ML20079M070
| ML20079M070 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/30/1983 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20079M054 | List: |
| References | |
| TAC-53417, NUDOCS 8401270151 | |
| Download: ML20079M070 (136) | |
Text
4 ATTACHMENT 2 PROPOSED CHANGES TO THE i
IN-SERVICE INSPECTION REQUIREMENTS l
I I
j B401270151 831230 PDR ADOCK 05000267 P
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.0-1 4
5.0 SURVEILLANCE REQUIREMENTS The surveillance requirements specified in this section define the tests, calibrations, and inspections which are necessary to verify the performance and operability of equipment essential to safety during all modes of operation, or required to prevent or mitigate the consequences of abnormal situations.
Implementation of the in-service inspection (ISI) surveillance requirements shall be per one of the following criteria, unless otherwise indicated:
ISI Criterion A.
The surveillance requirement shall be implemented before 90 days have elapsed
-t following the formal approval date of Amendment No. 33 by the Nuclear Regulatory 4
Commission.
ISI Criterion 8: The surveillance requirement shall be implemented before the beginning of fuel cycle 4, provided that fuel cycle 4 does not-begin within 90 days from the formal approval date of -Amendment No. 33 by the Nuclear Regulatory Commission.
Otherwise, the surveillance requirement shall be implemented before the end of the first scheduled plant shutdown following 90 days from the formal approval date of
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.0-2 Amendment.No. 33 by the Nuclear Regulatory Commission.
ISI Criterion C: The surveillance requirement shall be implemented before the beginning of fuel cycle 5.
ISI Criterion D: The surveillance requirement shall be implemented in the existing schedule of surveillance tests, following 90 days from the formal approval date of_ Amendment No. 33 by the' Nuclear Regulatory Commission.
l
.ISI Criterion E: Same' as ISI Criterion A but applicable to l
Amendment No.
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_ISI Criterion F:.Same as ISI Criterion B but applicable to fuel l
cycle 5 and Amendment No.
_l ISI Criterion G: Same as ISI Criterion D but applicable to
-l.
Amendment No.
K L
Fort St. Vrain #1 Technical Specifications L
Amendment Page 5.2-1 l
5.2 PRIMARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS Applicability Applies to the surveillance of the primary (helium) reactor coolant system, excluding the steam generators, l
and to.the surveillance of the reactor auxiliary systems.
Objective To ensure the capability of the components of the primary reactor coolant system to maintain the primary reactor coolant envelope as a fission product barrier and to ensure the capability to cool the core under all modes of operation.
Specification SR 5.2.1 - PCRV and PCRV Penetration Overpressure Protection Surveillance a) Each of the two overpressure protection assemblies protecting the PCRV shall be tested at intervals not to exceed five years, on an alternating basis, with one overpressure protection assembly tested during each refueling cycle.
The PCRV safety valve containment tank closure bolting shall be visually examined for absence of surface defects when the tank is opened for the above testing.
Tank closure flange leak cightness shall be determined following tank closure.
SR 5.2.1.a shall be implemented per ISI Criterion C.
Fort St. Vrain #1' Technical Specifications Amendment Page 5.2-2 b) Each of the two overpressure protection assemblies protecting a steam generator or-a circulator penetration interspace shall be tested at five calendar year intervals on an alternating basis, so that' one safety valve for each penetration interspace and one rupture disc of each type are tested at an approximate interval of two and a half years.
SR 5.2.1.b shall be implemented per ISI Criterion D.
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c) The ' instrumentation and controls associated with the
~ overpressure protection assemblies in a) and b) above shall be tested and calibrated as follows:
i
- 1) The pressure switch and alarm for each interspace between a rupture disc and the corresponding safety valve shall be functionally tested monthly and calibrated annually.
The pressure switch and alarm for the PCRV safety valve containment tank shall be functionally tested and calibrated annually.
SR 5.2.1.c.1 shall be implemented per ISI Criterion D.
- 2) The position indication circuits associated with 4
the PCRV overpressure protection system shut off valves shall be functionally tested and calibrated-when-testing either of the PCRV overpressure protection assemblies.
The pressure switch and
e Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-3 alarm for the PCRV safety valve bellows shall be functionally tested and calibrated in conjunction with its associated safety valve test.
SR 5.2.1.c.2 shall' be implemented per ISI Criterion C.
- 3) The control, interlock, and position indication circuits associated with each of the PCRV penetration overpressure protection system shut off valves'shall be functionally tested at five calendar year intervals.
SR 5.2.1.c.3 shall be impicmented per ISI Criterion D.
Basis for Specification SR 5.2.1 Testing of a PCRV overpressure protection assembiy can 4
only be performed when closing the corresponding manual 7
shut off valve, located upstream of the rupture disc.
LCO 4.2.7 does not allow isolation of such an assembly unless the primary pressure is less than 100 psia.
Consequently, testing and examinations will be performed at -shutdown.
One assembly will be isolated while the other one will remain in a fully operational condition during the testing procedure, thus ensuring overpressure protection of the PCRV.
The rupture disc is designed to be removed from the system for bench testing.
Verification is made of the correct
Fort St. Vrain #1 U
Technical Specifications Amendment Page 5.2-4 1
deflection of -the' disc at the set pressure level which would cause the membrane to be ruptured. The safety valve is tested-for setpoint activation without removing it from the system.
The pressurized portion of the assembly is monitored for leakage during plant operation.
Leakage examination of the containment tank cover seals and visual examination of the cover bolts provides assurance that containment tank integrity is restored after the tank cover has been re-installed.
Testing of a PCRV penetration overpressure protection assembly can be performed during plant operation since the assemblies are accessible and since LCO 4.2.7 requires only one assembly to be operable at any time.
The safety valve in each assembly is tested while in place i
to demonstrats that it opens at the correct set pressure.
The rupture discs are not provided with a testable design feature and, therefore, cannot be tested.
However, one
. rupture disc of each type assembly is visually examined to verify that the membrane is free of defects and that the knife blade remains sharp.
i The intervals specified for testing the overpressure protection assemblies are adequate to demonstrate the operability of the overpressure protection systems.
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k Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-5 The intervals specified for testing the associated instrumentation and controls are adequate to assure reliabi.lity of rupture disc and safety valve operation and to monitor the integrity of the PCRV safety valve piping
.and containment tank.
Specification SR 5.2.2 - Tendon Corrosion and Anchor Assemblies Surveillance The serviceability of the corrosion protection applied to and the condition of the prestressing tendons shall be monitored in accordance with paragraphs a) and b).
-Surveillance of the tendon end anchor assemblies shall be performed in accordance with paragraph c).
a) Corrosion protected wire samples of sufficient length (i.e., initially at least 15 feet where practical, or half the tendon length, whichever is shorter) shall be inserted with selected tendons (those tendons with load cells). Corrosion inspection of at least one of these wires shall be made at the end of the first and third calendar year after prestressing. Additional inspections shall be conducted at five calendar year intervals thereafter.
SR 5.2.2.a shall be implemented per ISI Criterion D.
b) A sample of the atmosphere contained in a representative number of tendon tubes (tendon tubes without load cells and tendon tubes with load cells
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-6 from which wire samples are examined) shall be drawn and analyzed for products of corrosion, in coordination with and at the same time intervals as for paragraph a) above.
c) Visual examination of 5's of the prestressing anchor assemblies shall be performed at five calendar year intervals.
This may include the anchor assemblies which can be visually examined while performing a) and b) above.
SR 5.2.2.c shall be implemented per ISI Criterion D.
Basis for Specification SR 5.2.2 The corrosion protection provided for the
'PCRV prestressing components is considered to be more than adequate to assure that the required prestressing forces are sustained throughout the operational life of the plant.
The details of the corrosion protection system are described in Section 5.6.2.5 of the FSAR.
Sampling tendon tube atmosphere for products will provide a secondary check on the adequacy of the cor osion protection provided for the stressing tendons.
Visual examination of tendon end anchor assemblies will provide additional assurance that the prestressing system has not degraded by checking the corrosion protection and integrity of the anchor assemblies.
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-7 Specification SR 5.2.3 - Tendon Load Cell Surveillance a) Checks on the possible shift in the load cell reference points for representative load cells shal!
be performed at the end of the first calendar year after initial prestressing and within 120 days prior to initial power operation. Additional checks shall be conducted at five calendar year intervals thereafter.
b) The load cell alarm circuit between the Data Acquisition System Room and the Control Room shall be functionally tested annually to assure that the operator in the Control Room is alerted when tendon load settings are exceeded.
SR 5.2.3.b shall be implemented per ISI Criterion A.
Basis for Specification SR 5.2.3 The PCRV tendons apply the force required to counteract the internal pressure.
Therefore, they are the PCRV structural components most capable of being directly monitored and of indicating the capability of the vessel to resist internal pressur-es.
Since the relation between effective prestress and internal pressure is directly and easily calculable, monitoring tendon loads is a direct and reliable means for assuring that the vessel always has capacity to resist pressures up to Reference Pressure.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-8 Monitoring of the tendon loads will assura that deterioration of structural components including progressive tendon corrosion, concrete strength reduction, excessive steel relaxation, etc., cannot occur undetected to a degree that would jeopardize the safety of the vessel.
Each of these phenomena would result in tendon load changes.
These changes, as reflected by the load cells, are monitored in the control room by an alarm system which alerts the operator when the tendon load settings are exceeded. The upper settings will be varied depending on the location of the tendon being monitored, while the lower settings for all load cells will be set to correspond to 1.25 times peak working pressure (PWP).
Specification SR 5.2.4 - PCRV Concrete Structure Surveillance a) Crack patterns on the visible surfaces of the PCRV 4
shall be mapped prior to and following the initial proof test pressure (IPTP).
Concrete cracks which exceed 0.015 inches in width shall be recorded.
Subsequent concrete surface visual inspections shall be performed after the end of the first and third calendar year following initial power operation.
Recorded cracks shall be assessed for changes in length and any new cracks will be recorded.
Additional inspections shall be conducted at ten calendar year intervals thereafter.
Fort St. Vrain #1 Technical Specifications.
Amendment Page 5.2-9 b) PCRV deformations and deflections at vessel midheight and at the center of the top head shall be monitored at five calendar year intervals during a vessel pressurization to operating pressure.
.SR 5.2.4.6 shall be implemented per ISI Criterion C.
c) The PCRV support structure shall be visually examined for evidence of structural deterioratica at ten cslendar year intervals.
SR 5.2.4.c shall be implemented per ISI Criterion C.
Basis for Specification SR 5.2.4 Cracks are expected to occur in the PCRV concrete resulting from shrinkage, thermal gradients, and local tensile strains due to mechanical loadings. The degree of cracking expected is limited to superficial effects and is not considered detrimental to the structural integrity of the PCRV.
Reinforcing steel is provided to control crack growth development with respect to size and spacing.
Model testing has also shown that severely cracked vessels contain the normal working pressure for extended periods of time as long as the effective prestressing forces are
-maintained.
Cracks up to about 0.015 inches (limits of paragraph 1508b, ACI 318-63) for concrete not exposed to weather are generally considered acceptable and corrosion
.of rebars at such cracks is of negligible consequence.
. Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-10 Large ' crack widths will require further assessment as to their significance, depending on the width, depth, length, and location of the crack on the structure, and must be considered with reference to the observed overall PCRV response.
Further discussion on the significance of concrete cracks in the PCRV is given in Section 5.12.5 of the FSAR.
Observed crack development with time during reactor operation will be related to the PCRV structural response as monitored by the. installed sensors and deflection measurements. Details of the PCRV structural monitoring provisions are given in Section 5.13.4 and Appendix E.17 of the FSAR.
The interval for surveillance after the fifth year following initial prestressing may be adjusted based on the analysis of* prior results.
Monitoring of overall PCRV deformations and deflections is the best indication of PCRV structural performance and verifies that the PCRV response is elastic and that no significant permanent strains exist.
Visual examination of the PCRV support structure will indicate that no structural deterioration has occurred.
Significant cracking patterns or sizes should be
-investigated with respect to their impact on the integrity of the PCRV.
3 Fort St. Vrain #1 a
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Technical Specifications
~
Amendment 4
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Page 5.2-11 rw w
Specification SR 5.2.5 - Liner Specimen Surveillance shall be placed adjacent to.the outside surface Specimens notch toughness of the top, head liner so that changes in due to' irradiation of the steel can be measured during the life of the reactor.
During._the fifth refueling cycle, three sets of 12 and weld material specimens of the PCRV liner materials be removed and tested to obtain Charpy impact data.
shall to provide The spectraen holders shall contain dosimeters Additional specimen integrated neutron flux measurements.
f tenth removal and testing shall be conducted during every s
refueling cycle thereafter.
4 I 4 Basis for Specification SR 5.2.5 test program will be performed to survey and assess the A
The testing shifts in HDTT-of the PCRV liner materials.
is to be accomplished by placing Charpy impact test the liner specimen:, made from the liner materials, near a
o and exposing them to appropriate neutron fluxes and 1
be
- The Charpy impact test specimens are to temperatures.
at a time, during the life of the vessel and s
removed, 36 tested to determine the condition of the vessel steel.
The total number of specimens placed in the reactor is
-approximately 750, which will allow the determination of a impact transition curve for the plate metal, the
-complete weld metal and the heat affected zone at each test interval.
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Fort St. Vrain #1 T@chnical Specifications Amendment Page 5.2-12 This testing program will meet the requirements of ASTM-E-185-70, with the following exceptions:
a) Tensile specimens are not included, since the liner is not a load carrying member, but only a ductile membrane.
b) No thermal control specimens have been provided, since there is no appreciable temperature cycling of the liner.
The liner materials will normally be kept at or below 150 degrees fahrenheit during all plant operations.
Tests performed on this liner material (see FSAR Section 5.7.2.2) have indicated that no observable changes in material characteristics developed during an exposure to a fluence equivalent to the first five years of full power operation.
- Further, these tests demonstrated no significant damage after a fluence equivalent to 30 years of full power operation.
The testing program prescribed for the Fort St. Vrain liner is in compliance with the ASME Boiler and Pressure Vessel Code,Section III N-110.
The interval for specimen removal and testing subsequent to the fifth refueling cycle may be adjusted based on the analysis of prior results.
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Fort St. Vrain #1 Technical Specifications c
.bendment Page 5.2-13 i
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Specification SR 5.2.6 - Plateout Probe Surveillance i
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One.
plateout probe sk11, be removed for evaluation 1
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,g coiteident with the first, third, and fifth refueling, and interais;
.f not to exceed five refueling cycles at t
thereafter.. If,,durtrg th'a second or fourth refueling
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- cycle, oV, as,'rgfueling cycle following the fifth
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_.kefueling, the [cipary coolant noble gas activity N
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i (gamma + beta) shoulcW1gcrease by 25% over the average I
activity. of the previous three months at the same reactor a
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onpe r,,l evel and the ' primary coolant activity is greater than 25% of desigh,,the r/ : pout probe shall be removed at de end of that refua ing cycle.
Tb3 probes shall be
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'..analyzei.for "Sr inventory,in, the reactor circuit.
The Q.y probes'rsoved saall aise,b.e analyzed for 2'2I.
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The plateotdprobes are located in penetrations extending
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.into steam generator shrouds and then into the gas stream o'f each, 'oolant,\\lcop.
One sample is accumulated by c
ie-e '% # continuously bypassing a small portion of the core outlet
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- (, coolant stream through diffusion tubes and scrption beds g
3 d ' f' < located in 'the probe body Another sample can be l
,yccumulated by continuously bypassing a portion of the t
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circulator outlet coolant stream through the probe.
The 1
dore outlet sample can be used to determine the concentrations of fission products in the coolant stream s
entering the steam generator; the circulator outlet sample
+
s
Fort St. Vrain #1-Technical Specifications Amendment Page 5.2-14 prvvides information about the amount of cleanup in each pass around the' circuit.
-The probes shall be analyzed for Sr and the results shall be used to establish the total Sr inventory in the reactor' circuit to determine compliance with LC0 4.2.8.
Results of probe analyses shall be compared with the calculated estimates of Sr which were made between probe removals.
The analysis-for 1'11 shall be made to
. determine the degree of conservatism of the assumptions made regarding the circulating and plated out iodine in the-primary coolant circuit.
The interval for probe removal and analysis subsequent to
-the fifth~ refueling cycle may be adjusted based upon the analysis of prior results.
Fort St. Vrain #1 z.
Technical Specifications Amendment Page 5.2 -Specification SR 5.2.7 - Water Turbine Drive Surveillance Components of the helium circulator water' turbine drive system shall be. tested as follows:
a) One circulator and the associated water supply valving in each loop will be functionally tested by operation on water turbine drive using feedwater, condensate, and boosted condensate (supplied to the firewater booster pumps at fire pump discharge pressure),
l' annually, or at the next scheduled plant shutdown if
.l the test was not performed during the previous year.
l' SR 5.2.7a shall be implemented per ISI Criterion G.
b) Safety valves (V-21522, V-21523, V-21542, and V-21543), located in'the water turbine supply lines, l
will be tested for relieving pressure annually, or at l
the next scheduled plant shutdown if the test was not l
performed during the previous year.
l SR 5.2.7b shall be impimented per ISI Criterion G.
c) Both turbine water removal pumps and the turbine water i
drain tank overflow to the reactor building sump shall l
be functionally tested quarterly.
l SR 5.2.7c shall be implemented per ISI Criterion G.
d) The instrumentation and controls associated with c) shall be functionally tested in conjunction with
c; Fort St. Vrain #1-Technical Specifications -
Amendment Page 5.2-Iti and at the same intervals as the turbine water removal
. pumps and shall be calibrated annually.
Basis for Specification SR 5.2.7 The circulator water turbine drives are normally operated during an extended shutdown.
Therefore the specified surveillance requirements are adequate to ensure water turbine operability.
Specification SR 5.2.8 - Bearing Water Pump And Makeup Pump Surveillance l
The circulator bearing water makeup pumps, bearing water l
makeup pumps, and associated instruments and controls shall be tested as follows:
a) The Normal Makeup Pump shall be operated in the l
recycle mode quarterly.
ll SR 5.2.8a shall be implemented per ISI Criterion G.
b) The Emergency Makeup Pump shall be functionally tested l
quarterly.
'l SR 5.2.8b shall be implemented per ISI Criterion G.
c) The associated instruuents and controls shall be functionally tested in conjunction with and at tile intervals specified in parts a) and b) above, and calibrated annually.
Fort St. Vrain #1
_ Technical Specifications Amendment-Page 5.2-17 l
d)l Each Bearing - Water
- Pump, and the associated l
instruments and controls shall be functionally tested l
-at each scheduled plant shutdown.
In addition, the l
instruments shall be calibrated annually, or at the lI next scheduled plant. shutdown if they.were not l.
calibrated during the previous year.
l SR 5.2.8d shall be implemented'per ISI Criterion F.
8 J
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Fort St.'Vrain #1 Technical Specifications Amendment Page 5.2-18 Basis for Specification SR 5.2.8 l
The bearing water pumps and bearing water' makeup pumps are l
required to operate for safe shutdown cooling of the l
reactor under accident conditions described in FSAR l
Section 10.3.9.
The specified tests and testing intervals.
l are sufficient to ensure adequate pump operation for the 1-performance of their required safety functions.
l Performance capability of the bearing water pumps is
.l verified by normal operatior;.
Performance capability of-l the-normal bearing water makeup pump is verified when-
- l operating-the pump in the recycle mode.
