 2.51 and include an allowance for an increase in the enthalpy rise hot channel factor at reduced power based on the expression:
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l 55 [1 + 0.2 (1 - P)] where P is the fraction of rated power l
")USAR Section 3.3.3 Proposed Amend:nent No.130 TS B2.1-1 09/19/95 I
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t These limiting hot channel factors are higher than those calculated at full power I for the range from all control rods fully withdrawn to maximum allowable control !
U rod insertion. The control rod insertion limits are given in J$ 3.10.d. Slightly higher hot channel. factors could. occur at lower power levels 6ecause additional !
control rods are in the core. However, the control rod insertion limits dictated '
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by Figure TS 3.10-3 insure that the DNBR is always greater at partial power than at full power. ,
The Reactor Control and Protection System is designed to prevent any anticipated I combinationoftransientconditionsthatwouldresultinaDNBRof[1.30. l REFERENCES I (1) WCAP 8092 !
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I Proposed Amendment No. 130 TS B2.1-2 09/19/95
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Applies to the maximum limit on Reactor Coolant System pressure.
$ECTIVE To maintain the integrity of the Reactor Coolant System.
SPECIFICATION The Reactor Coolant System pressure shall not exceed 2735 psig with fuel ve ergr.as sem,bl i.e_s i,nst_all 777 7 .; ed.,,.i.,n 7y the ,r,e,ac, tor,y,s se,l . ,. .,
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i Proposed Amendment No. 130 TS 2.2-1 09/19/95
4 Basis - Safety Limit. Reactor Coolant System Pressure (TS 2.2)
The Reactor Coolant Systemd serves as a barrier preventing radionuclides
. contained in the reactor coolant from reaching the atmosphere. In the event of a fuel cladding failure, the Reactor Coolant System is the primary barrier against' the release of fission products. By establishing a system pressure limit, the continued integrity of the Reactor Coolant System is assured. The maximum transient pressure allowable in the reactor. pressure vessel under .the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the Reactor Coolant System piping, valves and fittings under USASI B.31.1.0 is 120% of design pressure. Thus, the safetL)imit of 2735 psig (110% of design pressure, 2485 psig) has been established."
The nominal settings of the power-operated relief valves (2335-psig), the reactor high pressure trip (2385 psig) and the safety valves (2485 psig) have been established to prevent exceeding the~ safety limit of 2735 psig. The initial hydrostatic test was conducted at 3107 psig to assure the integrity of the Reactor Coolant System.
1 U)pSARSection4 (2)QSAR Section 4.3 Proposed Amendment No. 130 TS B2.2-1 09/19/95 l
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Leakage of Reactor Coolant (TS 3.1.d)na) ;
TS (TS 3.1.d.1) i Leakage from the Reactor Coolant System is collected in the containment or by the other closed systems. These closed systems are: the Steam and Feedwater System, the Waste Disposal System and the Component Cooling System. Assuming the existence of the maximum allowable activity in the reactor coolant, the rate of I gpm unidentified leakage would not exceed the limits of 10 CFR Part 20. This is shown as follows:
t If the reactor coolant activity is 91/8 Ci/cc (8 = average beta plus gamma energy per disintegration in Mev) and 1 gpm of leakage is assumed to be discharged through the air ejector, or through the Component Cooling System vent line, the yearly whole body dose resulting from this acgivity gt the site boundary, using an annual average X/Q = 2.0 x 10' sec/m, is 7 0.09 rem /yr, compared with the 10 CFR Part 20 limits of Q rem /yr.
With the limiting reactor coolant activity and assuming initiation of a 1 gpm leak from the Reactor Coolant System to the Component Cooling System, the radiation monitor in the component cooling pump inlet header would annunciate in the control room. Operators would then investigate the source of the leak ana take actions necessary to isolate it. Should the leak result in a continuous discharge to the atmosphere via the component cooling surge tank and waste holdup tank, the resultant dose rate at the site boundary would be 0.09 rem /yr as given above.
Leakage directly into the containment indicates the possibility of a breach in the coolant envelope. The limitation of I gpm' for an unidentified source of leakage is sufficiently above the minimum detectable leak rate to provide a reliable indication of leakage, and is well below the capacity of one charging pump (60 gpm).
