ML20085J694
| ML20085J694 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 09/28/1983 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20085J678 | List: |
| References | |
| NUDOCS 8310130284 | |
| Download: ML20085J694 (45) | |
Text
1 9
8 a g ATTACHMENT I PROPOSED REVISIONS TO SR 5.2.6 DO$ko$$$off, PDR
i Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-1 5.2 PRIMARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS Applicability Applies to the surveillance of the primary (helium) reactor coolant system, excluding the steam generators.
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 i
ensure the capability to cool the core under all modes of operation.
Specification 'SR 5.2.1 - PCRV and PCRV P'enetration 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 tightness shall be determined following tank closure.
SR 5.2.1.a shall be implemented per ISI Criterion C.
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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.
c) The instrumentation and controls associated with the overpressure protection assemblies in a) and b) above shall be tested and calibrated as follows:
(
- 1) The pressure switch and alarm for each interspace between a rupture disc and the corresponding e
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 the PCRV overpressure protection system shut off valves shall be functionally tested and calibrated when testing either of the PCRV' overpressure i
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Page 5.2-3 protection assemblies.
The pressure switch and 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 implemented per ISI Criterion D.
i Basis for Specification SR 5.2.1 Testing of a PCRV overpressure protection assembly can only be performed when closing the corresponding manual 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.
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Page 5.2-4 The rupture disc is designed to be removed from the system for bench testing. Verification is made of the correct 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 ext.aination 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 to demonstrate 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.
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page 5.2-5 The intervals specified for testing the overpressure protection assemblies are adequate to demonstrate the operability of the overpressure protection systems.
The intervals specified for testing the associated instrumentation and controls are adequate to assure reliability 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.
Fort St. Vrain #1 lechnical Specifications Amendment #
Page 5.2-6 b) A sample of the atmosphere contained in a representative number of tendon tubes (te-don tubes without load cells and tendon tubes with load cells 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% 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 corrosion protection provided for the stressing tendons.
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Page 5.2-7 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.
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 shall 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 pressures.
Since the relation between
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-8 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 Monitoring of the tendon loads will assure 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 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.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-9 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.
b) PCRV deformations and deflections at vessel midheight I
and at the center of the top head shall be monitored at five calendar year intervals during a vessel pressurization to operating pressure.
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SR 5.2.4.b shall be implemented per ISI Criterion C.
c) The PCRV support structure shall be visually examined for evidence of structural deterioration at ten calendar year intervals.
SR 5'.2.4.c shall be implemented per ISI Criterion C.
Basis for Specification SR 5.2.4 l
l 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 l
l cracking expected is limited to superficial effects and is not considered detrimental to the structural integrity of l
the PCRV.
Reinforcing steel is provided to control crack l
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 l
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Page 5.2-10 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.
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.
I 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 j
following initial prestressing may be adjusted based on j
the analysis of prior results.
l Monitoring of overall PCRV deformations and deflections is f
the best indication of PCRV structural performance and i
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Page 5.2-11 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.
Specification SR 5.2.5 - Liner Specimen Surveillance Specimens shall be placed adjacent to the outside surface of the top head liner so that changes in notch toughness due to irradiation of the steel can be measured during the life of the reactor.
During the fifth refueling cycle, three sets of 12 specimens of the PCRV liner materials and weld material shall be removed and tested to obtain Charpy impact data.
The specimen holders shall contain dosimeters to provide integrated neutron flux measurements. Additional specimen removal and testing shall be conducted during every tenth refueling cycle thereafter.
Basis for Specification SR 5.2.5 A test program will be performed to survey and assess the shifts in NDTT of the PCRV liner materials.
The testing is to be accomplished by placing Charpy impact te;t specimens, made from the liner materials, near the liner
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Page 5.2-12 and exposing them to appropriate neutron fluxes and temperatures. The Charpy impact test specimens are to be removed, 36 at a time, during the life of the vessel and 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 complete impact transition cuive for the plate metal, the weld metal and the heat affected zone at each test interval.
