ML20072B963

Proposed Mods to Tech Specs,Updating Inservice Insp & Test Programs

ML20072B963
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/22/1983
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20072B947	List: ML20072B946 ML20072B953 ML20072B963
References
NUDOCS 8303080101
Download: ML20072B963 (161)

Text

T e' ATTACHMENT 2 l PROPOSED MODIFICATIONS TO THE FORT ST. VRAIN TECHNICAL. SPECIFICATIONS 5

(SECTION 2) i i

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! M 2*o d R' S S d j h P

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Page 2-5 2.16 Refueling Shutdown The reactor is considerec shut down for refueling purposes when the reactor mode switch is locked in the " Fuel Loading" position simultaneous with either hot shutdown or the cold shutdown reactivity conditions.

2.17 Safe Shutdown Cooling Safe shutdown cooling refers to cooling of the core with Safe Shutdown Equipment providing for removal of core stored energy and for adequate sustained decay heat removal. The reactivity condition in the core is either J

hot or cold shutdown.

2.18 Surveillance Interval A surveillance interval is the interval of time between surveillance check, tests, or calibration. Unless otherwise stated, the surveillance interval can be adjusted by + 25% to accomodate normal operational l- schedules. Unless otherwise stated in these-specifications, surveillance may be terminated on those instruments or equipment not in normal use during reactor shutdown or refueling shutdown if the surveillance interval is one month or less.

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Page 2-6 2.19 Trip Trip is defined as the switching of an instrument or a device with two stable states from its normal state to its abnormal state. The result of a trip on a system level may be control rod scram, pressure relief, loop shutdown, etc..

2,20 Core Average Outlet Temparature The core average outlet temperature is defined as the arithmetic average of the individual refueling region outlet temperatures, 37 T avg = I Ti i=1 37 where Ti = individual refueling region outlet temperature.

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ATTACHMENT 3 PROPOSED MODIFICATIONS TO THE FORT ST. VRAIN TECHNICAL SPECIFICATIONS (SECTION 5 - INCLUDING IN-SERVICE INSPECTION CHANGES) 1

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Page 5.0-1 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.

l Implementation of the in-service inspection (ISI) surveillance l renuirements shall be per one of the following criteria, unless l otherwise indicated:

l~ ISI Crite ion A: The surveillance requirement shall be l implemented before 90 days have elapsed l following the formal approval date of l Amendment No. by the Nuclear Regulatory l Commission.

l ISI Criterion B: The surveillance requirement shall be l implemented before the beginning of fuel l cycle 4, provided that fuel cycle 4 does not l begin within 90 days from the formal approval l date of Amendment No. by the Nuclear l Regulatory Commission.

l Otherwise, the surveillance requirement shall

! be implemented before the end of the first l scheduled plant shutdown following 90 days l from the formal approval date of L >

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Page 5.0-2 l Amendment No. by the Nuclear Regulatory

[ Commission.

l ISI Criterion C: The surveillance requirement shall be l implemented before the beginning of fuel l cycle 5.

l ISI Criterion D: The surveillance requirement shall be l implemented in the existing schedule of l surveillance tests, following 90 days from the l formal approval date of Amendment No. by l the Nuclear Regulatory Commission.

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Page 5.1-1 5.1 REACTOR CODE AND REACTIVITY CONTROL - SURVEILLANCE REQUIREMENTS Applicabil<:y Applies to the surveillance of the reactor core and core reactivity control mechanisms.

Objective To ensure the capability to control the reactivity and temperature of the reactor core.

Specification SR 5.1.1 - Control Rod Drives Surveillance The surveillance of the control rod drives shall be as follows:

i a) All 37 control rod pairs will be scrammed from the full out to the full in position once a year and the scram time measured. Operable withdrawn control rods shall have a scram time less than 160 seconds.

b) All control rods which are withdrawn during power operation will be exercised a short distance (about 6 inches) once a month. Operation of position indicators, motion i ndi cator,s , _and the abs'ence of slack cable indication shall be verified.

Fort St. Vrain #1 Technical Specifications Amendment # l Page 5.1-2 Basis for Specification SR 5.1.1 Tests will be performed on the control rod drives to assess their capability to control the reactivity of the reactor core. On a yearly basis, the control rods will be scrammed from the full out position and the scram time measured. The drive mechanisms are designed for a normal scram time of 140 + 10 seconds. However, for safe reactivity control of the reactor, scram times of the drive mechanisms may be as great as 160 seconds without altering the kinetics of the scram.

The drive mechanism will be used to exercise sequentially, -

all withdrawn rods over a short distance (about 6 inches) once a month. This test will assess the operability of the control rods and drives and position indicating instrumentation. Any binding of the rods in their -

channels can be determined by a slack cable indication.

Specification SR 5.1.2 - Reserve Shutdown System l Surveillance The surveillance of the reserve shutdown system shall be as follows:

a) The ability to pressurize each of the 37 reserve shutdown hoppers to 10 psi above reactor pressure, as indicated by operation of the hopper pressure switch, shall be demonstrated every three months. Operable l reserve shutdown hoppers shall be capable of being

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Page 5.1-3 I pressurized. The ability to operate the ACM quick l disconnect valves, which provide an alternate means of l actuating the hopper pressurization valves, shall be l demonstrated every three months, and the ACM valve I actuation gas pressure shall be monitored weekly.

J l SR 5.1.2.a shall be implemented per ISI Criterion A.

3 b) The test pressurizing gas pressure indicator shall be calibrated annually.

c) An off-line functional test of a reserve shutdown assembly shall be performed in the hot service facility, or other suitable facility, following each of the first five refueling cycles and at two refueling cycle intervals thereafter. These tests will consist of pressurizing reserve shutdown hopper to the point of rupturing the disc and releasing the poison material. If a reserve shutdown hopper rupture disc does not rupture at a differential pressure less than 300 psi and release the poison material, the reactor shall be placed in a shutdown conditic ; until it can be shown that LCO 4.1.6 can be met.

d) The instrumentation which alarms a low pressure in the reserve shutdown actuating pressure lines shall be l functionally tested in conjunction with the test, and l at the same intervals, specified in part a) above, and calibrated once a year. Operable reserve shutdown l

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Page 5.1-4 hoppers shall have z.1 actuating bottle pressure greater than or equal to 1,500 psig.

e) The reserve shutdown hopper pressure switches shall be calibrated at the same , interval that they are removed from the reactor for maintenance.

l l f) Visual examination shall be performed of pipe sections l which require disassembly and reassembly within the l refueling penetrations, after they have been

-l disassembled as required for refueling or maintenance.

l SR 5.1.2.f shall be implemented per ISI Criterion B.

l g) Demonstration shall be made at each refueling outage

, l that each subsystem is operable by actuating .each l group of pressurizing valves from the Control Room.

l The capability of pressurizing the corresponding l hoppers need not be demonstrated during this test.

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l Valve position indication and fail safe operation l shall be observed during this test.

l SR 5.1.2.g shall be implemented per ISI Criterion B.

Basis for Specification SR 5.1.2 The reliability of the reserve shutdown system to perform its function will be maintained by a control system pressure test and actual off-line rupture tests conducted in the hot service facility or other suitable facili'.y.

The control system pressure test demonstrates the ability

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Page 5.1-5 to pressurize the hoppers and indicates the operability of the control system components. A successful test will increase the hopper pressure about 10 psi above reactor pressure. This differential is well below the minimum 115 psi differential required to burst the disc.

The off-line tests consist of actual disc ruptures and poison drops. These will be used to determine the reliability of the differential burst pressure of the disc, and the tendency of the poison material to hang up or deteriorate in the hoppers over ex'. ended periods of time.

This test information will be used to verify the capability to shut down the reactor in an emergency situation. The reserve shutdown system hoppers operate in two subsystems. The first consists of the seven hoppers in refueling regions 1, 3, 5, 7, 22, 28, and 34; the

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second subsystem is comprised of the remaining 30 hoppers in the remaining refueling regions. Safe contro, of the reactor by the reserve shutdown system can be accomplished with one of the seven hoppers inoperative, and one of the remaining 30 hoppers inoperative. A differential pressure from 585 to 315 psi is available from the helium supply k bottle with a pressure greater than or equal to 1,500 psig.

l ACM -valve actuation gas is provided by storage cylinders l which can be manually connected to each subsystem valve

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Page 5.1-6 l air header by means of quick-disconnect valves.

l Availability and operability of the ACM valve actuation is l demonstrated by testing.

l LCO 4.1.6 prevents performing an overall control system l operational test at power since it allows only one reserve l shutdown hopper to be inoperable in each subsystem when l the reactor is either at low power or at power. To l prevent the release uf su crve chutdown material in the l core, all hoppers of a subsystem must be rendered l inoperable when testing the centrol system. This can only l be performed when the reactor is shut down. Only valve l actuation has to be tested since the ability to pressurize l each hopper is demonstrated every three months.

Specification SR 5.1.3 - Temperature Coefficient Surveillance The reactivity change as a function of core temperature change shall be measured at the beginning of each refueling' cycle.

Basis for Specification SR 5.1.3 The major shifts in reactivity change as a function of core temperature change will occur following refueling.

The specified frequency of measurement following each major refueling will assure that the change of reactivity as a function of changes in core temperature will be I

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Page 5.1-7 measured on a timely basis to evaluate the limit specified in Specification LCO 4.1.5.

Specification SR 5.1.4 - Reactivity Status Surveillance A surveillance check of the reactivity status of the core shall be performed at each startup and once per week during power operation. If the difference betwen the observed and the expected reactivity, based on normalization to a base steady state core condition, reaches 0.01 AK, this discrepancy shall be considered an abnormal occurrence.

The initial base steady state core condition and changes of this base shall be approveo by the NFSC.

Basis for Specification SR 5.i.4 The specified frequency of the surveillance check of the core reactivity status will assure that the difference between the observed and expected core reactivity will be evaluated regularly.

This specification is designed to ensure that the core reactivity level is monitored to reveal in a timely manner the existence of potential safety problems or operational 4

problems. An unexpected and/or unexplained change in the observed core reactivity could be indicative of such problems.

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Page 5.1-8 The normalization to an initial base steady state core condition will eliminate discrepancies due to manufacturing tolerances, cnalytical modeling approximations and deficiencies in basic data at the beginning of operation. Changes of the base steady state core conditions are permissible to eliminate explainable discrepancies resulting from long-term reactivity burnup effects and core refuelings.

Comparison of predicted and observed reactivities in a base steady state configuration will ensure the comparison will be easily understood and readily evaluated.

i Any reactivity anomaly greater than 0.01 Am would be unexpected and its occurrence would be thoroughly investigated and evaluated. The value of 0.01 An is considered to be a safe limit since a shutdown margin of at least 0.01 Ar with the highest worth rod pair fully withdrawn is always maintained (see LCO 4.1.2).

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Page 5.1-9 Specification SR 5.1.5 - Withdrawn Rod Reactivity Surveillance The reactivity worth of the control rods which are withdrawn from the low power condition to the operating condition, in the normal withdrawal sequence, shall be measured at the beginning of each refueling cycle. The

, measured rod worths will be used to insure that the criteria for the selection of the rod sequence of Specification LCO 4.1.3 are met.

Basis for Specification SR 5.1.5 The measurement of control rod worths at the beginning of a refueling cycle will provide for an evaluation of calculational methods for control rod worths used in the prediction of the maximum worth rod in Specification LCO 4.1.3.

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Specification SR 5.1.6 - Core Safety Limit Surveillance During power operation the total operating time of the fuel elements within the core at power-to-flow ratios above the curve of Figure 3.1-2 will be e' valuated once per week when the plant operation is within the normal operating range, and as soon as practicable after any deviation from the normal operating range. These operating times will be compared to the allowable operating time of Specification SL 3.1 to assure that the

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Core Safety Limit has not been exceeded.

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Page 5.1-10 Basis for Specification SR 5.1.6 Only during operation of the plant outside of the normal operating range is there a potential for accumulating significant operating times at power-to-flow ratios greater than the curve of Figure 3.1-2. Therefore, weekly evaluations of the total accumulated operating time at power-to-flow ratios greater than the curve of Figure 3.1-2 is sufficient during normal operation.

Following any significant deviation from the normal operating range, the operation should be evaluated to determine the degree to which the actual total operation of the core approached the Core Safety Limit.

Specification SR 5.1.7 - Region Peaking Factor Surveillance The calculated region peaking factors (RPF's) used in dete-mining the individual region outlet temperatures for

( Regions 20 and 32 through 37 and percent RPF discrepancy (see LCO 4.1.7) for Regi,ons 1 through 19 and 21 through 31 shall be evaluated according to the following schedula for each refueling cycle:

a) Calculated RPF's: 1) Prior to initial power operation after refueling.

2) At the equivalent of 20

(+5) effective days at

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Page 5.1-11 rated thermal power after refueling.

3) At the equivalent of 40

(+5) effective days at rated thermal power after refueling.

4) At monthly intervals thereafter, provided that the core has accumulated an exposure of at least the equivalent of 10 effective days at rated thermal power since the previous evaluation. If the core has accumulated an exposure of less than the equivalent of 10 effective days at rated thermal power since the previous evaluation, the evaluation may be deferred until the next applicable interval.

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Page 5.1-12 b) Percent RPF Discrepancy: Within a total elapsed time of 10 calendar days at reactor power levels above 40% of rated thermal power after the completion of any of the " Calculated RPF" evaluations required above with the following qualifications:

1) A " Percent RPF Discrepancy" evaluation shall be performed prior to exceeding 40% of rated thermal power for the first time after refueling, but at a reactor power above 30%

of rated thermal power.

2) If the total elapsed time at reactor power levels above 40% of rated thermal power does not exceed 10 calendar days prior to the subsequent " Calculated RPF" evaluation, the

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Page 5.1-13

" Percent RPF Discrepancy" evaluation is not required, but the total elapsed time at reactor power levels above 40'4 of rated thermal power between

" Percent RPF .

Discrepancy" evaluations shall not exceed 45 calendar days.

Basis for Specification SR 5.1.7 The calculated region peaking factors for Regions 20 and 32 through 37 and their comparison regions will change during the refueling cycle as fission product inventories saturate, fissile ' material and burnable poison are depleted, and control rods are withdrawn from the core.

Evaluations based upon operating experience gained prior to completion of rise-to power testing (i.e., Cycles l'and 2 and part of Cycle 3) indicate that the ratio of the calculated region peaking factors in Regions 20 and 32 through 37 to the calculated region peaking factors in comparison regions as a function of control rod configuration, changes gradually in a predictable manner during a refueling cycle. A surveillance check of the calculated region peaking factors at the specified

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Page 5.1-14 frequency will assure that the appropriate region peaking factors continue to be used in determining the region outlet temperature for Regions 20 and 32 through 37.

1 The calculated and measured region peaking factors for Regions 1 through 19 and 21. through 31 (candidate comparison regions) will change during the refueling cycle as fission product inventories saturate, fissile material and burnable poison are depleted, control rods are withdrawn from the core, and region flow characteristics change. A surveillance check of the percent region peaking factor discrepancy will provide assurance that the requirements of LC0 4.1.7c are being met for comparison regions. The frequency for surveillance has been established based upon conservative evaluat.ons of potential fuel kernel migration, which could occur if a region with an excessively large, negative region peaking factor discrepancy were used as a comparison region.