Twelve (12) hours of operation before placing the reactor in the HOT SHUTDOWN condition are required to provide adequate time for determining t whether the leak is into the containment or into one of the closed systems and to identify the leakage source. ;
TS 3.1.d.2 The 150 gpd leakage limit through any one steam generator is specified to ensure tube integrity is maintained in the event of a main steam line break or under loss-of-coolant accident conditions. This reduced operational leakage rate is applicable in conjunction with the tube support plate ;
voltage-based plugging criteria as specified in TS 4.2.b.5. i l
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na>USAR Sections 6.5, 11.2.3, 14.2.4 Proposed Amendment No. 130 TS B3.1-10 09/19/95
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3.7 AUXILIARY ELECTRICAL SYSTEMS APPLICABILITY Applies to the availability of electrical power for the operation of plant auxiliaries.
DBJECTIVE To define those conditions of electrical power availability necessary to provide 1) safe reactor operation and 2) continuing availability of engineered safety features.
SPECIFICAT16
- a. The reactor shall not be made critical unless all of the following requirements are satisfied:
- 1. The geserve juxiliary ransformer is fully operational and :
energized to supply power [~ to the 4160-V buses.
- 2. A second external source of power is fully operational and energized to supply power to emergency buses 1-5 and 1-6.
- 3. The 4160-V buses 1-5 and 1-6 are both energized.
- 4. The 480-V buses 1-52 and 1-62 and their MCC's are both energized -
from their respective station service transformers.
- 5. The 480-V buses 1-51 and 1-61[ pare both energized from their respective station service transformtrs.
- 6. Both station batteries and both DC systems are DRERABEE, except during testing and surveillance as described in'~C6"~b.
- 7. Both diesel generators are PEMEE. The two underground storage tanks combine to supply at Tdiff~35,000 gallons of fuel oil for either diesel generator and the day tanks for each diesel generator contain at least 1,000 gallons of fuel oil.
- 8. At least one pair of physically independent transmission lines serving the substation is OPERARE. The three pairs of physically independent transmfHiBTTines are:
A,]7 R-304 B F-84 and Q-303 and Y-51 QjR-304andY-51 Proposed Amendment No. 130 TS 3.7-1 09/19/95 i
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- b. During power operation or recovery from inadvertent trip, any of the following conditions of inoperability may exist during the time intervals specified. If PREMBgq] is not restored within the time s
@.pecified, Jj g g within then thewithin next 61hours.
' hour action shall be initiated to achieve
- 1. Either Euxiliary transformer may be out of service for a period not exceeding 7 dEys provided the other~ iuxiliary [ransformer and r bothdieselgeneratorsaregPEBAgg.
- 2. One diesel generator may be inoperable for a period not exceeding ,
7 days provided the other diesel generator is tested daily to ensure Jy! and the engineered safety features associated j with t les51(( generator are ,M{
- 3. One battery may be inoperable for a period not exceeding 24 hours provided the other battery and two battery chargers remain Di W with one charger carrying the d-c supply system.
- 4. If the conditions in TS 3.7.a.8 cannot be met, power operation may continue for up to 7 days provided at least two transmission linesservingthesubstationarepf{gg[j.
- 5. Three offisite power supply transmission lines may be out of service for a period of 7 days provided reactor power is reduced '
to 50% of rated power and the two diesel generators shall be tested daily for DJ{R$8JQIX.
- 6. One service 4160)for 24 ho]urs provided The redundant bus and its loadsVor4!
remain RP,Eg g . !
- c. When its normal or emergency power source is inoperable, a system, train or component may be considered DPERABG for the purpose of '
satisfying the requirements of its applI65611I^[il!gJgGit0!iDt15...;4#ff704 pftRT,108,provided:
- 1. Its corresponding normal or emergency power source is gjggy; !
and
- 2. Its redundant system, trainy or component is QgE33B[{.