Performance l
capability o f-the emergency bearing water makeup pcmp is
_l verified when testing the associated check valves as I
required per SR 5.3.4.
Specification SR 5.2.9 - Helium Circulacor Bearing Water Accumulators Surveillance The helium circulator bearing water accumulators,
-instrumentation, and controls shall be functionally tested l
quarterly and calibrated annually.
I SR 5.2.9 shall be implemented per ISI Criterion E.
Basis for Specification SR 5.2.9 Heli t.m Circulator bearing water is normally supplied from the bearing water system and is backed up by the backup bearing water system supplied from the Emergency Feedwater
Fort St. Vrain'#1 Technical Sp@cifications Amendment Page 5.2-19 Header.
In the event of a failure-in both of these
- systems, the water stored in the bearing water accumulators is adequate to safety shut down both helium l
circulators in -a loop. The specified tests and testing
-l intervals are sufficient to ensure operability of the
' l --
-accumulator controls, should they be called upon to l
perform their required function.
Specification SR 5.2.10 - Fire Water System / Fire Suppression Water System Surveillace a) The fire water system shall be verified operable as follows:
1)-The motor driven and engine driven fire pumps shall be functionally tested monthly.
The associated instruments and controls shall be functionally tested monthly and calibrated annually.
- 2) The diesel engine fuel shall be inventoried monthly and sampled and tested quarterly.
- 3) The diesel engine shall be inspected during each refueling shutdown.
- 4) The diesel engine starting battery and charger shall be inspected weekly for proper electrolyte level and overall battery voltage. The battery electrolyte shall be tested quarterly for proper specific gravity.
Fort St. Vrain #1 Technical Specifications-Amendment' Page'5.2-20
- 5) The batteries, cell plates, and battery racks, shall be inspected each refueling cycle for evidence of-physical damage or abnormal degradation. The battery-to-battery and terminal connections shall be verified to be clean, tight, free of corrosion, and coated with anti-corrosion material each refueling cycle.
.b) The fire suppression water system shall be verified.
operable as follows:
- 1) Monthly ~ by verifying that each valve (manual,
.pcwer operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
-2) Semi-annually by performance of a fire suppression e
water system flush.
- 3) Annually by cycling each testable valve in the fire suppression water system flow path through at least one complete cycle of full travel.
- 4) Each refueling cycle by performing a fire suppression water system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
(a) Verifying that each automatic valve in the flow path actuates to its correct position.
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1 Fort St. Vrain #1 Technical Specifications-
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Amendment' Page 5.2-21
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'(b) Verifying' that each fire water pump develops
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no less 4
~ at least-1,425 gpm at a system -head
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_than 119 psig.
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SR 5.2.10.b4b shall. be implemented per ISI l-Criterion G.
l (c) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.
_(d) Verifying that each fire water pump starts sequentially _to maintain the fire suppression wat'er system pressure at greater than or equal to 275 feet water gauge.
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- 5) Each three years by performing a flow test.
Basis for Specification SR 5.2.10 The-fire water pumps are required to supply water for fire suppression.and safe shutdown cooling.
The specified interval is sufficient to ensure proper operation testing of 'the pumps and controls.
The motor driven pump routinely operates' intermittently.
l The operability of the fire suppression water system safe ensures that adequate fire suppression and emergency shutdown cooling capability is available. The specified operation testing interval is sufficient to ensure proper of_the system when required.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-22 Specification SR 5.2.11 - Primary Reactor Coolant Radioactivity Surveillance A grab sample of primary coolant shall be analyzed a minimum of once per week during reactor operation for its radioactive constituents and shall be used to calibrate the continuous primary coolant activity monitor.
If the continuous primary coolant activity monitor is inoperable, the primary coolant activity level reaches 25%
of the limits of LC0 4.2.8, or the primary coolant activity level increases by a factor of 25% over the previous equilibrium value of the same reactor power level, the frequency of sampling and analysis shall be increased to a minimum of once each day until the cctivity level decreases or reaches a new equilibrium value (defined by four consecutive daily analyses whose results are within + 10%) at which time weekly sampling may be resumed.
Basis for Specification SR 5.2.11 The design of the instrumentation is such that under normal operating conditions the activity of the prima ry coolant is measured and indicated on a continuous basis.
The weekly sampling interval provides an adequate check on the continuous monitoring equipment.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-23 Specification SR 5.2.12 - Primary Reactor Coolant Chemical Surveillance The primary coolant shall be analyzed for chemical constituents a minimum of once per week.
If the chemical impurity levels exceed 50 percent of the limits of LCO 4.2.10 or LCO 4.2.11, whichever is applicable, the frequency of sampling and analysis shall be increased to a minimum of once each day until the level decreases or reaches a
new equilibrium value (defined by four consecutive daily analyses whose results are within
+ 10%), at which time weekly sampling may be resumed.
Basis for Specification SR 5.2.12 The chemical constituents in the primary coolant are routinely measured on a
continuous basis.
Tha specification of an interval for surveillance allows for routine maintenance of the chemical impurity monitoring equipment.
The presence of higher than nominal impurity levels of chemical impurities is related to core materials corrosion which might occur only with very high levels for sustained periods of time.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-24 Specification
'SR 5.2.13 - PCRV Concrete Helium Permeability Surveillance The permeability of the PCRV concrete to helium shall be measured prior to the initial startup of the reactor and after the end of the third year following initial power operation. Additional measurements shall be made at five year intervals thereafter.
Basis for Specification SR 5.2.13 Measurements.
of the relative helium permeability throughout plant life provides, as a supplement to other surveillance efforts, information concerning the continued integrity of tne PCRV concrete.
The interval for surveillance after the fifth year following the initial power operation may be adjusted based on the analysis of prior results.
Specification SR 5.2.14 - PCRV Liner Corrosion Surveillance Requirement The PCRV liner shall be examined for corrosion induced thi.ining, using ultrasonic inspection techniques at the end of the third and fifth years following initial power operation. Additional examinat;ons shall be conducted at ten year intervals thereafter.
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v
. Fort St. Urain #1 Technical Sp@cifications Amendment Page 5.2-25 Basis of Specification SR 5.2.14 The ultrasonic inspection of the PCRV liner is provided to detect-the thinning of the liner due to corrosion or to detect defects within the liner at representative areas.
Although no corrosion is expected to
- occur, this specification allows for' detection of corrosion or liner defects in the event of some unexpected and unpredicted changes in the liner characteristics. The provisions are discussed in Section 5.13 of the FSAR.
The-interval for. surveillance after the fifth year following initial power operation may be adjusted based on the analysis of prior results.
Specification SR 5.2.15 - PCRV Penetration Interspace Pressure Surveillance 4
The instrumentation which monitors the pressure differential between the purified helium supply header to the PCRV penetration interspaces and the primary coolant system will be functionally tested once every month and calibrated annually.
r Basis for Specification SR 5.2.15-This calibration and test frequency is adequate to insure that the purified helita being supplied to the PCRV penetration interspaces shall be at a higher pressure than the primary coolant pressure within the PCRV.
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-26 Specification SR 5.2.16 - PCRV Closure
- Leakage, Surveillance Requirements The surveillance of PCRV closure leakage shall be as follows:
a) PCRV primary and secondary closure leakage shall be determined once each quarter, or as soon as practicable after an unanticipated increase in pressurization gas flow is alarmed.
SR 5.2.16.a shall_be implemented per ISI Criterion A.
b) The instrumentation monitoring PCRV penetration closure interspace pressurization gas flows, including alarms and high flow isolation, shall be functionally tested monthly and calibrated annually.
c) The instrumentation which monitors or alarms pressure in the core support floor and core support floor columns shall be functionally tested and calibrated arnually.
SR 5.2.16.c shall be implemented per ISI Criterion A.
d) The controls, position indication, and fail safe operation for remote manual isolation valves associated with pressurizing, purging, and venting PCRV closures shall be functionally tested at five calendar year intervals, and for automatic isolation valves, annually, or at the next scheduled plant
r Fort St. Vrain #1 Technical Specifications
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Amendmant Page-5.2-27 shutdown-if these valves have not been tested during the previous year.
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SR 5.2.16.d shall be implemented per ISI Criterion B.
the HTFA purge lines shall be e) The check valves on tested at five calendar year' intervals.
shall be implemented per ISI Criterion B.
SR 5.2.16.e f) -The check valves which are part of the HTFA or such refueling penetrations shall only be tested when f
a penetration is open for refueling or maintenance, f five the check valves have not been tested in the last years.
shall be implemented per ISI Criterion B.
SR 5.2.16.f helium purification cooler well closure shall be l
g) Each monitoring leak tested, and the well pressure during each
-l instruments shall be-calibrated, once refueling cycle. In addition, the instruments and
.l l
automatically isolate the controls used to functionally tested at
.l purification system shall be l
the same frequency.
_l SR 5.2.16g shall be implemented per ISI Criterion F.
l Basis for Specification SR 5.2.16
iY.
Fort St. Vrain #1
. Technical Specifications Amendment Page 5.2-28 The interval specified for determining the actual primary and secondary closure leakage is adequate to assure compliance with LC0 4.2.9.
In the determination of closure leakage at the reference differential pressure, laminar leakage flow shall.be conservatively
- assumed, therefore in correcting the determined closure leakage to reference differential
- pressure, the ratio of the reference differential pressure, and test differential pressure shall be used.
The interval specified for functional testing and calibration of the instrumentation and alarms monitoring the penetration closure interspace pressurization gas flow will assure sensing and alarming any change in pressurization gas flow.
The interval specified for functional test and calibration of the instrumentation and alarms monitoring the core support floor and columns will assure sensing and alarming any change in their structural integrity.
The interval specified for valve testing is adequate to assure proper valve operation when isolation of the closure auxiliary piping is required.
l The interval specified for testing the helium purification l
cooler wells is adequate to verify the well integrity, as I
well as that of primary coolant boundary components l
located therein.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-29 Specification SR 5.2.17 - Helium Circulator Pelton Wheels DELETE SPECIFICATION SR 5.2.17 IN ITS ENTIRETY Specification SR 5.2.18 - Helium Circulators Surveillance a) At the time of the first main turbine generator overhaul, one helium circulator unit shall be removed in its entirety from the PCRV and thoroughly inspected for signs of abnormal wear or component degradation.
- 1) Such inspection shall include examination of i
bearing surfaces, seal surfaces, brake system, i
I buffer seal system, and labyrinth seals.
- 2) The helium circulator compressor wheel
- rotor, turbine wheel, and Pelton wheel shall be inspected for both surface and subsurface defects in accordance with
'the appropriate
- methods, procedures, and associated acceptance criteria specified for Class I components in Article NB-2500,Section III, ASME Code.
b) Following the first complete helium circulator inspection, a previously uninspected helium circulator shall be removed and inspected at ten calendar year intervals.
The helium circulator compressor wheel rotor, turbine wheel, and Pelton wheel shall be inspected as specified in Paragraph a.2.
Other helium circulator components, accessible without further
~
Fort St. Vrain #1 Technical Specifications Amendment-Page 5.2-30 disassembly than required to inspect ~ these wheels, shall be_ visually examined.
Results Hof these examinations shall be submitted to.the to determine NRC~ staff for review and shall be evaluated
- the need for scheduling additional future inspections.
SR 5.2.18'shall be implemented per ISI Criterion D.
. Basis for Specification SR 5.2.18
-Experience with.the operation of single stage steam turbines as' prime movers is common throughout industry.
a machine is running satisfactorily, lit'.'r r
Once such no wear occurs to it.
designs of emergency systems of conventional Unlike most Shutdown nuclear power plants, the components of the Safe System of the Fort St.
Vrain plant are utilized and operated during normal operation of the plant.
This i
includes the helium circulators.
The performance of the helium circulators is monitored the during operation, i.e., instruments are provided with capability to measure compressor differential pressure and temperature and flow, bearing temperature, bearing water flow, buffer heliun flow, and shaft speed and vibration.
Examination at the time of the first turbine generator overhaul, and at approximately ten year intervals is sufficient to monitor the condition of the
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-31 helium circulator. The first turbine generator " tear-down" or overhaul usually occurs after one year running to check the total assembly. Only checks of components are performed during subsequent turbine. generator overhauls.
The helium compressor and steam. turbine blading'should experience minimal wear in its running environment, and, with.this. length of service before inspectio, will have undergone sufficient stress' cycling to accurately indicate service life.
Specification SR'5.2.19 - IACM Diesel-Driven Pumps Surveillance DELETE SPECIFICATION.SR 5.2.19 IN ITS ENTIRETY Specification SR 5.2.20 - ACM-Diesel Driven Generator Surveillance a) The diesel driven ACM generator shall be checked weekly by starting, and obtaining design speed and voltage.
b) The generator shall be tested monthly under load for a minimum of two hours. The load under this condition shall be at least 100% of design ACM equipment full load.
l l.
rort St. Vrain #1 Technical Specifications Amendment Page 5.2-32 Basis for Specification SR 5.2.20 A weekly check of the Alternate Cooling Method generator to demonstrate its capability to start and a monthly test of the generator under load provides adequate assurance that the Alternate Cooling Method generator will be available to supply electrical power under the highly degraded, loss of forced circulation situation.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-33 l
Specification SR 5.2.21 - ACM Transfer Switches, Valves,
[
and Instruments Surveillance l
a) Those valves and transfer switches that must be j
manually positioned for actuation of the Alternate l
Cooling Method (ACM) mode of operation shall be tested l
annually, or at the next scheduled plant shutdown if l
such test was not performed during the previous year.
[
SR 5.2.21a shall be implemented per ISI Criterion'F.
l b) Local indicators for the helium purification dryer l
inlet temperature, for the helium purification I
pumpdown line pressure and for the reactor plant l
cooling water surge tank cover gas pressure shall be l
calibrated at each plant shutdown for refueling.
l SR 5.2.21b shall be implemented per ISI Criterion F.
l Basis for Specification SR 5.2.21 l
In the event that the ACM mode of operation must be j
implemented, it is necessary to manually position valves l
(manual valves as well as valves which would normally be l
pneumatically or electrically operated) and to manually I
reposition electrical transfer switches.
Local l
instruments allow system monitoring during the l
depressurization phase of the PCRV and during the j
subsequent cooling phase of the reactor.
The specified l
tests and testing intervals are sufficient to assure
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2 l operability of these componenets should they be called l
upon for performance of their required safety functions.
Specification SR 5.2.22 - PGX Graphite Surveillance PGX graphite surveillance specimens shall be installed into five (5) bottom transition reflector elements of the Fort St.
Vrain core to provide a means for assessing the condition of the PGX. graphite support blocks during operation of the. reactor.
These specimens (16 per reflector element) will be installed in reflector elements as indicated in Table 1 and will be removed at subsequent refueling intervals, as indicated in Table 1, unless the progressive examination of
.the specimens dictate
.otherwise.
Upon
- removal, these specimens will be subjected to examination, and compared with laboratory -
control specimens in evaluating oxidation rates, oxidation profiles, and general dimensional characteristics.
The results of these tests and examinations shall be utilized to assess the condition of the PGX core support blocks in the reactor and shall also be utilized to modify, as necessary, the planned removal of subsequent PGX surveillance specimens.
The results of these examinations shall be submitted to the NRC staff for review.
m._
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-35 Basis for Specification SR 5.2.22 The PGX graphite specimens will be placed in modified coolant channels in five (5) transition reflector elements in the hottest columns of regions 22, 24, 25, 27, and 30.
The surveillance test specimens will be s.bjected to the primary coolant conditions, as well as other reactor parameters that are normally seen by the PGX core support blocks.
Examination and tests of the surveillance test specimens at regular intervals can readily be utilized to assess oxidation rates, oxidation profiles, as well as general degradation of the PGX core support blocks to adequately predict the structural integrity of the core support blocks over the operating life of the reactor.
Fort St. Vrain #1' Technical Specifications Amendment t
Page 5.2-36 SR 5.2.22 PGX-GRAPHITE SURVEILLANCE Table 1 TRANSITION ELEMENT ASSEMBLY WITHDRAWAL SCHEDULE l
l l
Withdrawal at l
l l
l-Refueling-1 l _ Fuel Region l
Column l
Number
- l l
l l
l l
25 l
7 l
2 l
.1 l
l l
l 30 l
3 l
4 l
l
'l i
l l
24 l
7 l
6 l
l l
l l
l' 22 l
6 l
9 l
1 I
I I
l 27 l
2 1
17-l
- Schedule would be adjusted to remove transition element assemblies at a faster rate should specimens at any withdrawal interval show a burnoff significantly greater than predicted.
1 l
Fort St. Vrain #1
~*
Technical Specifications Amendment Page 5.2-37
-Specification SR 5.2.23 - Firewater Booster Pump Surveillance Each firewater booster pump shall be tested annually by providing motive power to one water turbine drive in conjunction with the performance of SR 5.2.7.
In addition each pump shall be functionally tested quarterly.
The associated instruments and controls shall functionally be tested quarterly and calibrated annually.
Basis for Specification SR 5.2.23 During accident conditions described in Final Safety Analysis Report, Section 14.4.2.1, one of the firewater booster pumps and one firewater pump are required to provide adequate core cooling.
The specified testing interval is sufficient to ensure proper operation of the pump and associated controls.
l Speci fication SR 5.2.24 - Reactor Auxiliary Cooling Water l
System Survaillance l
The reactor auxiliary cooling water systems shall be l
tested as follows:
a) The circulating water makeup pond minimum inventory shall be verified daily.
The pond level instrumentation shall be functionally tested monthly and calibrated annually.
Fort St. Vrain #1 Technical. Specifications Amendment Page 5.2-38 I
l
'b) Each circulating water makeup pump and the associated l
-instruments and controls (including firewater pump' pit l
instruments and controls) shall be functionally tested l
monthly.
In addition, the instruments 'shall be l
l-calibrated, and the pump performance capability (flow l
and head) and mechanical condition (vibration l
amplitude and bearing temperature) shall be verified, l
annually or at the next scheduled plant shutdown if
'l this was not performed during the previous year.
l~
SR 5.2.24b shall be implemented per ISI' Criterion F.
c) The valve lineup of the. flow path between the circulating water storage ponds and the fire water pump pits'shall be verified correct monthly.
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~~
Fort St. Vrain #1 Technical Specifications g
~
Amendment Page 5.2-39
_ l-d) Alignment and settlement of the circulating water x-l.
makeup pond embankments shall be verified at five
- l calendar year. intervals.
The embankments and the
- l water structures shall be examined at the same l
intervals for. abnormal _ erosion, cracks,
- seepage, l
leakage, accumulation.of silt or debris (as l
. applicable) which might indicate a deterioration of l
- structural safety or operational adequacy of the l
storage ponds.
l SR 5.2,.24d shall'oe implemented per Id! Criterion G.
- l e) Each service water pump and the associated instruments l
and controls 'shall be functionally tested monthly.
In l
addition, the instruments shall be calibrated, and the l
pump ' performance (flow and head) ar.d mechanical l
condition (vibration amplitude and bearing temperature)?chall be verified, annually or at the l
s l
next scheduled plant shutdown if this was not l
performed during the previous year.
l SR 5.2.24e shall be implemented per ISI Criterion F.
o l
.f) Each reactor plant cooling water pump and the l
associated instruments and controls shall be l
functionally tested monthly.