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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 toe primary reactor coolant envelope as a fission preduct barrier and to ensure the capability to cool the core under all modes of operation.

l Specification SR 5.2.1 - PCRV and PCRV Penetration l Overpressure Protection Surveillance l a) Each of the two overpressure protection assemblies l protecting the PCRV shall be tested at intervals not l

~ to exceed five years, on an alternating basis, with l one overpressure protection assembly tested during l each refueling cycle.

l The PCRV safety valve containment tank closure bolting l shall be visually examined for absence of surface l defects when the tank is opened for the above testing.

l Tank closure flange leak tightness shall be determined l following tank closure.

l SR 5.2.1.a shall be implemented per ISI Criterion C.

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Page 5.2-2 l b) Each of the two overpressure protection assemblies protecting a steam generator or a circulator l penetration interspace shall be tested at five l calendar year intervals on an alternating basis, so

[ that one safety valve for each penetration interspace l and one rupture disc of each type are tested at an l approximate interval of two and a half years.

l SR 5.2.1.b shall be implemented per ISI Criterion D.

l c) The instrumentation and controls associated with the l overpressure protection assemblies in a) and b) above I shall be tested and calibrated as follows:

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.

l The pressure switch and alarm for the PCRV safety l valve containment tank shall be functionally l tested and calibrated annually.

l SR 5.2.1.c.1 shall be implemented per ISI l Criterion D.

2) The position indication circuits associated with l the PCRV overpressure protection system shut off l valves shall be functionally tested and calibrated I when testing either of the PCRV overpressure l

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Page 5.2-3 l protection assemblies. The pressure switch and l alarm for the PCRV safety valve bellows shall be l functionally tested and calibrated in conjunction l with its associated safety valve test.

l SR 5.2.1.c.2 shall be implemented per ISI l Criterion C.

l 3) The control, interlock, and position indication l circuits associated with each of the PCRV l penetrstion overpressure protection system shut l off valves shall be functionally tested at five l calendar year intervals.

l 1R 5.2.1.c.3 shall be implemented per ISI l Criterion D.

Bas's for Specification SR 5.2d l Tasting of a PCRV overpressure protection assembly can l only be performed when closing the corresponding manual l shut off valve, located upstream of the rupture disc.

l LCO 4.2.7 does not allow isolation of such an assembly l unless the primary pressure is less than 100 psia.

l Consequently, testing and examinations will be performed l at shutdown. One assembly will be isolated while the

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other one will remain in a fully operational condition l during the testing procedure, thus ensuring overpressure l protection of the PCRV.

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Page 5.2-4 l The rupture disc is designed to be removed from the system I for bench testing. Verification is made of the correct

! deflection of the disc at the set pressure level which l would cause the membrane to be ruptured. The safety valve l is tested for setpoint activation without removing it from l the system.

l The pressurized portion of the assembly is monitored for l leakage during plant operation. Leakage examination of l the containment tank cover seals and visual examination of l the cover bolts provides assurance that containment tank l integrity is restored after the tank cover has been l re-installed.

l Testing of a PCRV penetration overpressure protection l assembly can be performed during plant operation since the l assemblies are accessible and since LCO 4.2.7 requires l l only one assembly to be operable at any time.

l The safety valve in each assembly is tested while in place l

l to demonstrate that it opens at the correct set pressure.

[ The rupture discs are not provided with a testable design l feature and, therefore, cannot be tested. However, one l

l l rupture disc of each type assembly is visually examined to l

l verify that the membrane is free of defects and that the l knife blade remains sharp.

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Page 5.2-:i l The intervals specified for testing the overpressure l protection assemblies are adequate to demonstrate the l operability of the overpressure protection systems.

The intervals specified for testing the associated I instrumentation and controls are adequate to assure l reliability of rupture disc and safety valve operation and I to monitor the integrity of the PCRV safety valve piping l' and containment tank.

l Specification SR 5.2.2 - Tendon Corrosion and Anchor l Assemblies Surveillance ,

The serviceability of the corrosion protection applied to and the condition of the prestressing tendons shall be l monitored in accordance with paragraphs a) and b).

I Surveillance of the tendon end anchor assemblies shall be l performed in accordance with paragraph c).

l a) Corrosion protected wire samples of sufficient length l (i.e., initially at least 15 feet where practical, or

l. half the tendon length, whichever is shorter) shall be l inserted with selected tendons (those tendons with l load cells). Corrosion inspection of at least one of l these wires shall be made at the end of the first and l third calendar year after prestressing. Additional l inspections shall be conducted at five calendar year l intervals thereafter.

l SR 5.2.2.a shall be implemented per ISI Criterion D.

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Page 5.2-6 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 from which wire samples are examined) shall be drawn and analyzed for products of corrosion, in l coordination with and at the same time intervals as l for paragraph a) above.

l c) Visual examination of 5% of the prestressing anchor l assemblies shall be performed at five calendar year l intervals. This may include the anchor assemblies l which can be visually examined while performing a) and l b) above.

l SR 5.2.2.c shall be implemented per ISI Criterion D.

i-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 l

l plant. The details of the corrosion protection system are l

l described in Section 5.6.2.5 of the FSAR.

t Sampling tendon tube atmosphere for products will provide a secondary check on the adequacy of the corrosion l

l protection provided for the stressing tendons.

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Page 5.2-7 l Visual examination of tendon end anchor assemblies will l provide additional assurance that the prestressing system l has not degraded by checking the corrosion protection and l integrity of the anchor assemblies.

Specification SR 5.2.3 - Tendon Load Cell Surveillance l a) Checks on the possible shift in the load es11 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 d

thereafter.

l b) The load cell alarm circuit between the Data i

l Acquisition System Room and the Control Room shall be l functionally tested annually to assure that the

, l operator in the Control Room is alerted when tendon l load settings are exceeded.

l SR 5.2.3.b shall be implemented per ISI Criterion A.

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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 i

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Pege 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 4

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).'

t l Specification SR 5.2.4 - PCRV Concrete Structure l- Surveillance l a) Crack patterns on the visible surfaces of the PraV shall be mapped prior to and following the in'.ial proof test pressure (IPTP). Concrete cracks which l

exceed 0.015 inches in width shall be r corded.

Subsequent concrete surface visual inspect ons shall be performed after the end of the first and third calendar year following initial pov3r operation.

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Page 5.2-9 Recorded cracks shall be assessed for ch.nges in length and any new cracks will be recorded.

Additional inspections shall be conducted at ten calendar year intervals thereafter.

l b) PCRV deformations and deflections at vessel midheight l and at the center of the top head shall be monitored l at five calender year intervals during a vessel .

l pressurization to operating pressure.

l SR 5.2.4.6 shall be implemented per ISI Criterion C.

I c) The PCRV support structure shall be visually examined l for evidence of structural deterioration at ten l calendar year intervals.

l 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 michanical 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.

l Model testing has also shown that severely cracked vessels i

i contain the normal working pressure for extended periods I

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Page 5.2-10~

of time as long as the effective prsstressing 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 .'.ructure, 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 ir 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. DetailsofthePCRV structural monitoring provisions are given in Section 5.13.4 and Appendix E.17 of the FSAR.

I The interval for surveillance after the fifth year following initial prestressing may be adjusted based on the analysis of prior results.

l Monitoring of overall PCRV deformations and deflections is l the best indication of PCRV structural performance and

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Page 5.2-11 l verifies that the PCRV response is elastic and that no l significant permanent strains exist.

l Visual examination of the PCRV support structure will l indicate that no structural deterioration has occurred.

l Significant cracking patterns or sizes should be l investigated with respect to their impact on the integrity l 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.

l During the fifth refueling cycle, three sets of 12 l specimens of the PCRV liner materials and weld material shall be removed and tested to obtain Charpy impact data.

The specimen hotders shall contain dosimeters to provide integrated neutron flux measurements. Additional specimen

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l, removal and testing shall be conducted during every tenth I 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 test specimens, made from the liner materials, near the liner r

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Page 5.2-12 and expcsing 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 te 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 curve for the plate metal, the weld metz1 and the heat affected :one at each test interval.

This testtng 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 temperr.ture 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 l to a fluence equivalent to the first five years of full power operation. Further, these tests demonstrated na significant damage after a fluence equivalent to 30 years

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Page 5.2-13 l 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.

Specification SR 5.2.6 - Plateout Probe Surveillance One plateout probe shall be removed for evaluation coincident with the first, third, and fifth refueling, and at intervals not to exceed five refueling cycles thereafter. If, during the second or fourth refueling cycle, or any refueling cycle following the fifth 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 probes removed shall also be analyzed for 281I.

Basis for Specification SR 5.2.6 The plateout probes are located in penetrations extending 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

Fort St. Vrain #1 Technical Specifications Amendment #

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 reactor circuit to determine compliance with LCO 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 2'2I 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.

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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

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shall be functionally tested every three months.

d) The instrumentation and controls associated with c) shall be functionally tested in conjunct, ion with 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.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.2-16 Specification SR 5.2.8 - Bearino Water Makeup Pemo 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 ensure proper operation of the pumps and associated controls.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.2-17 Specification SR 5.2.9 - Helium Circulator Bearina Water Accumultors Su.~veillance 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 annual calibration interval will assure proper operation of the accumulator controls if they should ever be called upon to function. .

r Fort St. Vrain #1 Technical Sptcifications i Amendment # l Page 5.2-18 Soecification 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 ranthly 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.

5) The batteries, cell plates, and battery racks, shall be inspected each refueling cycle for evidence of physical damage or abnormal degradation. The batter-to-battery and terminal connections shall be verified to be clean, tight,

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.2-19 free of corrosion, and coated with anti-corrosion materia 1 'ch refueling cycle.

b) The fire suppression water system shall be verified operable as follows:

1) Monthly by verifying 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

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suppres('on water system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

(a) ' Veri fyi ng that each automatic valve in the i

flow path actuates to its correct position.

(b) Verifying that each fire water pump develops ,

at least 1,500 gpm at a system head of 290 feet.

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Fort St. Vrain #1 Technical Spocifications 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 te 125 psig.

5) Each three years by performing a flow test.

Basis for Specification SR 5.2.10 v

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 fire 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.

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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 ionitor.

l 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 prime,ry 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 analysis 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.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.2-22 Soecification 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 analysis 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 maintenance of the chemical impurity monitoring l equipment. The presence of higher than nominal impurity l

levels of chemical impurities is related to core materials corrosion which might occur only with very high levels for 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 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 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 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.

Fort St. Vrain #1 Technical Specifications 1 Amendment #

Page 5.2-24 i

i Basis of Specification SR 5.2.14 f

The ultrasonic inspection of the PCRV liner is provided to

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detect the thinning of the liner due to corrosion or to detect defects within the liner at representative areas.

f 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 b'e adjusted based on

< the analysis of prior results.

Specification SR 4.2.15 - PCRV Penetration Interspace Pressure Surveillance ,

The instrumentation which monitors the pressure

, differential between the purified helium supply header to the PCRV penetration int'erspaces and the primary coolant system will be functionally tested once every mont'h 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 panetration interspaces shall be at a higher pressure than the primary coolant pressure within the PCRV.

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Page 5.2-25 Soecification SR 5.2.16 - PCRV Closure Leakage, Surveillance Reouirements The surveillance of PCRV closure leakage shall be as follows:

a) PCRV primary and secondary closure leakage shall be l determined once each quarter, or as soon as

'l practicable after an unanticipated increase in pressurization gas flow is alarmed.

l 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 anndally.

l c) The instrumentation which monitors or alarms pressure l in the core support floor and core support floor l

columns shall be functionally tested and calibrated l annually.

l SR 5.2.16.c shall be implemented per ISI Criterion A.

I d) The controls, position indication, and fail safe l operation for remote manual isolation valves l associated with pressurizing, purging, and venting l

PCRV closures shall be fu.e.ionally tested at five l calendar year intervals, and for automatic isolation l valves, annually, or at the next scheduled plant 4

Fort St. Vrain #1 Technical Spacifications Amendment #

Page 5 2-26 l shutdown if these valves have not been tested during l the previous year.

l SR 5.2.16.d shall be implemented per ISI Criterion B.

I e) The check valves on the HTFA purge lines shall be l tested at five calendar year intervals.

l SR 5.2.16.e shall be implemented per ISI Criterion B.

l f) The check valves which are 'part of the HTFA or i refueling penetrations shall only be tested when such l a penetration is open for refueling or maintenance, if l the check valves have not been tested in the last five l years.

I 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.

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Fort St. Vrain #1 Technical Sp2cifications 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.

l The interval specified for functional test and calibration l of the instrumentatien and alarms monitoring the core l support floor and columns will assure sensing ano alarming l any change in their structural integ-ity.

l The interval specified for valve testing is adequate to l assure proper valve operation when isolation of the l closure auxiliary piping is required.

Specification SR 5.2.17 - Helium Circulator Pelton Wheels l DELETE SPECIFICATION SR 5.2.17 IN ITS ENTIRETY l Specification SR 5.2.18 - Helium Circulators Surveillance l 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.

l 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 l 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 criteri-specified for Class I components in Article NB-2500,Section III, ASME Code.

l b) Following the first co:nplete helium circulator inspectior., a previously uninspected helium circulator l shall be removed and inspected at ten calendar year l inter /als. The helium circulator compressor wheel l rotor, turbine wheel, and Pelton wheel shall be l inspected as specified in Paragraph a.2. Other helium l circulator components, accessible without further l disassembly than required to inspect these wheels, l 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.

l SR 5.2.18 shall be implemented par 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 Sprcifications 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 f

System of the Fort St. Vrain plant are utilized and operated during normal operation of the plant. This includes the helium circulators.

l The performance of the helium circulators is monitored l during operation, i.e., instruments are provided with the l capability to measure compressor differential pressure and l flow, bearing temperature, bearing water temperature and l flow, buffer helium flow, and shaft speed and vibration.

Examination at tne time of the first turbine generator 1 overhaul, and at approximately ten year intervals l thereafter, is sufficient to monitor the condition of the

-l helium circulator. The first turbine generator " tear- ,

down" or overhaul usually occurs after one year running to l check the total assembly. Only checks of components are l 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 inspection, will have undergone sufficient stress cycling to accurately indicate service life.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.2-30 Soecification SR 5.2.19 - IACM Diesel-Driven Pumos 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 minimura of two hours. The load under this condition shall be at least 100% of design ACM equipment full i

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 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) nonths, 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 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.

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Fort St. Vrain #1 Technical Sp1cifications Amendment #

Page 5.2-32 Upon removal, these specimens will be subjected to I 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 Soteification 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.

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The surveillance test specimens will be subjected to the primary coolant conditions, as well as other reactor i,

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 l l Withdrawal at l l l l Refueling l l l Fuel Region I Column l Number

• l l 1 I I l 25 l 7 l 2 l l l l l l 30 l 3 l 4 l l l 1 -l l 24 l 7 l 6 l l l l l l 22 l 6 I 9 1 1 I I I l 27 1 2 l 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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Technical Specifications Amenoment #

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 qtarterly 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 ccre 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:

1 a) The circulator 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 ci rculating 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 stora<;e ponds and the fire water pump pits shall be veri'ied correct monthly.

Basis for Specificaton 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.

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- Specification SR 5.2.25 - Core Support Block Surveillance l The top surface of the core support block for fuel regions l fitted with PGX graphite specimens shall be visually i l examined by remote TV for indication of cracks, in l particular in areas where analysis shows the highest l tensile stresses exist, at the refueling shutdown when the .

l PGX graphite specimens are scheduled to be removed from l the core in accordance with Technical Specification l SR 5.2.22.

l SR 5.2.25 shall be implemented per ISI Criterion D.