J Proposed Amendment No. 130 TS 3.7-2 09/19/95
'o - ; BASIS The intent of this 75 is to provide assurance that at least one external source and one standby souN^e of electrical power is always available to accomplish safe shutdown and containment isolation and to operate required engineered safety features equipment following an accident. t Plant auxiliary power is normally supplied by two segArate external power sources which have multiple off-site network connectionsR3 the [eserve $uxiliary
.[ransformer from the 138-Kv portion of the plant sEbstation, and a~ tertiary j winding on the substation auto transformer. Either source is sufficient to j supply all necessary accident and post-accident load requirements from any one ;
of four available transmission lines. l Each diesel generator is connected to one 4160-V safety features bus and has ; sufficient capacity to start sequentially and operate the engineered safety features equipment supplied by that bus. The set of safety features equipment ; items supplied by each bus is, alone, sufficient to maintain adequate cooling of l the fuel and to maintain containment pressure within the design value in the ; event of a loss-of-roolant accident. Each diesel generator starts automatically upon low voltage on its associated i bus, and both diesel generators start in the event of a safety injection signal. A minimum of 7 days fuel supply for one diesel generator is maintained by requiring 36,000 gallons of 'ruel oil, thus assuring adequate time to restore off8 site power or to replenish fuel. The diesel fuel oil storage capacity requirements are consistent with those specified in ANSI N195-1976/ANS-59.51, [ections5.2,5.[,and6.1. The plant 125-V d-c power is normally supplied by two batteries each of which will have a battery charger in service to maintain full charge and to assure adequate power for starting the diesel generators and supplying other emergency loads. A third charger is available to supply either battery.
$ ensure The arrangement that no single faultofcondition the auxiliary power sources will deactivate more and thanequipment one redundan and this {t set of safety features equipment items and will therefore not result in failure of the l
E.antpyg,tecg!pn_sgst o L a u w : x w ; =pms to.aw.~=ad rgspo g adeguate1 p s,aj ;a=:u- pss-op 21-.La cpoJan( i acyid,ent., WySARFigure8.2-2 Proposed Amendment No. 130 TS B3.7-1 09/19/95
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- m. Reactor Coolant Flow +
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'l. During steady-state power operation, reactor coolant flow rate shall be 2: 89,000 gallons per minute average per loop. .If l reactor coolant flow rate is < 89,000 gallons per m' Jte per ;
loop, action shall be taken in accordance with TS 3.10 n. i
.2. Compliance with this flow requirement shall'be demonstrated by-verifying the reactor coolant flow during initial power-escalation following each REFUELING, between 70% and 95% power ,
with plant parameters as constant as practical.
- n. If, during power operation any of the conditions of TS 3.10.k, TS 3.10.[, or TS 3.10.m.1 are not met, restore the parameter in 2 hours or less to within limits or reduce power to < 5% of thermal rated power within an additional 6 hours. Following analysis, thermal power may be raised not to exceed a level analyzed to maintain a minimum DNBR of 1.30.
1 1 i Proposed Amendment No. 130 TS 3.10-9 09/19/95
.c 3.12 $ $ Jg E M J g !Jj g MEgj!! @ g @ JI M]!!1 M APPLICARICITY Applies to the pfE@8]CJH of the Control Room Post [ Accident Recirculation System.
OBJECTIVE To specify QgRABIQJ{ requirements for the Control Room Post [ Accident RecirculationSysfem.
$PECIFICAY10ll
- a. The reactor shall not be made critical unless both trains of the Control Room Post [ Accident Recirculation System are gPjg]E.
- b. Both trains of the Control Room PostEAccident Recirculation System, including filters shall be DPERABLE or the reactor shall be shut down within 12 hou]rs, except WifThiin one of the two trains of the Control Room Post [ Accident Recirculation System is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding Edays,
- c. During testing the system shall meet the following performance requirements:
- 1. The results of the in-place cold D0P and halogenated hydrocarbon tests at design flows on HEPA filter and charcoal adsorber banks shall show a 99% DOP removal and a 99% halogenated hydrocarbon removal.
- 2. The results of the laboratory carbon sample analysis from the Control Room Post" ccident gecirculation [ystem carbon shall show
'h 90% r*5diofetiv@e methyl iodide removal at conditions of 66*C, and 95% RH.
- 3. Fans shall operate within 10% of design flow when tested. i I
Proposed Amendment No. 130 j TS 3.12-1 09/19/95 l i
8A51$NCihtF614A68ii465tMidst?Rs6ffsilitl66Ysistiii?fT5T3?!21 The gontrol goom Rost@ccident Recirculation System is designed to filter the Qontrol Room atmosphere during (ontrol Room is~olation conditions. TheControl Room gostgAccident gecirculation $ystem~is designed to automatically start upon
$1S or hicj6 radiation signal at i~nlet of unit.
