ML20137H566

Proposed Tech Specs Re Inclusion of Phase II of Inservice Insp & Testing Program

ML20137H566
Person / Time
Site:	Fort Saint Vrain
Issue date:	11/27/1985
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20137H505	List: ML20137H499 ML20137H532 ML20137H566
References
TAC-53417, NUDOCS 8512020424
Download: ML20137H566 (65)

Text

4 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGES 83120$$0 PDR P

Fort St. Vrain #1 Technical Specifications Amendment Page 5.0-2 Amendment No. 33 by the Nuclear Regulatory Commission.

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

ISI criterion 0: The surveillance requirement shall be implemented in the existing schedule of surveillance tests, following 90 days from the formal approval date of Amendment No. 33 by the Nuclear Regulatory Commission.

l ISI Criterion E: Same as ISI Criterion A but applicable to l

Amendment No.___.

l ISI Criterion F: Same as ISI Criterion 8 but applicable to fuel l

cycle 5 and Amendment No.

l ISI Criterion G: Same as ISI Criterion 0 but applicable to l

Amendment No.___.

Fort St. Vrcin #1 Technical Spscifications Amendment Page 5.2-1 j

5.2 PRIMARY COOLANT SYSTEM - SURVEILLANCE REQUIREMENTS Applicability Applies to the surveillance of the primary (helium) reactor coolant system, excluding the steam generators, l

and to the surveillance of the reactor auxiliary systems.

Objective To ensure the capability of the compenents of the primary reactor coolant system to maintain the primary reactor coolant envelope as a

fission product barrier and to ensure the capability to cool the core under all modes of operation.

Specification SR 5.2.1 - PCRV and PCRV Penetration Overpressure Protection Surveillance a)

Each of the two overpressure protection assemblies protecting the PCRV shall be tested at intervals not to. exceed five years, on an alternating basis, with one overpressure protection assembly tested during j

each refueling cycle.

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

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

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

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Fort St. Vrain #1 Technical Specifications Amendment 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 l

pumps at fire pump discharge pressure), annually, or at l

the next scheduled plant shutdown if the test was not l

performed during the previous year provided that the l

surveillance interval does not exceed 18 months.

l SR 5.2.7a shall be implemented per ISI criterion G.

b)

Safety valves (V-21522, V-21523, V-21542, and V-21543),

i located in the water turbine supply lines, will be tested l

for relieving pressure annually, or at the next scheduled l

plant shutdown if the test was not performed during the l

previous year provided that the surveillance interval does l

not exceed 18 months.

SR 5.2.7b shall be implemented per l

ISI Criterion G.

c)

Both turbine water removal pumps and the turbine water l

drain tank overflow to the reactor building sump shall be l

functionally tested once per 92 days.

SR 5.2.7c shall be I

l implemented per ISI Criterion G.

I, d)

The instrumentation and controls associated with c) shall be functionally tested in conjunction with i

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Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-16 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.

l Specification SR 5.2.8 - Circulator Bearing Water Pumps And Makeup Pump Surveillance l

The circulator bearing water pumps, bearing water makeup l

pumps, and associated instruments and controls shall be tested as follows:

a)

The Normal Makeup Pump shall be operated in the l

recycle mode every 92 days.

l SR 5.2.8a shall be implemented per ISI Criterion G.

b)

The Emergency Makeup Pump shall be functionally tested l

every 92 days.

l SR 5.2.8b shall be implemented per ISI Criterion G.

c)

The associated instruments and controls shall be

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functionally tested in conjunction with and at the intervals specified in parts a) and b) above, and calibrated annually.

1

Fort St. Vrain #1 Technien1 Spscifications Amendment Page 5.2-17 l

d)

Each Bearing Water

Pump, and the associated l

instruments and controls shall be functionally tested l

at each scheduled plant shutdown.

In addition, the l

instruments shall be calibrated annually, or at the l

next scheduled plant shutdown if they were not l

calibrated during the previous year provided that the j

surveillance interval does not exceed 18 months.

l SR 5.2.8d shall be implemented per ISI Criterion F.

Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-18 Basis for Specification SR 5.2.8 l

The bearing water pumps and bearing water makeup pumps are l

required to operate for safe shutdown cooling of the l

reactor under accident conditions described in FSAR l

Section 10.3.9.

The specified tests and testing intervals l

are sufficient to ensure adequate pump operation for the l

performance of their required safety functions.

l Performance capability of the bearing water pumps is l

Verified by normal operation.

Performance capability of l

the normal bearing water makeup pump is verified when l

operating the pump in the recycle mode.

Performance l

capability of the emergency bearing water makeup pump is l

verified when testing the associated check valves as l

required per SR 5.3.4.

Specification SR 5.2.9 - Helium Circulator Bearing Water Accumulators Surveillance The helium circulator bearing water accumulators, instrumentation, and controls shall be functionally tested l

every 92 days and calibrated annually.

l SR 5.2.9 shall be implemented per ISI Criterion E.

Basis for Specification SR 5.2.9 Helium Circulator bearing water is normally supplied fror i

the bearing water system and is backed up by the backup bearing water system supplied from the Emergency Feedwater

Fort St. Vrcin #1 Technical Specifications Amendment Page 5.2-19 Header.

In the event of a

failure in both of these

systems, the water stored in the bearing water accumulators is adequate to safely shut down both helium l

circulators in a

loop.

The specified tests and testing l

intervals are sufficient to ensure operability of the l

accumulator

controls, should they be called upon to l

perform their required function.

Specification SR 5.2.10 - Fire Water System / Fire Suppression Water System Surveillace a)

The fire water system shall be verified operable as follows:

1)

The motor driven and engine driven fire pumps shall be functionally tested monthly.

The associated instruments and controls shall be functionally tested monthly and calibrated annually.

2)

The diesel engine fuel shall be inventoried monthly and sampled and tested quarterly.

3)

The diesel engine shall be inspected during each refueling shutdown.

4)

The diesel engine starting battery and charger shall be inspected weekly for proper electrolyte level and overall battery voltage.