Fort St. Vrain #1 Technical Sptcifications Amendment #

Page 5.2-36 l Basis for Specification SR 5.2.25 l Visual examination of the core support blocks in those

-l regions chosen for insertion of PGX graphite specimens l will provide additional assurance that integrity of the l core support blocks does not degrade due to plant l operating conditions, since those regions were selected l because of their higher potential for PGX graphite l burnoff. Analysis shows that the highest tensile stresses l occur on the top surface of the core support blocks, at l the keyways, and at the web between reactor coolant l channels.

l Specification SR 5.2.26 - Region Constraint Devices l Surveillance l The region constraint devices (RCD's) shall be inspected I at each refueling outage using the fuel handling machine l from those regions being refueled as follows:

l a) The upper core plenum shall be visually examined by l- remote TV to verify that RCD's within visible range l are'in place on top of the core.

l b) As RCD's are removed, the fuel handling machine l location coordinates and lifting force shall be l monitored to verify that the RCD pins were engaged in l the fuel columns and that they disengage as expected.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.2-37 l c) Selected RCD's shall be visually examined by remote TV l in the fuel handling machine after removal to verify l their structural integrity.

l d) As RCD's are re-installed, the fuel handling machine l location coordinates shall be monitored to verify that l the RCD pins have engaged in the fuel columns.

I SR 5.2.26 shall be implemented per ISI Criterion B.

t Basis for Specificatjon SR 5.2.26

! R:gion constraint devices, located on to;, of fuel columns i of generally three adjacent fuel regions, restrain region l movements in relation to one another by means of centering l rins trserted in the hand?ing hole of the upper plenum l elements.

l Visual examination of the upper core plenum and comparison l of the as-installed /as-found RCD coordinates ,will . assure l that the RCD's remain in place and that no phenomenon is l occurring which could cause them to disengage from the l fuel columns. Comparison of RCD coordinates will require l correction to account for changes in fuel column height l due to irradiation of graphite and coordinate changes l which will occur when RCD's are removed from a different l refueling penetration than the one from which they were l installed.

Fort St. Vrain #1 Technical SpIcifications Amendment #

Page 5.2-38 l Monitoring the lifting force to remove the RCD's with the l fuel handling machine will provide early indications,

.., l should a phenomenon occur over time which might eventually l prevent them from moving with the fuel columns or prevent l their removal from the reactor. Removal and l re-installation will act as go/no go dimensional test of l the region constraint devices.

l Visually examining and photographing selected RCD's in the l fuel handling machine will assure that there are no l unacceptable deformations, loose or missing parts, or l other visible defects.

l Specification SR 5.2.27 - Helium Shutoff Valves l Surveillance l Proper closure of the helium shutoff valves shall be l monitored annually, or at the next scheduled plant l shutdown, if such monitoring has not been performed during l the previous year.

l SR 5.2.27 shall'be implemented per ISI Criterion C.

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Page 5.2-39 l Basis for Specification SR 5.2.27 l The helium shutoff v'alves are self-actuated check valves l which close when the corresponding circulators are l shutdown or tripped. Simultaneous long term failure of l both the circulator and its helium shutoff valve, under l very degraded conditions of remaining plant equipment, l could lead to a situation analogous to a loss of forced l circulation accident, due to the open recirculation path l between circulator outlet and inlet plenums.

l Verification that the helium shutoff valves close properly 1 will provide assurance that the residual heat remuval l capability would not be degraded by the malfunction of a l helium shutoff valve.

l Specification SR 5.2.28 - PCRV Penetrations and Closures l Surveillance l a) Accessible portions of 'PCRV penetration pressure l retaining welds shall be examined for indications of l surface defects as follows:

l 1) Surface examine (MT or PT) the following three l welds in one steam generator penetration in each l loop at five calendar year intervals:

l - the penetration shell to secondary closure weld, l - the secondary closure to upper bellows support

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.2-40

l. weld, and l - the lower bellows support to reheat header l sleeve weld.

l 2) Surface examine (MT cr PT) the following two welds 1 in the bottom access penetration at 10 calendar l year intervals:

l - the penetration shell.to spherical head held, I and l - the spherical head to closure flange weld.

l SR 5.2.28.a shall be implemented per ISI Criterion C.

l b) Accessible portions of the PCRV penetration closure l and flow restrictor restraint components shall be l examined for indications of defects as follows:

l 1) Visually examine the he.lium circulator restraint l system (cylinder, ring, and bolting) for one l

- penetration in each loop at five calendar year l intervals.

i l SR 5.2.28.b.1 shall be implemented per ISI l Criterion C. -

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Page 5.2-41 l 2) Visually examine the refueling penetration l holddown plate bolting at each refueling outage.

l SR 5.2.28.b.2 shall be implemented per ISI l Criterion B.

l 3) Visually examine the bottom access penetration l primary closure split ring assembly and its l secondary closure bolting at 10 calendar year l intervals.

l SR 5.2.28.b.3 shall be implemented per ISI l Criterion C.

I c) Accessible portions of the PCRV safety valve l penetration containment tank support components shall l be examined at 10 calendar year intervals for l indications of defects as follows:

I 1) Surface examine (MT or PT) the support skirt to l tank attachment weld.

l- 2) Visually examine the support skirt *between the l tank and PCRV outer wall.

l 3) Visually examine, torque, and tension test the l bolting attaching the support skirt to the PCRV l outer wall.

I SR 5.2.28.c shall be implemented per ISI Criterion C.

Fcrt St. Vrain #1 Technical Specifications Amendment #

Page 5.2-42 l Basis for Specification SR 5.2.28 l Structural integrity of Fort St. Vrain PCRV penetration l secondary pressure retaining boundaries is normally l verified by continuous leakage monitoring and by periodic l leakage testing of the penetration interspace. The l specified examinations of accessible circumferential welds l at structural discontinuities will provide additional l assurance concerning the continued integrity of the l secondary pressure boundary at these critical locaticns.

l Exanination of accessible penetration closures, flow l restrictors, and equipment restraint or support components l provides assurance that these components remain l structurally sound and capable of performing their safety l function under both normal and accident conditions.

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Fort St. Vrain #1 Technical Sptcifications .

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Page 5.3-1 5.3 SECCNDARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS Aoplicability 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.

l Soecification SR 5.3.1 - Steam / Water Du.ap System l Surveillance l a) The steam / water dump valves shall be tested l individually every three months.

l b) The steam / water dump tank level indicators shall be l checked daily, and functionally tested every three l months.

l SR 5.3.1.b shall be implemented per ISI Criterion A.

! l c) The steam / water dump tank level, pressure and l temperature instruments (including indicators, alarms, l and interlocks - where applicable) shall be l functionally tested and calibrated annually, or at the l next scheduled plant shutdown if such surveillance has l not been performed during the previous year.

l SR 5.3.1.c shall be implemented per ISI Criterion B.

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Page 5.3-2 Basis for Soecification 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 steam generator tube rupture.

The cump 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.

l The specified frequency for instrumentation functional l test and calibration is adequate to assure that the water l level in the steam / water dump tank does not exceed the l

l limits of LCO 4.3.3, and, in case of dump, to confirm that l the proper steam generator has been dumped, and to prevent l venting and draining of the tank to the radioactive l

l gaseous and liquid systems before the contents have been I adequately cooled.

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Page 5.3-3 Specification SR 5.3.2 - Main and Hot Reheat Steam Stop Check Valves Surveillance The main steam and hot reheat steam stop check valves l shall be full stroke tested in accordance with l specification SR 5.3.4 and partial stroke tested once per l week.

l SR 5.3.2 shall be implemented per ISI Critericn A.

Basis for Specification SR 5.3.2 i

The main steam stop che:k and hot reheat stop check valves will be partially stroked once a week during plant 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 be stroked during power operation by means- of special electrical circuitry in the hydraulic control system which 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 -

close when required, without causing severe pressure, temperature, flow, or power generation transients.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-4 l Specification SR 5.3.3 - Byoass and pressure Relief Valves l Surveillance l The main steam and hot reheat steam power operated l (electromatic) pressure relief valves, and the six hot l reheat steam bypass valves shall be tested once per year, l or at the next scheduled plant shutdown if the valves have l not been tested during the previous year. The main steam l bypass valves shall be tested in accordance with l specification SR 5.3.4.

l 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 l valves and pressure relief valves will be tested during l plant shutdown as follows:

l a) The main steam and hot reheat steam power operated l pressure relief valves will be tested by exercising l the valves.

b) The main steam bypass valves will be tested for operability by cycling the valves.

c) The six hot reheat steam hypass valves will be tested by exercising each valve to ensure freedom of movement.

4 e Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-5 The main steam bypass valves divert up to 77% 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.

l The main steam power operated relief valves divert the l remaining steam flow to atmosphere.

l The six hot reheat steam bypass valves and the power i operated pressure re11of valves ensure a continuous steam i flow path from the helium circulators for decay heat removal.

The tests r& quired on the above valves will demonstrate that each valve will function properly. Test frequency is considered adequate for assuring valve operability at all times.

Specification SR 5.3.4 - Safe Shutdown Cooling Valves, Surveillance Those valves that are pneumatically, hydraulical.ly, or electrically operated, that are required for actuation of the Safe Shutdown Cooling mode of operation, shall be l tested annually, or at the next scheduled plant shutdown l if these valves have not been tested during the previous l year.

l In addition, the above test shall include the normally l closed check valves which are required to open for

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-6 l actuation of the safe shutdown cooling mode of operation, I when such testing is practical.

l SR 5.3.4 shall be implemented per ISI Criterion B.

Basis for Specification SR 5.3.4 The safe shutdown cooling mode of operation utilizes systems or portions of systems that are in use during normal plant operation. In many cases, those valves required to initiate Safe Shutdown Coc'ing are not called upon to function during normal operation of the plant, except to stand fully closed or open.

l Testing of these valves will assure their operation if called upon to initiate the Safe Shutdown Cooling mode of operation.

During reactor operation, the instrumentation required tc 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 instrumentation not normally in use is tested at intervals j 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 the FSAR and are represented in FSAR, Figure 10.3-4.

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-7 i Valve testing will include, as applicable, full stroking l each valve, or an observation that the valve disc travels l from the valve normal operating position to the position l required to perform the safety function, an observation l that the remote position indicators accurately reflect l actual valve position, and a neasurement of the full l stroke time for the hydraulically actuated automatic l valves.

Specification SR 5.3.5 - Hydraulic power System Surveillance 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.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-8 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 Specification 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 instrunent 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 values.

Specification SR 5.3.7 - Secondary Coolant Activity Surveillance The secondary coolant system will be analyzed for l'1I, tritiua, 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 Technical Specifications Amendment #

Page 5.3-9 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 analysis 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 adeounte to sonitor the activity of the secondary coolant.

Soecification 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 i 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:

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-10 Number of Snubbers Found Inoperable During Inspection Next Required or Durine Inspection Interval Inspection Interval 0 18 Months 1 25%

1 12 Months 1 25%

2 6 Months 1 25%

3, 4 124 Days 1 25%

5,6,7 62 Days 1 25%

28 31 Days 1 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 demonstrated 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 shall be assumed that the facility '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 functionally

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-11 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 tested 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 vitually inspected for overall integrity and operability. The inspection will include verification of proper or.ientation, adequate hydraulic fluid level and proper attachment or snub 5er 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.

l l

l

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-12 Experience at operating facilities has shown that the required surveillance program should assure an acceptab':

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 snubbers 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.

l l Specificaton SR 5.3.9 - Safety Valves Surveillance l The steam generator superheater and reheater safety valves

! and the steam / water dump tank safety valves shall be l

l tested at five calendar year intervals to verify their i

• r l setpoint.

l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-13 l SR 5.3.9 shall be ,mplemented per ISI Criterion B.

l Basis for Specification SR 5.3.9 l The safety valves protect the integrity of the steam l generators, which are part of the reactor coolant l boundary, and of the dump tank, which may contain l radioactive fluids. Testing the safety valve setpoints l will assure that the pressure within the equipment remains l within design limits.

l When practical, testing of the safety valves will be i

l scheduled during the surveillance interval so that testing 4 l of one (or more) s fety m valve (s) of similar type and l operating conditions several times during the interval l will provide additional confidence in safety valve l reliability and adequate overpressure protection.

l

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Specification SR 5.3.10 - Secondary Coolant- System

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l Instrumentation Surveillance l The secondary coolant reheat steam instrumentation used l a) for control and indication of emergency condensate l flow to the reheaters and reheater backpressure, in l case of safe shutdown cooling, l b) to automatically open the reheater discharge bypass on l high pressure, and

I Fort St. Vrain #1

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Page 5.3-14 l c) to monitor reheater discharge bypass temperature, and l reheater inlet temperature, l shall be functionally tested and calibrated annually, or l at the next scheduled plant shutdown if such surveillance l was not performed during the previous year.

l SR 5.3.10 shall be implemented per ISI Criterion B.

l Basis for Specification SR 5.3.10 l The frequency specified for surveillance of the above l instrumentation will assure that they perform their l expected automatic actions, and that the operator will be l provided with accurate information which he can use for l safe shutdown cooling or to avoid abnormal equipment l operation.

l Specification 5.3.11 - Steam Generator Bimetallic Welds l Surveillance l The accessible portions of steam generator bimetallic l- welds shall be volumetrically examined for indications of l subsurface defects as follows:

l a) The main steam ring header collector to main steam .

] piping weld for one steam generator module in each l loop at five calendar year intervals.

Fort St. Vrain #1

. Technical Specifications Amendment #

Page 5.3-15 l b) The main steam' ring header collector to collector l drain piping weld for one steam generator module in l each loop at five calendar year intervals.

l c) The same two steam generator modules initially l selected shall be re-examined at each interval.

I d) The bimetallic welds described in a) and b) shall also l be inspected for two other steam generator modules in l each loop during the initial examination.

l SR 5.3.11 shall be implemented per ISI Criterion C.

l Basis for Specification 5.3.11 l The steam generator crossover tube bimetallic welds l between Incoloy 800 and 2 1/4 Cr-1 Mo materials are not l accessible for examination. The bimetallic welds between

, l the steam generator ring header collector, the main steam l piping, .and the collector drain piping are accessible, l involve the same materials and operate at conditions not l significantly different from the crossover tube bimetallic l

l welds. The collector drain piping weld is also l geometrically similar to .the crossover tube weld.

l Examination of selected bimetallic welds that are l accessible will provide additional assurance concerning l the continued integrity of steam generator bimetallic l l welds. Although no degradation is expected to occur, this l

l specifica'. ton allows for detection of defects which might l result from conditions that can uniquely affect bimetallic

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.3-16 l l welds made between these materials. Additional collector l welds are inspected at the first examination to establish l a baseline which could be used, should defects be found in l later inspections and additional examinations subsequently l be required.

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Page 5.3-17 THIS PAGE INTENTIONALLY LEFT BLANK f

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4 5.4 INSTRUMENTATION AND CONTROL SYSTEMS - SURVEILLANCE AND CALIBRATION REQUIREMENTS Applicability Applies to the surveillance and calibration of the reactor protective system and other critical instrumentation and controls.

Objective i

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.

c) Table 5.4.3 - Minimum Frequencies for checks, calibrations, and testing of Circulator Trip System.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-2 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 in 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 l refueling cycle calibration was selected for passive l

devices (thermo-couples, pressure transducers, etc.).

Devices incorporating semiconductors, particularly ampli fi ers , will be also calibrated on a once per i

! refueling cycle basis, and any drift in response or bistable setpoint will be discovered from the' test l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-3 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 sensitivity 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.