If the system is found to be inoperable, there is no immediate threat to the Control Room and reactor operation may continue for a limited period of time shile repairs are being made. Ifthesystemcannotberepairedwithingdays, the reactor is placed in MTj${@gX until the repairs are made. Proposed Amendment No. 130 TS B3.12-1 09/19/95
b. o 3.14 SHOCK SUPPRESSORS (SNUBBERS) APPLICABILITY Applies to the QPggjQg of shock suppressors which are related to plant safety. ,
@ E IYI To ensure that shock suppressors, which are used to restrain safety-related piping under dynamic load conditions, are functional during reactor operation.
SPECIFICATHN
- a. The reactor shall not be made critical unless all safety]related shock suppressors are QRERAB @ except as noted in 3.14.b
- b. power operation or recovery from inadvertent trip, if any Duringlrelated safety shock suppressorfound is inoperable one of the following actions shall be taken within 72 hours:
- 1. The inoperable shock suppressor shall be restored to an DPERABLE condition or replaced with a spare shock suppressor of"TilifiTi?
specifications; or
- 2. The fluid line restrained by the inoperable shock suppressor shall, if feasible, be isolated from other safetysrelated systems if otherwise permitted by the TS and thereafter operation may continue subject to any limitations by the Q for that fluid line; or
- 3. Actions shall be initiated to shut down the reactor and the reactor shall be in a $15@U]M condition within 36 hours.
Proposed Amendment No. 130 TS 3.14-1 09/19/95
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BASISM9hocl0I$sisiFssidilliiET5nubb6EsWITST3H41 Shock suppressors (snubbers) are designed to prevent unrestrained pipe motion under dynamic loads, as might occur during seismic activity or severe plant transients, while allowing normal thermal motion during startup or shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic event or other events initiating dynamic loads. It is therefore re uired that all snubbers designed to protect the reactor coolant and other safet related systems or components be operable during reactor operation. The intent f this JJ is to prohibit startup , or continued operation with defective safety [related shock suppressors. l Because the protection afforded by snubbers is required only during low i probability alternativeevents, action $~e$ 3.14.b
~ fore allows reactor a period shutdown is of 72 hours for repairs or feasible required.
t i i i Proposed Amendment No. 130 TS B3.14-1 09/19/95
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TABLE TS 3.1-2 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES
' NMIM)MIIMI SY_S_ TEN VALVEN_0? ALLWABLEiLEANAGEEBASEDIM
_ m _~ . TMESW Reactor Vessel, Core Flooding SI-304A s 5.0 jallons per isinute Line (Upper Plenum Injection) SI-303A s 5.0 Ballons per Binute SI-304B s 5.0 yallons per iiinute SI-303B s 5.0 jallons per sinute Loop B 12" Accumulator Discharge SI-228 s 5.0 jallons per @inute Line E[eakage rates $ 1.0 gpm are considered acceptable. Leakage rates F 1.0 gpm but [s 5.0 gpm are considered acceptable if the latest iiieasured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater. Leakage rates [ 1.0 gpm but Id 5.0 gpm are considered unacceptable if the latest iiieasured rate exceeded the rit^e determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater. [eakageratesgreaterthan5.0gpmareconsideredunacceptable.
@ Minimum test differential pressure shall not be 3150 psid.