In addition, the l
instruments shall be calibrated, and the pump l
performance (flow and head) and mechanical condition l
(vibration amplitude and bearing temperature) shall be l-
- verified, annually
'o r at the 'next scheduled plant
~
Fort St. Vrain #1 Technical Sptcifications Amendment Page 5.2 I shutdown if this was not performed during the previous l
year.
l 4
l SR 5.2.24f shall be implemented per ISI Criterion F.
l g) Each purification cooling water pump and the l'
associated instruments and controls shall be l'
functionally tested monthly.
In addition, the l
instruments shall be calibrated, and the pump-
- l-performance (flow and head) shall
'e
- verified, u
a l.
annually or at the next scheduled plant shutdown if l
this was not performed during the previous year.
l-SR 5.2.24g shall be implemented per ISI Criterion F.
l h) Instruments and valves, used for automatic isolation' l
of portions of the reactor plant cooling water system
' l and of the-purification cooling water system, that may l
be required for confinement of reactor coolant, shall l
be functionally tested onces during each refueling l
cycle.
In addition, the instruments shall be l-calibrated at the same interval.
= l SR 5.2.24h shall be implemented per ISI Criterion F.
Basis for Specification SR 5.2.24 l
The reactor auxiliary cooling water systems (including l
water makeup system, service water system, reactor plant l'
cooling water system, ar.d purification cooling water l
system) are required to operate for reactor cooling under 9
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-41 I
postulated loss-of forced circulation cooling accident l
conditions.
Except for the purification cooling water
'~
i-system, they are also required for safe. shutdown cooling l
of the reactor under other postulated accident conditions.
l The circulating water makeup system also supplies water l
for fire suppression.
These systems routinely operate l
during normal plant operation.
Routine operation in l
conjunction with the specified tests and testing intervals l
are sufficient to a,sure adequate system and/or component l
operation for the performance of their required safety
.l functions.
l.
Measuring the position of survey markers and evaluating l
the c.hanges in position of these markers will allow l
changes in embankment alignment and settlement to be l
determined, as well as their possible impact on the
~l structural integrity of the storage pond.
Examination of i
the embankments and of the water structures will provide l
for an additional verification that no phenomenon occurs l
which might be detrimental to the ability of the storage
.l pond to perform its safety function. Measurement of the l
silt accumulation in the storage pond will allow a l
verification that the minimum water inventory required by l
LCO 4.3.5 is available for Safe Shutdown Cooling of the l
reactor.
l The interval specified for instruments and valves is l
adequate to assure their automatic isolation function, if l
degradation were to occur in the integrity of the reactor
,, -.. ~.
E Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-42 l
coolant boundary, resulting in primary coolant leakage l
into the system.
Specification SR 5.2.25 - Core Support Block Surveillance The top surface of the core support block for fuel regions fitted w'ith PGX graphite specimens shall be visually examined by remote TV for indication of cracks, in particular'in areas where analysis shows the highest tensile stresses exist, at the refueling shutdown when the PGX graphite specimens are scheduled to be removed from the core in accordance with Technical Specification SR 5.2.22.
SR 5.2.25 shall be implemented per ISI Criterion D.
Basis for Specification SR 5.2.25 Visual examination of the core support bloc'.s in those regions chosen for insertion of PGX graphite specimens will provide additional assurance that integrity of the core support blocks does not degrade due to plant operating conditions, since those regions were selected because of their -higher potential for pGX graphite burnoff. Analysis shows that the highest tensile stresses occur on the top surface of the core support blocks, at the keyways, and at the web between reactor coolant channels.
- r g
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-43 Specification SR 5.2.26 - Region Constraint Devices Surveillance The region constraint devices (RCD's) shall be inspected j
at each refueling outage using the fuct handling machine from those regions being refueled as fo11cws:
a) The upper core plenum shall be visually examined by remote TV to verify that RCD's within visible range are in place on top of the core, b) As RCD's are removed, the fuel handling machine location coordinates and lifting force shall be monitored to verify that the RCD pins were engaged in the fuel columns and that they disengage as expected.
c) Selected RCD's shall be visually examined by remote TV in the fuel handling machine after removal to verify their structural integrity, d) As RCD's are re-installed, the fuel handling machine location coordinates shall be monitored to verify that the RCD pins have engaged in the fuel columns.
SR 5.2.26 shall be implemented per ISI Criterion B.
r Fort St. Vrain #1 Technical Specifications Amendment-Page 5.2-44 Basis for Specification SR 5.2.26 Region constraint devices, located on top of fuel columns of generally three adjacent fuel regions, restrain region movements in relation to one another by means of centering pins inserted in the handling hole of the upper plenum elements.
Visual examination of the upper core plenum and comparison of the as-installed /as-found RCD coordinates will assure that the RCD's remain in place and that no phenomenon is occurring which could cause them to disengage from the fuel columns. Comparison of RCD coordinates will require correction to account for changes in fuel column height due to irradiation of graphite and coordinate changes which will occur when RCD's are removed from a different refueling penetration than the one from which they were installed.
Monitoring the lifting force to remove the RCD's with the fuel handling machine will provide early indications, should a phenomenon occur over time which might eventually prevent them from moving with the fuel columns or prevent their removal from the reactor.
Removal and re-installation will act as go/no go dimensional test of the region constraint devices.
Visually examining and photographing selected RCD's in the
-fuel handling machine will assure that there are no
f.
Fort St. Vrain #1 f
Technical Specifications Amendment Page 5.2-45 unacceptable deformations, loose or missing parts, or other visible defects.
Spacification SR 5.2.27 - Helium Shutoff Valves Surveillance Proper closure' of the helium shutoff valves shall be monitored annually, or at the next scheduled plant shutdown, if such monitoring has not been perfctmed during the previous year.
SR 5.2.27 shall be implemented per ISI Criterion C.
Basis for Specification SR 5.2.27 The helium shutoff valves are self-actuated check valves which close when the corresponding circulators are shutdown.or tripped.
- imultaneous long term failure of both the circulator and its helium shutoff valve, under very degraded conditions of remaining plant equipment, could lead to a situation analogous to a loss of forced circulation accident, due to the open recirculation path between circulator outlet and inlet plenums.
Verification that the helium shutoff valves close properly will provide assurance that the residual heat removal capability would not be degraded by the malfunction of a helium shutoff valve.
Fort St. Vrain #1 I
Technical Specifications Amendment Page 5.2-46 Specification SR 5.2.28 - PCRV Penetrations and Closures S_urveillance a) Accessible partions of PCRV penetration pressure retaining welds shall be examined for indications of surface defects as follows:
- 1) Surface examine (MT or PT) the following three welds in one steam generator penetration in each loop at five calendar year intervals:
the penetration shell to secondary closure
- weld, the secondary closure to upper bellows support weld, and the lower bellows support to reheat header sleeve weld.
21 Surface examine (MT or PT) the following two welds in the bottom access penetration at 10 calendar year intervals:
the penetration shell to spherical head weld, and the spherical head to closure flange weld.
SR 5.2.28.a shall be implemented per ISI Criterion C.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-47 b) Accessible portions of the PCRV penetration closure and flow restrictor restraint components shall be examined for indications of defects as follows:
- 1) Visually examine the helium circulator restraint system (cylinder, ring, and bolting) for one penetration in each loop at five calendar year intervals.
SR 5.2.28.b.1 shall be implemented per ISI Criterion C.
- 2) Visually examine the refueling penetration holddown plate bolting at each refueling outage.
SR 5.2.28.b.2 shall be implemented per ISI Criterion B.
- 3) Visually examine the bottom access penetration primary closure split ring assembly and its secondary closure bolting at 10 calendar year intervals.
SR 5.2.28.b.3 shall be implemented per ISI Criterion C.
c) Accessible portions of the PCRV safety valve penetration containment tank su7 port components shall be examined at 10 calendar year intervals for indications of defects as follows:
l
' Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-48
tank' attachment weld.
4
- 2) Visually examine the support skirt between the tank and PCRV outer wall.
'3) Visually examine,. torque, and tension test the bolting attaching the support skirt to the PCRV outer wall.
SR 5.2.28.c shall be implemented per ISI Criterion C.
Basis for Specification SR 5.2.28 Structural integ*ity of Fort St. Vrain PCRV penetration secondary pressure retaining boundaries is normally verified by continuous leakage monitoring and by periodic leakage testing of the penetration interspace.
The specified examinations of accersible circumferential welds at structural discontinuities will provide additional asstrance concerning the continued integrity of the secondary pretsure boundary at these critical locations.
Examination of accessible penetration closures, flow restrictors, and equipment restraint or support components provides assurance that these components remain structurally sound and capable of performing their safety function under both normal and accident conditions.
t
Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-49 THIS PAGE INTENTIONALLY LEFT BLANK
t St..Vrain #1
.hnical' Specifications Amendment Page 5.3-1 5.3 SECONDARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS E
Applicability i
Applies to the surveillance of the secondary (steam) coolant system, including the steam generators and turbine plant.
Objective To ensure the core cooling capability of the components of the steam plant system.
n-Specification SR 5.3.1 - Steam / Water Dump System Surveillance a)
The steam / water dump. valves shall be tested individually every three months.
b)
The steam / water dump tank level indicators shall be checked daily, and functionally tested every three months.
SR 5.3.1.b shall be implemented per ISI Criterion A.
c)
The steam / water dump tank level, pressure and temperature-instruments (including indicators, alarms, and interlocks - where applicable) shall be functionally tested and calibrated annually, or at the next scheduled plant shutdown if such i
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Fort St. Vrain #1 Technical Sp:ctfications.
Amendment Page 5.3-2 surveillance has not been performed during the previous year.
SR 5.3.1.c shall be implemented per ISI Criterion B.
l
4
' Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-3 Basis'for Specification SR 5.3.1
~
The steam / water dump system' is. provided to minimize water in-leakage into the core as a. result of a steam generator tube rupture (FSAR, Section 6.3).
Satisfactory operation of the dump valves, as is sufficiently demonstrated by testing every.three months, will minimize core damage and primary coolant system pressure rise in the event of a f
steam generator tube rupture.
The dump valve test will be accomplished by closing the (normally. locked open) block valve downstream of the dump valve to be tested. After operation of the dump valve, the block valve will again be locked open, returning the dump valve to service.
The specified frequency for instrumentation functional test and calibration is adequate to assure that the water
-level in the steam / water dump tank does not exceed the limits of LCO 4.3.3, and, in case of dump, to confirm that the proper steam generator has been dumped, and to prevent venting and draining of the tank to the radioactive gaseous and liquid systems before the contents have been adequately cooled.
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4 Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-4
['
Specification SR 5.3.2 - Main and Hot Reheat Steam Stop Check Valves Surveillance The main steam and hot reheat steam stop check. valves shall be full stroke tested in accordance with 2
specification SR 5.3.4 and partial stroke tested once per week.
SR 5.3.2 shall be implemented per ISI Criterion A.
Basis for Specification SR 5.3.2 The main steam stop check and hot reheat stop check valves j
will be partially stroked once a week during plant i
2 operation.
Full stroking tests are impractical because complete closure of any one valve would automatically shut down one or more circulators. Therefore, the valves will N
be stroked during power operation by means of special
' electrical circuitry in the hydraulic control system which 4
limits closure to 10% without interfering with emergency closure action called for by the plant protective system.
This test will demonstrate that the valves are free to i
close when required, without causing. severe pressure, temperature, flow,.or power generation transients.
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Fort-St. Vrain #1 Technical Specifications Amendment Page 5.3-5 Specification SR 5.3.3 - Bypass and Pressure Relief Valves Surveillance The main steam and hot reheat steam power operated (electromatic) pressure relief valves, and the six hot reheat steam bypass. valves shall be tested once per year, or at the next scheduled plant' shutdown if the_ valves have not been tested during the previous year. The main steam bypass valves shall be tested in accordance with specification SR 5.3.4.
SR 5.3.3 shall be implemented per ISI Criterion A.
Basis for Specification SR 5.3.3 The specified secondary (steam) coolant system bypass valves and pressure relief valves will be tested during plant shutdown as follows:
a)
The main steam and hot reheat steam power operated pressure relief valves will be tested by exercising the valves.
b)
The main steam bypass valves will be tested for operability by cycling the valves.
c)
The six hot reheat steam bypass valves will be tested by e>.ercising each valve to ensure freedom of movement.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-6 The main steam bypass valves divert up to 77?s steam flow (via desuperheaters) to the bypass flash tank on turbine trip or loop isolation, so that the steam is available for driving helium circulators, boiler feedpump turbines, etc.
The main steam power operated relief valves divert the remaining steam' flow to atmcsphere.
The six hot reheat steam bypass valves and the power operated pressure relief valves ensure a continuous steam f'ow path from thn helium circulators for decay heat removal.
The tests _ required on the above valves will demonstrata that each valve will function properly. Test frequency is considered adequate for assuring valve operability at all times.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-7 Specification SR 5.3.'4 - Safe Shutdown Cooling Valves Surveillan_ce l
The following valves shall be tested annually, or at the l
next scheduled plant shutdown if these valves have not l
been tested during the previous year:
Pneumatically, hydraulically, or electrically l
l l
operated valves that 9 required co operate for l
actuation of the safe shutdown cooling mode of I
l operation I
l (implemented per ISI Criterion B);
l Normally closed check valves that are required to l
open for actuation of the safe shutdown cooling l
mode of operation l
(implemented per IS Criterion B); and Valves (including normally power operated valves) l l
that must be manually positioned for actuation of l.
the safe shutdown cooling mode of operation l
(implemented per ISI Criterica G).
l
Fort St. Vrain #1 Technical Specifications-Amendment Page 5.3-8 Basis for Specification SR 5.3.4 1
The safe shutdown cooling mode of operation utilizes systems or portions of systems that are in use during
-normal plant operation.
In many case;, those valves required to initiate Safe Shutdown Cooling are not called upon. to function during normal operation of the plant, except to. stand fully closed or open.
Testing of these valves will assure their operation if f
called upon to initiate the Safe Shutdown Cooling rode of 1
operation.
During reactor operation, the instrumentation required to monitor and control the Safe-Shutdown mode of cooling is normally in use'and any malfunction would be immediately brought to the attention of the operator.
That t-instrumentation not normally in use is tested at intervals specified by other surveillance requirements in this Technical Specification.
Safe Shutdown. Cooling, the systems or portions of systems involved, are discussed in Sections 10.3.9 and 10.3.10 of 4
the FSAR and are represented in FSAP tigure 10.3-4.
s Valve testing will include, as appliu_
e, full stroking l
each valve, or an observation that the valve stem or disc
-travels from the valve normal operating position to the 4-position required to perform the safety function, an
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-9 l
observation that the remote position indicators accurately I
reflect actual valve position, and a measurement of the full stroke time for the hydraulically actuated automatic valves.
Specification SR 5.3.5 - Hydraulic Power System hrveillance The pressure indicators and low pressure alarms on the hydraulic oil accumulators pressurizing gas and on the hydraulic power supply lines shall be functionally tested once every three months and calibrated once per year.
Basis for Specification SR 5.3.5 The hydraulic power system is a normally operating system.
Malfunctions in this system will normally be detected by failure of the hydraulic oil pumps or hydraulic oil accumulators to maintain a supply of hydraulic oil at or above 2500 psig.
Functional tests and calibrations of the pressure indicators and low pressure alarms on the above basis will assure the actuation of these alarms upon a malfunction of the hydraulic power system which may compromise the capability of operating critical valves.
L
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-10 Specification SR 5.3.6 - Instrument Air System Surveillance The pressure indicators and low pressure alarms on the instrument air receiver tanks and headers shall be functionally tested monthly and calibrated annually.
Basis for Soecifi:ation SR 5.3.6 The instrument air system is a normally operating system.
Malfunctions in this system will be normally detected by failure of the instrument air compressors to maintain the instrument air receiver tanks at a pressure above the alarm setpoint.
Functional tests. of the pressure indicators and low pressure alarms on a monthly basis and calibration on' an annual basis will assure the actuation of these alarms upon a malfunction of the instrument air system which may compromise the capability of operating critical valves.
Specification SR 5.3.7 - Secondary Coolant Activity Surveillance The secondary coolant system will be analyzed for 182I, tritium, and gross beta plus gamma concentration once per week during reactor operation.
If the secondary coolant activity level reaches 25% of the limit of LCO 4.3.8 or the secondary coolant activity level increases by a factor of 25% over the previous
Fort St-Vrain #1 Techni< ' Specifications Amendme..
Page 5.3-11
. equilibrium value at the same reactor power level, the frequency of sampling and analysis shall'be increased to a minimum of once each day until the activity level decreases or reaches a new equilibrium value (defined by four consecutive daily analyses whose results are within
+10%), at which time weekly sampling may be resumed.
Basis for Specification SR 5.3.7
.The specification surveillance interval is adequate to monitor the activity of the secondary coolant.
Specification SR 5.3.8 - Hydraulic Snubbers Surveillance
-The following surveillance requirements apply to all Class I piping system hydraulic snubbers:
a)
All hydraulic snubbers whose seal material has been demonstrated by operating experience, lab testing or analysis to be compatible with the operating
-environment shall be visually inspected.
This inspection shall include, but not necessarily be limited to, inspection of the hydraulic fluid reservoir, fluid connections, and mechanical linkage connections to the piping and anchor to verify snubber operability in accordance with the following schedule:
Fort St. Vrain #1' Technical Specifications Amendment Page 5.3-12 Number of Snubbers Found Inoperable During Inspection Next Required or During-Inspection Interval Inspection Interval 0
18 Months plus or minus 25%
4 1
12 Months plus or minus 25%
2 6 Months plus or minus 25%
-3, 4
124 Days plus or minus 25%
5,6,7 62 Days plus or minus 25%
greater than or equal to 8 31 Days plus or minus 25%
The required inspection interval shall not be lengthened more than one step at a time.
b)
All hydraulic snubbers whose seal materials are other than ethylene propylene or other material that has been demonsteated to be compatible with the operating environment shall be visually inspected for operability every 31 days.
c)
The initial inspection shall be performed within 6 months from issuance of this Technical Specification.
For the purpose of entering the schedule in a) above, it sha'l be assumed that the facili ty had been on -a six (6) month inspection interval.
d)
Once each refueling cycle, starting with the first refueling, a representative sample of 10 hydraulic snubbers or approximately 10 percent of the hydraulic snubbers, whichever is less, shall be
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-13 functionally tested for operability including verification of proper piston movement, lock up and
. bleed.
For each unit and subsequent unit found inoperable, an. additional 10 percent or ten snubbers shall be so teited until no more' failures are found or all units have been tested.
Snubbers of rated capacity greater than 50,000 pounds need not be functionally tested.
Basis for Specification SR 5.3.8 All Class I hydraulic snubbers are visually inspected for overall integrity and operability.
The inspection will include verification of proper orientation, adequate hydraulic fluid level and proper attachment of snubber to piping and structures.
The inspection frequency is based upon maintaining a constant level of snubber protection.
Thus, the required inspection interval varies inversely with the observed snubber failures.
The number of inoperable snubbers found during a required inspection determines the time interval for the next inspection. However, the results of such early inspection _ performed before the original required time interval has elapsed (nominal time less 25 percent) may not be used to lengthen the required inspection interval. Any inspection where results require a shorter inspection-interval will override the previous schedule.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-14 Experience at operating facilities has shown that the required surveillance program should assure an acceptable level of snubber performance provided that the seal materials are compatible with the operating environment.