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-4 l TABLE 5.4-1 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF SCRAM SYSTEM l

l l Channel l l l l ll Description i Function IFrequency(1)I Method l l l 1. Manual (Con- la. Test l R ia. Manually trip sys- l l l trol Room) l l l tem. l l l 2. Manual (I-49) la. Test l M la. Manually trip each l ll l l l channel. l l l 3. Startup Chan- la. Check l 0 la. Comparison of two l ll nel

• l l l separate channel l l l l l l indicators. l l l lb. Test l P lb. Internal test sig- l ll l l l nal to verify trips l ll l l l and alarms. I ll Ic. Calibrate l R lc. Internal test sig- l l l 1 l l nal shall be l l l l l l checked and cali- l ll l l l brated to assure l l l l l l that its output is l ll l l l in accordance with l ll l l l the design require-l l l l l l ments. This shall l ll l l l be done after com- l ll l l l pleting the ex- l ll l l l ternal test signal l ll l l l procedure by l l l l l l checking the output l l l l l l indication when l ll 1 1 I turning the in- l l l l l. l ternal test signal l , .

ll l l l switch. l-l l 4. Linear Power la. Check l 0 la. Comparison of six l l l Channel l l l separate channel l

. l l l l l indicators. l l l -l ib. Test l M lb. Internal test sig- l ll l l l nal to verify trips l 1I I I I and alarms. l l l lc. Calibrate l D lc. Channel adjusted tol l l l l l agree with heat l ll l l l balance calcu- l l l l l l l lation. l ll ld. Calibrate l R ld. Internal test sig- l ll l l l nals to adjust l l1 l l 1 trips and indica- l l ll 1 l l tions. l l l S. Wide Range la. Check l 0 la. Comparison of threel ll Power Channel l l l separate indica- l ll l 1 I tors. I l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-5 l TABLE 5.4-1 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF SCRAM SYSTEM (Cont'd) l ll Channel l l l l l l Description i Function IFrequency(1)] Method l l l S. (Cont'd) lb. Test l P lb. Internal test sig- l l1 l l l nals to verify l ll l l l trips and alarms. l ll lc. Calibrate l M lc. Channel adjusted tol ll I l l agree with heat l l l l l l balance calcu- l ll l l l lation. l ll ld. Calibrate l R ld. Internal test sig- l ll l l 1 nals to adjust l ll l l l trips and indica- l ll l l l tions. l l l 6. Primary la. Check l D la. Comparison of two l ll Coolant l l l separate high level l ll Moisture (All l l l channel mirror l ll Channels) l l l temperature indica-]

ll l l l tions. l ll lb. Check l D lb. Comparison of six l ll l l l separate low level l ll l l l channel mirror l ll l l l temperature indica-l ll l l l tions. l ll lc. Calibrate l R lc. Inject moisture l ll l l l laden gas into sam-l ll l l l pie lines. l Il Id. Check l D ld. Verification of l l l l l l eight separate mon-l l l l l l itor's sample flow,l l l l l l per Item (t) of l ll l l l Notes for Tables l ll l l l 4.4-1 through l l -l l l l' 4.4-4. l ll le. Test l M le. Verify that each ofl ll l l l the eight monitors l ll l l l will alarm on low l l l l l l and high sample l ll l l 1 flow. l l l 7. Primary la. Test i M la. Trip one high l ll Coolant l l l level, one low l ll Moisture (High; l l level channel, l ll Level Chan- l l l pulse another low l l[ nels) 1 I l level channel. l l l 8. Reheat Steam la. Check l D la. Comparison of the l ll Temperature l l l averaged thermo- l l l l l l couple channel in- l ll l l l put indications. l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-6 l TABLE 5.4-1 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF SCRAM SYSTEM (Cont'd) l ll Channel l l l l ll Description l Function IFrequency(1)I Method l l l 8. (Cont'd) lb. Test l M lb. Trip channel, l ll l l l verify alarms and l ll l l l indications. In- l ll l l l ternal test signal l ll l l l to verify trips andl ll l l l alarms. l ll lc. Calibrate l R lc. Compare each l ll l l l thermocouple output l ll l l l to an NBS traceable l ll l l l standard. Internall ll l l l test signal to ad- l ll l l l just trips and in- l ll l 1 I dicators. l l l 9. Primary la. Check l D la. Comparison of six l ll Coolant l l l separate channel l ll Pressure l l l indicators. l ll lb. Test l M lb. Trip channel, in- l ll l l [ ternal test signal l ll l l l to verify trips andl ll l l l alarms. l ll lc. Calibrate l R lc. Known pressure ap- l ll l l l plied to sensor. l ll l l l Internal test sig- ]

ll l l l nel to adjust tripsi

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ll l l I and indicators. l l l10. Circulator la. Check l D la. Comparison of eightl ll Inlet l l l

• separate indica- l ll Temperature l l l tors. l ll lb. Test l M lb; Trip channel, in- l

, ll l l l ternal test signal l l 'l I l l to verify trips andl lI l l l alarms. l ll lc. Calibrate l R lc. Compare each l ll l l l thermocouple output l ll l l l to an NBS traceable l ll l l l standard. Internall ll l l l test signal to ad- l ll l l l Just trips and in- l ll l l [ dicators. l

, l l11. Hot Reheat la. Test l M ' la. Reduce pressure at l ll Header l l l sensor to trip l ll Pressure l l l channel, verify l ll l l l alarms and indica- l ll l i l tions. l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-7 l TABLE 5.4-1 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF SCRAM SYSTEM (Cont'd) l ll Channel l l l l ll Description l Function IFreouency(1)] Method l l l11. (Cont'd) lb. Calibrate l R lb. Known pressure ap- l ll l l l plied at sensor to l ll l 1 l adjust trips. l l l12. Main Steam la. Test l M la. Reduce pressure at l ll Pressure l l l sensor to trip l ll l l l channel, verify l ll l l l alarms and indica- l ll l l l tions. l ll lb. Calibratel R lb. Known pressure ap- l ll l l l plied at sensor to l ll l 1 I adjust trips. l l l13. Two Loop la. Test l M la. Special test modulel ll Trouble l l l used to trip chan- l ll l l l nel by energizing l ll l l l each of four appro-l ll l l l priate pairs of l ll l l l , two-loop trouble l ll l l l relays. l ll lb. Test l R lb. Trip logic to causel ll l l l two loop trouble l ll .

I I I scram. l l l14. Plant 400 V la. Test l M la. Trip each channel l l l Power Loss l l l by applying simu- l ll l l l lated loss of vol- l ll l l l tage signal; verify l ll l l l alarms and indica- l ll l l l tions. l l l15. High Reactor la. Check l D la. Comparison of threel

, ll Building l l l separate channel l l ll Temperature l l l indicators. l (Pipe Cavity) 16. Test kM lb. Trip channel, v,eri-l l l -l I ll fy alarms and indi-l l l l l l l l l cations. Internal l l l l l l l test signal to l l ll l l l verify trips and l

! ll I I I ala rms . l l

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-8 l TABLE 5.4-1 1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS. AND TESTING l OF SCRAM SYSTEM (Cont'd)

I ll Channel l l l l ll Description l Function IFrequency(1)l Method l l l15. (Cont'd) Ic. Calibrate l R lc. Compare each l ll l l l thermocouple output l ll l l l to an NBS traceable l ll l l l standard to adjust l ll l l l temperature trip l ll l l l point. l t

l NOTE 1: D - Daily when in use.

l M - Monthly.

l R - Once per refueling cycle.

[ P - Prior to each startup, if not done the previous week.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-9 l TABLE 5.4-2 l MINIMUM FREQUENCIES FOR CHECKS. CALIBRATIONS, AND TESTING l OF LOOP SHUTDOWN SYSTEM I

ll Channel l l l l ll Descriotion i Function IFrequency(1)! Method l ll 1. Steam Pipe la. Check l D la. Comparison of l ll Rupture (Pipe l l l separate ultrasonici l1 Cavity) l l l channel indicators /l ll l l l loop. l ll lb. Test l M lb. Pulse test one l ll l l l temperature and l ll l l l pressure channel l ll l l l with another temp- l ll l l l erature and pres- l ll l l l sure channel l ll l l l tripped, while l ll l l l simultaneously l ll l l l having two ultra- l ll l l 1 sonic channels l ll l l l tripped. l ll 1 l l 1 l l lc. l lc. DELETED l ll l 1 l l l l ld. Test l M ld. Pressure switch l l l l l l actuated by pres- l l l l l l sure applied at l ll l l 1 sensor. I ll le. Test l M le. Temperature switch l ll l l l actuated by heat l ll -

l l l applied at sensor. l ll [f. Test l M lf. Internal test sig- l ll l l l nal to adjust l ll l l l ultrasonic trip. I ll lg. Test l M lg. Trip test signal l ll l l l solenoid valves to l l1 l l l verify loop l ll l l l integrity. l ll lh. Calibrate l R lh. Known pressure l ll l l l applied at sensor l ll l l l to verify response.l ll l1. Calibratel R l1. Known temperature l ll l l l applied at sensor l ll l l l to verify response l l l lj. Calibrate l R lj. Known sound applied l ll '

l l l at sensor to verify l ll I I I response. l l l 2. Steam Pipe la. Check l D la. Comparison of l t

ll Rupture (Underl l l separate ultrasonici l

ll PCRV) l l l channel indicators /l

-l l I l l loop. l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-10 l TABLE 5.4-2 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS. AND TESTING l OF LOOP SHUTDOWN SYSTEM (Cont'd) l l l Channel l l l l l l Description i Function IFrequency(1)I Method l l l 2. (Cont'd) lb. Test i M lb. Pulse test one l ll l l l temperature and l l l l l l pressure channel l ll l l l with another l l l l l l temperature and l ll l l l pressure channel l ll l l l tripped, while l ll l l l simultaneously l ll l l l having two l ll l l l ultrasonic channels l ll l l l tripped. l ll l l l l ll lc. l lc. DELETED l l l l l l l ll ld. Test l M ld. Pressure switch l ll l l l actuated by l l l l l l pressure applied atl ll l l l sensor. l ll le. Test l M le. Temperature switch l l l l l l actuated by heat l l l l l l applied at sensor. I ll lf. Test l M If. Internal test l ll l l l signal to adjust i ll l l l ultrasonic trip. l l l Ig. Test l M lg. Trip test signal l l l l l l solenoid valves ,l ll l l l to verify loop l ll l l l i nteg ri ty.. l ll lh. Calibrate l R lh. Known pressure l l l l l l applied at sensor l ll l l l to verify response.l ll l1. Calibrate l R l1. Known temperature l ll l l l applied at sensor l ll l l l to verify response.l ll lj. Calibratel R lj. Known sound applied l

-l l l l l at detector to l ll l l l verify response. l l l 3. Circulator 1A la. Test l M la. Pulse test and l l l and IB Trippedi l l verify proper indi-l ll l l l cations. l l l lb. Test l R lb. Trip both circula- l ll l l l tors to test loop l ll l l l shutdown. l l l 4. Circulator 1C la. Test 1 M la. Pulse test and l ll and 10 Tripped [ l l verify proper indi-l ll l l l cations. l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-11 l TABLE 5.4-2 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF LOOP SHUTOOWN SYSTEM (Cont'd)

I ll Channel l l l l ll Description l Function IFrecuency(1)l Method l l l 4. (Cont'd) lb. Test l R lb. Trip both circula- l ll l l l tors to test loop l ll l l l shutdown. l l l 5. Steam la. Test l M la. Pressure switches l ll Generator l l l actuated by l ll Penetration l l l pressure applied. [

ll Pressure Ib. Test l M lb. Pulse test each l ll l l l channel with l ll l l l another chan,el l ll l l l tripped and verify l ll l _l l proper indications.]

ll lc. Calibrate l R lc. Known pressure l ll l l l applied at sensor l ll l l l to adjust trip. l l l 6. Reheat Headerla. Cneck l D la. Comparison of threel ll Activity l l l separate indicators l ll l l l in each loop. l ll [b. Test l M lb. Pulse test each l ll l l l channel with l ll l l l another channel [

ll l l l tripped and verify [

ll l l l proper indications.l ll lc. Calibratel R lc. Expose sensor to l ll l l l known radiation l ll l l l source and adjust l ll l l l trips and indica- l ll l l l tors. l l l 7. Superheat la. Check l D la. Comparison of threel ll Header l l l separate temper- l ll Temperature l l l ature indicator's l ll l l l per loop l ll lb. Check l D . [b. Comparison of threel ll l l l separate temper- l l l l l l ature differential l ll l l l indicators. l ll lc. Test l M lc. Pulse test one l ll l l l channel with l ll l l l another channel l ll l l l tripped and verify l ll l l l oroper indications.l

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Page 5.4-12 l TABLE 5.4-2 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF LOOP SHUTDOWN SYSTEM (Cont'd) 1.

ll Channel l l l l ll Description l Function IFrecuency(1)l Method l l l 7. (Cont'd) Id. Calibrate l R ld. Compare each l ll l l l thermocouple output l '

ll l l l to an NBS traceable l ll l l l standard. Internal l ll l l l test signal to l ll l l l adjust trips and l ll l l indicators. l l l 8. Primary la. Test l M a. Trip each channel, l ll Coolant l l l verify proper indi-l ll Moisture (Low l l l cations. l ll Level Ib. Test l M lb. Trip each channel, l ll Channels) l l l pulse test other l ll l l l loop to check loop l ll l l l identification. l l l 9. Primary la. Test l M la. Pulse test one l ll Coolant l l l channel with l ll Pressure l l l another channel l ll l l l tripped and verify l ll l l l proper indications,l ll l l l both channels. l s

l NOTE 1: D - Daily when in use.

l M - Monthly.

l R - Once per refueling cycle.

l Prior to each startup, if not done the previous week.

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Page 5.4-13 l TABLE 5.4-3 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF CIRCULATOR TRIP SYSTEM i

ll Channel l l l l l l Description i Function IFrequency(I)l Method [

l l 1. Circulator la. Check l 0 la. Comparison of six l ll Speed, Steam, l l l separate speed l l l and Water l l l indications per l l l l l l circulator. l ll lb. Test l M lb. Internal test sig- l ll l l l nal to verify trip l ll l l l setting and indi- l ll l l l cators. l ll lc. Test l M lc. Pulse test one l i l I i l channel with l l l l l l another channel l l l l l l tripped, and verify l l l l l l proper indications.l ll [d. Calibratel R ld. Known pulse l l l l l l frequency applied l l l l l l at sensor to adjust l ll l l [ trips and l ll l l l indicators. l l l 2. Feedwater Flowla. Check l D la. Comparison of six l ll l l l separate indicators l ll l l l per loop. l ll [b. Test l M lb. Internal test l ll l l l signal to verify l l l l l l trip setting and I l l l l l indications. l ll lc. Test l M lc. Pulse test one l ll l l l channel with l ll l l l another channel l ll l l l tripped, and verify l

, ll l l l proper indications.l l l 'l ld. Calibrate l R [d. Apply known AP at l l l l l l flow transmitter. l l l l l l Internal test sig- l l l l l l nal to adjust trips l ll l l l and indicators. [

t l l 3. Circulator la. Check l D ja. Comparison of threel l ll Bearing Water l l l separate indi- l ll Pressure l l l cators/ circulator. l ll lb. Test i M lb. Pulse test one l l l l l l channel with l l l l l l another channel l

, l l l l l tripped and verify l l l l l l proper indications.l l

l ll lc. Calibrate l R lc. Known pressure l l ll l l l applied to adjust l ll l l l trip setting. l

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Page 5.4-14 l TABLE 5.4-3 l MINIMUM FREQUENCIES FOR Ci!ECKS, CALIBRATIONS, AND TESTING l OF CIRCULATOR TRIP SYSTEM (Cont'd) l ll Channel l l l l ll Description i Function IFrequency(1)I Method l l l 4. Circulator la. Test l M la. Pressure switches l ll Penetration l l l actuated by l ll Pressure l l l pressure applied. l ll lb. Test l M lb. Pulse test one l ll l l l channel with l ll l l l another channel l ll l l l tripped and verify l ll l l l proper indications.l ll lc. Calibrate l R lc. Known pressure l ll l l l applied at sensor l ll l l l to adjust trip l ll l l I setting. l l l 5. Circulator la. Check l D la. Comparison of threel ll Drain Pressurel l l separate indi- l ll l l l cators/ circulator. l ll lb. Test l M lb. Pulse test one l ll l l l channel with l ll l l l another channel l ll l l l tripped and verify l ll l l l proper indications.l ll lc. Calibrate l R lc. Known pressure l ll l l l applied at sensor l ll l l l to adjust trip l l l l l I setting. l l l 6. Tirculator la. Check l D la. Comparison of threel ll Seal l l l separate indi- l l ll Malfunction l l l cators/ circulator. l l ll lb. Test l M lb. Pulse test one l ll l l l channel with l ll l l l another channel l

, l.l l l l tripped and verify l l ll l l l proper indications.l ll lc. Calibrate l R lc. Known pressure l

! ll l l l applied at sensor i

ll l l l to adjust trip l ,

Il l l l setting. l l

l l

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Page 5.4-15 l TABLE 5.4-3 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF CIRCULATOR TRIP SYSTEM (Cont'd) l ll Channel l l l l ll Description I Function lFreouency(1)I Method l

, l l 7. Circulator la. Test l R la. Trip steam turbine l ll Trip (Manual) l l l drives. Verify l ll l l l water turbine l ll l l 1 automatic start. l l

l NOTE 1: D - Daily when in use.

l M - Monthly.