The battery electrolyte shall be tested quarterly for proper specific gravity..

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Fort St. Vrain $1 Technical Spacifications Amendment Page 5.2-20 5)

The batteries, cell

plates, and battery racks, shall be inspected each refueling cycle for i

evidence of physical damage or abnormal degradation.

The battery-to-battery and terminal connections shall be verified to be clean, tight, free of corrosion, and coated with anti-corrosion material each refueling cycle, b)

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

1)

Monthly by verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

2)

Semi-annually by performance of a fire suppression water system flush.

3)

Annually by cycling each testable valve in the fire suppression water system flow path through at least one complete cycle of full travel.

4)

Each refueling cycle by performing a

fire suppression water system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

(a) Verifying that each automatic valve in the flow path actuates to its correct position.

FCrt St. Vrain #1 Tschnical Spacifications Amendment Page 5.2-20a (b) Verifying that each fire water pump develops l

at least 1,425 gpm at a system head no less l

than 119 psig.

SR 5.2.10.b4b shall be l

implemented per ISI Criterion G.

(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 l

to 275 feet water gauge.

5)

Each ti?ree years by performing a flow test.

Basis for Specification SR 5.2.10 The fire water pumps are required to supply water for fire suppression and safe shutdown cooling.

The specified testing interval is sufficient to ensure proper operation of tne 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 l

of the system when required.

f Fort St. Vrain #1 Technical Specifications Amendment Page 5. 2-26 i

shutdown if these valves have not been tested during the previous year.

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

ei The check valves on the HTFA purge lines shall be tested at five calendar year intervals.

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

f)

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

SR 5.2.16.f shall be implemented per ISI Criterion B.

l g)

Each helium purification cooler well closure shall be l

leak

tested, and the well pressure monitoring l

instruments shall be calibrated, once during each l

refueling cycle.

In

addition, the instruments and l

controls used to automatically isolate the l

purification system shall be functionally tested at

~l the same frequency.

l SR 5.2.16g shall be implemented per ISI Criterion F.

Basis for Specification SR 5.2.16

Fort St. Vrain #1 Tachnical Spacifications Amendment Page 5. 2-27' The interval specified for determining th. actual primary and secondary closure leakage is adequate to asstare compliance with LCO 4.2.9.

In,the determination of closure leakas a at the reference differential

pressure, laminar leakage flew shall be conservatively

assumed, therefore in correcting the determined closure leakage to reference differential pressure, the ratio of the reference differential

pressure, and test differential pressure shall be used.

The interval specified for functional testing and calibration of the instrumentation and alarms monitoring the penetration closure interspace pressurization gas flow will assure sensing and alarming any change in pressurization gas flow.

The interval -specified for functional test and calibration of the instrumentation and alarms monitoring the core support floor and columns will assure sensing and alarming any change in their structural integrity.

The interval specified for valve testing is adequate to assure proper valve operation when isolation of the closure auxiliary i

piping is required.

l The intarval specified for testing the helium purification l

cooler wells is adequate to verify the well integrity, as well l

as that of primary coolant boundary components located therein.

k b

Fort St. Vrain $1 Tcchnical Specificationo Amendment Page 5.2-27a Specification SR 5.2.17 - Helium Circulator Pelton Wheels DELETE SPECIPICATION SR 5.2.17 IN ITS ENTIRETY Specification SR 5.2.18 - Helium Circulators Surveillance a)

At the ties of the first main turbine generator overhr:1, one helium circulator unit shall be removed in its entirety from the PCRV and thoroughly inspected for signs of abnormal wear or component degradation.

1)

Such inspection shall include examination of bearing surfaces, seal

surfaces, brake

system, buffer seal system, and labyrinth seals.

I I

Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-31 l Specification SR 5.2.21 - ACM Transfer

Switches, Valves, and l Instruments Surveillance l a)

Those valves and transfer switches that must be manually l

positioned for actuation of the Alternate Cooling Method (ACM) l mode of operation shall be tested for operability by partial l

stroking of the valve twice annually at an interval between l

tests to be not less than four (4) months, nor greater than l

eight (8) months.

A full functional test shall be perfo rmed l

annually, or at the next scheduled plant shutdown if such test l

was not performed during the previous year provided that the l

surveillance interval does not exceed 18 months.

SR 5.2.21a l

shall be implemented per ISI Criterion F.

l b)

Local indicators for the helium purification dryer inlet l

temperature, for the helium purification pumpdown line pressure l

and for the reactor plant cooling water surge tank cover gas l

pressure sha11 be calibrated at each plant shutdown for

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l refueling.

SR 5.2.21b shall be implemented per ISI Criterion F.

l Basis for Specification SR 5.2.21 l In the event that the ACM mode of operation must be implemented, it l is necessary to manually position valves (manual valves as well as l valves which would normally be pneumatically or electrically l operated) and to manually reposition electrical transfer switches.

l Local instruments allow system monitoring during the depressurization l phase ~of the PCRV and during the subsequent cooling phase of the l reactor.

The specified tests and testing intervals are sufficient to l casure

Fort St. Vrain il Technical Specifications d

Amendment Page 5.2-32 l oparability of these components should they be called upon for l parformance of their required safety functions.

Spncification 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 cupport blocks during operation of the reactor.

These specimens (16 per reflector element) will be installed in reflector elements as indicated in Table 1 and will be removed at subsequent refueling intervals, as indicated in Table 1, unless the progressive examination of the specimens dictate otherwise.

Upon

removal, these specimens will be subjected to examination, and compared with laboratory control specimens in evaluating oxidation rates, oxidation profiles, and general dimensional characteristics.

i 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.

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Fort St. Vrain #1 Tcchnical Spscificctions Amendment 1

Page 5.2-32a l

Basis for Specification SR 5.2.22 The PGX graphite specimens will be placed in modified coolant channels in five (5) transition reflector elements in the hottest columns of regions 22, 24, 25, 27, and 30.

The surveillance test specimens will be subjected to the primary coolant conditions, as well as other reactor parameters that are normally seen by the PGX core support blocks.