Snubbers containing seal material which has not been demonstrated by operating experience, lab tests, or analysis, to be compatible with the operating environment should be inspected more frequently (every month) until material compatibility is confirmed or an appropriate changeout is completed.
To further increase the assurance of snubber reliability, functional tests should be performed once each refueling cycle.
These tests will include stroking of the snub' ers o
to verify proper piston movement, lock up and bleed.
Ten percent or ten snubbers, whichever is less, represents an adequate sample for such tests. Observed failures on these samples should require testing of additional units.
Snubbers in high radiation areas or those especially difficult to remove need not be selected for functional tests provided operability was previously verified.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-15 Specification SR 5.3.9 - Safety Valves Surveillance l
Safety. valves shall be tested as follows:
l a)
The steam generator superheater and reheater safety l
valves and the steam / water dump-tank safety valves shall be tested at five calender year intervals to verify'their set' point.
-l' SR 5.3.9_
shall be implemented per ISI Criterion B.
l b)
All cther Class I safety valves (not covered by l
.other surveillance requirements) shall be tested at I
ten calendar year intervals to verify their l-setpoint.
l SR 5.3.9b shall be implemented per ISI Criterion F.
Basis for Specification SR 5.3.9 l
Safety valves protect the integrity of the plant l
components which are part of the primary or secondary l
reactor coolant boundary, and also the integrity of l
systems required to safely shutdown and cool the reactor l
under accident conditions.
Testing the safety valve setpoints' will assure that the pressure within the equipment ccmains within design limits.
When practical, testing of the safety valves will be scheduled during the surveillance interval so that testing of one (or more) safety valve (s) of similar type and
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-16 operating conditions several times during the interval l
will provide additional confidence in safety valve reliability and adequate overpressure protection.
Specification SR 5.3.10 - Secondary Coelant System Instrumentation Surveillance The secondary coolant reheat steam instrumentation used a) for control and indication of emergency condensate flow to the reheaters and reheater backpres.;ure, in l
case of safe shutdown cooling, b)
.to automatically open the reheater discharge bypass on high' pressure, and i
c)'
to monitor reheater discharge bypass temperature, i
and reheater inlet temperature, shall be functionally tested and calibrated annually, or
}
at the next scheduled plant shutdown if such surveillance s
was not performed during the previous year.
4 i
SR 5.3.10 shall be implemented per ISI Criterion B.
i 4
1 1
i i
. -... ~,....... _., - _ _ _ _. ~,
,.. -, ~
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-17 Basis for Specification SR 5.3.10 The frequency specified for surveillance of the above instrumentation will assure that they perform their automatic actions, and that the operator will be expected can use for provided with accurate information which he or to avoid abnormal equipment safe shutdown cooling operation.
Generator Bimetallic Welds Specification 5.3.11_- Steam Surveillance generator bimetallic of steam accessible -portions The of welds shall be volumetrically examined for indications subsurface defects as follows:
main steam ring header collector to main steam a)
The each piping weld for one steam generator module in loop at five calendar year intervals, steam ring header collector to collector b)
The main drain piping weld for one steam generator module in each loop at five calendar year intervals.
modules initially c)
The same two steam generator selected shall be re-examined at each interval.
welds described in a) and b) shall d)
The bimetallic two other steam generator also be inspected for
i Fort St. Vrain #1 Tochnical Specifications Amendment Page 5.3-18 modules in each loop during the initial examination.
SR 5.3.11 shall be implemented per ISI Criterion C.
Basis for Specification 5.3.11 The steam generator crossover tube bimetallic welds between Incoloy 800 and 2 1/4 Cr-1 Mo materials are not accessible for examination. The bimetallic welds between the steam generator ring header collector, the main steam piping, and the collector drain piping are accessible, involve the same aaterials and operate at conditions not significantly different from the crossover tube bimetallic welds.
The collector drai r-piping weld is also geometrically similar to the crossover tube weld.
Examination of selected bimetallic welds that are accessible will provide additional assurance concerning the continued integrity of steam generator bimetallic welds. Although no degradation is expected to occur, this specification allows for detection of defects which might result from conditions that can uniquely affect bimetallic welds made between these materials. Additional collector welds are inspected at the first examination to establish a baseline which could be used, should defects be found in later inspections and additional examinations subsequeritly be required.
I
Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-19 THIS PAGE INTENTIONALLY LEFT BLANK 1
Fort St. Vrain #1 Technical Spscifications Amendment #36 - 10/13/83 Page 5.4-1 5.4 INSTRUMENTATION AND CONTROL SYSTEMS - SURVEILLANCE AND CALIBRATION REQUIREMENTS Applicability Applies to the surveillance and calibration of the reactor protective.., em and other critical instrumentation and controls.
Objective To assure the operability of the reactor protection system and other critical instrumentation and controls by specifying their surveillance and calibration frequencies.
Specification SR 5.4.1 - Reactor Protective System and Other Critical Instrumentation and Control
- Checks, Calibrations, and Tests The surveillance and calibration tests of the protective instrumentation shall be as given in Tables 5.4.1 through 5.4.4:
a)
Table 5.4.1 - Minimum Frequencies for
- checks, calibrations, and testing of scram system.
b)
Table 5.4.2 - Minimum Frequencies for
- checks, calibrations, and testing of Loop Shutdown System.
-.- -. a
Fort St. Vrain #1 Technical Specifications
~
Amendment #36 - 10/13/83 L
Page 5.4-2 c)
Table 5.4.3 -' Minimum Frequencies for
- checks, calibrations, and testing of Circulator Trip System.
d)
Table 5.4.4 - Minimum Frequencies for-checks, calibrations, and testing of Rod Withdrawal Prohibit System.
Basis for Specification SR 5.4.1 The specified surveillance check and test minimum frequencies are based on established industry practice and operating experience at conventional and nuclear power plants. The. testing is in accordance with the IEEE Criteria. for Nuclear Power Plant Protection Systems, and
~1n accordance with accepted industry standards.
Calibration frequency of the instrument channels listed in Tables 5.4.1, 5.4.2, 5.4.3. 5.4.4 are divided into three categories:
passive type indicating devices that can be
-compared with like units on a
continuous basis; semiconductor devices and detectors that may drift or lose sensitivity; and on-off sensors which must be tripped by an ~ external source to determine their setpoint. Drift tests by GGA on transducers similar to the reactor pressure transducers (FSAR Section 7.3.3.2) indicate insignificant long term drift.
Therefore a once per refueling. cycle calibration was selected for passive devices (thermo-couples, pressure transducers, etc.).
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-3 Devices incorporating sem' conductors,
.particularly amplifiers, will be also calibrated on a once per refueling cycle 'b' asis, and any drift in response or bistable setpoint_ will be discovered from the test program.
Drift of electronic apparatus is not the only consideration in determining a calibration frequency; for example, the change in power distribution and loss of detector chamber ser.sitivity require that the nuclear power range. system be calibrated every month. On-off sensors are calibrated and tested on a once per refueling cycle basis.
)
Fort St. Vrain #1
~
Technical Specifications Amendment #36 - 10/13/83 Page 5.4-4 TABLE 5.4-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF SCRAM SYSTEM l
Channel l
1 l
l l
Description i
Function Frequency (1)I Method l
l 1. Manual la. Test R
la. Manually trip l
l (Control Room)I I
I system.
l-l 2. Manual (I-49)-la. Test l
M la. Manually trip each l l
l l
l channel.
l l 3. Startup la. Check l
0 la. Comparison of two !
l Channel
.I l
l separate channel l
l.
l l
l indicators.
l l
lb. Test l
P lb. Internal test l
l l
l l
signal to verify l
l l
l l
trips and alarms.
l
'l Ic. Calibrate l-R lc. Internal test l
l l
l l
signal shall be
[
[
l l
l checked and l
l l
-l l
calibrated to l
l l
l l
assure that its l
l l
l l
output is in l
l-l l
l accordance with thel I
l l
i design l
l-l l
l requirements. Thisl l
'l l
l shall be done afterl l
l l
l completing the l
l l
l l
external test l
l l
l l
signal procedure byl l
l l
l checking the output l l
l l
l indication when l
l l
l l
turning the l
l l
l l
internal test l
l l
l l
signal switch.
_1 l.4. Linear Power la. Check l
D la. Comparison of six l
l Channel l
l l
separate channel l
I I
l l
indicators.
l
.l lb. Test l
M lb. Internal test l
l l
l l
signal to verify l
-l l
l l
trips and alarms.
l l
lc. Calibratel D
lc. Channel adjusted tol l
l l
l agree with heat l
l l
l l
balance l
l l
l l
calculation.
l l
ld. Calibrate l R
ld. Internal test l
l l
l l
signals to adjust l l
l l
l trips and l
l l
l l
indications.
l l S. Wide Range la. Check l
D la. Comparison of threel l'
Power Channel l l
l separate l
l l
l l
indicators.
_[
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-5 TABLE 5.4-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF SCRAM SYSTEM (Cont'd) l Channel l
l l
l l
Description i
Function IFrequency(1)l Method l
l S. (Cont'd)
Ib. Test l
P lb. Internal test l
l l
l l
signals to verify l l
l l
l trips and alarms.
l l
lc. Calibrate l M
lc. Channel adjusted tol l
l l
l agree with heat l
l l
l l
balance l
l l
l l
calculation.
l l
ld. Calibrate l R
ld. Internal test l
l l
l l
signals to adjust l l
l l
l trips and l
l l
l l
indications.
l l 6. Primary la. Check l
D la. Comparison of two l l
Coolant l
l l
separate high levell l
Moisture (All l l
l channel mirror l
l Channels) l l
l temperature l
l l
l indications.
l l
lb. Check l
D lb. Comparison of six l
l l
l l
separate low level l l
l l
l channel mirror l
l l
l l
temperature l
l l
l l
indications.
l
Fort St. Vrain #1 Technical Sp cifications Amendment #36 - 10/13/83 Page 5.4-6 TABLE 5.4-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING 0F SCRAM SYSTEM (Cont'd) l-Channel l
l l
l
.l Description l
Function l Frequency (1)l Method l
l 6. (Cont'd) lc. Calibrate l R
lc. Inject moisture l
l l
'l l
laden gas into l
l l
l l
sample lines.
l l
ld. Check l
0 ld. Verification of I
l l
l l
eight separate l
l l
l l
monitor's sample l
l l
l j
flow, per Item (t) l l
l l
l of Notes for Tables l l
l
-l l
4.4-1 through l
l l
l l
4.4-4.
l l
le. Test l
M le. Verify that each ofl l
l l
l the eight monitors l l
l l
l will alarm on low l l-l l
l and high sample l
7 l_
l l
l flow.
l l 7. Primary
.la. Test l
M la. Trip one high l
l
. Coolant l
l l
level, one low l
l Moisture (Highi l
l level channel, I
l Level l
l l
pulse another low l l
Channels)
'l
..]
l level channel.
l l 8. Reheat Steam ja. Check l
0 la. Comparison of the l l
Temperature l
l l
averaged l
l
.l l
l thermocouple l
l l
l l
channel input l
l l
l l
indications.
l l
lb. Test l
M lb. Trip channel, l
l 1
l l
verify alares and l l
l l
l indications.
l
.l l
l l
Internal test l
l l
l l
signal to verify l
l l
l l
trips and alarms.
l l
lc. Calibrate]
R lc. Compare each
(
l l
l l
thermocouple output l l
l l
l to an NBS traceable l l
l l
l standard.
Internall l
l l
l test signal to l
l l
l l
adjust trips and l
l l
l l
indicators.
l l 9. Primary la. Check l
D la. Comparison of six l
(~
Coolant l
l l
separate channel l
l Pressure l
l l
indicators.
l l
lb. Test l
M lb. Trip channel, l
l l
l l
internal test l
l l
l l
signal to verify l
l l
l l
trips and alarms.
l
Fort St. Vrain #1 Technical Specif1 cations-Amendment #36 - 10/13/83 Page 5.4-7 TABLE 5.4-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF SCRAM SYSTEM (Cont'd) ll Channel l
l l
l l
Description l
Function IFrequency(1)I Method l
l-9. (Cont'd)
[c. Calibrate l R
lc. V own pressure l
l l
1 l
applied to sensor. l
-l l
.l l
Internal test l
.l l
l l
signal to adjust l
l
_ l l
l trips and l-l 1
I l
indicators.
l l10. Circulator la. Check l
D la. Comparison of eightl l~
Inlet l
l l
separate l
l
. Temperature l
l l
indicators.
l l
lb. Test l
M-lb. Trip channel, I
l l-l l
internal test l
l 1
-l l
signal to verify
-l l
l l
l trips and alarms.
l l
lc. Calibrate l R
Ic. Compare'each l
.l l
l l
thermocouple outputi l
l l
l_
to an NBS traceablel l
l l
l standard.
Internal l
-l :
l l
l test signal to l
l l
l l
adjust trips and l
l-1 i
I indicators.
l l11. Hot Reheat la. Test l
M la.-Reduce pressure at l l
Header l
l l
sensor to trip l
l Pressure I
l
[
channel, verify l
l l
l l
alarms and l-l l
l l
indications.
l l
lb. Calibrate l R
lb. Known pressure l
l l
l l
applied at sensor ]
l I
I I
to adjust trips.
l
_l12. Main Steam la. Test l
M la. Reduce pressure at l l
Pressure l
l l
sensor to trip l
l l
l l
channel, verify l.
l l
l l
alarms and l
l l
l l
indications.
l l
lb. Calibrate l R
lb. Known pressure l
.l l
l l
applied at sensor [
l l
l l
to adjust trips.
l l13. Two Loop la. Test i
M la. Special test modulel l
Trouble l
l l
used to trip l
,l l
l l
channel by l
l l
l l
energizing each of l l
l l
l four appropriate l
l l
l l
pairs of two-loop l
l l
l 1
trouble relays.
l
Fort St. Vrain #1 t
Technical Specifications Amendment #36 - 10/13/83 Page 5.4-8 TABLE'5.4-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF SCRAM SYSTEM (Cont'd) l Channel l
l l
[
.l Description l
Function l Frequency (1)l Method l
l13. (Cont'd) lb. Test l
R lb. Trip logic-to causel l
l l
l two loop trouble l
l l
l l
l l14. Plant 400 V la. Test l
-M la. Trip each channel l
l Power Loss-
-l l
l by applying l
-l l
l l
simulated loss of l
~l l
l l
voltage signal; l
l l
l l
verify alarms and l l~
l l
l indications.
[
l15. High Reactor la. Check l
D la. Comparison of threel
-l_
Building l
l l
separate channel l
l Temperature
-l.
l l
indicators.
l l-(Pipe Cavity) lb. Test l
M lb. Trip channel, l
l l
l l
verify alarms and ]
l l
l l
indications.
l
-l l
l l
Internal test-l l
l l
l signal to verify l
l l
l l
trips and alarms.
l l
lc. Calibrate l R
lc. Compare each l
l
'l l
l thermocouple output]
- l l
l l
to an NBS traceable l l
l l
l standard to adjust l l
l l
l temperature trip l
l:
l l
l point.
l NOTE 1: D - Daily when in use.
M - Monthly.
R - Once per refueling cycle.
P - Prior to each startup, if not done the previous week.
e-
.~
n Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/33 Page 5.4-9 TABLE 5.4-2 MINIMUM FdE00ENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF LOOP SHUTDOWN SYSTEM
.l Channel
[
l l
l l
Description l
Function IFrequency(1)l Method l
l 1. Steam Pipe la. Check l
0 la. Comparison of l
l Rupture (Pipe ]
l l
separate ultrasonic l l
Cavity) l l
l channel indicators /l l
l l-l loop.
l l
lb. Test i
M lb. Pulse test one l
l l
l l
temperature and l
l l
l l
- p. essure channel l
l l
l l
with another l
l l
1 I
temperature and l
l l
l
{
oressure channel l
l l
l l
tripped, while l
l l
l l
simultaneously l
l-l l
l having two l
l l
l l
ultrasonic channels l l
l l
l tripped.
l l
l l
l l
l Jc.
l lc. DELETED l
l l
l l
l ld. Test l
M ld. Pressure switch l
l l
l l
actuat3d by l
l l
l l
pressure applied atl l
l l
l sensor.
I l
le. Test l
M le. Temperature switch l l
l l
l actuated by heat.
l l-l l
l applied at sensor. l l
lf. Test i
M If. Internal test l
l l
l l
signal to adjust
'l l
l l
l ultrasonic trip.
I l
lg. Test l
M
[g. Trip test signal
[
l l
l l
solenoid valves to ]
l I
l i
verify loop l
l l
l l
integrity.
l l-lh. Calibrate l R
lb. Known pressure l
l l
l l
applied at sensor l l
l l
l to verify response.l l
l1. Calibrate l R
l1. Known temperature l
l l
{
l applied at sensor l l
l l
l to verify response l l
lj. Calibrate l R
lj. Known sound applied]
l l
l l
at sensor to verify l
-l l
l l
response.
l l 2. Steam Pipe la. Check l
D la. Comparison of I
l Rupture (Underl l
l separate ultrasonici l
PCRV) l l
l channel indicators /l l
l l
l loop.
l
Fort-St. Vrain 01 Technical Spscifications Amendment #36
-10/13/83 Page 5.4-10 TABLE 5.4-2 MINIMUM FREQUENCIES.FOR CHECKS, CALIBRATIONS, AND TESTING OF LOOP SHUTDOWN SYSTEM (Cont'd) l Channel l
l l
l Description l
Function IFrequency(1)l Method l-l 2.:(Cont'd) lb. Test l
.Y.
lb. Pulse test one l
I l
1 l
temperature and l
l=
l l
l pressure channel l
l l
l l
with another-l l
l l
l temperature and l
l l-l l
pressure channel l
1 l
l l
l tripped, while l
l l
l l
simultaneously l
l l
l l
having two l-l l
l l
ultrasonic' channels l
.l l
l l
tripped.
l 1
1 1.
I l
Ic.
l lc. DELETED l
l t
Fort St. Vrain #1
~,- ' -
1 Technical Spzcifications Amendment #36 - 10/13/83 Page 5.4-11 TABLE 5.4-2 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING 0F LOOP SHUTDOWN SYSTEM (Cont'd) l-Channel l
l l
l
-l Description l
Function IFrequency(1)I Method l
l 2._(Cont'd) ld. Test l
M ld. Pressure switch
-l l
l l
l actuated by l
i l
l l
l pressure applied atl I
l l
I sensor.
l l
le. Test l
M le. Temperature switch-l l
l l
l actuated by heat l
l l
l
[
applied at sensor. l l
jf. Test l
M If. Internal test l
l
'l l
l signal to adjust l
l l
l l
ultrasonic trip.
l l
- 19. Test l
M
- 19. Trip test signal l
l l
l' l
l l
l l
to verify loop l
l l
l l
integrity.
l l
lb. Calibrate l R
lb. Known pressure l
l l
l l
applied at sensor l l
l l
l to verify response.l l
l1. Calibrate l R
l1. Known temperature l
l l
l l
applied at sensor l l
l l
l to verify response.l l
lj. Calibrate l R
lj. Known sound appliedl l
l l
l at d-tector to l
l l
l l
verify response.
[
l 3. Circulator 1A la. Test l
M la. Pulse test and l
l and IB Tripped]
l l
verify proper l
l l
l l
indications.
l l
lb. Test l
R lb. Trip both l
l.
l l
l circulators to testl l-l l
l loop shutdown.
l l 4. Circulator 1C la.-Test l
M la. Pulse test and l.