I R - Once per refueling cycle.

l P - Prior to each startup, if not done the previous week.

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Page 5.4-16 l . TABLE 5.4-4 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF ROD WITHDRAWAL PROHIBIT SYSTEM i

ll Channel l l l l ll Description l Function IFrequency(1)l Method l l l 1. Startup la. Check l D la. Comparison of two l ll Channel l l l separate channel l ll l l l indicators. l ll lb. Test l P lb. Internal test l ll l l l signal to verify l ll l l l all trips and l ll l l l alarms. l ll lc. Calibrate l R lc. The internal test l ll l l l signal shall be l ll l l l checked and cali- l ll l l l brated to assure l ll l l l that its output is l ll l l l in accordance with l ll l l l the design require-l ll l l l ments. This shall l ll l l l be done after com- l ll l l l pleting the exter- l ll l l l nal test signal l ll l l l procedure by l ll l l l checking the output l l l l l l indication when l ll l l l turning the l ll l l l internal test l ll I I I signal switch. l l l 2. 11near Channella. Check l D la. Comparison of six l ll l- l l separate level l ll l l l indicators. l ll lb. Test l M lb. Internal test l ll l l l signal to verify l

ll l l l trips and alarms. l l~l lc. Calibrate l D lc. Channel adjusted tel ll l l l agree with heat l ll l .l l balance l ll l l l calculation. l l l ld. Calibrate l R ld. Internal test l ll l l l signals to adjust l ll l l l trips and l i ll l t i indications. l l l 3. Wide Range la. Check l 0 la. Comparison of threel ll Power Channel l l l separate l l l l l l l indicators. l ll lb. Test l P lb. Internal test l ll l l l signals to verify l ll l l l trips and alarms. l

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Page 5.4-17 -

l TABLE 5.4-4 l MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TESTING l OF R0D WITHDRAWAL PROHIBIT SYSTEM (Cont'd) l ll Channel l l l l ll Description l Function IFrequency(1)I Method l l l 3. (Cont'd) lc. Calibra- l M lc. Channel adjusted tol ll l tion l l agree with heat l ll l l l balance calcu- l ll l l l lation. l ll ld. Calibra- l R ld. Internal test sig- l ll l tion l l nals to adjust l ll l l l trips and indi- l ll l l l cations. l l l 4. Multiple Rod la. Test l P ja. Attempt two rod l ll Pair l l l pair withdrawal. l ll Withdrawal lb. Check l R lb. Simulate current l ll l l l through sensor to l ll l l l verify trip and l ll l l l alarms. l I

l NOTE 1: D - Daily when in use.

l M - Monthly.

l R - Once per refueling cycle.

l P - Prior to each startup, if not done the previous week.

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Page 5.4-18 Specification SR 5.4.2 - Control Room Smoke Detector The control room smoke detectors and alarms will be functionally tested once per year.

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 3 + 75 F, a calibration shall be 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.

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Page 5.4-19 Basis for Specification 5.4.3 The long-term thermocouple drift is estimated to be g 15"F per year and this drift was included in the measurement uncertainty of 1 50'F used to establish LCO 4.1.7. With t'hi s measurement uncertainty, a root mean square difference of g 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).

Specification SR 5.4.4 - PCRV Cooling Water System Temperature Scanner - Surveillance Requirement PCRV Cooling System temperature scanner readings shall be checked by comparison of representative liner cooling tube thermocouple outputs to their respective subheader temperatures and associated alarms tested once per month during power operation.

All thirty-six (36) outlet subheader temperature i indicators shall be calibrated annually. In addition, ninety-seven (97) liner cooling tube outlet thermocouples shall be calibrated annually.

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Page 5.4-20 Basis for Specification SR 5.4.4 The temperature scanner for the PCRV cooling system provides for continuous temperature monitoring of the outlet water temperature of each individual liner cooling tube and alarming of high outi.et temperatures.

The surveillance interval specified is sufficient to detect any drift in the output of the individual thermocouples or scanner electronics to assure the temperature limitations of the PCRV cooling system are not exceeded.

The ninety-seven (97) thermocouples shall be distributed among the thirty-six (36) subheaders so that between 16.7%

and 21.5% of the total in each subheader are calibrated each year. Thus, the maximum time between calibration of any one thermocouple, or any complete subheader, shall not exceed six (6) years. The overall percentage of thermocouples calibrated per year exceeds 18%.

l l .

The surveillance interval for calibration, combined with that for checking, assures sufficient accuracy of temperature measurement to adequately protect the PCRV l

concrete.

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Page 5.4-21 Specification SR 5.4.5 - PCRV Coolina Water System Flow Scanner - Surveillance Reauirement A PCRV Cooling System flow scanner readout shall be taken and alarns functionally checked monthly. The scanner and alarms, and six (6) subheader flow meters shall be calibrated annually.

Basis fo- Specification SR 5.4.5 The flow scanner acts as a backup to the temperature scanner and initiates no automatic protective action, only an ala m. Because a restriction or a leak in the system would develop over a period of time, the monthly interval for comparing scanner readouts is sufficient to detect any long term change in the system.

Specification SR 5.4.6 - Core Ap Indicator - Surveillance Requirement The core AP instrumentation shall be calibrated 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.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.4-22 Specification SR 5.4.7 - Control Roon Temperature -

Surveillance Requirement The control room temperature control thermostat shall be functionally tested monthly and calibrated annually.

Basis for Soetification SR 5.4.7 The surveillance interval specified for functional testing and calibration of the control room thermostat will assure its abil,ity 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 Flow Instrumentation -

Surveillance Requirement The power to flow indication shall be verified daily and shall be calibrated once per refueling cycle.

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Basis for Specification SR 5.4.8 The power to flow ratio indication is an indication of the balance between the heat generation and removal within the I

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 indicatica. A l

lack of response by this instrumentation would be noticed by the operator. Calibration of the instrumentation on a 1

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Page 5.4-23 once per refueling cycle basis is acceptable by industry standards for this type of instrumentation.

Scecification SR 5.4.9 - Area and Miscellaneous Process 4adiation 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 i

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 every six months and calibrated every two years.

Basis for Specification 5.4.10 The intervals specified for testing and calibration of the Seismic Instrumentation are recommended by the manufacturer to assure the instruments operate as intended.

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Page 5.4-24 Soecification SR 5.4.11 - PCRV Surface Temoerature Indication - Surveillance Requirement i

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 t.he 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 l Coolant Moisture Instrumentation - Surveillance Requirements

'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.

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Page 5.4-25 Specification SR 5.4.13 - 480 V Switchaear 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.

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Page 5.5-1 5.5 CONFINEMENT SYSTEM - SURVEILLANCE nE0VIREMENTS 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 building 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 nct a significant parameter as is shown in FSAR Section 6.1.4.2.

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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 on a monthly basis and calibrated annually.

The louver groups shall be individually exercisec' quarterly.

Quarterly louver testing may be performed while the reactor is in operation only if the fo' lowing 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 sy. tem is operating per Technical Specifications.

d) No radioactive gas waste reltases are in progress, nor is fuel handling being per'ormed.

j e) No airberne activity at ove background as indicated by the building activity monitors.

f) Area radiation moritors and local alarms are operable per Technical Specifications.

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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.

1) Communication shall' exist betwen 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.

Basis for Specification SR 5.5.2 l

The reactor building pressure relief device is designed to protect the building in the event that pressure in the Reactor Building exceeds the turbine building pressure by 3 inches of water. The device consists of louvers installed in a number of individual modules operated by t

mechanical linkages to pneumatic actuators (see FSAR

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Page 5.5-4 Section 6.1.3.4). The specified test frequency shall ensure the operability of the reactor building relief system.

Specification SR 5.5.3 - Reactor Buildino Exhaust Filters, Surveillance The exhaust filters in the reactor building ventilation system shall be tested as follows:

a) A l a.boratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shall be performed after 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 g90% radioactive methyl iodide removed when tested in accordance with ANSI N510-1975 (130*C, 95%

R.H.).

• Defined as any material which could reasonably be expected to interfere with the charcoal to adsorb methyl iodide, b) A halogenated hydrocarbon test shall be performed once per calender year or after each replacement of a charcoal adsorber bank or after structural maintenance on the filter housing. Halogenated hydrocarbon

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Page 5.5-5 removal by the charcoal filters shall be g 99% when conducted at normal flow conditions in accordance with the applicable portions of ANSI N510-1975.

c) The HEPA filters shall be leak tested in place once per calendar year, after each complete or partial replacement of a HEPA filter bank, or after any structural maintenance on the filter housing, using cold DOP. Cold DOP removal by the HEPA filters shall be g 99% when tested in accordance with the applicable portions of ANSI N510-1975.

d) Flow distribution across the HEPA and enarcoal filters will be tested with initial operation of the system and following any structural modification to the filter housings. Air distribution shall be demonstrated within 1 20% across the HEPA and charcoal filters when tested in accordance with ANSI N510-1975.

e) Total pressure drop across the combined HEPA filter and charcoal adsorber banks shall be less than 6" HO 2

at filter design flow 1 10%.

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 norma 1 and accident conditions of operation. The system consists of three 50%

Fort St. Vrain 01 Technica1 Specifications Amendment #

page 5.5-6 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 radiciodine to the atmosphere. Bypass leakage for the charcoal adsorbers, and particulate removal efficiency for HEPA filters are determined by halogenated hydrocarben and DOP respectively. The laboratory carbon sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. The survetlalnce 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.

Fort St. Vrain #1 Technical Sp cifications Amendment #

Page 5.5-7 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.

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Page 5.6-1 5.6 EMERGENCY POWER SYSTEMS - SURVEILLANCE REQUIRMENTS Apolicability Applies to the surveillance of the equipment supplying electrical power to the essential plant services.

Objective To establish the minimum frequency and type of surveillance for equipment supplying electric power to the plant auxiliaries to ensure that the motive power sources required to safely shut down the plant is available.

Specification SR 5.6.1 - Standby Diesel Generator Surveillance The surveillance of the standby diesel generators shall be as follows:

a) Each standby generator set will be started and loaded to at least 50% of rated full load capacity once every week. The test shall continue for at least two hours to enable the engine (s) and the generator to attain their normal operating temperature.

b) A loss of outside source of power and turbine trip shall be simulated twice annually with the interval between tests to be not less than four (4) nor greater than eight (8) months to demonstrate that the standby generators, automatic controls, and load sequencers are operable.

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Page 5.6-2 c) The diesel engine protective functions shall be calibrated annually, d) The diesel engine exhaust temperature " shutdown" and "declutch" shall be functionally tested monthly and calibrated annually.

Basis for Specification SR 5.6.1 The weekly test of the standby diesel generator is to exercise the engine by operating at design temperature and to demonstrate operating capability. These tests will allow for detection of deterioration and failure of equipment.

Tests once a year during refueling will functionally test the standby generator system.

1 Specification SR 5.6.2 - Station Battery Surveillance

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The surveillance of the station and PPS batteries shall be as follows:

a) The specific gravity and voltage of the pilot cell and temperature of adjacent cells and overall battery voltage shall be measured every week.

b) The specific gravity and voltage to the nearest 0.01 volt, temperature of every fifth cell and height of electrolyte shall be measured every three months.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.6-3 c) The station and PPS batteries shall be load-tested to partial discharge once a year during plant shutdown.

Basis for Specification SR 5.6.2 The type of battery surveillance called for in this specification has been demonstrated through experience to provide a reliable indication of a battery cell initial breakdown well before it becomes unserviceable. Since batteries will deteriorate with time, these periodic tests will avoid precipitious failure.

The manufacturer's recommendation for equalizing charge is vital to maintenance of the ampere-hour capacity of the battery. As a check upon the effectiveness of this charge, each battery will be load tested to determine its ampere-hour capacity. In addition, its voltage shall be monitored as a function of time. If a cell has deteriorated or if a connection is loose, the voltage under load will drop excessively, indicating need for replacement or maintenance.

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Page 5.7-1 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 handling and storage systems.

Specification SR 5.7.1 - Fuel Handling Machine Surveillance l

l The surveillance of the fuel handling machine will be as follows:

a) Prior to refueling, the fuel handling machine cooling l

water leak detector will be functionally tested, b) A functional test of the Fuel Handling Machine and Isolation Valve movements, interlocks, limit switches, and alarms will be performed or simulated prior to annual refueling periods.

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Page 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 zooling 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 The surveillance of the fuel storage facility will be as follows: -

a) The fuel storage facility helium pressure indicators and alarms will be calibrated and functionally tested annually.

b) The 'uel storage facility cooling system fiow indicators, and flow and temperature alarms shall be calibrated and functionally tested annually.

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Page 5.7-3 Basis for Specification SR 5.7.2 The fuel storage wells are provided for safe storage of new and irradiated fuel elements. The basic design 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.

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Page 5.8-1 5.8 RADIDACTIVE EFFLUENT DISPOSAL SYSTEMS - SURVEILLANCE REQUIREMENTS Applicability Applies to surveillance of the Radioactive Effsuent Disposal Systems.

Objective To establish the minimum frequency and type of surveillance on the equipment of the Radioactive Effluent Disposal Systems to assure that releases of radioactivity are within those specified in Section 4.8.

Specification SR 5.8.1 - Radioactive Gaseous Effluent System Surveillance The surveillance of the radioactive gaseous waste disposal system shall be as follows:

a) Automatic vent high activity blocking and transfer functions of the gaseous waste system shall be tested prior to each controlled release or once a month, whichever is more frequent.

b) Automa'ic t gaseous waste header high activity transfer to the gas waste vacuum tank shall be tested once per month.

Fort St. Vrain 01 Technical Specifications Amendment #

Page 5.8-2

. c) The gas waste header activity monitors shall be functionally tested once per month and calibrated quarterly.

d) The vent monitor system shall be functionally tested weekly, calibrated quarterly, and following maintenance on the detector system.

e) Flow recorders shall be calibrated annually, f) The vent iodine / particulate monitor filter shall be analyzed once per week.