Examination and tests of the surveillance test specimens at regular intervals can readily be utilized to assess oxidation rates, oxidation profiles, as well as general degradation of the PGX core support blocks to adequately predict the structural integrity of the core support blocks over the operating life of the reactor.

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Technical Specifications Amendment Page 5.2-34 Specification SR 5.2.23 - Firewater Booster Pump Surveillance Each firewater booster pump shall be tested annually by providing motive power to one water turbine drive in conjunction with the performance of SR 5.2.7.

In addition each pump shall be functionally tested quarterly.

The i

associated instruments and controls shall functionally be tested quarterly and calibrated annually.

Basis for Specification SR 5.2.23 During accident conditions described in Final Safety Analysis Report, Section 14.4.2.1, one of the firewater booster-pumps and one firewater pump are required to provide adequate core cooling.

The specified testing interval is sufficient to ensure proper operation of the pump and associated controls.

l Specification SR 5.2.24 - Reactor Auxiliary Cooling Water l

Systems Surveillance i

!l The reactor auxiliary cooling water systems shall be

~l tested as follows:

a) -The circulating water makeup pond minimum inventory shall be verified daily.

The pond level instrumentation shall be functionally tested monthly and calibrated annually.

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Fort St. Vrain #1 Tschnical Spscifications Amendment Page 5. 2-34a l

b)

Each circulating water makeup pump and the associated l

instruments and controls (including firewater pump pit l

instruments and controls) shall be functionally tested l

monthly.

In

addition, the instruments shall be l

calibrated, and the pump performance capability (flow l

and head) and mechanical condition (vibration l

amplitude and bearing temperature) shall be verified, l

annually or at the next scheduled plant shutdown if l

this was not performed during the previous year.

l SR 5.2.24b shall be implemented per ISI Criterion F.

c)

The valve lineup of.the flow path between the circulating water storage ponds _

and the fire water pump pits shall be verified correct monthly.

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Fort St. Vrain #1 Technical Specifications Amendment Page 5. 2-34b l

d)

Alignment and settlement of the circulating water makeup I

pond embankments shall be verified at five calendar year l

intervals.

The embankments and the water structures shall l

be examined at the same intervals for abnormal

erosion, l

cracks, seepage, leakage, accumulation of silt or debris l

(as applicable) which might indicate a

deterioration of l

structural safety or operational adequacy of the storage l

ponds.

SR 5.2.24d shall be implemented per ISI Criterion l

r.

l e)

Each service water pump and the associated instruments and l

controls shall be functionally tested monthly.

In l

addition, the instruments shall be calibrated, and the l

pump performance (flow and head) and mechanical condition l

(vibration amplitude and bearing temperature) shall be l

verified, annually or at the next scheduled plant shutdown l

if this was not performed during the previous year.

l SR 5.2.24e shall be implemented per ISI Criterion F.

l f)

Each reactor plant cooling water pump and the associated l

instruments and controls shall be functionally tested l

monthly.

In

addition, the instruments shall be l

calibrated, and the pump performance (flow and head) and l

mechenical condition (vibration amplitude and bearing l

temperature) shall be verified, annually or at the next l

scheduled plant snutdown if this was not performed during l

the previous year.

SR 5.2.24f shall be implemented per l

ISI Criterion F.

Fort St. Vrain #1 Technical Specifications Amendment Page 5.2-34c l g)

Each purification cooling water pump and the associated l

instruments and controls shall be functionally tested monthly.

l In

addition, the instruments shall be calibrated, and the pump l

performance (flow and head) shall be verified, annually or at l

the next scheduled plant shutdown if this was not performed l

during the previous year.

SR 5.2.24g shall be implemented per l

ISI Criterion F.

l h)

Instruments and valves, used for automatic isolation of portions l

of the purification cooling water system and the reactor plant l

cooling water

system, that may be required for confinement of l

reactor coolant and that are capable of being tested, shall be l

tested for operability by partial stroking or full stroking of l

the valves, as appropriate, twice annually at an interval l

between tests to be not less than four (4) months, nor greater l

than eight (8) months.

Additionally, these instruments and l

valves shall be functionally tested annually or at the next l

scheduled plant shutdown if such test was not performed during l

the previous year, provided that the surveillance interval does l

not exceed 18 months. SR 5.2.24h shall be implemented per ISI Criterion F.

l Basis for Specification SR 5.2.24 l The reactor auxiliary cooling water systems

(' including water makeup l system, service water system, reactor plant cooling water system,-and l purification cooling water system) are required to operate for l reactor cooling under i

J

Fort St. Vrain #1 Tachnical Specifications Amendment Page 5.2-34d 1

l postulated loss of forced circulation cooling accident conditions.

l Except for the purification cooling water

system, they are also l required for safe -shutdown cooling of the reactor under other l postulated accident conditions.

The circulating water makeup system l also supplies water for fire suppression.

These systems routinely l operate during normal plant operation.

Routine operation in l conjunction with the specified tests and testing intervals are l sufficient to ensure adequate system and/or component operation for l the performance of their required safety functions.

i l Measuring the position of survey markers and evaluating the changes l in position of these markers will allow changes in embankment l alignment and settlement to be determined, as well as their possible i

l impact on the structural integrity of the storage pond.

Exa.mination-l of the embankments and of the water stractures will provide for an l additional verification that no phenomenon occurs which might be I

l detrimental to the ability of the storage pond to perform its safety l function.

Measurement of the silt accumulation in the storage pond l will allow a verification that the minimum water inventory required l by LCO 4.3.5 is available for Safe Shutdown Cooling of the reactor.

l The interval specified for instruments and valves is adequate to l assure their automatic isolation function, if degradation were to l occur in the integrity of the reactor coolant boundary, resulting in l primary coolant leakage into the system.

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Fort St. Vrcin $1 Tschnical Sp;cificationa Amendment Page 5.2-35 Specification SR 5.2.25 - Core Support Block Surveillance The top surface of the core support block for fuel regions fitted with PGX graphite specimens shall be visually examined by remote TV for indication of cracks, in particular in areas where analysis shows the highest tensile stresses exist, at the refueling shutdown when the PGX graphite specimens are scheduled to be removed from the core in accordance with Technical Specification SR 5.2.22.