I and ID Tripped l l
l verify proper l
.[.
l l
l indications.
l
.l lb. Test l
R lb. Trip both l
l.
l l
l circulators to testl l
l l
l loop shutdown.
l l 5. Steam la. Test l
M la. Pressure switches l
l~
. Generator l
l l
actuated by l
l Penetration l
[
l pressure applied.
l l
Pressure Ib. Test l
M lb. Pulse test each l
l l
l l
channel with l
l l
l l
another channel l
l l
l l
tripped and verify [
a J
l l
l l
proper indications.l t
~
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83' Page 5.4-12 TABLE 5.4-2 MINIMUM FREQUENCIES FOR' CHECKS, CALIBRATIONS, AND TESTING OF LOOP SHUTDOWN SYSTEM (Cont'd) l Cnarnel l
l l
l l
'Descri3 tion l
Function IFrequency(1)l Method l
- l S. (Cont'ii).
Ic. Caiibratel R
lc. Known pressure l
l l
l-l applied at sensor l l
i I
I to adjust trip.
l l 6.
Reheat Headerla. Check l
-0
-la. Comparison of threel l
Activity' l
l l
separate indicators l l
l l
l in each loop.
l
-l lb. Test l
M lb. Pulse test each l
l l
l'
.l channel with l
l l
l l
another channel l
l l
l l
tripped and verify l l
l_
l l
prope-indications.l-l lc. Calibrate l R
lc. Expose sensor-to l
l
.l l
l known-radiation l
l l
l l
source _and adjust l l
l
.l l
trips and l
l l
l l
indicators.
l
7 Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-13 TABLE 5.4-2 MINIMUM FREQUENCIES FOR_ CHECKS, CALIBRATIONS, AND TESTING
-0F LOOP SHUTDOWN SYSTEM (Cont'd) l' Channel l
l l
l
.l Description l
Function l Frequency (1)l Method l
- l 7. Superheat la. Check l
D la. Comparison of threel l
Header l
l l
separate l
l
-Temperature l
l l
temperature l
l l
l l
indicator's per l
I l
1 l
loop.
l l
lb. Check l
D lb. Comparison of threel l
l
-l l
separate l
l l
l l
temperature l
l l
l l
differential l
~l l
l l
indicators.
l l
lc. Test l
M
[c. Pulse test one
[
).
l~
l l
channel with l
j
-l l
l another channel l
l l
l l
tripped and verify l
.l l
l l
proper indications.l l
ld. Calibrate l R
ld. Compare each l
l.
l l
l thermocouple output l l
l l
l to an NBS traceable l l
l l
l standard.
Internall l
l l
l test signal to l
l l
l l
adjust trips and l
l i
i l
indicators.
l l. 8. Primary.
la. Inst l
M la. Trip each channel, l l
Coolant l
l l
verify proper l
l Moisture (Low l l
l indications.
l l
Level Ib. Test i
M lb. Trip each channel, I l
Channels) l l
l pulse test'other l
l l
l l
loop to check loop l
'l' l
I I
identification.
l l 9. Primary la. Test l
M la. Pulse test one l
l Coolant l
l l
channel with l
l Pressure l
l l
another channel l
l l
l l
tripped and verify l l
t l
l proper indications,l l
i l
l both channels.
l t
t NOTE 1:
D - Daily when in use.
M - Monthly.
R - Once per refueling cycle.
P - Prior to each startup, if not done the previous week.
~
4 Fort St. Vrain #1
- Technical Specifications
. Amendment'#36 - 10/13/83 Page 5.4-14 TABLE 5.4-3 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND T 4
OF CIRCULATOR TRIP SYSTEM l
. l7 Channel l
l l
Method
_l Description 1
Function IFreque_cy(1)]
n l-1. Circulator la. Check l
0 la. Comparison of six l l_
l l
-separate speed l
l-and Water l
l.
-l indications per l
-.l
. Speed, Steam, I l
l J
circulato.
l l
. l' lb. Test-l M
lb. Internal test j
l l
signal sto verify l
l l
l l
trip setting and
.i' l
l l
l indicators.
l l
l
'l lc. Test l
M lc. Pulse test one l
l l
channel with l
l l
l l
another channel l
l tripped, and verifyl l
l l
l
'l. proper indications.l l
1 l
l R
ld. Known pulse l
ld. Calibrate l frequency applied. l l
-l-l l
'at sensor to adjustl l
l l
l l
l l
l trips and l
_l l-l l
l indicators.
l 0
la. Comparison of six l
-l-j 2. Feedwater Flowla. Check l
separate indicatorsi l
l l
l l
l per loop.-
-l l
l lb. Test l
M lb. Internal test 4
l l
l signal to verify l
l trip setting and l
l l
l l
l l
}
indications.
l
. l l
l lc. Test l
M lc. Pulse test one l
_l l
]
cnannel with l
f.
l l
l l
another channel l-tripped, and verify l 4
l l
l l
proper indications.l l
l ld. Apply known delta Pl l
l R
l l
}
at flow I
l l
[d. Calibrate l l
l l
l transmitter.
l 4
l l
l l
Internal test l
l l
l signal to adjust l
l l
l l
l trips and l
_l l
indicators.
ila. Comparison of threel l
I 1
l^3. Circulator-la. Check l
0 l
l l
separate l
Pressure l
l l
indicators /
l l
l Bearing _ Water l l
_I circulator.
_l l
l_
y 9
+w w -
y<
p,y
-m.<-
-,13,.y n.,
r w m y.,,yy.y
-w
,y,--,,3-yw.,,,yy,ww-.vw,,.me,-y-,
9
--y-wp, ye,..w..
.+wy,9 g.-w-~
,m.,.qs.y.~..
a Fort St. Vrain #1 Technical Specifications Amendment ~#36 - 10/13/83-Page 5.4-15 TABLE 5.4-3.
MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF CIRCULATOR TRIP SYSTEM (Cont'd) l Channel l
l.
l l
l Description-l Function IFrequency(1)I Method l
l 3. (Co :t'd) lb. Test.
l M
- 16. Pulse test one l
l-l l
l channel with
-l l
l l
l another channel l
l l
l l
tripped and verify i l
l l
l proper indications.1 l
lc. Calibrate l R
lc. Known pressure l
l l
l applied to adjust ~ ll l
l l
l trip setting.
l l 4. Circulator.
la. Test l
M la. Pressure switches
[
l Penetration l
l l
actuated by l
l Pressure l-l l
pressure applied.
l l
lb. Test l
M lb. Pulse test one l
l l
l l
channel with l
l l
l l
another channel l
l l
l l
tripped and verify l l
l l
l proper indications.l l
lc. Calibratel R
lc. Known pressura l
l l
l l
applied at sensor l l
i l
l to adjust trip l
l l
~l I
setting.
l l S. Circulator ja. Check l
D la. Comparison _ of threel l
Drain Pressurel l
l separate l
l l
l l
indicators /
l l
l l
circulator.
l l
lb. Test l
M lb. Pulse test one l
l l
l l
channel with-l l
l l
l another channel I
l l
l l
tripped and verify l l
l l
l proper indications.l l
lc. Calibrate]
R lc. Known pressure l
l l
l l
applied at sensor l l
l l
l to adjust trip l
l l
I setting.
l l 6. Circulator la. Check l
D la. Comparison of threel l
Seal l
l l
separate l
l Malfunction l
l l
indicators /
l l
l l
l circulator.
l l
lb. Test i
M lb. Puls test one l
l l
l channel with l
l l
l l
another channel l
l l
l l
tripped and verify [
l l
l l
proper indications.l
Fort'St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-16 TABLE 5.4-3
. MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF CIRCULATOR TRIP SYSTEM (Cont'd) l Channel:
l l
l l
l Description l
Function IFrequency(1)I Method l
l 6. (Cont'd)
Ic. Calibrate l R
lc. Known pressure l
l l
.l l
applied at sensor l l
l l.
l to adjust trip l
l l
l l
setting.
l l 7. Circulator la.. Test l
R la. Trip steam turbine l l
Trip (Manual) l l'
l drives.
Verify l
l l
l l
water turbine l
l
[
l l
automatic start.
l NOTE 1: D - Daily when in use.
M - Monthly.
R - Once per refueling cycle.
P - Prior to each startup, if not done the previous week.
e Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-17 TABLE 5.4-4 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS. AND TESTING OF ROD WITHDRAWAL PROHIBIT SYSTEM l
. Channel l
l l
l l
Descrintion l
Function IFrequency(1)I Method l
l11. Startup la. Check l
D la. Comparison of two l l
Channel-l l
l separate channel l
l l
l l
indicators.
l l
lb. Test l
P lb. Internal test l
l l
l l
signal to verify l
l l
l l
all trips and l
l l
l l
alarms.
l l
lc. Calibrate l R
lc. The internal test l l-l l
l signal shall be l
l l
l l
checked and l
l l-l l
calibrated to i
l l
l I
assure that its l
l l
l l
output is in l
l l
l l
accordance with thel l
l l
l design l
l l
l l
requirements. Thisl l
l l
l shall be done afteri l
l l
completing the l
l l
l l
external test l
l l
l l
signal procedure byl l
l l
l checking the output l l
l l
l indication when l
l l
l l
turning the l
l l
l l
internal test l
l l
l l
signal switch.
l l 2. Linear Channella. Check l
D la. Comparison of six l
l l
l l
separate level l
l-l l
l indicators.
l l
lb. Test l
M lb. Internal test l
l l
l l
signal to verify l
l l
l l
trips and alarms.
l l
lc. Calibrate l D
lc. Channel adjusted tol l
l l
l agree with heat l
l l
l l
balance l
l l
l l
calculation.
. l l
ld. Calibrate l R
ld. Internal test l
l l
l l
signals to adjust l 1
l l
I trips and l
'l l
l 1
indications.
l l 3. Wide Range la. Check l
D la. Comparison of threel l
Power Channel l l
l separate l
l l
l l
indicators.
l 4
.t i
a 9-
+-
%y
- + +
4.-,
,,,,--,c
,.y
--v.,
c,ww,,-- -. -me
-,-7
-e r
,---,g-
~ ~,,.
,-,-y e-+-,w--
,.,,e--4--m m y.,,,...vwv---<e,--,
- - + -
?
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-18 TABLE 5.4-4 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING OF ROD WITHDRAWAL PROHIBIT SYSTEM (Cont'd) l Channel l
l l
l l
Description I -Function IFrequency(1)]
Method l
- l 3. (Cont'd) lb. Test l
P lb. Internal test l
l l
l-l-
signals to verify l_
{
]
l l
trips and alarms.
l l
lc. Calibra-l M
lc. Channel adjusted tol l
l tion l
l agree with heat l
l l
l l
balance l
l l
l l
calculation.
l l
[d. Calibra-l R
ld. Internal test l
l l
tion l
l signals to adjust l l
l l
l trips and l
l l
l l
indications.
l l 4. Multiple Rod la. Test l
P la. Attempt two rod l
l Pair I.
l l
pair withdrawal.
l
. l Withdrawal lb. Check l'
R lb. Simulate current-l l
l l
l through sensor to l
l l
l l
verify trip and l
l l
l l
alarms.
l NOTE 1: D - Daily when in use.
M - Monthly.
R - Once per refueling cycle.
P - Prior to each startup, if not done the previous week.
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-19 Specification SR 5.4.2 - Control Room Smoke Detector The control room smoke detectors and alarms will be functionally tested once per year.
l Basis for Specification SR 5.4.2 The control room smoke detectors provide for sensing of the smoke in the outlet air ducts from both the control room and the auxiliary electrical room.
In the event of any fire or smoke in the control panels, alarms will be initiated.
Specification SR 5.4.3 - Core Region Outlet Temperature Instrumentation The output of two thermocouples measuring each region outlet temperature will be checked daily during power operation.
If the indicated temperatures for a region differ by 2 + 75 F, a calibration shall De made and the faulty thermocouple replaced by an operable thermocouple.
The core region outlet thermocouple shall be calibrated once per year during power operation by traversing a calibrated thermocouple along each of the seven coolant thermocouple assemblies.
i Fort St. Vrain #1 Technical Spscifications Amendment #36 - 10/13/83 p
Page'5.4-20 Basis for Specification 5.4.3
- The long-term thermocouple drift is estimated to be s 15 F per year and this drift.was included in the measurement uncertainty of.1 50*F used to establish LC0 4.1.7. -With this measurement uncertainty, a
. root mean square difference of 2 1 75 F would be an indication of a faulty reading.
Daily. checks and yearly calibrations are considered adequate since the expected _ drift in calibration is small and has been included in establishing LCO 4.1.7 (See FSAR-Section 7.3.3).
a m
mai n m
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-21
~
l Specification SR 5.4.4 - PCR'!
Cooling Water System l
Temperature Instruments Surveillance
_l The PCRV cooling water system temperature instruments l
shall be tested as-follows:
l a).
Once' a month during plant operation at power a l
' scanner reading shall be taken of the inlet header l
and tube outlet temperatures.
The inlet l'
temperature readings shall be compared to the l
corresponding temperature indicators.
The
. l associated temperature alarms shall also be l
functionally tested at the same frequency.
l SR 5.4.4a shall be implemented per ISI Criterion G.
l b)-
The
- scanner, the inlet-and outlet header i
l temperature indicators, and the outlet subheader
.l temperature indicators shall be calibrated l
annually.
l SR 5.4.4b shall be implemented per ISI Criterion G.
ll c)
The inlet header and tube outlet thernocouples, l
which provide input to the scanner, shall be l
calibrated at five calendar year intervals.
l SR 5.4.4c shall be implemented per ISI-Criterion G.
4 6
mm-n-.-
ava. -,
,,o nw4
,s
.--v
,,.-y,--y,e-.
-y-ws y-y,,-p,+
.,v,.
e,
,n--,,p
, - - -, + - - - < - - -
-e--r-,
,w..wn,,
y
L'
' Fort St-. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-22 Basis for Specification SR 5.4.4-l A. scanner is used~for monitoring the PCRV cooling system l
water inlet temperature and iridividual tube water outlet l
temperatures, and for alarming high-outlet temperatures.
l Periodic scanner readout provides the information l
necessary to evaluate the water temperature increase in l
individual tubes.
A comparison of inlet temperature l'
scanner reading to corresponding inlet temperature l
-indicators assures that unacceptable drift in the scanner l
electronics does not occur.
4 l
Calibration of the scanner and temperature indicators l
assures the accuracy of temperature mea suremer.ts,
in l
particular for verifying compliance with LCO 4.2.15.
l To the extent practical, thermocouples in individual l
subheaders will be calibrated at various times during the l
interval, to assure that unacceptable thermocouple drift
]
does not occur.
l The specified intervals for checks and calibrations are l-sufficient to provide accurate temperature measurements to l
adequately protect the PCRV concrete and to monitor the l
integrity of the thermal barrier.
A
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-23
.l
' Specification SR 5.4.5 - PCRV Cooling Water System Flow l-Instruments Surveillance
-l A PCRV Cooling System scanner flow readout shall be taken l
and normal mode alarms functionally checked monthly.
The l
scanner and alarms, and six (6) subheader flowmeters shall l
be calibrated annually, or at the next scheduled plant l
shutdown if they were not calibrated during the previous l
year.
l SR 5.4.5 shall be implemented per ISI Criterion G.
Basis for Specification SR 5.4.5 l
Flow scanning acts as a backup to temperature scanning and l
initiates no automatic protective actions, only an alarm.
l Because a restriction or a leak in the system would l
develop over a period of time, the specified interval for
]
comparing flow readouts is sufficient to detect any long l
term change in the system.
Specification SR 5.4.6 - Core AP Indicator - Surveillance Requirement The core AP instrumentation shall be ca'.. orated on a once per refueling cycle interval.
Basis for Specification SR 5.4.6 Core differential pressure is an indication of gross blockage of flow in the core.
l.
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-24 Specification SR 5.4.7 - Control Room Temperature -
Surveillance Requirement The control room temperature control thermostat shall be functionally tested monthly and calibrated annually.
Basis for Specification SR 5.4.7 The surveillance interval specified for functional testing and calibrati]n of the control room thermostat will assure its ability to not only control the room temperature as desired, but to also indicate the correct room temperature within the accuracy of the instrument.
Specification SR 5.4.8 - Power to ' low Instrumentation -
Surveillance Requirement The power to flow indication shall be verified daily and shall be calibrated once per refueling cycle.
Basis for Specification SR 5.4.8 The power to flow ratio indication is an indication of the balance between the heat generation and removai within the primary coolant system.
A verification of the power to flow indication on a daily basis is adequate to assure the instrument is indicating properly.
In addition, any change in reactor power level no matter how small, should produce a change in the power to flow ratio indication. A lack'of response by this instrumentation would be noticed
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-25 by the operator. Calibration of the instrumentation on a once per refueling cycle basis is acceptable by industry standards for this type of instrumentation.
Specification SR 5.4.9 - Area and Miscellaneous Process Radiation Monitors - Surveillance Requirement The area radiation monitors shall be functionally checked weekly and calibrated annually.
Basis for Specification SR 5.4.9 The surveillance interval specified for functional testing and calibration are adequate to assure the proper operation of these detectors.
Specification SR 5.4.10 - Seismic Instrumentation -
Surveillance Requirement The Seismic Instrumentation shall be functionally tested l
every six months and calibrated every two years.
f Basis for Specification 5.4.10 l
l l
The intervals specified for testing and calibration of the i
Seismic Instrumentation are recommended by the manufacturer to assure the instruments operate as intended.
l:
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-26 Specification
-SR 5.4.11 - PCRV-Surface Temperature Indication - Surveillance Requirement The PCRV surface-temperature indicators shall be functionally tested monthly and calibrated annually.
Basis for Specification SR 5.4.11 The PCRV surface temperature indicators provide for continuous monitoring of surface concrete temperatures to assure the proper temperature gradient is maintained through the PCRV. wall and heads.
.The surveillance interval specified is adequate to detect any drift or malfunction of this instrumentation.
Specification SR 5.4.12 - Analytical System Primary Coolant Moisture Instrumentation - Surveillance
' Requirements 4
The analytical
' system primary coolant moisture
. instrumentation shall be calibrated on a once per refueling cycle basis.
Basis for Specification SR 5.4.12 The surveillance interval specified for calibration of this instrumentation will assure the proper operation of these detectors.
M
- @ h/
O b
////p
/g$'4
+ '{,3 IMAGE EVALUATION
/jp//
"g,, f///
TEST TARGET (MT-3)
,4 #4 W
%,,,,I g,//
l.0 l#MM 5l fha
!? m u
!.8 1.25 1.4_ 'l i
-I 1.6 4-150mm 6"
+-
,,f I
- li >
- ?
//
l
Fort St. Vrain #1 Technical Specifications Amendment #36 - 10/13/83 Page 5.4-27 Specification SR 5.4.13 - 480 V Switchgear Room Temperature Indication - Surveillance Requirement The 480 V switchgear room temperature indicator and alarm shall be functionally tested monthly and calibrated annually.
Basis for Specification SR 5.4.13 The surveillance interval specified for this instrumentation assures its proper operation on a
continuous basis.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.5-1 5.5 CONFINEMENT SYSTEM - SURVEILLANCE REQUIREMENTS Applicability Applies to the surveillance of the reactor building (confinement) and the reactor building ventilation system.
Objective To ensure that the structure and components of the reactor building and ventilation systems are capable of minimizing the release of radioactivity to the atmosphere during potential abnormal conditions.