Specification SR 5.8.2 - Radioactive Liquid Effluent System Surveillance The surveillance of the radioactive liquid waste disposal system shall be as follows:

1 a) The level alarms and pump interlocks on the two liquid waste receiver tanks and monitoring tank shall be l

tested once per year.

b) The liquid effluent discharge blocking valve shall be functionally tested prior to each release or once a month, whichever is more frequent.

c) The activity monitors of the liquid waste disposal line and the low cooling water blowdown flow switch shall be functionally tested prior to the controlled -

discharge of any liquid wastes or once a month,

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~ Fort St. Vrain #1 Technical Spscifications A endment #

Page 5.8-3 whichever is more frequent. The activity monitors shall be calibrated quarterly and following maintenance on the detector system.

Basis for Specification SR 5.8.1 and SR 5.8.2 The frequency specified above is based upon industry experience and minimal disposal requirements of the plant.

Tests prior to discharge using the installed check source mounted in the instrument will provide both a check on the calibration as well as a dynamic test of the various monitors, alarms, and protective functions.

~

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Fort St. Vrain #1 Technical Sp cifications Amendment #

Page 5.8-4 THIS PAGE INTENTIONALLY LEFT BLANK r .

1 1

I l

l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-1 5.9 ENVIRONMENTAL SURVEILLANCE - SURVEILLANCE REQUIREMENTS Acolicability Applies to sampling for environmental radioactivity in the vicinity of the plant.

Objective To establish a sampling schedule which will recognize changes in radioactivity in the environs and assure that effluent releases are kept as low as practicable and within the limits of Appendix B, Table II, 10 CFR 20.

Specification SR 5.9.1 - Environmental Radiation, Surveillance Requirements a) Gaseous Release Sampling of air, external gamma, milk, forage and crops sha De conducted in accordance with Action Guide 3 during the first three years of operation and thereafter in accordance with Table 5.9-1 and l

l ,

Table 5.9-2, as specified below:

• I l 1) If releases from the plant vent produced concentrations or exposures less than 3% of those specified in 10 CFR 20 for unrestricted areas and the general population during the previous quarter, the environmental survey shall be

conducted in accordance with Action Guide 1 for

_ the current quarter.

Fort St. Vrain (11 Technical Sptcifications Amendment #

Page 5.9-2

2) If the concentrations or exposures during the previous quarter were greater than 3% but less than 10% of those specified in 10 CFR 20 for unrestricted areas and the general population, the environmental survey shall be conducted in accordance with Action Guide 2 for the current quarter. If the samples taken under Action Guide 2 do not indicate any significant increase in environmental radioactivity, the survey shall revert to Action Guide 1.

3) If the concentrations or exposures during the previous quarter were greater than 10% of those specified in 10 CFR 20 for unrestricted areas and the general population, the environmental survey shall be conducted in accordance with Action Guide 3 for the current quarter. If the samples taken under Action Guide 3 do not indicate any significant increase in environmental radioactivity, the survey shall revert to Action Guide 2.

i l

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-3 b) Liouid Release Sampling of water and silt, potable water, and aquatic biota shall be conducted in accordance with Action Guide 3 during the first three years of operation and thereafter in accordance with Table 5.9-1 and Table 5.9-2 as specified below:

1) If the gross beta gamma activity released from the station during the previous quarter was less than 3% of MPC,, the environmental survey ' hall be conducted in accordance with Action Guide 1 for the current quarter.

2) If the gross beta gamma activity released from the station during the previous quarter was greater than 3% MPC, but l e.s s than 10% MPC,, the environmental survey shall be conducted in accordance with Action Guide 2 for the current quarter. If the samples taken under Action Guide 2 do not indicate any significant increase in environmental radioactivity, the survey shall revert to Action Guide 1.

3) If the gross beta gamma activity released from the station during the previous quarter was greater than 10% of MPC,, the environmental survey shall be conducted in accordance with Action Guide 3 for the current quarter. If samples taken under

Fort St. Vrain #1 Technical Specifications Amendment # )

Page 5.9-4 Action Guide 3 do not indicate any significant increase in environmental radioactivity, the survey shall revert to Action Guide 2.

4) Results of the aquatic biota sampling program will be reviewed with appropriate agencies after one year of sampling following commercial operation to establish the required extent of future sampling.

Basis for Specification SR 5.9.1 Programs for monitoring the environment in the vicinity of Ft. St. Vrain will be conducted by Colorado State University under a contract from Public Service Company of Colorado (the licensee) and by the Colorado Department of Health with assistance by the Environmental Protection Agency's Western Environmental Radiation Laboratory. The Colorado Department of Health program includes sampling and analyses of air, water and milk. In addition, they will have special programs for sampling tritium in surface water and atmospheric concentrations of "Kr.

A preoperational radiological monitoring program has been conducted since March 1969. This program has established an adequate baseline to which operational environmental data can be compared.

% operational environmental surveillance program will be maintained on a continuous basis to verify that projected and anticipated concentrations of radioactive materials in t

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-5 the environment are not exceeded. The extent to which environmental monitoring programs are conducted should depend on the actual release of radioactivity into the environment.

When the quantity of material released is small the environmental monitoring program may be minimal. For larger releases of radioactive material, a more comprehensive environmental monitoring program is appropriate. The surveillance levels specified in Action Guide 1 and Action Guide 2 are comparable to intake Range 1 and Range 2 as given in Federal Ra.diation Council Report No. 2.

The operational surveillance program provides for collection and analyses for samples within an area extending to a twenty mile radius from the reactor. A concentrated area of sampling within a one mile radius is designated t'he facility zone; the area from one to ten miles is called the adjacent zone, and the reference zone is from ten to twenty miles.

Table 5.9-2 gives the location of each sampling station and the types of samples to be taken at each station.

Table 5.9-3 gives the minimum sensitivities for the various analyses and/or measurements made on the samples.

Figure 5.9-1 and Figure 5.9-2 indicate the sample station locations.

l L -

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-6 The aquatic biota sampling program is a supplemental part of the Environmental Surveillance Monitoring Program and was not a factor in the design of the basic sampling program which was designed on the basis of critical pathways to man.

It is felt that sampling during the preoperational phase and for a representative period following operation will adequately demonstrate any potential effect of plant operation on aquatic biota. Therefore, it is planned that the results of the aquatic biota sampling program will be reviewed with representatives from interested agencies such as the Bureau of Sport Fisheries and Wildlife and the Colorado Game, Fish and Parks Department following one year of commercial operation to establish the extent to which sampling should be continued beyond that point.

For further information, see FSAR Section 2.7.

O f

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-7 l TABLE 5.9-1 l ENVIRONMENTAL RADIATION SURVEILLANCE PROGRAM SCHEDULE l

ll l SAMPLING FREQUENCIES AND ANALYSES BY ACTION l l l EXPOSURE ROUTES ORI LEVELS (BASED UPON ACTUAL EMISSIONS AS PERCEN- l l l MEDIA & SAMPLE l TAGES OF RELEASE RATES AUTHORIZED BY 10 CFR 20)l l l TYPES (NUMBER OF l Action Level 1:l Action Level 2:l Action Level 3: l ll LOCATIONS) I Less Than 3% l 3% to 10% IGreater Than 10%l ll l Average mR/ day l Average mR/ day l Average mR/ day l l l EXTERNAL EXPOSURE l determined by l determined by l determined by l ll l QUARTERLY cumu-l QUARTERLY cumu-[ MONTHLY analysis l l lTLD Chips (36 Lo- llative expo- llative expo- lof all TLD's. l l l cations) Isures; collec- lsures; collec- l l ll l tion and analy-l tion and analy-l l ll l sis in rotation l sis in rotation l l ll lof 1/3 of all lof 1/3 of all l l ll lTLD's MONTHLY. ITLD's MONTHLY. l l ll l l l l ll ATMOSPHERE l Gross beta, lSame as for l Gross alpha and l ll levery filter, l Level 1, plus l beta, every fil-l l l Membrane Filters l WEEKLY; gamma l gross alpha on lter; gamma spec-l l jfor Particulates; l spectrum of lone weekly set ltrum of filter l l l Charcoal Car- l filter and lof filters, land cartridge l l ltridges for l cartridge com- l MONTHLY. l composites, all l l l Iodine. (7 Loca- lposites, l l WEEKLY. l l ltions) IMONTHLY. l l l ll l l 1 l l l Tritium Oxide (2 l Specific activ-ISpecific activ-l Specific activ- l l l Locations: F1 andlity of tritium lity of tritium lity of tritium l l lF4) ~ lin atmospheric lin atmospheric lin atmospheric l ll l water vapor by l water vapor by l water vapor by l

, , ll l passive absorp-l passive absorp-l passive absorp- l l ll l tion and liquid l tion and liquid l tion and liquid l l ll l scintillation l scintillation l scintillation l ll l counting, l counting, l counting, l ll IQUARTERLY. IMONTHLY. IWEEKLY. l i l'l l l l l l

ll WATER l Gross beta, IGross beta, l Gross beta, l ll l tritium, and l tritium, and l tritium, and l l l Potable Water (2 l gamma spectrum [ gamma spectrum l gamma spectrum l l l Locations) lanalyses; l analyses; l analyses; I ll l facility area l facility area l facility area l ll land nearest land nearest land nearest off-l ll loff-site supply loff-site supply l site supply l ll l(shallow wells l(shallow wells l(shallow wells l ll lat town of Gil-lat town of Gil-lat town of Gil- l ll l crest, 6 miles l crest, 6 miles ] crest, 6 miles l ll lnorthe st), l northeast), l northeast)', l ll lQUARTELLY. l MONTHLY. l MONTHLY, plus l ll l l lSr 89 and 90 l ll l l lanalyses. l

1 Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-8 l TABLE 5.9-1 l ENVIRONMENTAL RADIATION SURVEILLANCE PROGRAM SCHEDULE (Cont'd) l ll l SAMPLING FREQUENCIES AND ANALYSES BY ACTION l l l EXPOSURE ROUTES ORI LEVELS (BASED UPON ACTUAL EMISSIONS AS PERCEN- l ll MEDIA & SAMPLE l TAGES OF RELEASE RATES AUTHORIZED BY 10 CFR 20)l l l TYPES (NUMBER OF l Action Level 1:l Action Level 2:lAction Level 3: l ll LOCATIONS) l Less Than 3% 1 3% to 10% l Greater Than 10%l ll l 1 l l l l Precipitation (2 lNo collection l Gross beta, l Gross beta, l l l Locations: F1 andlor analyses of l MONTHLY. l tritium, and l l lF4) l precipitation l lSr 89 and 90, l ll lat Level 1. l l MONTHLY; gamma l ll l l l spectrum of com-l ll l l lposite, l ll I i 10VARTERLY. __ _ l ll 1 I l l l l Surface Water and l Gross beta, lSame as for lSame as for l l l Silt (7 Locations)l tritium, and l Level 1, but l Level 2, plus l ll Igamma spectrum,lMONTHLY. lSr 89 and 90 l ll l QUARTERLY. l l analyses, l ll l l l MONTHLY. l ll 1 l l l ll FOOD CHAINS ITritium and l Tritium and ITritium and l ll l gamma spectrum l gamma spectrum l gamma spectrum l l l Soil, Forage, and l analyses of l analyses of l analyses of l l l Crops (13 Loca- l forage and l forage and l forage and crops l l ltions) l crops in the lcrops in the lin the most l ll lmost probable lmost probable [ probable routes l ll l routes to man. l routes to man. lto man. Same asl ll l QUARTERLY, as l MONTHLY during l Level 2, plus l ll lavailable l growing season lSr 89 and 90, l ll l(i.e., spring, l(i.e., approxi-lplus concurrent l ll l summer, and lmately April tolsoil samples l ll l fall). 10ctober). l analyzed for thel ll l l lsame nuclides. l l l 'l I l l MONTHLY during l l ll 1 lorowino season. l l ll l l l l

' l l Beef Cattle (1 Lo-lN) analysis of [ Gamma spectrum,lSame as for i l l cation: Facility lFeef at Level l tritium, and llevel 2, plus l l l Area) l1. lSr 89 and 90 l total body count l ll l l analyses on onelof 2 to 4 ani- l ll l l meat sample lmals from l ll l [from beef l facility area, l ll l l rased in ] QUARTERLY. l ll l Ifacility area; l l ll l l ANNUALLY, at l l ll l lend of grazing l l ll l l season (i.e., [ l ll I Ilate fall). l l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-9 l TABLE 5.9-1 l ENVIRONMENTAL RADIATION SURVEILLANCE PROGRAM SCHEDULE (Cont'd) l ll l SAMPLING FREQUENCIES AND ANALYSES BY ACTION l l l EXPOSURE ROUTES ORl LEVELS (BASED UPON ACTUAL EMISSIONS AS PERCEN- l ll MEDIA & SAMPLE l TAGES OF RELEASE RATES AUTHORIZED BY 10 CFR 20)l l l TYPES (NUMBER OF l Action Level 1:l Action Level 2:l Action Level 3: l ll LOCATIONS) I Less Than 3% l 3% to 10% l Greater Than 10%l ll l l l l l l Milk (13 Loca- l Tritium, gamma l Tritium, gamma l Tritium, gamma l l ltions) l spectrum, and [ spectrum, and lspectru, and l ll lSr 89 and 90 lSr 89 and 90 lSr 89 and 90 l l1 l analyses on l analyses on l analyses on com-l ll l composite. l composite. lposite. Same asl ll l Facility area l Facility, adja-lfor Level 2, buti ll lonly, l cent and ref- l WEEKLY during l ll l QUARTERLY. lerance areas; l pasture season, l ll l l MONTHLY during lotherwise l ll l l pasture season,l MONTHLY. l ll l lotherwise l l ll 1 IQUARTERLY. l l ll 1 l l l ll AQUATIC BIOTA l Gross beta and l Gross beta and l Gross beta and l ll [ gamma spectrum l gamma spectrum l gamma spectrum l l l2 Streams, Above l analyses of l analyses of l analyses of com-l l land Below Dis- l composites of l composites of lposites of each l l l charge Point leach of 4 cate-leach of 4 cate-lof 4 categories:l ll lgories: Igories: l1) suspended l ll l1) suspended l1) suspended l organisms, l ll lorganisms, l organisms, l2) benthic l ll

~

12) benthic l2) benthic l organisms, l ll lorganisms, l organisms, l3) vascular I il 13) vascular l3) vascular l plants, and l ll l plants, and l plants, and l4) fish. Same l ll l4) fish. l4) fish. las for Level 2, l ll lQUARTERLV, as l MONTHLY during lplus Sr 89 and l l'l latailable. l summer; other- 190 analyses. l ll l l wise QUARTERLY,l l ll l las available. l l

- - _ , - ,-. ,m__

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-10 l TABLE 5.9-2 l SAMPLE TYPES AND LOCATIONS FOR ENVIRONMENTAL SURVEILLANCE IN THE l VICINITY OF THE FORT ST. VRAIN NUCLEAR GENERATING STATION l (Revised April 1971) l ll 1 l l l ll l Media Sampled At Location l l l ll l l l l l l l Compass l l l l LOC.I l l l l l lDirec- l Description of Sampling l b c l lI .D. l AirlM&FITLDlH,0l SEDI AQB I tion i Location g l l F-1 l

• l l*l l l l N l Potato cellar at W end of l ll l l l l l l 1 l farm site just across creek l Il I l l l l l l lN of reactor; TLD on pole atl ll l l I I I I I INE corner. l l lF-2 l

• l l*I l l l E lA cabin .3 mile E and l ll l l l l l l l 1.4 mile N of site #1. l l lF-3 l