SR 5.2.25 shall be implemented per ISI Criterion D.

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Fort St. Vrain el 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.

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

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

The tests required on the above valves will demonstrate that each valve will function properly.

Test frequency is considered adequate for assuring valve operability at all times.

Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-6 Specification SR 5..'_',- safe Shutdown Cocling Valves Surveillance l

The following valves shall be tested for operability by partial l

stroking every 92 days unless they cannot be operated during l

normal plant operation.*

A full functional test shall be l

performed annually, or at the next scheduled plant shutdown if l

such test was not performed during the previous year provided l

that the surveillance interval does not exceed 18 months:

l Pneumatically, hydraulically, or electrically operated l

valves that are required to operate for actuation of the l

safe shutdown cooling mode of operation l

(implemented per ISI Criterion B);

l Normally closed check valves that are required to open for l

actuation of the safe shutdown cooling mode of operation "l

(imp'.emented per ISI Criterion B) ; and l

Valves (including normally power operated valves) that l

must be manually positioned for actuation of the safe l

shutdown cooling mode of operation l

(implemented per ISI Criterion G).

l

• (implemented per ISI Criterion F)

Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-6a 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 Cooling are not called upon to function during normal operation of the plant, except to stand fully closed or open.

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

During reactor operation, the instrumentation required to monitor and control the Safe Shutdown mode of cooling is normally in use and any malfunction would be immediately brought to the attention of the operator.

That instrumentation not normally in use is tested at intervals specified by other surveillance requirements in this Technical Specification.

Safe Shutdown Cooling, the systems or portions of systems involved, are discussed in Sections 10.3.9 and 10.3.10 of the FSAR and are represented in FSAR, Figure 10.3-4.

Valve testing will include, as applicable, full stroking l

each valve, or an observation that the valve stem or disc travels from the valve normal operating position to the position required to perform the safety

function, an

Fort St. Vrain #1 Technical Sp;cifications Amendment Page 5.3-7 observation that the remote position indicators accurately reflect actual valve position, and a measurement of the full stroke time for the hydraulically actuated automatic valves.

Specification SR 5.3.5 - Hydraulic Power System 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 Soecification SR 5.3.5 The hydraulic power system is a normally operating system.

Malfunctions in this system will normally be detected by failure of the hydraulic oil pumps or hydraulic oil accumulators to maintain a supply of hydraulic oil at or above 2500 psig.

Functional tests and calibrations of the pressure indicators and low pressure alarms on the above basis will assure the actuation of these alarms upon a malfunction of the hydraulic power system which may compromise the capability of operating critical valves.

l

Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-18 to verify proper piston movement, lock-up, and bleed.

The number of each type of snubber represented by use of either plan presented in Section 5.3.8.c) of this specification is an adequate sample for such tests.

Observed failures on these samples should require testing of additional units.

The required surveillance program will assure a higher degree of snubber functional reliability.

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Fort St. Vrain #1 Technical Specifications Amendment Page 5.3-18a Specification SR 5.3.9 - Safety Valves Surveillance l

The steam generator superheater and reheater safety valves l

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

tested at five calendar year intervals to verify their l

setpoints.

'l SR 5.3.9 shall be implemented per ISI Criterion F.

l Basis for Specification SR 5.3.9 l

Safety valves protect the integrity of the plant l

components which are part of the primary or secondary l

reactor coolant

boundary, and also the integrity of l

systems required to safely shutdown and cool the reactor l

under accident conditions.

Testing the safety valve setpoints will assure that the pressure within the equipment remains within design limits..

When practical, testing of the safety valves will be scheduled during the surveillance interval.so that testing v

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Fort St. Vrain #1 Technical Specifications Amendment Page 5. 4-13 l

Specification SR 5.4.4 - PCRV Cooling Water System l

Temperature Instruments Surveillance l

The PCRV cooling water system temperature instruments l

shall be tested as follows:

l a)

Once a

month during plant operation at power a l

scanner reading shall be taken of the inlet header l

and tube outlet temperatures.

The inlet l

temperature readings shall be compared to the l

corresponding temperature indicators.

The l

associated temperature alarms shall also be l

functionally tested at the same frequency.

l SE 5.4.4a shall be implemented per ISI Criterion G.

l b)

The

scanner, the inlet and outlet header l

temperature indicators, and the outlet subheader l

temperature indicators shall be calibrated l

annually.

l SR 5.4.4b shall be implemented per ISI Criterion G.

l c)

The inlet header and tube outlet thermocouples, l

which provide input to the

scanner, shall be l

calibrated at five calendar year intervals.

l SR 5.4.4c shall be implemented per ISI criterion G.

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- - - - - + - - -

Fort St. Vrain #1 Technical Specifications Amendment Page 5. 4-13a Basis for Specification SR 5.4.4 l

A scanner is used for monitoring the PCRV cooling system l

water inlet temperature and individual tube water outlet l

temperatures, and for alarming high outlet temperatures.

l Periodic scanner readout provides the information l

necessary to evaluate the water temperature increase in l

individual tubes.

A comparison of inlet temperature l

scanner reading to corresponding inlet temperature l

indicators assures that unacceptable drift in the scanner l

electronics does not occur.

l Calibration of the scanner and temperature indicators l

assures the accuracy of temperature measurements, in l

particular for verifying compliance with LCO 4.2.15.

l To the extent practical, thermocouples in individual l

subheaders will be calibrated at various times during the l

interval, to assure that unacceptable thermocouple drift l

does not occur.

l The specified intervals for checks and calibrations are l

sufficient to provide accurate temperature measurements to l

adequately protect the PCRV concrete and to monitor the l

integrity of the thermal barrier.

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Fort St. Vrain $1 Technical Specifications Amendment Page 5.4-13b l

Specification FR 5.4.5 - PCRV Cooling Water System Flow l-Instruments Surveillance l

A PCRV Cooling System scanner flow readout shall be taken and l

normal mode alarms functionally checked monthly.