PAGE 1 0F 1 Proposed Amendment No. 130 09/19/95
- -,-..s 2 am.: g4m. m ne ._A,a. _ . . 4.a5 dr BASIS The Auxiliary feedwater System (AFW) mitigates the consequences of any event that ,
causes a loss of normal feedwater. The design basis of the AFW System is to remove decay and residual heat by delivering the minimum required flow to at least one steam generator until the Reactor Coolant System (RCS) is cooled to the point of placing the Residual Heat Removal System into operation. In accordance with ASME Code Section XI, Subsection IWP, an in-service test of each auxiliary feedwater pump shall be run nominally every 3 months (quarterly) during normal plant operation. It is recommended that this test frequency be : maintained during shutdown periods if this can be reasonably accomplished, i although this is not mandatory. If the normally scheduled test is not performed i during a plant shutdown, then the motor-driven pumps shall be demonstrated OPERABLE within 1 week exceeding 350*F; and the turbine-driven pump shall be demonstrated OPERABLE within 72 hours of exceeding 350*. Quarterly testing of the AFW pumps is used to detect degradation of the component. This type of testing may be accomplished by measuring the pump's developed head at one point of the pump characteristic curve. This verifies that tFe measured performance is within an acceptable tolerance of the original pump baseline performance. TS 3.4.b requires all three AFW pumps be OPERABLE prior to heating the RCS average temperature > 350*F. It is acceptable to heat the RCS to > 350*F with the turbine-driven pump inoperable for a limited time period of 72 hours. The wording of TS 3.4.b.2.B and TS 4.8.b allows delaying the testing until the steam flow is consistent with the conditions under which the performance acceptance criteria were generated. The discharge valves of the two motor-operated pumps are vormally open, as are the suction valves from the condensate storage tanks and tho two valves on a cross tie line that directs the turbine-driven pump discharge to either or both i steam generators. The only valve required to function upon initiation of i auxiliary feedwater flow is the steam admission valve on the turbine-driven pump. ' Proper opening'of the steam admission valve will be demonstrated each time the turbine-driven pump is tested. l Proposed Amendment No. 130 TS B4.8-1 09/19/95
=.. -. - .. . .:4 4.9 REACTIVITY ANOMALIES PPLICABICICY Applies to potential reactivity anomalies.
08dECitVf To require evaluation of reactivity anomalies within the reactor. SPEcfMC4ficN Following a normalization of the co.nputed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be periodically compared with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of )% in reactivity, an evaluation as to the cause of the discrepancy sha'11 be made and reported to the Commission within 30 days. 4 i 1 Proposed Amendment No. 130 TS 4.9-1 09/19/95 1
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BA51$NMACTIVITY?AN0MA01FMM To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel
- burn-up and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. .As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burn-up and reactivity is compared with that predicted. This process of normalization should be completed after about 10% of the total core burn-up. Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously ,
evaluated. ' Any reactivity anomaly greater than 1% would be unexpected, and its ' occurrence would be thoroughly investigated and evaluated. The value of 1% is considered a safe limit since a shutdown margin of at least l 1% with the most reactive rod in the fully withdrawn position is always ined. ma , int...a.e7 7 ,7 m g 7,7y,.y777 7 .y ,7 7 7a . ,..; m -.m._ m _. am m am: mh_7.a.7,,.c._ ;n._ _ 7yg_ 7.. .. ~g m _ . .a m _ .<..- _o; .au_.e
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l I mySAR Section 3.2 Proposed Amendment No. 130 TS B4.9-1 09/19/95
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4.12 SPENT FUEL POOL SWEEP SYSTEM , APPLICAblCITY Applies to testing and surveillance requirements for the spent fuel pool sweepsystemin[]3.8.a.9.
,0BJECTIVE To verify the performance capability of the spent fuel pool sweep system.
$PSTFICAfteN
- a. At least once per operating cycle or once every 18 months, whichever occurs first, the following conditions shall be demonstrated:
- 1. Pressure drop across the combined HEPA filters and charcoal 10 inches of water and the pressure drop adsorber across banks any HEPA bankis @ is [ 4 inches of water at the system design flow rate (i 10%).
- 2. Automatic initiation of each train of the system.
- b. 1. The in-place D0P test for HEPA filters shall be performed (1) at least once per 18 months and (2) after each complete or partial replacement of a HEPA filter bank or after any maintenance on the system that could affect the HEPA bank bypass leakage.
- 2. The laboratory tests for activated carbon in the charcoal filters shall be performed (1) at least once per 18 months for filters in a standby status or after 720 hours of filter operation, and ,
(2) following painting, fire, or chemical release in any l ventilation zone communicating with the system.
- 3. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or 1 I
after any maintenance on the system that could affect the charcoal adsorber bank bypass leakage. i Proposed Amendment No. 130 TS 4.12-1 09/19/95
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- c. Perform an air distribution test on the HEPA filter bank after any
. maintenance or testing that could affect the air distribution within the system. The test shall be performed at design flow rate !