Specification SR 5.5.1 - Reactor Building, Surveillance Requirements The instrumentation which monitors the reactor Failding sub-atmospheric pressure will be functionally tested once every month and calibrated once a year.
Basis for Specification SR 5.5.1 The reactor building atmosphere is normally maintained slightly below atmospheric pressure by the ventilation system (see FSAR Section 6.1.3.2).
This requirement minimizes the amount and consequences of airborne activity released from the plant under most conditions (see FSAR Section 14.12.8). The leak rate of the building itself is not a
significant parameter as is shown in FSAR Section 6.1.4.2.
Fort St. Vrain #1-Technical Specifications Amendment Page 5.5-2 Specification SR 5.5.2 - Reactor Building-Pressure Relief
- Device, Surveillance The reactor building overpressure relief system differential pressure-switches shall be functionally tested.en a monthly basis and cc.ibrated annually.
The louver groups shall be individually exercised.
quarterly.
Quarterly louver testing may be performed while the reactor is in operation only if the fo'ilowing prerequisites are adhered to:
a)
Reactor shall be under normal steady state operating conditions.
b)
Primary coolant pressure is within the normal envelope for existing conditions.
c)
Reactor building ventilation system is operating per Technical Specifications, d)
No radioaccive gas waste releases are in progress, nor is fuel handling being performed.
e)
No airborne activity above background as indicated by the building activity monitors.
f)
Area radiation monitors and local alarms are operable per Technical Specifications.
s
--m--
4,
,e-n.
---c------e.-
--c-+-----.-r,.e--,
. ~. - - -
-,----x
--*-,---e-
Fort St. Vrain #1 Technical Specifications Amendment Page 5.5-3 g)
No surveillance testing is being performed on the reactor ventilation system or the radiation monitoring systems.
h)
Only one segment (group of louvers) of the louver system shall be tested at any given time.
i)
Communication shall exist between personnel performing the tests and the control room operators.
j)
Capability shall exist to manually shut the louver panels.
k)
Testing of the louver system shall not exceed a total duration of six (6) hours in any one quarter.
1)
Non-compliance with any of the above conditions will require testing to be discontinued and the louver system will be returned to normal.
The reactor building relief (louver) system shall be exercised annually.
Fort St. Vrain #1 Technical Specifications Amendment
-Page 5.5-4 P
1 -
Basis for Specification SR 5.5.2 The reactor building pressure relief device is designed to.
protect the building in the event that pressure in the i
Reactor Building exceeds the' turbine building pressure.by 3 inches of water.
The device consists of louvers irstalled in a number of individual modules operated by mechanical linkages to pneumatic actuators (see FSAR' Section 6.1.3.4).
The specified test frequency shall i
ensure the operability of the reactor building relief s
system.
l Specification SR 5.5.3 --Reactor Building Exhaust System l'
Surveillance.
l
- The exhaust filters and fans in the reactor building ventilation system shall be tested as follows:
a)
A laboratory analysis of a representative carbon j.
sample obtained in accordance with Regulatory i
Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shall be performed after i
i each 4400 hours0.0509 days <br />1.222 hours <br />0.00728 weeks <br />0.00167 months <br /> of operation of the unit, or following painting, fire or chemical
- release in any ventilation zone communicating with the unit.
The results of laboratory carbon sample analysis from the unit shall show 2 90% radioactive methyl c
iodide removed when tested in accordance with ANSI N510-1975 (130 C, 95% R.H.).
f i
,m.-
,.m.,
.,y.
_n..,
ee,
_v.
,-,,e-,.,,..,..,m.,,-y.,,,
m_,,.,,
.,rr-m-
,-r,,
._,_y,.
Fort St. Vrain #1 Technical Specifications Amendment Page 5.5-5 i
- Defined as any material which could reasonably be expected to interfere with the charcoal to adsorb methyl iodide.
4 i
b)
A halogenated hydrocarbon test shall be performed once per calendar year or after each replacement of a
charcoal adsorber bank or after structural maintenance on the filter housing.
Halogenated i'
hydrocarbon removal by-the charcoal filters shall be 2 99% when conducted at normal flow conditions i
in accordance with the applicable portions of ANSI N510-1975.
j c)
The HEPA filters shall be leak tested in place once v
per calendar year, after each complete or partial i
j replacement of a HEPA filter bank, or after any structural maintenance on the filter housing, using cold 00P.
Cold DOP removal by the HEPA filters shall be 2 99% when tested in accordance with the applicable portions of ANSI N510-1975.
d)
Flow. distribution across the HEPA and charcoal filters will be tested with initial operation of the system and following any structural i
modification to the filter housings.
Air h
distribution shall be demonstrated within + 20%
across the HEPA and charcoal filters when tested in
^
accordance with ANSI N510-1975.
i i
b
r Fort St. Vrain #1 Technical Specifications Amendment Page 5.5-6 e)
Total pressure drop across the combined HEPA filter l
and charcoal adsorber banks shall be verified I
weekly to be less than 6" H O at filter design flow 2
l 10%.
I SR 5.5.3e shall be implemented per ISI Criterion G.
l f)
The performance capability (capacity and total l
pressure) and mechanical condition (vibration l
amplitude) of each reactor plant ventilation l
exhaust fan shall be verified annually, or at the i
next scheduled shutdown if such verification was i
not performed during the previous year.
SR 5.5.3f shall be implemented per ISI Criterion F.
I g)
Instrumentation associated with the above filters l
and fans shall be calibrated annually, or at the i
next scheduled shutdown if calibration was not l
performed during the previous year.
SR 5.5.39 shall be implemented per ISI Criterion F.
Basis for Specification SR 5.5.3 The reactor building exhaust filter system is designed to filter the reactor building atmosphere prior to release to the facility vent stack during both normal and accident conditions of operation. The system consists of three 50%
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Fort St. Vrain #1 Technical Specifications Amendment Page 5.5-7 capacity units, two of which are in continuous operation, with the third on standby.
High efficiency particulate air (HEPA) filters are installed before the charcoal adsorbers to remove particulate matter from the air stream to prevent clogging of the iodine adsorbers.
The charcoal adsorbers are installed to reduce the potential release of radioiodine to the atmosphere.
Bypass leakage for the charcoal adsorbers and particulate removal efficiency for HEPA filters are determined by halogenated hydrocarbon and DOP respectively.
The laboratory carbon sample test results indicate a radioactive methyl iocide removal efficiency for expected accident conditions. The surveilalnce test frequencies specified establishes system performance capabilities.
If system conditions are as specified, the calculated doses will be less than the guidelines stated in 10 CFR 100 for the accidents analyzed, as indicated in Sections 14.8 and 14.12 of the FSAR.
Pressure drop across the combined HEPA filter and charcoal adsorber of less than 6 inches of water at the filter design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.
The activated carbon adsorber in the affected unit should be replaced if a representative sample fails to pass the iodine removal efficiency test. Any HEPA filters found defective should be replaced.
r Fort St. Vrain #1 Technical Specifications Amendment Page 5.5-8 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 surveillance.
l In addition to routine operation, each reactor plant ventilation exhaust fan is functionally tested, and its l
capacity verified, when performing the required filter l
tests.
The specified performance test and testing l
. interval is sufficient to ensure adequate fan operation l
for the performance of their required safety. functions l
under postulated loss of= forced circulation cooling l
accident conditions.
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- Fort St. Vrain #1 Technical Specifications Amendment Page 5.7-1 f
5.7 FUEL HANDLING AND STORAGE' SYSTEMS - SURVEILLANCE REQUIREMENTS Applicability Applies to surveillance of the fuel handling and fuel storage systems during irradiated fuel handling and storage.
Objective To ensure the prevention of any uncontrolled release of radioactivity during fuel handling and fuel storage by establishing the minimum frequency and type of surveillance on the equipment for the fuel hancling and storage systems.
i Specification SR 5.7.1 - Fuel Handling Machine Surveillance The surveillance of the fuel handling machine will be as follows:
1
- j a)
Prior to refueling, the fuel handling machine cooling water leak detector will be functionally tested.
b)
A functional test of the Fuel Handling Machine and
?
Isolation Valve movements, interlocks, limit I
switches, and alarms will be performed or simulated prior to annual refueling periods.
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? ge 5.7-2 Basis for Specification SR 5.7.1 The fuel handling machine provides for the safe refueling of the reactor.
To assure the reliability of the fuel handling machine during the refueling operation, the machine and its associated interlocks, limit switches and alarms will be tested prior to refueling. All motions of the machine should be cycled, including the pick-up and release of a dummy element. A test of the helium system and the cooling system will be made.
These checks will assure the capability. to maintain the proper atmosphere environment within the machine to prevent any uncontrollable release of activity, proper purging and back filling capabilities, and the capability to maintain temperature of fuel elements within the machine below 750 F.
Specification SR 5.7.2 - Fuel Storage Facility Surveillance l
The surveillance of the fuel storage facility shall be as follows:
a)
The fuel storage facility helium pressure indicators and alarms will be calibrated and functionally tested annually.
I Fort St. Vrain #1
. Technical Specifications Amendment Page 5.7-3 b)
The fuel storage facility _ cooling system flow indicators, and flow and temperature alarms -shall be calibrated and functionally tested annually.
.l c)
The fuel storage facility emergency ventilation-l system shall be functionally tested annually for
[
each fuel storage vault.
l SR 5.7.2c shall be implemented per ISI Criterion E.
Basis for Specification SR 5.7.2 The fuel storage wells are provided for safe storage of new and irradiated fuel-elements. The basic desigh of the wells is to provide a low temperature dry helium environment.
All conditions connected with this requirement are monitored by pressure, temperature, and-flow sensitive devices.
The temperature and flow detecting devices maintain surveillance of the wells' two independent cooling systems and are set to alarm at previously determined maximum or minimum values. The pressure sensitive device is available to guard against any _over pressurization of the wells.
The specified annual surveillance interval is sufficient to insure proper operation of the instrumentation.
l Emergency ventilation provides adequate backup spent fuel l
cooling should both water cooling systems become l
inoperable for one fuel storage well.
The specified test
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- Fort St. Vrain #1-Technical Specifications Amendment Page 5.7-4
- _ l and testing frequency are sufficient to demonstrate 'the l~
operability of the fuel storage vault emergency l
ventilation system, should it-be called ' upon for l
performance of its required safety function.
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ATTACHMENT 3 SAFETY ANALYSIS REPORT FOR THE FORT ST. VRAIN IN-SERVICE INSPECTION AND TESTING PROGRAM
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e A*+= h t 3 Page 1 SAFER / ANALYSIS REPORF EUR ME FSV IN-SERVICE INSPECTICE AND TESTING PROGRAM UPDAIE 1.
BACIGROCEO
'Ihe Ebrt St. Vrain in-service ireps.Licn and testing program is apar 4 fied by the Plant Tar +mir al 3=-4ficatim Surveillance Requiranants (Ref.1).
In response to a ccanitment in the 1972 Safety Evaluation Report (Ref. 2) Public Service Ccupany has been review 2ng, as a ccn+4n"4ng effort, the in-service inspection and testing program for Fort St. Vrain to f==*=-k the acquired operating experience with the plant, and to update the program in light of more recent rules and regulaticms.
9 nal 1972 Safety Evaluation Report (Ref. 2) included a
'!he avi 4
ccumitment to review the ir.-avice ir% Lien program for the primary coolant system after five years of reactor operation.
'Ihe status of the review effort was neigin=11y described by Public Service Ctsnpany, together with the planned approach to follow in ermfn=4w with the 1972 Safety Evaluaticn Report cannitment (Ref. 3). A review of Public Service renpany plans was performed by the Nuclear Regulatory re=ni==icn, who also identified priority itens to be addressed beyond the smpe of the original Safety Evaluation Report connitment (Ref. 4). 'Ihm general in-service ir %h and testing program review plan and the priccity items were fatt kr discussed in letters and at meetings between the 1+v laar Regulatory cr=ai==4on and Public Service Ccupany until a basic agrasenant was reached between both parties (Ref. 5 ther-h 10). A achark'le was es*ah14 ahad for the review of surv=411=nce requia. Ats for all major plant systans and equignent by subdividing them in priority categories as requested by the Nuclear Regulatory ce=n4=aica (Ref.11).
In the first phase of the 914-review and update, implemented by Jaendment No. 33 to the Operating License (Ref.12), the surveillance requiranents pertaining to plant systans and equipnent identified as priority category I were addressed (namely the g-L = sed concrete reactor vessel (PCRV), the reactor in*=vna1=, the reactor primary and secondary coolant systens, and the PCRV a# 14ary system).
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Page 2 Public Service W ny's plan to W ed with a review and update of the surved11ance requirements for other reactor
="v4?4=vy process systens and equignant knoortant to safety were outlined in a letter to NRC (Ref.13). Diis second and last phase of the in e vice inspection and testing Wu-u review and update crupletes Public Service n,nnany's crzanitments regarding this subject.
2.
EmmnrtYW AND REVIDi PW9 2 e adequacy of the surv=411=nce requirements for individual systems and e-----
Ls was reviewed in light of the inscrtance of their safety functicms to prevent or mitigate the -eequences of postulated = M dants that could cause undue risks to the health and ufety of the public.
S e highest S L-x;e to safety was ammigned to those systems and n=vvients which are critical for the mitigaticn of a postulated ;----- L loss of forced circulaticn cooling Wdant where potential for fuel damageexists (Design Basis Wdant No.1), under the Alternate Cmling.%thod (EN) mode of operation.
Following were those systans and cruponents which are critical for the mitigaticn of postulated loss of nozmal forced circulats.cn cooling -4d-nts, under the Safe Shutdown Cooling node of operation. Diis cooling method applies to accidents initiated by envis-ital dis +'% (earthquake, tornado) or by aq"4==nt failures such as faarhater or steun pipe rupture.
Next those systems.and ccamonents required to mitigate the cxmsequences of a PCRV Asaisurization accident were ocnsidered (M=v4== Credible Mr-4 dant and Design Basis Wdant No. 2).
1 Finally, the review included those systerns and u@wnts required to mitigate the consequences of a loss of normal cooling of a spent fuel storace well.
For each =Mdant cutlined above, the various auxiliary systems and ccupenents wers systenatically evaluated for identificz. tion of their critical safety functions. Exa. sting Technical Spar-i*icaticn Surveillance Bequ2.rements were then reviewed to a==a== their adequacy for ensuring operability of these identified systems and Pts, shculd they be called upon for the performance of their required safety functions. Credit was taken, as appropriate, for reutine operaticn of systems and
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Ai.6.2==nt 3 Pace 3 emp,nents during rv, mal plant opmetion as a factor in the dem:nstration of their operability. System operational readiness as well as operability of pmps, ccmoressors, fans, valves, controls. and instrunentation were all addressed.
In addition, systank and empcments with a primary or e7 reactor coolant pressure M=4=7 function, or with a containment isolation function, were also reviewed for
=<i===7 of the related surveillance requiranants. As previously indicated in Amendment No. 33 to the Operating License, the @=4= was placed on nonitoring t% structural integrity of the reactor ocolant pressure boundaries wherever prac+i c-a1
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3.
ETALUATICN AND CDICLUSICNS
' Die y--si W to the Tar +ml<-=1 Ap=<-4 4'4'-= tion Survaillarce aequirements generally expand the scope of in-service - 4 6 and testing that is currently performed at the Fort St. Vrain Nuclear Generating Staticn. 'Ihis, in essence, provides greater assurance of plant safety and ral4=h414ty.
Individual surv=i11mww requa.runents have been evaluated in detail by Public Service Ccapany. 'Iha results of these reviews revealed that existing surv=411mu m requirements were generally adequate in light of plant operating experience, iwimse to safety, unique design features and limitations, and ASME Code dev=1&t for large IIIGR designs. Minor Wi*ications to i
survaillance intervals were made to reflect w atig experience, and to provide operating flav 4h414ty. Additional tests were included to assure the operability and accaracy of instrumentaticn which can be used for mcnitoring the structural integrity of i
major plant aqnimt. Additicmal
--:--rt testing was rea:mnanded, as a result of detailed reviews of plant systems, either when ccupenants inwimd to safe plant shutdown and cooling were not in the scope of the current Technical Specifica-tiens, or when the testing methcxi could be inproved to provide additicnal c.ssurance of mutuw:st ra14=hi14ty.
Since the r W changes to the T e kle-al Specifications do not result frcan nodifications to plant equignent, nor do they involve reductions in the margins of safety, it can be ccerluded that no unreviewed safety question is raised. Besides the pu 0==' changes, no changes are required to other parts of the Technical Specificat:.ons.
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REFERENCES 1.
Plant Ttachnic'11 Specifications 2.
Safety Evaluation Report of January 20, 1972, Secticn 3.3 3.
Public Service Czcany letter dated Cub.i.=r 13, 1978 i
(P-78169), In-service Inspection - Ibrt St. Vrain 4.
Nuclear Regulatory h4==im let'ar dated January 15, 1979, In-service Irmipin. Lion and Testing Pl.ws-u for Fort St.
Vrahl 5.
Public Service Capany letter dated Mrch 15, 1979 (P-79058), In-service Inspection Progran for Fort St. Vrain 6.
Nuclear Regulat.my e-4==icn letter dated June 5,1979, Sumary of Meeting Held cm May 2,1979, to Discuss In-service Inspecticn 7.
Public Service Ompany Pi:Wi=5s Report. Meeting held on August 20, 1979, between the Nuclear Regulatory h4*=4m and Public Service cmtvwly.
8.
Public Service Ompany letter dated August 22, 1979 (P-79176), Fort St. Vrain In-service Insp iico and Testing Prws==
9.
Nuclear Regulatory C - i==4cn letter dated cui.d. 5, 1979, Proposed Plan of In-service Ir-po.iicn and Testing for Fort St. Vrain
- 10. Public Service Capany Prwu=55 Report. Meeting held on Ncm:mber 1,1979, between the Nuclear Regulatory F-4==4m and Public Service empely
- 11. Public Service Cmpany letter dated Novmber 30, 1979 (P-79289), Fort St. Vrain In-service Inspecticm and Testing Pr@umu
- 12. Fort St. Vrain Nuclear Generating Station, 7mendment No. 33 l
to Facility Operating License CPR-34, dated March 8, 1983.
- 13. Public Service Capany letter dated June 1,1983 (P-83195)
In-service Inspec?h and Thesting Status Update i
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.MM4G 6;
ATTACHMENT 4
SUMMARY
AND RESULTS OF A REVIEW 0F THE ADEQUACY OF THE APPLICABLE ISI SURVEILLANCE REQUIREMENTS L
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IN19000CTICN 9 tis Attactment 4 sumarizes, ard presents the results of, a detailad review of the applicahle Tar-hnir-=1 F 4fication Surv=411mmu Baquirements for the Fort St. Vrain reactor "wi14a7 systens iuyo.Liant to safety ard not addressed in Anunichent No. 33 to the Operating T f e=naa.
Die systens and ocmptr.ents under the scope of this review are those which have a critical safety funct. ion in case of postulated M aants that could cause undue risks to the health and safety of the public, namely:
reactor ecoling using U1e
- Icss of Forced Circulaticm CTliM (A4 =r*'===d in section 2; Alternate Cboling Method - N N), as
- Ioss of Nnmal Forced Cirr1=+4at Cooling (reactor cooling d in using the Safe Shutdown t'mlig method), as dia
'===
section 3;
- Icss of Beactor Pr==='+2ticn (reactor cooling using the normal shutdown cooling method with feedwater), as d4=rm===d in sectica 4; and
- Ioss of Spent Fuel Storage Well Water Cooling (spent fuel cooling using the boosted ventilation method), as di====4 in Sec?Jcn 5 In additicn, as discussed in Sectiat 6, the.