The scanner l

and

alarms, and six (6) subheader flowmeters shall be l

calibrated annually, or at the next scheduled plant shutdown l

if they were not calibrated during the previous year provided that the surveillance interval does not exceed 18 months.

l SR 5.4.5 shall be implemented per ISI Criterion G.

Basis for Specification SR 5.4.5 l

Flow scanning acts as a backup to temperature scanning and l

initiates no automatic protective

actions, only an alarm.

l Because a

restriction or a leak in the system would develop l

over a period of time, the specified interval for comparing l

flow readouts is sufficient to detect any long term change in l

the system.

Specification SR 5.4.6 - Core Delta P

Indicator -

Surveillance Requirement The core Delta P

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 01 Technical Specifications Amendment Page 5.4-14 Specification SR 5.4.7 Control Room Temperature - Surveillance Requirement The control room temperature control thermostat shall be functionally tested monthly and calibrated annually.

Bacis for Specification SR 5.4.7 The surveillance interval specified for functional testing and calibration of the control room thermostat will assure its ability to not only control the room temperature as

desired, but to also indicate ths correct room temperature within the accuracy of the l

instrument.

4 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.

F

._.._m_.

.y.

Fort St. Vrcin #1 Technical Specifications Amendment Page 5.5-3 Basis for-Specification SR 5.5.2 The reactor building pressure relief device is designed to protect the building in the event that pressure in the 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 mechanical linkages to pneumatic actuators (see FSAR Section 6.1.3.4).

The specified test frequency shall ll i

ensure the operability of the reactor building relief system.

j l Specification SR 5.5.3 - Reactor Building Exhaust System l

Surveillance

) l The exhaust filters and fans in the reactor building ventilation system shall be tested as follows:

a)

A laboratory analysis of a representative carbon 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 > 90% radioactive methyl iodide removed when tested in accordance with ANSI N510-1975 (130 degrees C, 95% R.H.).

Fort St. Vrcin #1 Technical Specifications Amendment Page 5.5-4

• 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 calendar year or after each replacement of a

charcoal adsorber bank or after structural maintenance on the filter housing.

Halogenated hydrocarbon removal by the charcoal filters shall be > 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 > 99% when tested in accordance with the applicable portions of ANSI NS10-1975.

d)

Flow distribution across the HEPA and charcoal filters will be tested with initial operation of the system and following any structural modification to the

?ilter housings.

Air distribution shall be demonstrated within + 20%

across the HEPA and charcoal filters when tested in accordance with ANSI N510-1975.

Fort St. Vrain #1 Technical Specifications Amendment i

Page 5.5-5 l

e)

Verify a minimum flow rate of 15,390 cfm per train during l

system operation when tested in accordance with ANSI N510-1975 l

at least once per 18 months or (1) after any structural l

maintenance on the HEPA filter or charcoal adsorber

housings, l

or (2) following painting with other than low solvent paints, l

fire, or chemical release in any ventilation zone l

communicating with the system.

SR 5.5.3e shall be implemented l

per ISI criterion G.

l f)

The performance capability (capacity and total pressure) and l

mechanical condition (vibration amplitude) of each reactor l

plant ventilation exhaust fan shall be verified annually, or l

at the next scheduled shutdown if such verification was not l

performed during the previous year provided that the l

surveillance interval does not exceed 18 months.

SR 5.5.3f l

shall be implemented per ISI Criterion E.

l g)

Instrumentation associated with the above filters and fans l

shall be calibrated

annually, or at the next scheduled l

shutdown if calibration was not performed during the previous l

year provided that the calibration interval does not exceed 18 l

months.

SR 5.5.3g shall be implemented per ISI Criterion F.

Basis for Specification SR 5.5.3 The reactor building exhaust filter system is designed to filter the reactor building atmosphere prior to release to the facility vent stack during both normal and accident conditions of operation.

The system consists of three 50%

Fort St. Vrain #1 Technical 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 hydrocarbon and DOP respectively.

The laboratory carbon sample test results indicate a

radioactive 1

methyl iodide removal efficiency for expected accident conditions.

The surveillance test frequencies specified establish 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.

l l

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.

4 1

Fort St. Vrcin #1 Technical Specifications 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.

l In addition to routine operation, each reactor plant ventilation l

exhaust fan is functionally tested, and its capacity

verified, l

when performing the required filter tests.

The specified l

performance test and testing interval is sufficient to ensure l

adequate fan operation for the performance of their required l

safety functions under postulated loss of forced circulation l

cooling accident conditions.

l l

I 4

7 y.-

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- - + -

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.__.-.---.,----.-.m.-o-

.,,_y.

,. -,,, _, -. _,, _ ~,.. _ - _.. - - - - -. - -.,,, _. -

Fort St. Vrain #1 Technical Specifications Amendment Page 5.7-1 1

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 i

~

To ensure the prevention of any uncontrolled release of radioactivity during fuel handling and fuel storage by establishing the minimum frequency an'd type of surveillance on the equiprent for the fuel handling and storage systems.

4 Specification SR 5.7.1 - Fuel Handling Machine Surveillance 4

I The surveillance of the fuel handting machine will be as follows:

l a)

Prior to refueling, the fuel handling machine cooling water leak detector will be functionally tested.

j b)

A functional test of the Fuel Handling Machine and Isolation valve movements, interlocks, limit switches, and alarms will i

be performed or simulated prior to annual refueling periods.

9

.. - _ _ _ _. _ _. -. ~, _,. -

. __ -- - _ _,_,.-._,_.~,-.._..._.

Fort St. Vrcin #1 Technical Specifications Amendment 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 cooling system will be made.

These checks will assure the capability to maintain the proper atmosphere environment within the machine to prevent any uncontrollable release of

activity, proper purging and back filling capabilities, and the capability to maintain temperature of fuel elements within the machine below 750 degrees F.

Specification SR 5.7.2 - Fuel Storage Facility Surveillance l

The cooling water, purge, and ventilation systems for each l

Fuel Storage Well containing Irradiated Fuel shall be l

demonstrated Operable:

l a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by:

l 1.