( 10%). The results ofathe test shall show the air distribution . is uniform within 20dN. ' [E$$$$$$$5$$$$$5EkIN$$$$EEIi$E$Id!Edhf2EEIbid!$$$$$[Eib[$UdhES ; t t i i i ("In WPS letter of August 25, 1976 to Mr. Al Schwencer (NRC) from Mr. E. W. James, we relayed test results for flow distribution for tests I performed in accordance with ANSI N510-1975. This standard refers to flow l distribution tests performed upstream of filter assemblies. Since the test ; results' upstream of filters were inconclusive due to high degree of turbulence, ! tests for flow distribution were performed downstream of filter assemblies with ' acceptable results (within 20%). The safety evaluation attached to Amendment 12 references our letter of August 25, 1976 and acknowledges acceptance of the test results. Proposed Amendment No. 130 1 TS 4.12-2 09/19/95 l l
t ILAsn Pressure drop across the combined HEPA filters and charcoal adsorbers of
# 10 inches of water and 4 inches across any HEPA filter bank at .the system design flow rate ( 10%) will indicate that the filters and adsorbers are not ~
- t. logged by excessive amounts of foreign matter. A test frequency of once per operating cycle establishes system performance capability. This pressure drop is approximately 6 inches of water when filters are clean.
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52 dated June 1973. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly, and obtaining at least two samples. Each sample should be at least 2 inches in diameter and a length equal to the thickness of the bed. The use of multi-sample assemblies for test samples is an acceptable alternate to mixing one bed for a sample. If the iodine removal efficiency test results are unacceptable, all adsorbent in the system should be replaced. Any HEPA filters found defective should be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52 (Rev.1) dated June 1976. If painting, fire, or chemical release occurs such that the charcoal adsorbers become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use. Degradation of the HEPA filters due to painting, fire or chemical release in a communicating ventilation zone would be detected by an increased pressure drop across the filters. Should the filters become contaminated, engineering judgment would be used to determine if further leakage and/or efficiency testing was required. Demonstration of the automatic initiation capability is necessary to assure system performance capability. In-place testing procedures will be established utilizing applicable sections of ANSI N510 - 1975 standard as a procedural guideline only. Proposed Amendment No. 130 TS B4.12-1 09/19/95
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4.13 RADIOACTIVE MATERIALS SOURCES , APPLICABILITY ! Applies to the possession, leak test, and record requirements for
- radioactive material sources required for operation of the facility.
OBJECTIVE' To ensure that radioactive material sources which are beneficial to ! facility operation are available to the facility and these sources are , verified to be free from leakage. ! I SPECIFICATION
@ Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the .
Commission or the State. !
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If Sources which contain by-product material that exceeds the quantities listed in 10 CFR 30.71, Schedule B, and all other sources containing > 0.1 microcuries shall be leak tested in accordance with ' this[$. Q Any source specified by TS 4.13.2 which is determined to be leaking shall be immediately withdrawn from use, repaired or disposed of in accordance with the Commission's regulations. Leaking is defined as the presence of .005 microcuries of the source's radioactive : material on the test sample. l M Each sealed source with a half-life > 30 days, and in any form other than gas, shall be tested for leakage at intervals not to exceed 6 months, except for: D Startup sources inserted in the reactor vessel, 23 Fission detectors following exposure to core flux, 34 Irradiation sample sources inserted in the reactor vessel, : Sources enclosed within the Eberline Model 1000 Multi-Source p~ Gamma Calibrator, ' M Sources enclosed within the Shepherd Model 89-400 Self-Contained
*~" Calibrator, and l
@ Hydrogen-3 sources. j iD!