- included an evaluation of the adequacy of in avice impacticm and testing requirements for verifying:
- the structural integrity of reactor coolant primary and
==<vwwh7 bourdad== am-* ated with the helian purification system;
- the integrity of the PCRV liner tbamal barrier using instumentaticn in the reacra plant cooling water system;
- the isolation Th414ty of auv414=7 systems which may act as reactor coolant==mndan boundary;
- the stzuctural integrity of the circulating water makeup 6L.* pcrids; and
- the operability of system annhha_n required to maintain system structural integrity.
Die scope Mhad above covers those itens that were left open fran the first phase of the review, as well as the other items not previously addressed in the first phase but which should be cerisidered for in-service inspectica and testing.
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IDSS CF EURCED CIRCIIATICN CCXLING ACCIDDTI 2.1 SYSTEM SA51!TY ETNCTIWS
'Ihe arv idarit cxxxiiticms postulate that a ramanent loss of forced cizm1=+4rn cooling occurs mncurrent with, or caused by, a loss of the three-roca ccmtwl eczuplex. 'Ihm various nuclear safety functions are amnplished as described below.
0:mtrol of reactivity is achieved by reactor scram, and by marnal inserticm into the. reactor core of the reserve shutdown (RSS) poisen material. 'Ihe reactor Wicctial system is not included in this review and surv=411awa requirements for the reserve shutdown systen were previously reviewed as part of.
License Amendment No. 33.
<moa it is determined that restaption of forced circulation is not ===ih1=, it har m== m== 7 to depressurize the primary coolant system in order to prevent er--
cL=ble damage to reactor internal ---,---- d., Miich might be caused by natural
-v Lion of hot gas to m= ally moler regicos of the reactor. Depr=-emtion occurs to the atnesphere after treatznant of the reactor coolant ilwudi both the hellun purificatism systen and the reactor plant ventilation exhaust sytam. Effluent releases are mcmitored for radiew-tivity.
Cooling required for operation of the hallun purification system is provided by the reactor plant cooling water system for the high tangerature filter adanrhars, by the purificaticn cooling water system for the heliun curification coolers, and by the liquid nitrogen systen for the low *armarature adsorbers.
Beactor r==4^=1 and decay heat renoval is achieved by thermal radiation to the liner cooling water system. 'Ihe cooling water flow is redistributed to account fcr relative changes in heat i
loads between the various reactor regicms. 'Ihe cooling water system pressure is increased as nar===7 to prevent boiling due to the potentially higher than normal cooling water t
t- - -r8tures. 'Ihe heat load is transferred frczn the liner /
reactor plant moling water systen to the service water system.
'Ihe <-==h414ty exists to operate the service water systen in its normal closed loop configuratien using the cooldrg tower and the return puups. However, enough water inventory is available in the circulating water makeup iL.uinge ponds to operate the service water system for a sufficient time in a once through i
ccmfiguraticn, in sez"es with the circulating water makeup system. As an ultimate backup, firewatei could be cirralated directly through the PCRV liner cooling e we.
However, under the postulated accident conditiens, firewater is dedicated to fire suppression.
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Page 3 Icng tem confinement of radioactivity is achieved by the PCRV, and any leakage which might occur by diffusion through the PCRV cmcrete wall is collected, filtered and released by the reactor plant ventilation exhaust systam.
Spent fuel cooling is achieved as during -1 operation by the reactor plant coolim water systam.
1he mly required support function is an alternate electrical power supply which is provided by a dadir-ated ACM dimaal generator and distributicn system. Transfer switches are provided to disemnoct the -1 power supply frt:m, and connect the ACM power supply to, individual cereponents. With a few eixceptions, transfer switches and valves are repositioned manually.
2.2 SYSTDi OPERATICNAL RDCINESS All of the reactor a=414=7 systems addressed in secticn 2.1 above are used during mmal plant operation. System operatimal readiness is assured by this mode of operation and by testing of individual ocgenents as discussed below.
2.3 SURVEILIANCE CF PQ1PS AND FANS l
Iong tem operaticn of certain pues and fans is required follow 2ng a postulated loss of forced circulaticn ecoling
-4 dant, in order to cool the reactor and to cmfine radio-activity, th 2,y protecting the health and safety of the pihlit.
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1he critical ptsaps and fans required for long term w ation are:
- the circulating water makeup pmps,
- the service water pmps,
- the reactor plant cooling water punps, and l
- the reactor plant ventilation exhaust fans.
The purification cooling water pwps require short tem operability for depressurization of the primary coolant system, during the initial phase of ~4 dant mitigation.
The current survM11ance requirements and imlementing surveillance rcradwes were reviewed to verify if they provide for:
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- a monthly functional test that demonstrates west ava41nh414ty;
- an annual verification of mt perfomance "= Phi 14 ty -
(flow and hood):
- an annual verification of mar +umie=1 condition (vibration amplitude and k== ring 1-pc&ture) for those ccanpanents requiring long tem operability; and
- a monthly functional test and an annual calibration of the associated instrunents and centrols.
Since the <=prvients under consideration routinely operate for um==1 plant operation, ':he tests and testing im=nr4== outlined above are censicerad sufficient to demcnstrate continued
,vn,ah414ty of these tvunprwients When Called upm to pairdoEm their required safety functions under merid=rit conditions resulting fran unav=41=h414ty of forced e4vr'i1= tion cooling.
Serv =411anca requiranents for the circulating water==1r=_m pap are currently included in Tar +mical Specificaticn SR 5.2.24.
+
1here are no existing surv=411=ne= requirunants for the other ptmps. Based m this review, m-= i changes to SR 5.2.24 have been identified as follows:
'1he scope was changed to include all of the reactor auxiliary cooling water systams.
'Ihe existing functicnal test interval for the circulating water makeup pings was e+ume=d frtan weekly to monthly. Past test
==="4 anr*, in ocnjunction with normal coeraticn, indicates that a mcnthly functional test is sufficient to derrrnstrate l'
pmp av=41=h414 ty
- A requirement fcr a monthly functir. 31 test of the other peps was added, includi:xy the===e4=ted instrunents and centrols.
1hese instrunents and controls will also be r-alibrated annually.
- Boquirunants for an annual performance test of the =mr414="I cooling water pimps were addad. 'Ihese include a verificatim of performance r-apahility for all the pups (head and flow) and an a=aa=amarit of mectumie=1 candition for all except the purificaticn cooling water punps (vibratim and bearing tangerature).
Surveillance.%wts for the reactor plant exhaust systen are currently included in Techrdcal Specification SR 5.5.3.
'Ihe exhaust fans and==amiated instrunents and controls are functionally tested weekly in cr:njunction with existing exhaust filter pressure drop testPig. Two pwW changes to SR 5.5.3 have been identified by Sis review:
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- Requirements for an annual perfoz: nance test of the exhaust fans were added. 'Ibese include a verificatim of perfor: nance r==h414ty (capacity and total pressure) and an assessnant of *ical condition (vibration). Due to fan design, bearings are in=-> = ikle for tanperature measurenent durig operation.
- A requirenant for an annual calibration of the instrunents associated with the exhaust fans and filters was added.
Operational flMk414ty was provided for the annual tests by allowing centinued plant opezation until the next scheduled plant shutzbwn, when one year has elapsed frcm the previous tast.
l 2.4 SURVEIIIANCE T VALVES Actuatim of the n mode of coeration requires operation of manual isolatim valves and manual operation of m11y power operated valves. Only the notor operated haliun purifie= tion system isolaticn valves (HV-2301/2302) can be l
powered by the n electrical supply; however, these valves can also be manually operated using a== r4=1 extensicn red.
Tarimical W4 #ie=*4m SR 5.2.21 currently requires a semi-annual test of the manual actuaticn r==h414ty of sneuna*4cally and electrically operated valves. However, bara - these valves cannot be fully stroked during plant operation, only partial stroke tests are perfr - d.
Test experience to data shows that an annual test interval is sufficient to exercise manually operated valves, Wtile the igu. e of the safety function indicates that testing should demonstrate that each valve is capable of noving to the required position. In most cases this requires the valve to be fully A M. 'Iherefore, SR 5.2.21 has been changed to require that valve testing be acccnolished annually or at the next scheduled shutdown so that valve testing can fully denonstrate their operability.
In addition, a review of systen operating procedure SCP-48-01 identified a number of manual isolation valves that also aust be positicmed to actuate the n mode of operation. Since these valves are not normally exercised during operation, SR 5.2.21 was changed to require that manual valves also be tested l
at the same frequency.
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Attachnent 4 Ps.ge 6 2.5 SURVEIIDNCE CE' INSnUENfS Only local instrunants are available for monitoring syste operations under AC4 ccmditions. Irrportant functions to be monitored include the progress of PCRV depr===M2ation, cooling for adequate operatim of the hallun ptr4#4 cation syste, r =v-Aicm of boiling in the PCRV liner cooling water syst e,
and M4=-tivity of gaseous effluents.
T % eture instruments are supplied with n electrical power to monitor heliuta pe4f4<-= tion dryer inlet t=um ture and e
pcwide information about the adequacy of cooling for r -: =4ng of reactor coolant by the purificaticn systs. Pressure instrunents in the helium purificaticn syste p=rhn line allow monitoring of the progress of PCRV depressurization and also :: Emir.or PCRV pressure subsequently. Each PCRV liner cooling outlet subheader is==4=-4 with a local s-.ture indicator to indicate the Wilal need to increase syste pressure to prevent M414ng; that pressura is monitored at the surge tanks where the cover gas l
pressure can be adjusted as required. A radiation :ronitor (PING-1) l is prwided in the plant exhaust stack and sucolied with Act electrical power.
l Surveillance of the PCRV liner cooling water outlet subheader l
t s--
ture ir wits is adequately v4fied in Technical W fir-=ticm SR 5.4.4.
Surv=411=nce of the radiation numitars is currently specified in Technical %=r474 rations SR 5.4.9 and SR 5.8.1.
Tb assure that the indicators can accurately disolay the required information, Tachnical i=r 4*icaticn SR 5 2 21 has been changed to require that the other instrunents ida*4*ied atsve be e=14Mted at each plant shutdown for refueling, since this plant conditicm is required for =-- - ='h414ty to scme of these instruments.
2.6 SURVEIIDNCE CF C2FINEMENT Performance of helium purificaticn syst e @ s ts required to depr===n-49e the PCRV is normally monitored during plant operaticn by the analytical instrtsnentaticn systs. Performarre l
of the reactor plant ventilation syste exhaust filters is verified as. W by Technical Specificaticn SR 5.5.3.
1 No additicmal survaillance requmanents are needed to provide assurance that caseous effluents will be adequately treated prior to release.
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3.
IDSS CF NOR91 PORC2D CIRCtXATICM CXEDG ACCIDmr 3.1 OYSE!M SAFE 3Y PINCTICNS I
2 tis amidarit is chare *=d by the postulated failure of all class II plant =T'4; d following an envisu-a tal disetrhance such as an earthT'=le.
Core reactivity is centrolled by reactor scran, and the reserve shutdown system is available as a backup. 9m integrity of the rear: tor coolant pressure boundary is unaffected and the PCRV runnins pr==@=d.
Cooling is perfozzned using firewater i
in a ence +hre=v'h mnfiguraticn to drive at least ena circulator and to supply the um +. ling staan generator. Water is supplied frczn the circulating water stcrage pcmds to the firewater puup pits by the circulating water nakeup system. Diis systen also supplies water for mee through cooling of service water users idnich have to operata.
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l Ccnfinenent of.hvity is assured by the primary and h7 reactor coolant b W es. Die reactor bi41 ding con-finenent is not relied upon since tlm ventilation exhaust systan is Class II and thus not available.
Required swii.ng functions include operaticn of the m=ary halim circulator =~414=*4== (bearing water: supply to the circulator, makeum, and emergency supp.ly by the==1= tor systan; water turbina drive: supply of boosted firewater, and water removal). Also, energency electrical power is supplied by the i. 27y dia=m1 generators, which require operation of the
==arv-4= tad ccoling system and fuel oil transfer systen. Cartain valve positicming requires cperation of the==-v-4ated actuator (hydraulic, pnetznatic, electric notor). Other valve positioning is done manually, in particular for the firewater supply, and for isolation of critical Class I/ Class II bourviari==.
l 1
1 3.2 SYSTEM OPERATICNAL READINESS Sana of the reacter atwiliary systens addressed in section 3.1 above are used during nozznal plant cperation. For these, systen wot.ional raadina== is assured by this node of opention and by testing of individual etmponents as d4-aad below.
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.-.n A++=4=mt 4 Page 8 For those systes, or particus of systems, whose OF-Tdion is not required during remnal plant coeration, the plant Tachnic=1 an=e4 Fir-=ticms include various requirements which demcmstrata system operational r==d4=== as A--er4h=4 below.
T=chnical Saarifie= tion SR 5.2.7a requires that one circulator in each loop and the a=eated water supply valving be func*imally tasted anr==11y by operaticn en water turbine drive using feed-water, ocmdensata, and boosted mndensate (where cmdensata at reduced pressure is supplied to the amaroency water booster pumps to einnlate firewater). Min 4== performance r==h41ity i
of the hallun circulators is verified f.or each water supply.
Since the above tests require that La reactor be shui.Jc n, SR 5.2.7a has been nrwi4 74-9 to provide operaticmal f1=vih414ty by =11-4ng ccmtinued plant operation until the next scheduled plant shutdom, when cza year has =1=na-4 frcm the previous test.
T=chnical apacif4r-= tim SR 5.2.9 requires that the helium circulator t
h==*4ncr water accumulators and===v 4=ted instrumentaticn and controls be func*4==11y tasted monthly and, as applicable, a=14hrated annually. 'Iha =na-4 74 =ri test %=*=1y demonstrata the cgerational readiness of the = m =n1= tor syst e.
A review of test results indicates that a quarterly test interval would provide sufficient assurance of syst e operability. 'Iberefore,
SR 5.2.9 has been W _ accordingly.
J 3.3 SURVEILUNCE T PtNPS AND CINPRESSORS I
l CDen loss of nozmal means of forced circulaticm, safe shutdown cooling of the reactor is achieved by using firewater to supply a staan genera::or, and boosted firewater to drive a heliun circulater water *nchina.
'Ihis rrode of operation requires operaticn of the following putps:
-(*)the circulating water makeup pumps to supoly a source of firewater and service water; the firewater motor and engina driven purps to supply the varicus plant users (steam generators, circulators, bearing water==k=?);
the amargency booster purps for circulator water drive; the turbine water renoval pumps for circulator water drive; the h== ring water purps for circulator operation; the normal and energency beanng water makeup purps for circulator operation;
-(*)the service water putps for cooling of essential ccuperants;
-(*)the reactor plant cooling water pungs for PCRV and spent fuel cooling; the hydraulic oil purips; and the instrument air cmpressors.
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Attachrumt 4 Pace 9 Some of these ptmps (marked by an aste. risk *) have been reviewed in Section 2.3 above and are not addressed further. Die currunt surv=411=nca requirments and inclementing Arvaill=nea yu-bres were reviewed to verify if they provide for:
- a quarterly functicmal test that darnnnstratas ptmp av=41=h41ity;
- an annual varification of performance r==hi14ty (flow and head) for those ptmps that do not na==11y operate; and
- a queu.wly functional test and an annual. calibration of the maanciated instrments and controls.
Vari *4 e= tion of ptmo==rh=nir=1-condition is not considered necessary, dus to the short term operability requiranents following the =rv-idartt.
Sie above test requiremerts are considered to be sufficiarit to demcmstrate adequate ptmp operability when callad uoen for the performance of their required safety functions under amid-nt ocnditions resulting fra loss of==1 forced cirm1=+ 4m cooling e=a=hi14ty.
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S n:v=411=ne= requirunants for the firewater pungs are currently included in T=chn4"=1 t="4 F4caticn SR 5.2.10.
SR 5.2.10a.1 I,
prwides for a monthly functional test of the pimps and associated instrunantatiat, and for an annual irs.st calibration.
i SR 5.2.10b.4b provides for a var 4*4 ration of performance c=a=hi14ty at each refueling. Diese requirements are considered sufficient to denrmstrata ptzup apar=h414ty. However, the specified ministat
'ormance ("at least 1500 gcm at a system head of 290 feet")
is based cn the design conditicus for the pumps, and experience indicates that mane d.c..:stias of performance in service should be =11 W. Sie min 4== performance requirement for the firewater ptmps is detaminmi by their safe shutdown cooling function, when they have to supply 1100 gpn Marived fran data in FSAR Section 10.3.9) at a d4=rharge pressure of 128 psig required to achieve =4n4== oer-formance of the haliun circulator (darived frcm test experience with SR 5.2.7).
Siis mini== performance reauirement for safe shutdown cooling is equivalent to a 5 percent degradation of ne=4nal ptmp l
performance (frca 1500 gpn at 125 psig to 1425 gpn at 119 psig). Such degradation of ptmp performance in service is acceptable under Secticn XI of the ASME Code which allows for a 6 percent degradation before entering the required alert. range, and a 10 percent degradation before a pung is declared inoperable.
Die requirements of SR 5.2.23 provide for all the required *m for the emergency bocater pungs. Ptmp performance c=r=hi14 ty is verified by achieving min 4== helitri circulator speed, as recuired per SR 5.2.7a.
21erefore, no changes to SR 5.2.23 are considered necessary.
SR 5.2.7c/d provides for a functional test of the turbine water reroval punos vid associated instruments, and for instrument calibration. e4 water removal when operating the circulators d
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cm water drive per SR 5.2.7a provides a demonstration of adequate ping performance Th414ty. Dierefore, no changes to SR 5.2.7c/d
. are considered necessary.
SR 5.2.8 provides for a ftuicticmal test of the normal and emergency i
l bearing water makeup ptmps and==~42 tad instrunants, and for instrunant calibration. Die rmmal==%y pung surv=411=nce test g - - S e also provides for v=r4 74e= tion of the ptmp perinmance c== h414ty. Performance Thi14ty of the amargency makeup ptmp is verified idian testing the associated discharge line check valve as required per SR 5.3.4.
Dierefore, no changes to SR 5.2.8 are considered necessary for these two pumps.
1here are no current surveillance requirements for the bearing water ptmps. De of the three ptmps in each loco are normally-l cperating.
n==% water pung W is a sensitive operation which may involve transient fluctuation of system pa ws, potentially resulting in circulator trip. In order not to disrupt normal plant operation, SR 5.2.8 has ben changed to provide for a functional test of the pungs and===nc4ated instrunents and controls at each scheduled plant shutdomn and for instrtunant calibration.
annually, or at the next scheduled plant shutdown if cme year has elapsed fran the previous test.
i Instrunant functional testing and calibraticn specified in SR 5.3.5 for the hydraulic oil systen result in a quarterly functicmal test and v=Hf4e=+4rn of performance e==h414ty for the l
hydraulic oil pumps. 1herefore, no changes to SR 5.3.5 are ccmaidered necessary for these ptmps.
i operabiHty of the normal instrunant air ocapressors is demonstrated by continuous operation. Instrunant functicnal testing and calibrations
=="4 74=d in SR 5.3.6 for the instrunant air system result in a quarterlf functional test of the standby air ocupressor. 1herefore, no changes to SR 5.3.6 are considered maary for the instrunant air ocaprospors.