Verifying that the outlet cooling water temperature l

of operating cooling coil (s) is 150 degrees F

or l

less, with a

flow rate of greater than 7 gpm, and l

either:

Fort St. Vrain #1 Technical Specifications Amendment Page 5.7-3 l

a)

Both cooling water coils are Operable with at least one l

coil'in operation, or l

b)

One cooling water coil is Operable and in operation, and l

the Fuel Storage Well Emergency Booster Fan is Operable.

l 2.

Verifying that the pressure within the well is at l

atmospheric pressure or slightly below.

SR 5.7.2a shall be l

implemented per ISI Criterion E.

l b.

At least once per 31 days by verifying that the Fuel Storage l

Well Emergency Booster Fan operates upon manual initiation.

SR l

5.7.2b shall be implemented per ISI Criterion E.

l c.

At least once per 18 months by:

l 1.

Performing a

Channel Calibration of the Fuel Storage j l Facility helium pressure indicators and alarms.

l 2.

Performing a

Channel Calibration of the Fuel Storage 4

l Facility cooling water system flow indicators, and flow and l'

temperature alarms.

l 3.

Verifying the capability of the Fuel Storage Well Emergency I

Booster Fan to draw a minimum of 9000 cfm of air through the l

Fuel Storage Facility.

SR 5.7.2c shall be implemented per

_ l ISI Criterion E.

l d.

Verifying during fuel handling that no Irradiated Fuel elements l

are inserted into the central position of Fuel Storage Wells.

l SR 5.7.2d shall be implemented per ISI Criterion E.

Fort St. Vrain #1 TOchnical Sp;cifications Amendment Page 5.7-4 l

Basis for Specification SR 5.7.2 l

The storage well cooling water system is designed with two l

100% capacity cooling coils supplied from independent water l

sources (See FSAR Section 9.1.2).

l The accident conditions described in the FSAR postulate the l

total loss of water cooling to one of the nine Fuel Storage l

Wells.

If this were to occur, adequate cooling could be l

achieved by an increase in the normal ventilation air flow l

to cool the well by convection on the external surface.

The l

increase in air flow is supplied by the Fuel Storage Well' l

Emergency Booster Fan.

This Specification is based on the l

analysis in FSAR Section 14.6.3.2 which uses the l

conservative assumption that a total flow of only 9000 cfm l

would be drawn (equally divided) through all three vault l

compartments of the Fuel Storage Facility, thus adequately l

protecting the affected storage well and fuel within it from l

damaging temperatures.

(There are three Fuel Storage Wells l

in each of the three Fuel Storage Vaults).

The specified l

test and testing frequency are sufficient to demonstrate the l

operability of the Fuel Storage Well Emergency Booster

Fan, l

should it be called upon for performance of its required l

safety function.

l To prevent significant oxidation of the irradiated fuel, the l

Fuel Storage Wells are designed to maintain the irradiated l

l fuel cool and under a nominally dry atmosphere of helium.

l All conditions connected with this requirement are monitored l

by

pressure, temperature, and flow sensitive devices.

The

Fort St. Vrain #1 TGehnical Sp;cifications Amendment Page 5.7-5 l

temperature and flow detecting devices maintain surveillance l

of the wells' two independent cooling water systems and are l

set to alarm at previously determined maximum or minimum l

values.

The 150 degrees F or less outlet temperature of l

water from a

single cooling coil assures the hottest l

irradiated fuel elements will be maintained below 750 l

degrees F, preventing any significant graphite oxidation in l

the event of air inleakage into the storage well.

l The Helium Storage System provides purified helium for this l

service, giving sufficient protection against a

moist l

atmosphere.

Over-pressurization of a

storage well is l

alarmed to the reactor operator and additional protection is l

provided by relief valves.

The specified surveillance l

intervals are sufficient to ensure proper operation of the l

instrumentation and to verify adequate fuel storage

.l conditions.

l The Action Statements provide for corrective actions within l

an adequate time to prevent the hottest stored fuel element l

from reaching 750 degrees F, thus obviating any significant l

oxidation in the unlikely combined events of air inleakage l

plus loss of normal cooling.

5 ATTACHMENT 3 SIGNIFICANT HAZARDS CONSI0ERATION FOR THE FSV INSERVICE INS 0ECTION AND TESTING PROGRAM UPDATE y,_y.

y-

-w._

SIGNIFICANT HAZARDS CONSIDERATION FOR THE FSV INSERVICE INSPECTION AND TESTIN2 PR03 RAM U?CATE 1.

BACKGROUND The Fort St.

Vrain Inservice Inspection and Testing Program is specifiec by tne Plan: Technical Specification Surveillance Requirements (Ref.1).

In resoonse to a commitment in the 1972 Safety Evaluation Pecort (Ref. 2) Public Service Company has been reviewing, as a continuing effort, the Inservice Inspection and Testing Program for Fort St. Vrain to feedback the acquired operating experience with the olant, and to uodate the program in light of more recent rules and regulations.

The original 1972 Safety Evaluation Report (Ref. 2) included a commitment to review the Inservice 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 approacn 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 commitment (Ref.

4).

The general Inservice Inspection and Testing Program review plan and tne priority items were further discussed in letters and at meetings between the Nuclear Regulatory Ccmmission and Public Service Company until a basic agreement was reached between both parties (Ref. 5 through 10). A schedale was established for tne review of surveillance requirements for all major plant systems and equipment by subdividing them in priority categories as requested by the Nuclear Regu? tory Commission (Ref.11).

In tne first phase of the program review and update, implemented by Amendment No. 33 to the Operating License (Ref.

12),

the surveillance reouirements pertaining to plant systems and equipment identified as priority category I were addressed (namely the crestressed concrete reactor vessel (PCRV), the reactor internals, tne reactor prima ry and secondary coolant systems, and the PCRV auxiliary system).

Public Servi:5 0 2,. -, ' ; ;':- t: proceed with a review and ucdate of the surveillance requirements for other reactor auxiliary process systems and equipment important to safety were outlined in a letter to NRC (Ref. 13).