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Sources specified by TS 4.13.2 which are in storage and not being used are exempt from the testing of TS 4.13.4. Prior to use or I transfer to another licensee of such a source, the leakage test of l TS 4.13.4 shall be current. p Startup sources and fission detectors shall be leak tested prior to initial insertion into the reactor vessel or prior to being subjected to core flux. Proposed Amendment No. 130 TS 4.13-1 09/19/95
p'] A complete inventory of radioactive materials sources shall be maintained current at all times. 9 f Proposed Amendment No. 130 TS 4.13-2 09/19/95
3 , l i E ! i Ingestion or inhalation of source material may give rise to total body or organ j irradiation. This specification assures that leakage from radioactive material ! sources does. not exceed allowable limits. In the unlikely event that those quantities of radioactive by-product materials of interest to this specification- t which are exempt from leakage testing are ingested or inhaled, they represent ! less than one maximum permissible body burden for total body irradiation. The ! limits for all other sources (including alpha emitters) are based upon ; 10 CFR 70.39(c) limits for plutonium. The Eberline Model 1000 Multi-Source Calibrator and the J. L. Shepherd i Model 89-400 are totally enclosed instrument calibrating assemblies for which , leak testing of the enclosed sources is not practical. Leak testing of these ! sources would require disassembly of the calibration assembly shield, controls, etc., resulting in personnel exposure without corresponding benefits. 4 i f r 1 Proposed Amendment No. 130 l TS B4.13-1 09/19/95 ! 1
6 ., 0 U, i 4.17 ,Colgit0QR00 PO$ M 19G EJ M R @ TION 3 SYSTEM APPLICABILITY ' Applies to testing and surveillance requirements for the Control Room : Postaccident Recirculation System in [S; 3.12. OBJECTIVE To verify the performance capability of the Control Room Postaccident ! Recirculation System. SPECIFICATION
- a. At least once per operating cycle or once every 18 months, whichever .
occurs first, the following conditions shall be demonstrated: l
- 1. Pressure drop across the combined HEPA filters and charcoal i adsorber banks is R 6 inches of water and the pressure drop i acrossanyHEPAbanlIis@4inchesofwateratthesystemdesign flow rate ( 10%).
- 2. Automatic initiation of the system on a high radiation signal at t the inlet of the unit and a safety injection signal.
- b. 1. The in-place D0P test for HEPA filters shall be performed (1) at least once per 18 months and (2) after each complete or partial '
replacement of a HEPA filter bank or after any maintenance on the system that could affect the HEPA bank bypass leakage.
- 2. The laboratory tests for activated carbon in the charcoal filters shall be performed (1) at least once per 18 months for filters in a standby status or after 720 hours of filter operation, and (2) following painting, fire, or chemical release in any ventilation zone communicating with the system.
- 3. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or after any maintenance on the system that could affect the charcoal adsorber bank bypass leakage.
- 4. Each train shall be operated at least 10 hours each month.
Proposed Amendment No. 130 TS 4.17-1 09/19/95
i e BASIS Control Room PostEAccident Recirculation System Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water and 4 inches across any HEPA filter bank at the system design flow rate (i 10%) will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. A filter test frequency of once per operating cycle establishes system performance capability. The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52, dated June 1973. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly, and obtaining at least two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. The use of multi-sample assemblies for test samples is an acceptable alternate to mixing one bed for a sample. If the iodine removal efficiency test results are unacceptable, all adsorbent in the system should be replaced. i i Proposed Amendment No. 130 1 B4.17-1 09/19/95 l l i 1
TABLE TS 4.1-3
, MINIMUM FREQUENCIES FOR EQUIPMENT TESTS EQUIPMENT TESTS TEST FREQUENCY
- 1. Control Rods Rod drop times of all Each REFUELING outage full length rods Partial movement of all Every 2 weeks when at or rods not fully inserted above HOT STANDBY in the core la. Reactor Trip Breakers Independent test"' shunt Monthly and undervoltage trip attachments Ib. Reactor Coolant Pump OPERABILITY Each REFUELING outage Breakers- Open-Reactor Trip lc. Manual Reactor Trip Open trip reactor
- trip Each REFUELING outage and bypass breaker
- 2. Deleted
- 3. Deleted ,
- 4. Containment Isolation OPERABILITY Each REFUELING outage
- Trip
- 5. Refueling System OPERABILITY Prior to fuel movement Interlocks each REFUELING outage
- 6. Deleted
- 7. Deleted I
- 8. RCS Leak Detection OPERABILITY Weekly N
- 9. Diesel Fuel Supply Fuel InventoryW Weekly
- 10. Deleted
- 11. Fuel Assemblies Visual Inspection Each REFUELING outage i
- 12. Guard Pipes Visual Inspection Each REFUELING outage l
- 13. Pressurizer PORVs OPERABILITY Each REFUELING cycle l
- 14. Pressurizer PORV Block OPERABILITY QuarterlyN Valves
- 15. Pressurizer Heaters OPERABILITYW Each REFUELING cycle
- 16. Containment Purge and OPERABILITYW Each REFUELING cycle Vent Isolation Valves
")following maintenance on equipment that could affect the operation of the equipment, tests should be performed to verify OPERABILITY.