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 or.ly 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
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-13 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 fiftn refueling cycle may be adjusted based on the analysis of prior results.
Specification SR 5.2.6 - Plateout Probe Surveillance One plateout probe shall be removed for evaluation l
coincident with the second, fourth, and sixth refueling, and at intervals not to exceed five refueling cycles I
thereafttr.
If, during the third or fifth refueling l
- cycle, or any refueling cycle following the sixth refueling, the primary coolant noble gas activity (gamma + beta) should increase by 25% over the average activity of the previous three months at the same reactor power level and the primary coolant activity is greater than 25% of design, the plateout probe shall be removed at the end of that refueling cycle. The probes shall be analyzed for Sr inventory in the reactor circuit.
The I
probes removed shall also be analyzed for 282I.
Basis for Specification SR 5.2.6 l
t The plateout probes are located in penetrations extending I
into steam generator shrouds and then into the gas stream of each coolant loop.
One sample is accumulated by continuously bypassing a small portion of the core outlet i
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Page 5.2-14 coolant stream through diffusion tubes and sorption beds located in the probe body.
Another sample can be accumulated by continuously bypassing a portion of the circulator outlet coolant stream through the probe.
The core outlet sample
~can be used to determine the concentrations of fission products in the coolant stream entering the steam generator; the circulator outlet sample provides 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 t
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
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removals.
The analysis for 282I shall be made
'o determine the degree of conservatism of the assumptions made regarding the circulating and plated out iodine in the primary coolant circuit.
i The interval for probe removal and analysis subsequent to l
the sixth refueling cycle may be adjusted based upon the analysis of prior results.
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Page 5.2-15 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),
annually.
b) Safety valves (V-21522, V-21523, V-21542, and V-21543), located in the water turbine supply lines, will be tested for relieving pressure annually.
c) Both turbine water removal pumps and the turbine water removal tank overflow to the reactor building sump shall be functionally tested every three months.
d) The instrumentation and controls associated with c) shall be functionally tested in conjunction with and at the same intervals as the turbine water removal pumps and shall be calibrated annually.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-16 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 Makeup Pump Surveillance The circulator bearing water makeup pumps and associated instruments and controls shall be tested as follows:
a) Normal Makeup Pump shall be operated in the recycle mode every three months.
b) Emergency Makeup Pump shall be functionally tested every three months.
c) The associated instruments and controls shall be functionally tested in conjunction with and at the intervals specified in parts a) and b) above, and calibrated annually.
Basis for Specification SR 5.2.8 During accident conditions described in FSAR Section 10.3.9, the circulator bearing water makeup pump is required to operate intermittently to make up bearing water. The specified testing interval is sufficient to er
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Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-17 ensure proper operation of the pumps and associated controls.
Specification SR 5.2.9 - Helium Circulator Bearing Water Accumulators Surveillance The helium circulator bearing water accumulators, instrumentation, and controls shall be functionally tested monthly and calibrated annually.
Basis for Specification SR 5.2.9 Helium 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 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 circulators in a loop.
The monthly test interval and i
annual calibration interval will assure proper operation of the accumulator controls if they should ever be called upon to function.
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Page 5.2-18 Specification SR 5.2.10 - Fire Water System / Fire l
Suppression Water System Surveillance 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 elect. 5 /te shall be tested quarterly for proper 2:.t f gravity.
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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 j
connections shall be verified to be clean, tight, l
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Page 5.2-19 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 veri fying that each valve (manual, power 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 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.
(b) Veri fying that each fire water pump develops at least 1,500 gpm at a system head of 290 feet.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-20 (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 water system pressure at greater than or equal to 125 psig.
- 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 testing interval is sufficient to ensure proper operation of the pumps and controls.
The motor driven pump routinely operates intermittently.
The operability of the fi re suppression water system ensures that adequate fire suppression and emergency safe shutdown cooling capability is available. The specified testing interval is sufficient to ensure proper operation of the system when required.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-21 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 LCO 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 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 5.2.11 The design of the instrumentation is such that under normal operating conditions the activity of the primary coolant is measured and indicated on a continuous basis.