2.

METHODOLOGY AND REVIEW PROCESS The acequacy of the surveillance requirements for incivicual systems anc components was reviewed in light of the imoortance of their safety functions to prevent or mitigate tne consequences of postulated accidents that could cause undue risks to the health and safety of the public.

The highest importance to safety was assignec to those systems anc components whicn are critical for the nitigation of a postulated permanent loss of forced circulation cooling accident where potential for fuel damage exists (Design Basis Accident No.

1), uncer tne PCRV liner Cooling mode of operation.

Following were those systems and components which are critical for the citigation of postulated loss of r.o rma l forced circulation cooling accicents, under the Safe Shutdown Cooling moce of operation.

This cooling metn:d applies t: a:cicents initiated by environmental disturbances (earthquake. tornaco) or by equipment failures sucn as feedwater or steam pipe rupture.

Next those systems anc components retuirec to mitigate the ccnsecueaces of a FOT,' :e:ressuri:ation acc' cent were consicered (Paximum Credible Accicent and Design Basis Accident No. 2).

Fi e s 11:,.

the review incluced those systems and components required to mitigate the consequences of a loss of normal cooling of a s:ent fuel storage well.

For eacn accident outlinec accve, tne various auxiliary systens and ccmoonents were systematically evaluatec for icentification of treir critical safety functions.

Existing Tecnnical Specification Surveillance Requirements were then reviewed to assess their adecuacy for ensuring operability of these icentifisc sjstems anc c;r;;nents, ince;c the, ce called u; n for the performance of their required safety functions.

Crecit was taken, as appropriate, for routine operation of systems anc components during normal plant operation as a factor in the cemonstration of neir operacility.

System operational readiness as well as coerability of pumps, compressors, fans, valves, controls anc instru entation were all accressec.

-4 In adcition, systems and components with a primary or secondary reactor coolant pressure boundary function, or with a containment isciation function, aere also reviewea for acegwacy of tn-related surveillance requirements.

As previously indicated in Amendment No.

33 to the Operating License, tne em:nasis oas placed on monitoring the structural integrity of the reactor coolant oressure bouncaries wherever practical.

3.

EVALUATIONS AND CONCLUS Q S Ine proposed cnanges to the Technical Specification Surveillance Recuirements generally exoand the scope of inservice examination and testing that is currently performed at the Fort St. Vrain Nuclear Generating Statten.

These proposed changes were initially submitted to the NRC in December, 1933, (Re f.

14).

Additional information was requested by the NRC in August,

1955, (Ref.

15) and many of the NRC comments have been incorporated in the Inservice Examination and Testing Program.

This, in essence, provices greater assurance of plant safety and reliability.

Incivicual surveillarce recuirements have teen evaluated in detail by public Service Company.

The results of these reviews revealec that ex' stir; surveillance recuirements generally were adecuate in light of plant operating experience, importance to

safety, unique cesign features and limitations, and ASME Ccce cevelopment for large HTGR designs.

Mirce modifications to surveillance intervals were made to reflect eserating experience, anc to provide cceratirg flexibility.

Accitional tests were incluced to assure tne operability and accuracy of instru eatation ani:5 can be used for monite 'n; the structural integrity of major piant ecuipment.

Accitional component testing was recom enced, as a result of cetailec revieas of plant systems, eitner when co corents important to safe plant shutdown and cooling were nct in tre scope of the cur-e-Technical Specifications, or when tne testing metnoc could ce improved to provice accitional as surance of component reliabil'ty.

Since the proposed cnanges to the Technical Specifications do net result from modifications to plant ecuipment, insteac :ney reflect acditional surveillances as recuired by inservice faspection and testinc. it c3n be concluded that tne proposed amencment will nct:

1)

Involve a

significant increase in the probability or Consecuences of in acc' cent crevicusly evaluated, o" 2)

Create the possibility of a new or diffe ert kind of accident from any accident previously evaluated, or 3)

Involve a significant reduction in a margin of safety.

Therefore, no significant hazards consideration is raised.

5 4.

REFERENCES 1.

Plant Technical Specifications 2.

Safety Evaluation Report of January 20, 1972, Secticn 3.3 3.

Public Service Company letter dated 0 tccer 13,1975 (P-78169), Inservice Inspection - Fort St. Vrain 4.

Nuclear Regulatory Commission letter dated January 15, 1979.

Inservice Inspection and Testing Program for Fort St. Vrain.

5.

Public Service Company letter dated Marcn 15, 1979 (P-79053), Inservice 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 neld en August 20, 1979, between the Nuclear Regulatory Cc mission and Public Service Company.

l l

8.

Public Service Com:any letter cated August 22,1979 (P-79176). Fort St.

Vrain Inservice Inscection and Testing Drogram.

3.

Nuclear Regulato y Ccmmission letter cate: Oct;ber 5, 1979, Pro:csed Plan of Inservice Inspection a d Testing for Fort St. Vrain.

10. Public Service Company Progress Recort.

Meeting neid on Novem er 1, 1979, between tne Nuclear Regulatory Co--ission and Public Service Company.

11.

Public Service Company letter dated Nove-cer 30, 1979 (F-79239), Fort St.

Vrain Inservice Inscection and Testing Program.

12.

Fort St. Vrain Nuclear Generating Station, Ameac ent No. 33 to Facility Operating License DPR-34, dated var:n 9. 1953.

13.

Public Service Comoany letter dated Jure 1.1983 (D-S3195) i Inservice Inscection and Test ng Status Update.

i la.

Puolic Service Company letter datec Cecemeer 30,1933 (P-33416) Proposed Tecnnical Specificati:n Cnanges -

Inservice Inspection and Testing Requirements.

15. Nuclear Regulatory Commission letter cated August 27, 1985 i

(G-35361) Inservice Inspection and Testing Requirements.

ATTACHMENT 4 ITEMIZED RESPCNSES TO THE NRC CC>MMENTS FROVIDEO IN (G-e5361) - 8/27/35 d

l i

CC.MMENT #1 1.