(2' Verify OPERABILITY of the bypass breaker undervoltage trip attachment prior ; to placing breaker into service. ' (3)Using the Control Room push-buttons, independently test the reactor trip breakers shunt trip and undervoltage trip attachments. The test shall also verify the undervoltage trip attachment on the reactor trip bypass breakers. I
$55555?5??b SPW W$$5?SI5 W SNS$
EJMjpt~oij of fuel required in all plant modes. N Not required when valve is administratively closed. E est T will verify OPERABILITY of heaters and availability of an emergency power supply. NThis test shall demonstrate that the valve (s) close in s 5 seconds. PAGE 1 0F 1 Proposed Amendment No. 130 09/19/95
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- 8. Ef@ggpHJ@ba$p~~6i&tld(e iif g6Knnumb Mifdr~y
@erfion, o(~^ita?g{gMFig tabulation on an annuai sis 6f ufTlity, and other personnel (including contractorsk receiving exposures
> 100 mrem /yr and their associated pH rem exposure according .
to work and job functions,U) e.g., reactor operations and surveillance, in-service inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and REFUELING. The dose assignment to various dut functions may be estimates based on pocket dosimeter iiH TL Small exposures totaling < 20% of the individual total d6"se nee not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific , major work functions. C. Challenges to and failures of tge pressurizer power operated relief valves and safety valves.' D. This report shall document the results of specific activity - analysis in which the reactor coolant exceeded the limits of TS 3.1.c.1. A during the past year. The following information , shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours prior to the first sample in which the limit was exceeded; > (4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the reactor coolant exceeded the radiciodine limit.
'O This tabulation supplements the requirements of Section 20.Q)f @ of 10 CFR Part 20.
(2) Letter from E. R. Mathews (WPSC) to D. G. Eisenhut (U.S. NRC) dated January 5, 1981. Proposed Amendment No. 130 TS 6.9-2 09/19/95
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..o. C 6.13 HIGH RADIATION AREA
- a. In lieu of the " control device" or " alarm signal" required by Paragraph which 20.160f(ity of radiationilisof> 10 the inliiis 100CFR mremPart
/hr,20, buteach high radiation area
< 1000 mrem /hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)"). Any individual or group of individuals permitted to enter such areas shall be provided
, with or accompanied by one or more of the following.
- 1. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- 2. A radiation monitoring device which continuously integrates the radiation dose in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
- 3. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the RWP.
U) Health Physics personnel or personnel escorted by Health Physics personnel l shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas. Proposed Amendment No. 130 TS 6.13-1 09/19/95
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- b. In addition to the requirements of 6.13.a., areas accessible to personnel with radiation levels such that a major portion of the body could receive in 1 hour a dose > 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in 1 hour a dose > 1000 mrem' ' that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.
)
i l l tz) Measurement made at 30 centimeters from source of radioactivity. Proposed Amendment No. 130 TS 6.13-2 09/19/95
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.6.19 MAJOR CRAW:ES TO RADI0 ACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS
Licensee initiated major changes to th,e radioactive waste systems ! (liquid, gaseous and solid): . l
- a. Shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the ev}{aluation was reviewed by the PORC. The discussion of each change shall contain:
- 1. A summary of-the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59. ;
- 2. Sufficient information to support the reason for the change l without benefit of additional or supplemental information;
- 3. A description'of the equipment, components and processes involved i and the interfaces with other plant systems; 2
- 4. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or ;
quantity of solid waste that differ from those previously j predicted in the license application and amendments thereto, i i
- 5. An evaluation of the change, which shows the expected maximum i exposures to individuals in the UNRESTRICTED AREA and to the -
general population that differ from those previously estimated in I the license application and amendments thereto; j
- 6. A comparison of the predicted releases of radioactive materials, .
- in liquid and gaseous effluents and in solid waste, to the actual i' releases for the period prior to when the changes are to be made;
- 7. An estimate of the exposure to plant operating personnel as a i result of the change; and
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- 8. Documentation of the fact that the change was reviewed and found v acceptable by the PORC. j
- b. Shall become effective upon review and acceptance by the PORC.
f 1 i i i 0)Licemes may choose to submit the information called for in this TS as part of th9/p g g USAR update. I Proposed Amendment No. 130 TS 6.19-1 09/19/95' o l}}