3.4 SURVEIIIANCE CF VALVES Surv illance requirements for the Safe Shutdown Cm ling valves, specified by Technical Specification SR 5.3.4, were previously reviewed, and mndified per License Anandment No. 33. 1 hey l,
currently address power operated valves and check valves.
Die review of reactor aw414=vy systans indicates that there are also valves which nust be manually positioned for actuation of the Safe Shutdown Cooling node of operaticn. 21is requires operaticn of certain manual isolaticn valves, and manual coeration of certain normally power operated valves (these whose controls are Class II and are postulated to have failed). In carticular, firewater sucply to the energency headers is achieved manually.
Also, certain Class I/ Class II boundary valves recuire manual isolaticn (since Class II portions of systems are postulated to have failed) in order to allow procer performance of systen safety hJ.cns.
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Attachnent 4 Page 11 Since these valves are not normally exarcinad during operation, Wir=1 P 4fication SR 5.3.4 was changed to include a requirement for testing the operability of those valves that nust i
be manually positioned for actuation of the Safe Shutdown Cooling mode of operation.
3.5 SURVEIWQG T INSTIGENTS Att instrunantation required to monitor and control the Safe Shutdown Cooling mode of operation is em=11y used during reactor operation, and any nalfuncticm wo.dd be rapidly brought to the operator's attention. Surveillarx:e requirunants are specified in section 5.3.10 and 5.4 of the Technical 3=rificaticms for instrunantatial not nomally in use.
Surveillance requinsnants for punp, compressor and valve instrunantation and otatrols are also g=c4 fled in the - ; - 44=y T=ehnie=1 W F4citions, as w e iate.
Tm@nical==e4 74-=*4mm SR 5.3.5 and SR 5.3.6 provide for monthly testing and annual e=14hraticm of m-lator pressure instrunants in the hydraulic oil system and the instrunant air systen, A W vely.
na reviw did not identify any additional surv=411=nce requirements to be included in the Technical *;a-474catims for Safe Shutdown Caoling instrunanta*h.
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IDSS T RE2CIOR PRESSURIZATICN ACCIDENT 4.1 SYS'HM SAFM FUNC:'ICNS L
2e Mdarit ocnditicms postulate a double failure of the primary and "7 reactor coolant boundaries resulting in depres-surization of the PCRV. 'IWo cases are considered:
- the >wi== Credible Accident, for which the safety analysis takes into account filtration of the reactor coolant by the reacter plant ventilation exhaust filters prior to release to a %, and
~
- the Mig Basis kridarst, EBA2, dure unfiltered release is postulated following a rapid depr--*ition of the PCRV l
caused by the postulated failure of a large PCRV p Laticn.
i Ccattrol of reactivity is achieved by reactor scran.
I l
Safe reactor cooling requires operaticm of the helim circulators on water turbina drives osing high pressure fes&etter. Other plant systems essentially runnin in their normal operating ocnfiguration.
i In the case of DBA2, mfiltered plant rsleases may be limitad by operation of the reactor bi41 dim overpressure protection
~j systen, which prevents failure of the reactor bs41d4% sidiaq and therefore allows the ocnfinsmant fm etion to be. h i
after the transient peak pressure has occured.
4.2 SYS'HM OPERATIGEL READINESS System operational readiness does not differ auch fran that discussed in Section 3.2 above. S e boiler f e m operate a:ntinuously during plant operation at power. Operaticnal readiness of the helita circulator water turbine drive is demonstrated as required by Technical W4fication SR 5.2.7.
~414=7 systen functicns are-dentmstrated as discussed previously.
Operaticnal readiness of the reactor bi41A4w cww.gessure pet,tectim features is adequately denemstrated by the testing required in Technical Specificaticn SR 5.5.2.
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As indicated above, these pays are used for plant coeration at power and any deficiencies in hydraulic perfomance or mechanical
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a:mdition will be noticed without delay. 1herefore, no surveillance reauirements are included in the Mie=1 *=cif4+
tims for these pays.
4.4 SURVEILGNCE & VALVl!S i
1he valves which require actuation for circulator water turbine drive are also used for safe shutdows moling and their surveillance requmanents are discussed in Sectim 3 above.
4.5 SURVEIT.IANCE & INSTRLMDF15 Instrunants are either used during norial plant operation, or are required for safe shutdown cooling as discussed in Section 3 above.
1he. mly additimal instrunents involved relate to reactor M41d45 overpresaire protection and adequate surveillance requirunants l
are specified in SR 5.5.1.
l
Paos 14 5.
I4SS CF SPDir PEL STORME WELL WATER COOLING ACCIDCFF 5.1 SYSIEM SAFETY ftNCTICNS
. 1ha arvidarte conditions postulata the total loss of water cooling to cna spent fuel storage well in cna of the three fuel eLu.-ge vaults. If this were to occur, adequata cooling would be achieved by an amargency ventilation system to cool the well by convecticri art the external surface.
5.2 SYS15( OPGATIGEL READINESS 1hc emergency ventilation system does not m-11y wei.e, and there are currently no surv=411m== requirunants. Technical
- =-4 4'4r-= tion SR 5.7.2 was W= to require an annual functional test of the systen, including operation of the boostar fan and actuatial of the ventilation dangers for endt fuel storage vault. Since this test will include una various ccaponents in the systen, no additional surveillance is required. 1here are no instruments which require <-alibration.
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Page 15 6.
AUXILIARY SYSTD4 INTEGRITY 6.1 REACIOR CIXIANT CDWIAINMENT l
6.1.1 HELIIM PURIFICATICN SYSTEM ammetor coolant circulates thew
- the helium purificaticm system during rx:rmal plant operatim where it is g-@ ssively purified.
Two identical pMfie= tion trains are installed. The high tangerature filter adsorbers, at the front and of the trains, are located '.n PCRV -g Lations. Surv=411ance regairenants for the PCRV p i L tions were addressed in the first phase of this review and were eyed by License Amancknant No. 33.
Primary coolant isolation valves are installed between the adsorbers and the helim purification molars. Each of these valves is located in a PCRV well to:,.0-r with the cooler.
1he well is designed to act as h4 ry con *=4==nt in the event leaks develop in the piping or cooler and nnrmally operates at a W e ic pressure. Other helim purification ecp14pn=nt em +=4ning sian4F4rane amounts of activity are also housed in similar PCRV walls. These other wells are maintained at a high vacuum to provide tbarmal 4n-i1=*4m.
a) Pressure boundaries 1he PCRV well design is essentially the same as the other PCRV top head p.eLatims (refueling and filter adsorbers). As was the case for the por Latim *="1 cm*=4-t boundary, t.%re are no pressure re*=4ning ccuponents that would require exaninaticri.
1here are no ciretzuferential welds located at structural discen-tirn.tities, no integral support attactznant welds and no bolti.w over 2 inches in diameter. The bolted well closure plates are also secured in. place by the PCRV holddown plates.
1he purificaticn cooler well is also similar to the PCRV safety valve pe Lation, where non-isolable primary coolant piping is housed in a==rm4=nf con *=4==nt tank. In that case, it was cmcluded that cmtinuous leakage monitori.w of the primary coolant boundary and periodic leakage testing of the secondarf boundary provided adequate verificaticxt of structural integrity.
1he eccler wells are provided with instrumentaticn which continuously monitors pressure in the wells. Any pressure bi41 dun due to piping or equipment leakage is =1=M in the centrol recm. The well itself is not subject to any loadi.g
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Attachnent 4 Page 16 corx11tions whidi could ocmceivably '+.g h the shall or closure plata. 14 tightness of the gasketed secondary closure is not currently checked on a Mrvi4c basis.
Since theos instrunants are used to nonitor the integrity of reactor melant ocntaiment, surv=411ance requis d.s have been established to verify their ===_ h414ty. Technical WF4eation SR 5.2.16 has been W to require that the cooler well pressure nonitoring instruments be calibrated once during each refueling cycle. Access to the well may be required, and this could be dona during a rrmnal refueling outage. In addition, a leak test of the well closure has been specified during the same interval to verify the integrity of the well closure.
2a other walls are operated at near full vacum ocnditions'and loss of vacum is alamari in the cx:ntrol roca Bis provides i
both leakage nonitoring of the equipnarst in the walls avi also cr=*4===1y variF4== the leak ti@L-ss of the well closures.
Piping or==4--t leaks in these walls is also isolable, either
.-w --
=11y, or autcmatically in the case of large leaks.
No additional surv=411ance is considered m aary for'the other 7"If4r'atiCm systER wells or liieb. wits.
b) anactor Coolant Boundary Isolation 2e primary coolant isolatias valves upstrean of the cooler can be operated autanariem1ly, runotely fran the centrol roan,or locally using a reach rod through.the well closure. % = h414ty of the valves remotely is demcmstrated by routina operaticn when l
the trains are switched for regeneraticm. In additial, SR 5.2.21 requires that each valve be exercised once a year, either using ACM power or 1mc=11y with the reach rod.
Although seat leakage is not critical, adequate seat ti@L-ss is also demonstrated rm*4=1-y, whenever a train is isolated and depr===whed for regeneration.
Instrunentation is prtvided to autcmatically isolate a train if leakage is detected in the purificaticn cooler heat exchanger or if large leaks were to develop in the downstream piping or equianent. A seccmd isolaticB valve, outside the PCRV, also l
c1m=== autcmatically for large leaks and provides double valve isolaticn if the leakage is downstream of the seccnd valve.
2ere are no current survaillance requirenants for this instru-mentation. SR 5.2.16 has been changed to require that the instrunents used for autanatic isolaticn be funcF m=11y tested 4
L and calibrated once during each refueling cycle. 21s tasting is adequate to verify the coerability of the aurcznatic isolation feature.
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Attactsnent 4 Page 17 6.1.2 RDCTOR PIAlff AIC PURIFICATICN 000LHC WATER SYS'mtS d) PCRV 'Ihermal Barrier Itmitoring
'Ihm PCRV thermal ber4=e and tha liner cooling system operate
'ww.0.er to limit and renove the heat load across the PCRV liner.
'Ihermal barrier integrity is demonstrated by continuous nonitoring of its thennal p-- f
---we using taperature and flow instrunants in the liner cooling system. In the first phase of this review, approved by License Amendment No. 33, it was concluded that provisions to test and r=14hente these instruments would assure that sufficiant information is available to assess thermal barrier integrity.
'Ihm water taperature at the outlet of individual liner cooling tuben is scanned during m==1 plant Op miicn and ala==d if high taperatures are experienced. 'Ibn water r-s- ei.ure at the main supply headers (tube inlet taperature) is scanned as well.
Flaw in individual liner cooling system subheadsrs is also scanned and a1===d during normal plant operaticn.
e=hin4= flow and temperature rise data provides M - --iicn about the heat load and, therefore, about the integrity of the tbamal harr4=r.
Baview of plant experience with surveillarce of the PCRV liner cooling system tauperature instrunants indicates that emparison of w sentative em - =gle outputs to their rWdve subheader tem-parature is inpractical and provides little assurance of instrunant SR 5.4.4 was therefors nodified to verify scanner reading accuracy.
accuracy on a nonthly basis by mei=m of inlet header weter
^w.i. ace scanner readings to the wus--g-:- Ming Mt temperature indicators, and to require an annual scanner catihration.
'Ihe inlet header and the outlet header tenperature indicators are also used to verify ocupliance with the requirements of ICO 4.2.15 relative to PCRV cooling water system h=Taratures. A new requirement was included in SR 5.4.4 for an annual calibration of these instrunants.
Lack 'of direct accessibility to certain th==rmales during operation makes it inpractical to e=14hrate on a yearly basis the percentage of instrunents currently specified in SR 5.4.4.
Because all instrunents are of =4=41ar design, and because operating cxmdi-ticms are not severe, it is not anticipated, as errd4==d by experience to date, that any substantial differences in instrunent behavior will occur over time between the various surbaadars.
SR 5.4.4 was therefore cl w- @ to provide more flexibility in the selection of thw-sgles to be calibrated in any particular year, while specifying a five year calibraticn cycle for all the therne-couples which provide input to the scanner. Currently, the therne-couples are calibrated on a six year cycle. 'Ihe remaining requirenants cf SR 5.4.4 (monthly scanner tenperature alarm check and annual calibration of subbandar outlet tarperature indicators) were found to be adecuate.
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Page 18 Technical Ap="4fication SR 5.4.5 currently includes requirements for a matthly reakut of flow, a monthly check of the alama, and an annual em14hraticn of the flow scanner and alama. It also requires an annual calibration for six of the thirty-six subheader flow transnitters. Since the reactor mry have to be shut domi (to ctzply with Ico 4.2.14) to renove and replacia the flow meters, SR 5.4.5 has been changed to allow the calibration interval to extend to the next scheduled shutdown.
Diese existing surveillance requirements, as modified above, are sufficient to assure that heat loads can be accurately determined and that abremal flow and tamperature conditions I
are alamad.
I b) Reactor Coolant Boundary Isolation Instrumentation and controls are provided to detect leakage and l
aub wdcally isolate parts of the line: cooling system or of the p_M fie= tion cooling water system, in case a leak occurs in i
j the primary 6=4=7 between the reactor coolant and these systens.
Die PCRV lher em145 tubes and subheaders, up to the isolation j
valves, have been designed for reactor coolant pressure and *=npara-ture conditions so that they can assure ccnfinement, should a leak i
occur at a v4fic location through both the PCRV liner and the cooling tube wall. Ingress of primary coolant in the moling water system during normal operation is detected by pressure switches in the affected loop header which alarm high pressure in the castrol rocan and autcmatically cicae all the loop subheader isolation I
valves.
Die purification ecoling water system is also designed to ef4=
reactor coolant, should a leak occur in one of the haliun puri-ficaticzi coolers. Ingress of primary coolant in the cooling water system is detected by pressure switches in the affected subheader, which alarm high pressure in the control roczu, autanatically isolate the subheader, and sutcmatically isolate the purification train.
1here are currently no surveillance requirements for these matrunents and cxxitrols. Since these instrunents mcnitor the integrity of portions of the reactor coolant boundary and provide autcmatic isolatias functions if leakage develops, surveillance requirements have been established to verify their operability. SR 5.2.24 has been changed to require that the above instruments and valves used for autcantic isolation be func*4 mally tested and calibrated once during each refueling cycle. 1hese tests and calibrations provide the nac====7 assurance that the required safety functions can be aemmlished.
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i Attacinent 4 Page 19 6.2 CIRCUEATDG MATER MMGUP SIGWE PCROS 1he circulatit4 water==%; storage ponds provide the ultimate source of czaoling water used for safe shutdown cooling, and Yor service water makeup in case of a permanent loss of forced circu-lation ace-4 dant.
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pcmds are shallow reservoirs, installed at a location of a natural dip. 1herefore, the water level is essentially at natural grade level, except in the su.u. wt part of the west pand where the natural grade level is only about two feet above the bottan. Cet fill embankments are provided wherever
.i nac====7 These embankments are designed with a 1:3 external slope and a 1:4 internal slooe. She internal slope and bottcm are lined with a 3 ft. thick clay layer. 1he top of the embank-ment is lined with concreta. 1he height of the erbankment is cnly about 5 ft. above grade level, except en the west and north sides of the west pcmd, where the height gr=dn=11y increases to about 18 feet. Settling basins provide an additicmal embankment that separates the pcmds fran grade level, en the north side.
Regulatory Guide 1.127 (Rev.1, March 1978) - Inspecticm of wetar structures =~4= tad with nuclear power plants
_%,ns that inservice ir-p Lion be perfnrmed, at periodic intervals, to check the condition of these structures and evaluate their structural safety and operaticmal adequacy. 1he cmsite inspection program addresses the ccmcreta structures, the embankment structures, the spillway structures and outlet works, the reser-i voirs, the intake and diachame structurcs, the safety and
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perfnrmance ir Letation, as well as operaticm and maintenance features, and post - Lucticm changes.
I 1he shu. gs pcmds have only minor concrete structures. Survey markers (brass caps) have been installed to allow a datazmination of signi74c=nt changes in aligreent or settlement of the embank-monts. Technical 9w-ification SR 5.2.24 has been modified to i
l require that alignment and settlement of the enbankments be verified at five calendar year intervals, and that the enbankments and water structures (i.e. spillways and outflows) be examined at the same intarval for a b rmal erosion, cracks, seepage, leakage, =er-1= tion of silt or debris (as applicable) which
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might indicate a datarloration of stnx:tural safety or operational adequacy of the storage ponds. Since no noticeable changes have occured frta the time the storage ponds were built ever l
eight years ago, the Sw-ifled inspection frecuency is consistent with the requirenents of Regulatory Guide 1.127.
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-_ em Page 20 6.3 SYSTEM STRUCIURAL IhTEGRITY As indicated previously, most of the Fort St. Vrain a=414 uy systems are rm*inaly used for normal plant operaticn. Process monitoring, as well as unlimited access to most parts of the plant at all times, provide for early d t iicn of incipient equignent failure and s@equent orderly pL1 tnt ah'+drm if required.
Cerrosion probes have been installed in cooling water syst es that could be subject to corrosion where chemistry centrol is more difficult. Mees probes are par 4M4r-=11y monitored as part of the plant water ciemistry control sws.u and there have been no indications of aw-aaaive corrosion rates. Experience to date indicates that no simi#4 cant deterioration has occurred in any 1
of these systans. m erefore, the above methods are = =4darsi adequate to assure the cmtinued structural integrity of the reactor =~414='f systes.
To svwt the systen kr=d=*ies frcm overload ccmditions, safety valves and hydraulic snubbers are installed in the =~414=nf syst es. Boy provide overpressure psei Ricn and r t. tion e
against -=4ve motion, resp i.ively.
Surveillance requirements for hydraulic snubbers are soecified in Technical Apar-4 74 cation SR 5.3.8.
2ese recuirements have been recently reviewed and updated by separate wus- =;- V=ce, and are pending approval.
Surv=411mm-a requirements for the PCRV =~414atf systen safety valves are included in Technical 9=r-4*4 cation SR 5.2.1 and were updated by License Jaarzknent No. 33. SR 5.2.7 included requirements for testing safety valves in the circulator water
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drive supply lines. SR 5.2.7 was changed to provide operational fleh414ty by allowing czmtinued plant operation until the next scheduled plant shutdown if one year has =1=W frczn the previous test. ' here are currently no surveillance requiremarr s for other i
=~414="f systen safety valves.
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Safety valves in these systens are generally small in size and not subject to high 'W&ture or other harsh envircratental l
canditions. Berefore, the set points are epid to renain l
stable over langer pa+4a of time. Only those safety valves l
which protect portions of systens that nave safety functions for the accident conditions discussed prev 2.cusly are con =4r'aved incertant to safety. Technical Specification SR 5.3.9 has been civinged to j
require that all Class I safety valves, not covered by other SR's be tested to verify their setpoints crice every ten years. Tests will be scheduled throughcut the interval so that ra14=h414ty is l
samoled more frequently. Se specified testina is censistent with the latest ASME requirenents for Code Class 2 and 3 safety valves per ANSI standard ANSI /ASME OM-1-1981 and will provide l
adequate assurance that safety valve operahi'2 4ty is not degraded.
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