The weekly sampling interval provides an adequate check on the continuous monitoring equipment.
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Page 5.2-22 Specification SR 5.2.12 - Primary Reactor Coolant Chemical S_urveillance.
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 LC0 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.
The specification of an interval for surveillance allows for routine mainter.ance of the chemical impur'.tv 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 i
sustained periods of time.
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Page 5.2-23 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 neasurements shall be made at five year intervals thereafter, i
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 the 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 i
The PCRV liner shall be examined for corrosion induced thinning, using ultrasonic inspection techniques at the end 'of the third and fifth years following initial power operation. Additional examinations shall be conducted at ten year intervals thereafter.
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Page 5.2-24 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.
l Specification SR 5.2.15 - PCRV Penetration Interspace Pressure Surveillance 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.
Basis for Specification SR 5.2.15 This calibration and test frequency is adequate to insure that the purified helium being supplied to the PCRV penetration interspaces shall be at a higher pressure than the primary coolant pressure within the PCRV.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-25 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 annually.
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 A
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page 5.2-26 shutdown if these valves have not been tested during the previous year.
SR 5.2.16.d shall be implemented per ISI Criterion B.
e) The check vilies on the HTFA purge lines shall be tested at five calendar year intervals.
SR 5.2.16.e shall be implemented per ISI Criterion B.
f) The check valves which are part of the HTFA or refueling penetrations shall only be tested when such a penetration is open for refueling or maintenance, if the check valves have not been tested in the last five years.
SR 5.2.16.f shall be implemented per ISI Criterion B.
Basis for Specification SR 5.2.16 The interval specified for determining the actual primary and secondary closure leakage is adequate to assure compliance with LCO 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.
Fort St. Vrain #1 Technical Spscifications Amendment #
Page 5.2-27 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.
Th,e interval specified for valve testing is adequate to assure proper valve operation when isolation of the closure _ auxiliary piping is required.
Specification SR 5.2.17 - Helium Circulator Pelton Wheels DELETE SPECIFICATION SR 5.2.17 IN ITS ENTIRETY l
Specification SR 5.2.18 - Helium Circulators Surveillance a) At the time of the first main turbine generator l
overhaul, one helium circulator unit shall be removed l
in its entirety from the PCRV and thoroughly inspected for signs of abnormal wear or component degradation.
- 1) Such inspection shall include examination of bearing surfaces, seal surfaces, brake system, buffer seal system, and labyrinth seals.
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Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-28
- 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 disassembly than required to inspect these wheels, shall be visually examined.
Results of these examinations shall be submitted to the NRC staff for review and shall be evaluated to determine 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.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-29 Once such a machine is running satisfactorily, little or no wear occurs to it.
Unlike most designs of emergency systems of conventional nuclear power plants, the components of the Safe Shutdown System of the Fort St.
Vrain plant are utilized and operated during normal operation of the plant.
This includes the helium circulators.
The performance of the helium circulators is monitored during operation, i.e.,
instruments are provided with the capabiitty to measure compressor differential pressure and flow, bearing temperature, bearing water temperature and flow, buffer helium flow, and shaft speed and vibration.
Examination at the time of the first turbine generator
- overhaul, and at approximately ten year intervals thereafter, is sufficient to monitor the condition of the 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 overhat The helium compressor and steam turbine blading should experience mininal wear in its running environment, and, l
with this length of service before inspection, will have undergone sufficient stress cycling to accurately indicate service life.
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Page 5'.2-30 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 a
shall be at least 100% of design ACM equipment full load.
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 j
available to supply electrical power under the highly degraded, loss of forced circulation situation.
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Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-31 Specification SR 5.2.21 - Hand Valve and Transfer Switch Surveillance Those pneumatically and electrically operated valves and electrical transfer switches that must be manually positioned to implement the ACM shall be tested twice annually at an interval between tests to be not less than four (4) months, nor greater than eight (8) months.