• l l*l l l l SE l0ld dairy barn on Robert l ll l l l l l l l l Bruce farm 1/2 mile SE of l II l l l l l l l Ireactor; TLD on first pole Nl ll l l 1 I I I I lof driveway. l l [F-4 l

• l l*l l l l S ISmall shed behind the Maul l ll l l l l l l l [ farmhouse 1/2 mile SW of re-l ll l l l l l l 1 lactor; TLD on pole W of l ll 1 I l I l l I lhouse. l l [F-7 l l l*l l l l NE l.3 mile E of #1; on the polel ll l l l l l l l lby gate where road turns S. l l lF-8 l l l "l l l l NE l.3 mile S of #7; on second l ll l l 1 l  ! l l Ipole S of oate or, hill. l l lF-9 l l l*l l l l SE l.2 mile S and .3 mile W of l ll l l l l l l l l#3; on 7th pole W of corner.l l lF-101 l l*l l l l S lOn 8th pole W of #9; di- l l l l l l l l l l lrectly across road from pumpi ll l l 1 l l 1 I thouse. [ ,

Il l l l l l l l l l l lF-111 l l *I I I I S 10n 3rd pole W of #10. l l 'l l I I i i i l I l l lF-121 l l *I I l 1 SW 10n 4th pole N of #4. l t

l1 I I l l l l~ l I I l lF-131 l l*l l l l SW 10n 5th pole N of #12. l l lF-14l l l*l l l l NW 15.5 miles S and .9 mile E ofl Il l l l l l l l lJohnstown. On pole at l ll l l l l l l l Icorner in road. l l IF-441

• l l*l l l l N lTed Horst farm, due N of re-l l ll l l l l l l l l actor site. l l b - With the exception of the winter months.

l c - Sampling location description provided for information only. Lo-l cations shall be changed as conditions warrant.

l

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-11 1 TABLE 5.9-2 i SAMPLE TYPES AND LOCATIONS FOR ENVIRONMENTAL SURVEILLANCE IN THE l VICINITY OF THE FORT ST. VRAIN NUCLEAR GENERATING STATION i (Revised April, 1971) (Cont'd) i Ii l i l l l l l Media Sampled At Locationi l l ll l l l l l l l Compass l l l I LOC. l l l l l l lDirec- l Description of Sampling l b C l lI.D.lAirlM&FITLDlH,0ISEDIAQB I tion I Location l 1 lA-5 l

• l

• l

• l l l l NE l Air sampler at Lloyd Rumsey I ll l l l l l l l Ifarm. TLD on 1st pole N of l ll l l l l l l l l Ernest Wolfrum farm. Milk I ll l l l l l l l land forage at Floyd Nicols l ll I I l I l 1 I Ifarm. I l lA-6 l

• l

• l

• l  !

l l SE l Calving shed on the Clifton I ll 1 l l l l l lWissler dairy farm. TLD on [

ll l l l l l l 1 list pole E of drivevay. l l lA-15l l*l*l l l 1 NW l Paul Bader dairy, 4.2 miles I i l l I l l l l l lW of Hwy 87 on Hwy 60; TLD l ll l l l l l l l lon pole on W side of drive- l ll l l l 1 I I l lway.

l lA-18] l l*l l l l N l3 miles S and 1.5 milesl W ofl Il l l l 1 l l l l#19, on SE pole of intersec-l ll l 1 l  ! l l l l tion, across street from oldl ll I I I I -l l I Ischool. l l lA-27l l l*l l l l NW l1 mile E of Hwy 87 on Hwy l l1 l l l l l l [ 156, then 1 mile S on 1st l ll l l l l l l l Ipole on NE side of corner. l l l A-281 l*l*l l l l NW l Norman Carlson diary, 2 l l l l l l l l l l Imiles E of Hwy 87 on Hwy 60 l l l l l l l l l l land 1 mile S. TLD on pole l Il l l 1 l l 1 I ldirectly W of driveway. l l IA-29l l l*l l l l N l3 miles S and 1.6 miles E ofl 1I l l 1 1 I I I IJohnstown, on E pole on S l l 'l 1 1 I I I i 1 Iside of road. .

I I lA-301 l l*l l l l NE l1 mile S on Hwy 256 on Hwy l ll l l I l l l l 160 on middle pole at NE sidel ll l l l l l l l Iside of intersection. l l lA-31l l*l*l l l l E l Oliver Loren dairy,1.5 l ll l l l l l l l Imiles S of Peckham; TLD on l ll 1 1 I I I I I list pole S of driveway. l l IA-32l l l*l l l l E l2.5 miles S and 2 miles W ofl l l l l l l l l l l#31, on NW pole at intersec-1 ll l 1 I I I l I ition. [

l b - With the exception of the winter months.

I c - Sampling location description provided for information only. Lo-l cations shall be changed as conditions warrant.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-12 l TABLE 5.9-2 l SAMPLE TYPES AND LOCATIONS FOR ENVIkONMENTAL SURVEILLANCE IN THE l VICINITY OF THE FORT ST. VRAIN NUCLEAR GENERATING STATION l (Revised April, 1971) (Cont'd) l l i i l i l ll l Media Sampled At Locationi l l

!l l l l l l l l Compass] l l l LOC.l l l l l l lDirec- l Description of Sampling l b

l lI.D. l AirlM&FITLDlH,0lSEDI AQB I tion i Location' l l lA-33l l*l*l l l l E l Miller Brothers dairy, .2 l ll l l l l l l l l mile S of junction of Hwy's l ll l l l l l l l l66 and 85 and .5 mile E; TLDl ll l l l l l l 1 Ion 1st pole W of driveway. l I lA-34l l l*l l l l 5 l1 mile E of junction of l ll l l l l l l l l Hwy's 87 and 254 on SW pole l ll l l l l l l l lat intersection. l l lA-35l l l*l l l l SW [2 miles E of Hwy 87 on Hwy l ll l l l l l l l 166 on SW pole at intersec- l ll l l 1 I I l l Ition. l l lA-36l l*l *l l l lW l Milk and forage at Marvin l ll l l l l l l l lCoonts farm. TLD at Phillipl

. l l l 1 I I I I I IBall dairy. l l lR-16l l l*l l l l NW l Going S of Hwy 87, on pole l ll l l l l l l l lbeside yield sign on Love- l ll l 1 I I I I I Iland off-ramp. l l lR-17l 1*l"l l l l N [Ed Borensen dairy, 8 miles El l -l l l l l l l l lof Hwy 87 on Hwy'34, then .3l ll l l l l l l l l mile S; TLD on pole in frontl l l l l l l l l l lof trailer house on W side l ll l l l l l l l lof road, i.e., 10 poles froml ll 1 I I I I I I IHwy 34. l l lR-191 l l*l l l l N l4 mi.les E of #17 on Hwy 34, l ll l l l l l l l lon pole at SW corner of in- l ll 1 l l l l l l ltersection (golf course l.

I 'l l 1 l l l l l lacross street from pole). l l lR-20l l*l*l l l l NE l Wally Kaufman diary, .5 milel ll l l l l l l l lE of LaSalle and 1.6 miles l l1 l l l l 1 1 1 lS; TLD on pole W of dairy l ll l l l l l l l lbarn on W side of road. l l lR-21l l l*l l l l E l2.5 miles S and 2 miles E ofi Il I l l l l l l l#20, on SW pole at intersec-l l l l I l 1 l I I Ition. l l lR-22l l*1*l l l l SE lHagan Brothers Dairy, 4.3 l ll l l l l l l l l miles E of Ione; TLD on 1st l ll l 1 I I i i l Ipole E of driveway. l l b - With the exception of the winter months.

l c - Sampling location description provided for information only. Lo-l cations shall be changed as conditions warrant.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-13 l TABLE 5.9-2 l SAMPLE TYPES AND LOCATIONS FOR ENVIRONMENTAL SURVEILLANCE IN THE l VICINITY OF THE FORT ST. VRAIN NUCLEAR GENERATING STATION l (Revised April, 1971) (Cont'd) l l1 1 I I I ll l Media Sampled At Location l l l ll l l l l l l l Compass l l l l LOC.l l l l l l lDirec- l Description of Sampling l c

l !I.D.lAirlM&FITLDlH,0lSEDIAQBD l tion I location g i lh '23l l*l*l l -l l S l Ernest Terry dairy, 2.3 l ll l l l l l l l Imiles W of Ft. Lupton on Hwyl ll l l l l l l 152; TLD on 1st pole W of l ll n. I I I I I I idriveway. l l lR-24l l l*l l l l SW l Going S on Hwy 87, on pole l ll l l l l l l l lby stop sign on Frederick / l ll l l l 1 I I I IDacono off-ramp. l l lR-25l l*l*l l l [ SW l Milk and forage at Angelo l ll l l l l l l l lBendegna farm on Oxford l ll l l l l l l l l Road. TLD near Paul Knutsonl ll l l l l l l I tfarm. l l lR-26l l l*l l l lW 14.7 miles N of Hwy 66 on Hwyl ll l l l l l l l l287 on SE pole at intersec- l ll 1 I I I I I I ltion. l l lE-38l l l l*l*l* l NE lGoosequill Pond, .3 mile E l ll l l l 1 l 1 land .4 mile N of site #1. l ll l l l l l l l l l lE-411 I I I*I*I lW ISlouch. Due W of reactor. l l lU-42l l l l*l*l* I SW lSt. Vrain from bridge; .7 l ll 1 I I I I I I imile W of site #4. l l lU-43l l l l*[*l* l NE l South Platte, 1 mile up- l ll l l l l l l l l stream from Goosequill dis- l ll 1 I l l l I l icharge. l l lU-42l l l l l l* l E lSt. Vrain at pumping station l ll l l l l l 1 I idue E of reactor. l l'lD-451 l l l

• l *.l

• l NW lSt. Vrain Creek S of bridge l ll 1 l l l l l l Inear cauging station. l l lD-39l l l l*l l l NE lIn town of Gilcrest, Co-op l ll l 1 l l l l l Istore tao water. l l lD-40l l l l*l*l* l N l1/2 mile below junction of l ll l l l l I l i Ithe St. Vrain and Platte. l l lD-371 l l l*l*l l NE l Lower Latham Reservoir; 2.5 l ll l l l l 1 I I Imiles E of LaSalle. l l lD-43l l l l l l* l N ISouth Platte 1/2 mile down- l ll l l l l l l l l stream from Goosequill dis- l ll 1 I I I I i 1 icharge. l l b - With the exception of the winter months.

i l c - Sampling location description provided for information only. Lo-i

! cations shall be changed as conditions warrant.

Fort St. Waia 01 Technical Specifications Amendment #

Page 5.9-14

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%,py,.% (8'y~g hi, F"igure 5 9-1 Sampling stations within the facility area (on site)

Revised 5-6-71.

a

l Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-15

. l nyse00ft w

..I .

l ,

j7 9 -

gm 37

• h 29

[. 20 wer Lotham Reservoir l /

e O" o.- .. , ig@ @

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{~ f Figure 5.9-2 Sampling stations in the adjacent and reference areas (off site). Revised S-6-71.

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-16 l TABLE 5.9-3 l SENSITIVITIES FOR ENVIRONMENTAL RADIATION MEASUREMENTS l

ll l l Counting l Analytical l M.D.A. l l l Media l Isotope l Efficiency l Technique i 99% C.L. l ll l l l l l l l Air l Gross a l 48%  ! Int. Prop. Ctr. l (a) l ll l l l l l ll l Gross S l 26% l Low Beta G. M. l 0.01 pCi/m2 l ll l l l l l ll l 2Cs l 21% l Gamma Spect. l 0.03 pC1/m l8 ll l l l l l ll l 'Zr 'Nb l 30% l Gamma Spect. l 0.02 pCi/m' l ll l l l l l ll l 2"Ru l 4.3% l Gamms Spect. l 0.20 pCf/m' l ll l l 1 1 I ll l 2" Ce l 6.8% l Gamma Spect. l 0.10 pCi/m' l ll l l l l l ll l 2'2I l (a) l Gamma Spect. l (a) l II I l l l l ll l 'H l 25% l Liquid Scint. ] 6.8 pC1/m 8 l ll l l 1 -

1 (d) l l l External l l l Thermolumines- l l l l Gamma l l l cent Read-out i 40 mR (b) l ll l l l l l l l Forage l Gross 6 l 8% l Low Beta G.M. l 3.0 pC1/g l ll l l l l l ll l 2Cs l (a) l Gamma Spect. l 35 pCi l ll l l l l l ll l 'H I 25% i Liquid Scint. I 2000 pCi/1 l ll l l l l . l l l Milk l 2Cs l 6.7% l Gamma Spect. l 2.0 pCi/1, l ll l l l l l l ll l 2'2I l (a) l Gamma Spect. l (a) l ll l l l l l ll . l ' " , "Sr l 8% l Chem. Separation l 10.0 pCi/1 l l 'l l I I I I ll l 'H I 25%  ! Liquid Scint. I 2000 pCi/1 l l

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.9-17 i TABLE 5.9-3 l SENSITIVITIES FOR ENVIRONMENTAL RADIATION MEASUREMENTS (Cont'd}

I Il l l Counting i Analytical l M.D.A. l l1 Media l Isotope i Efficiency-l Technique  ! 99% 0.L. l l1 I l 1 1 I I I Water i 2Cs l 6.7% l Gamma Spect. I 2.0 pCi/1 l ll l l 1 1 I II I Gross S I (c) l Low Beta G.M. l 1.3 pC1/1 l ll l l l 1 I II l 8H I 25% 1 Liquid Scint. I 2000 pCi/1 1 II I I I f l 1 l Sediment l Gross S I 8% i Low Beta G.M. i 3.0 pCi/g l

l'(a) To be determined. .

I i I (b) Radiation exposure necessary to produce a response equal to l l 3 sigma of the background as determined over a three month per-l iod.

l (c) Dependent upon amount of dissolved and suspended solids.

1 (d) Depends upon relati ve humidity.

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Fort St. Vrain #1 Technical Spocifications Amendment #

Page 5.10-1 5.10 FIRE SUPPRESSION SYSTEMS - SURVEILLANCE REQUIREMENTS Applicability Applies to the surveillance of the fire suppression and protection systems and equipment.

Objective To establish the minimum frequency and type of surveillance on the equipment of the fire suppression and protection equipment to assure that the capability exists for suppressing any fire involving safety related equipment.

Specification SR 5.10.1 - Three Room Control Complex HVAC System, Surveillance Requirement The HVAC isolation dampers and associated fans of the control room, auxiliary electric room, and the 480 volt switchgear room, shall be tested annually to verify correct response to a simulated actuation signal from the Halon Fire Suppression System.

Basis for Specification SR 5.10.1 .

Annual testing of the room isolation dampers and associated fans of the three-room control complex is sufficient to demonstrate capability to operate when required by the Halon Fire Suppression System. The dampers in the ventilation systems of the control and auxiliary electric rooms, and of the 480 volt switchgear

Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.10-2 room automatically close upon actuation of the Halon Fi-e Suppression System for that room. At the same time, the various ventilation fans associated with these areas a e also tripped off or prevented from starting. Tne isolation of the ventilation system concurrent with Halon discharge is required to maintain effective concentrations of Halon in the area of the fire.

Specification SR 5.10.2 - Halon Fire Suppression Systeg2 Surveillance Requirements Operability of the Halon fire suppression system for the control room, auxiliary electric equipment room, and 480 volt switchgear room shall be demonstrated as follows:

a) Quarterly, verify that Halon Storage Cylinder weight is at least 95% of full rated charge.

b) Quarterly, verify that Halon Storage Cylinder pressure is at least 90% of full rated charge.

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c) Annually, verify that distribution headers and nozzles are open by flowing clean air at low pressure through the system.

d) Annually, verify response of system to an actuation signal by disconnecting each solenoid coil and measuring the voltage created by the actuating signal.

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Forg St. Vrain #1 Technical Specifications Amendment #

Page 5.10-3 Basis for Specification SR 5.10.2 Quarterly, check-weighing and pressure verification of the Halon Fire Suppression System cylinders meets the requirements of NFPA Code Section 12A for fire suppression system operability tests. Annual verification that the distribution headers are not plugged demonstrates their ability to distribute Halon when needed to suppress a fire, while measurement of an actuation signal to the solenoid coil of each cylinder release mechanism gives extra assurance that the system will be capable of performing its design function when required.

Specification SR 5.10.3 - Smoke Detectors and Alarm, Surveillance Requirement The smoke detectors and alarms listed in Table 4.10-3 shall be demonstrated operable as follows:

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a) Monthly by functional test of the non-supervised circuits between the local panels and the main control panel.

b) Semi-annually by functional test of the smoke detectors and alarms, and by functional test of supervisory circuits associated with the smoke l

detector alarms.

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Fort St. Vrain #1 Technical Specifications Amendment #

%ge 5.10-4 Basis for Specification SR 5.10.3 Operability of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages.

Testing at the specified intervals is sufficient to ensure operability of the system.

Specification SR 5.10.4 - Fire Barrier Penetration Seal, Surveillance Requirements Fire Barrier Penetration Seals shall be visually inspected:

a) Annually, to verify each remains intact, or b) Immediately following any maintenance which disturbs the retardant material, to verify seal is returned to its previous condition.

Basis for Specification SR 5.10.4 i

Inspection of penetration seals either annually or following maintenance assures that the Fire Barrier Penetration Seals remain unchanged.

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.10-5 Speci fication SR 5.10.5 - Breathing Air System, Surveillance Requirement The operbility of the Breathing Air System shall be demonstrated annually, as follows:

a) Functionally test the compressors and air supply piping.

b) Test the quality of the air supplied.

Basis for Specification SR 5.10.5 The Breathing Air System and associated piping supplies compressed air for recharging air cylinders of self-contained breathing apparatus and to the control room for personnel use with approved breathing equipment, in a Halon or toxic gas atmosphere. By functionally testing the system annually, assurance of breathing air availability and quality is provided.

Specification SR 5.10.6, Fixed' Water Spray System, I Surveillance Requirement

( Each of the fixed water spray systems listed in LCO 4.10.5 shall be verified operable as follows:

a) Annually by cycling each valve in the flow path through one complete cycle of full travel.

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.10-6 b) Each refueling cycle by inspection of the flow headers to verify their integrity and inspection of each no::le to verify no blockage.

c) At least once per three years by performing an air flow test through each open head spray / sprinkler header and veri fying each open head spray / sprinkler nozzle is unebstructed.

d) The temperature instruments and controls associated with the reactor plant exhaust filters shall be functionally tested semi-annually.

Basis for Specification SR 5.10.6 Operation of the valves and verification of the flow path at the specified intervals is sufficient to demonstrate capability to operate if required.

Semi-annually testing the temperature instruments and controls associated with the reactor plant exhaust filters

.is sufficient to ensure operability of the system.

Specification SR 5.10.7 - Carbon Dioxide Fire Suppression i

System, Surveillance Reauirement The carbon dioxide fire suppression system shall be j demonstrated operable as follows:

i a) Weekly - verify storage tank level and pressure.

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.10-7 b) Annually:

1) Verify operation of system valves and associated dampers upon actuation signal.

2) Verify flow from each nozzle during a " puff test."

Basis for Specification SR 5.10.7 A weekly check of level and pressure in the carbon dioxide storage tank insures sufficient carbon dioxide for fire suppression and the support equipment is operating properly.

An annual flow check and simulated automatic actuation of the system along with the regular calibration of the system instrumentation provides adequate assurance that the system will be ready to suppress any fire that could occur in the emergency diesel generator rooms.

S*pecification SR 5.10.8 - Fire Hose Stations, Surveillance Requirement Each of the fire hose stations listed in Table 4.10-7 shall be checked monthly to insure all required equipment is at the station.

The fire hoses at these stations shall be removed for inspection, repacking, and refurbishing as required once per refueling cycle.

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.10-8 These fire hose stations shall be tested for flow and the fire hoses hydrostatically tested once every 3 years.

Basis for Specification SR 5.10.8 These checks of the fire water hose system will demonstrate the system's ability to operate if required.

Specification SR 5.10.9 - Yard Fire Hydrants and Hydrant Hose Houses, Surveillance Requirement Each of the yard fire hydrants and associated hydrant hose houses listed in LCO 4.10.8 shall be verified operable as follows:

a) Monthly by visual inspection of the hydrant hose house to assure all required equipment is at the hose house.

b) Semi-annually (once during March, April, or May, and once during September, October, or November) by l

l visually inspecting each yard fire hydrant and i

verifying that the hydrant barrel is dry and that the hydrant is not damaged.

c) Annually by conducting a hose hydrostatic test at a l

l pressure at least 50 psig greater than the ra:ximum l

pressure available at any yard fire hydrant and by replacement of all degraded gaskets in couplings.,

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Fort St. Vrain #1 Technical Specifications Amendment #

Page 5.10-9 Basis for Specification SR 5.10.9 Inspection and . testing at the specified intervals is sufficient to ensure operability of the hydrants and hydrant hose houses when required.

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ATTACHMENT 4 SAFETY ANALYSIS REPORT FOR THE FORT ST. VRAIN IN-SERVICE INSPECTION AND TESTING PROGRAM UPDATE 6

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AttachmenB 4 Page 1 SAFETY ANALYSIS REPORT FOR THE FSV IN-SERVICE INSPECTION AND TESTING PROGRAM UPDATE

1. BACKGROUND The Fort St. Vrain in-service inspection and testing program is specified by the Plant Technical Specification Surveillance Requirements (Ref.1).

In response to a commitment in the 1972 Safety Evaluation Report (Ref. 2) Public Service Company has been reviewing, as a continuing effort, the in-service inspection and tasting program for Fort St. Vrain to feedback the acquired operating experience with the plant, and to update the program in light of more recent rules and regulations.

The original 1972 Safety Evaluation Report (Ref. 2) includes a commitment to review the in-service inspection program for the primary coolant system after five years of reactor operation.

The status of the review effort was originally described by Public Service Company, together with the planned approach to follow in conforming with the 1972 Safety Evaluation Report commitment (Ref. 3). A review of Public. Service Company plans was performed by the Nuclear Regulatory Commission, who also identified priority items to be addressed beyond the scope of the original Safety Evaluation Report ccmmitment (Ref. 4). The general in-service inspection and testing program review plan and the priority items were further discussed in letters and at meetings between the Nuclear Regulatory Commission and Public Service Company until a basic agreement was reached between both parties (Ref. 5 through 10). A schedule was established for the review of surveillance requirements for all major plant systems and equip:,ent by subdividing them in priority categories as requested by the Nuclear Regulatory Commission (Ref. 11).

This Safety Analysis Report addresses t: ? changes to Technical Specification Surveillance Requirements . proposed by Public Service Company as a result of priority category I reviews.

2. METHODOLOGY FOR REVIEW OF ISI PROGRAM 2.1 Rules A set of general rules governing the Fort St. Vrain in-service inspection and testing program review was established by Public Service Company in agreement with the Nuclear Regulatory Commission, as discussed above. Specific important features of these rules were that Section XI Division 2 of the ASME Code, then in form of a draft, could l not be applied directly to Fort St. Vrain but was to be used

. as a guide; Public Service Company would provide the Nuclear l Regulatory Commission with c comparison of recommended surveillancs requirements with these proposed Code i

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Attachment 4 Page 2 requirements and justify the differences. It was also agreed that Public Service Company would rank the equipment by their importance to plant safety and specify surveillance requirements commensurate with that importance to safety.

Detailed definitions and criteria for examinations and tests were developed accordingly and transmitted to the Nuclear Regulatory Commission (enclosure 3 to Ref. 12).

2.2 Review Process 2.2.1 PSC Preliminary Submittals The results of all priority Category I reviews were transmitted to the Nuclear Regulatory Commission (Ref. 12, 13,14). These reviews covered the major systems and components that are important to safety, including the prestressed concrete reactor vessel, the reactor internals, the reactor primary coolant system, the reactor secondary coolant system, and the PCRV auxiliary system. Each of the above packages contained, for Nuclear Regulatory Commission review, draft modifications to the Fort St. Vrain Technical Specification Surveillance Requirements along with an evaluation of the existing and proposed inspections and tests, an identification of ASME Code Section XI requirements, and a discussion of the differences.

2.2.2 Independent Review of PSC Preliminary Submittals At the request of the Nuclear Regulatory Commission, the Public Service Company preliminary submittals were independently reviewed under the direction of the Los Alamos National Laboratory (Ref. 15, 16, 17). A first meeting was held at the Fort St. Vrain plant on November 20, 1981, between Los Alamos National Laboratory, Public ' Service Company, and their consultants (Ref. 18). The result of the independent review were subsequently included in a

. report, which was transmitted to both the Nuclear Regulatory Commission and Public Service Company Public Service Company's (Ref. . Most of reg w e- ed modifications to the Technical b ft .cion Surveillance Requirements were

@ ud ad to be adequate and acceptable, and a few ddit w el investigations were recommended. The recommendations resulting from the independent review were subsequently analyzed by Public Service Company who prepared a response to agree with or clarify these recommendations, or further support Public Service Company's position (Ref. 20). The responses were also revitmed independently, and the remaining open items were resolved at a final meeting held on July 29, 1982, at the Fort St. Vrain plant between the Nuclear Regulatory Commission, Los Alamos National Laboratory, Public Service

Attachment 4 Page 3 Company, and their consultants (Ref. 21, 22). At this meeting, it was agreed that Public Service Company would submit final proposed changes to the Technical Specification Surveillance Requirements, together with an implementation schedule.

2.2.3 Additional PSC Preliminary Submittals As a result of the independent review, Public Service Company has prepared additional preliminary submittal s to the Nuclear Regulatory Commission (Ref. 23) which include draft modified Technical Specification Surveillance Requirements, as well as an evaluation of the additional examinations and tests.

2.2.4 Final Proposed Changes to Technical Specification Surveillance Requirements ,

The final proposed changes to Technical Specification Surveillance Requirements were prepared by Public Service Company to reflect the draft proposed changes included in the preliminary submittals, modified as appropriate to take into account the results of the independent review, and the changes included in the additional preliminary submittals.

3. EVALUATION AND CONCLUSIONS The proposed changes to the Technical Specification Surveillance Requirements generally expand the scope of in-service examination and testing that is currently performed at the Fort St. Vrain Nuclear Generating Station. This, in essence, provides greater assurance of plant safety and reliability.

Individual surveillance requirements have been evaluated in detail in Public Service Company preliminary submittals covering the PCRV, the primary coolant systems, and the secondary coolant system, and these evaluations are incorporated herewith by reference. The results of these reviews revealed that existing surveillance requirements were generally adequate in light of the newly defined methodelogy, which accounts for plant operating experience, importance to safety, unique design features and limitations, and ASME Code development for large HTGR designs.

Minor modifications to surveillance intervals were made to reflect operating experience, and to provide operating flexibility. Additional tests were included to assure the operability and accuracy of instrumentation which can be used for monitoring the structural integrity of major plant equipment.

Additional component testing was recommended, as a result of

• detailed reviews of plant systems, either when components important to safe plant shutdown and cooling were not in the scope of the current Technical Specifications, or when the testing method could be improved to provide additional assurance of component reliability. Additional component examinations were

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Attachment 4 Page 4 also recommended, where appropriate, to provide further assurance concerning the continued structural integrity of critical welded joints, bolted connections, and load bearing features.

Through the independent review process, it was verified that: (1) The Public Service Company evaluations were technically correct; (2) uncovered deficiencies were resolved to the Nuclear Regulatory Commiss'on's satisfaction; (3) sufficient examination and testing is planned, using the state of the art within limitations inherent to the Fort St. Vrain design, to assure the long term safety of the plant.

Since the proposed changes to the Technical Specification do not result from modifications to plant equipment, nor do they involve reductions in the margins of safety, it can be concluded that no unreviewed safety question is raised. Besides the proposed changes, no changes are required to other parts of the Technical Specifications.

4. REFERENCES

1. Plant Technical Specifications

2. Safety Evaluation Report of January 20, 1972, Section 3 3.

3. Public Service Company letter dated October 13, 1978 (P-78169), In-service Inspection - Fort St. Vrain.

4. Nuclear Regulatory Commission letter dated January 15, 1979, In-service Inspection and Testing Program for Fort St.

Vrain.

5. Public Service Company letter dated March 15, 1979 (P-79058), In-service Inspection Program for Fort St. Vrain.

6. Nuclear Regulatory Commission letter dated June 5,1979, Summary of Meeting Held on May 2, 1979, to Discuss In-service Inspection. ,

7. Public Service Company Progress Report. Meeting held on .

August 20, 1979, between the Nuclear Regulatory Commission and Public Service Company.

l 8. Public Service Company letter dated August 22, 1979 (P-79176), Fort St. Vrain In-service Inspection and Testing Program.

9. Nuclear Regulatory Commission letter dated October 5, 1979, Proposed Plan of In-service Inspection and Testing for Fort St. Vrain.

, 10. Public Service Company Progress Report. Meeting held on l November 1, 1979, between the Nuclear Regulatory Commission i and Public Service Company.

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Attachment 4 Page 5

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11. Public Service Company letter dated November 30, 1979 (P-79289), Fort St. Vrain In-service Inspection and Testing Program.

12. Public Service Company letter dated February 8, 1980 (P-80014), Fort St. Vrain In-service Inspection and Testing - PCRV Auxiliary System.

13. Public Service Company letter dated March 3, 1980 (P-80034),

Fort St. Vrain In-service Inspection and Testing (PCRV and PCRV Internals).

14. Public Service Company letter dated March 31, 1980 (P-80064), Fort St. Vrain In-service Inspection and Testing (Reactor Primary and Secondary Coolant Systems).

15. Los Alamos National Laboratory letter dated OcteSer 30, 1981 (Q-13:81:365).

16. Los Alamos National Laboratory letter dated November 2,1981 (Q-13:81:369) (Proposed Agenda for a Meeting November 20,1981).

17. Public Service Company letter dated November 9, 1981 (P-81285), Los Alamos National Laboratory Evaluation of Fort H St. Vrain ISI Program.

18. Los Alamos National Laboratory letter dated '

December 10, 1981 (Q-13:81:420), Fort St. Vrain ISI Program Review Meeting.

19. Los Alamos National Laboratory letter dated January 5,1982 (Q-13:82:5) (Review of the Public Service Company Proposed In-service Inspection Program).

k_ 20. Public Service Company letter dated March 29, 1982 (P-82061), Fort St. Vrain In-service Inspection and Testing (Response to the Recommendations of Los Alamos National Laboratory Report Q-13:82:5).

21. Los Alamos National Laboratory letter dated June 30, 1982 (Q-13:82:228) (Comments Regarding Public Service Company Response).

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'22. Meeting of July 29,1982, between Nuclear Regulatory Commission, Los Alamos National Laboratory, Public Service Company, and their Consultants.

23. Public Service Company letter dated September 30, 1982 (P-82430), Fort St. Vrain In-service Inspection and Testing l

Program Additional Surveillance Requirements.

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