Basis'for Specification SR 5.2.21 In the event that the ACM must be implemented, it is necessary to position pneumatically and electrically operated valves manually and to reposition electrical transfer switches.
The test frequency and interval specified will assure operability in the event such operation is required.
Specification SR 5.2.22 - PGX Graphite Surveillance PGX graphite surveillance specimens shall be installed into five (5) bottom trancition 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.
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Technical Specifications Amendment #
Page 5.2-32 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.
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 subjected 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 #
Page 5.2-33 SR 5.2.22 PGX GRAPHITE SURVEILLANCE Table 1 TRANSITION ELEMENT ASSEMBLY WITHDRAWAL SCHEDULE l
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Number
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1 l
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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.
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Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-34 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.
Specification SR 5.2.24 - Circulating Water Makeup System Surveillance The circulating water makeup system shall be verified operable 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.
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Fort St. Vrain #1' Technical Specifications Amendment #
Page 5.2-35 b) The circulating water makeup pumps shall be functionally tested weekly.
The pump controls and instrumentation including the fire water pump pits shall be functionally tested monthly and calibrated annually.
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.
Basis for Specification SR 5.2.24 The circulating water makeup system is required to supply water for fire suppression and safe shutdown cooling. The specified testing interval is sufficient to ensure proper operation of the pumps and controls. The system routinely operates during normal plant operation.
Specification SR 5.2.25 - Core Support Block Surveillance The top surface of the core support block for fuel regions fitted with PGX graphite specimers 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.
Fort St. Vrain #1
-Technical Specifications Amendment #
Page 5.2-36 Basis for Specification SR 5.2.25 Visual examination of the core support blocks 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.
Specification SR 5.2.26 - Region Constraint Devices Surveillance The region constraint devices (RCD's) shall be inspected at each refueling outage using the fuel handling machine from those regions being refueled as follows:
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.
Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-37 c) Selected RCD's shall be visually examined by remote TV in the fuel handling machine after removal to veri fy 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.
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.
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Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-38 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 unacceptable deformations, loose or missing parts, or other visible defects.
Specification 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 performed during the previous year.
Fort St. Vrain #1 Technical Specifications AmendT.ent #
Page 5.2-39 SR 5.2.27 shall be implemented per ISI Criterion C.
Basis for Specification SR 5.2.27 The helium shutof' valves are self-actuated check valves which close when the corresponding circulators are shutdown or tripped.
Simultaneous 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 maltunction of a helium shutoff valve.
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Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-40 Specification SR 5.2.28 - PCRV Penetrations and Closures e
Surveillance i
a) Accessible portions of PCRV penetration pressure retaining welds shall be examined fo~r 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.
- 2) 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.
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Fort St. Vrain #1 Technical Specifications Amendment #
Page 5.2-41 b) Accessible -portions of the PCRV penetration closure and ' flow restrictor restraint components shall be examined for indications of defects as follows:
- 1) Visuall.y 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 support components shall be examined at 10 calendar year intervals for indications of defects as follows:
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Page 5.2-42
- 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 integrity of Fort St. Vrain PCRV penetratior, secondary pressure retaining boundaries is normally verified by continuous leakage monitoring and by periodic leakage testing of-the penetration interspace.
The specified examinations of accessible circumferential welds at structural discontinuities will provide additional assurance concerning the continued integrity of the secondary pressure 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.
Fort St. Vrain #1 0 0 Technical Specifications Amendment #
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O Q ATTACHMENT 2 SIGNIFICANT HAZARDS CONSIDERATIONS EVALUATIONS
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-t SIGNIFICANT HAZARDS CONSIDERATIONS ANALYSIS Since the proposed revision to Specification SR 5.2.6 has no operational impact and provisions remain in the specification to remove the plate-out probe for evaluation should the primary coolant noble gas activity increase by 25% over the average activity of the previous three months at the same reactor power level, it is apparent that no significant hazards considerations are involved.
Based on the above, operation of Fort St. Vrain in accordance with the proposed changes will not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.