SR 5.2.7 - Water Turbine Drive Surveillance a)

The proposed change would exterd the annual test interval for one circulator aad the associated water supply valving in each torp to the next scneculed plant shutdown if tne test was not performed curing the previous year. We find tnis proposed change acceptable provided that the surveillarce interval does not exceed 18 months on the basis tnat (1) operating experience has illustrated sufficiently satisfactory performance of this sy stem such that no significant hazard is created by an extension of this test

interval, (2) t potential hazards from an additional shutdown and startup transients would be avoided and (3) decay heat can still be removed via steam driven circulators or the Liner Cooling System.

The provision to not have the surveillance interval exceed 18 months is consistent with NRC Standard Tecnnical Specifications for LWRs and ensures a minimum surveillance interval, b)

The proposed change would entend the annual test interval for safety valves in the water turbine supply lines to the next seneduled clant shutdown if the test was not performed during the D-evious year.

We find this eco:0 sed change acceptacle, provided that the surveillance interval does not exceed 13 months, based on the same reasons as given in comment 1 (a) above.

c)

The phrase "every three months" has been changed to read " quarterly" for the functional testing of both turbine water removal pumps and the turbine removal tank overflow to the reactor building sump.

We recommend that the standard technical specification terminology of 92 days be used rather than that proposed.

RESPCNSE TO CCYMENT =1 PSC concurs witn the NRC comment (s) and will implement the above requirement (s) in applicable Surveillance Protedares and their associated schedules.

COMMENT d2 2.

SR 5.2.S - Searing Water Pump and Makeup Pump Surveillance The bearing water makeup pumps have been added to this surveillance. We recommend that circulating bearing water pumps ce added to the title for consistency.

a)

The phrase "every three months" has been changed to read

" quarterly" for tne operation of the Normal Makeup Pump in the recycle mode. We recommend that the terminology of STS of 92 days be used rather tnan that proposed.

c)

No NRC ccmment.

b)

The phrase "every three months" has been cnanged to read

" quarterly" 'or the functional testing of the Eeergency Makeup Pump. We recommend that the terminology of STS of 32 days be used rather than that proposed.

d)

The pr0 posed surveillance for the tearing pumps would provide for a functional test of the pumps and associated instruments and controls at each scheduled plant shutdown, or at the next scheduled plant shutdown if less than a year has elapsed from the previous test.

This schedule would not disrupt normal plant operation and provides a test not Oreviously required.

We find this surveillance acceptable as it is consistent with current practice in the application of the ASVE Ccce to current plant, r-svided tne surveillance interval does not exceed IS months.

RESPONSE TO CC* MENT 82 PSC concurs with tne NQC comment (s) and will ircle ent tf e above requirement (s) in applicable Surveillance Procedures and their associated schedules.

4 COMMENT #3 3.

SR 5.2.9 - Helium Circulator Bearing Watee Accumulators The proposed test interval for testing of tne helium circulator bearing wate-accumule. tors, instruments and centrols would be exter. cec trem mentniy to quarterly.

Tne licensee justfries tnis change on a review of prior test results whicn shows satisfactory perf rmance.

Easec on nis justification oe finc tne proposea change acceptable. We recommend that the quarterly interval be stated as 90 days.

RESPCNSE TO CCFMENT =3 PSC concurs with the NRC ccmment(s) and will implement the above requi remen t( s )

in apolicaole Surveillance Precedures and their asscciated schedules.

I 9

I d

CCM.YENT #4 4

SR 5.2.10 Fire Water System / Fire Suppression Water System Surveillance a)

No NRC comment, b)4)(b)

A reduction by 5 percent in the flow and head testing requirements for the firewater pumps nas been proposed to account for pump cegradation.

Degradation to this degree is acceDtacle under the ASME Ccde and the pump performance continues to exceed the minimum perfomance requirements by a sufficient margin.

We find the proposed changes acceptable.

b)4)(d)

The fire suppression water system pressure is changed to read "275 feet water gauge" from "125 psig."

This is an acceptable change.

RESPCNSE TO CCvvENT e4 The above recuire ents will be incor;; rated into applicable Surveillaa.:e Pr:cecures aac t*eir assectatec :: eeut es.

i

(

l v

l COMMENT #5 5.

SR 5.2.16 - PCRV Closure Leakage Surveillance Requirements a)-f)

No NRC comments, g)

The proocsed change would require once during each refueling cycle a leakage test and for eacn nelium curification cooler well, a calioration of the well pressure mcnttoring instruments and a functiceal test of the instruments and controls used to automatically isolate the purification system.

The addition of this surveillarce reautre ent verifies the operability of instru ents used to monitor containment integrity.

To make a judgement on the acceptability of tre proposed leakage test a description of how this requirement 4

meets the intent of the ASME Code, Article

IGB,

" Examination and Inspection," should be provided.

RESDCNSE TO CCvvENT =5 The proposed leavage test is in accordance with ASME Section XI Article IGB-2000 " Examination and Testing".

IGS-2510 in this article re:uires confor ance witn IGA-5000 and IGB-5C00. Conditions reautred by IGA-5320, "Peeumatic Testing Procedures", will be included in the test peccedures issued to perform tne leak test.

IGB-5300 re:uires testing at or near the erd of each inspection interval as defined either in Iaci) l33-2til-1 or Table ICE-2412-1.

The inspection l

interval specified in proposed SR 5.2.16 (g) 1.e., "occe during each refueling cycle" is well witnin the requirements of Section XI.

i i

.m

m COMMENT #6 l

t i

6.

SR 5.2.21 ACM Transfer Switches, Valves, and Instrument i

Surveillance.

The surveillance requirement has been retitled from the previous title of SR 5.2.21 - Han: valve and Transfer

5. itch Surveillance.

1 a)

For those valves and transfer switches that must be manually positioned for actuation of the Alternate Cooling Method (ACM) moce of operation, the licensee proposes to change the surveillance interval to annually or at the next scheduled plant shutdown if such a test was not performed during the previcus year.

While we understand that full operation of these valves ard switches is not possible during plant oceration, we nevertheless believe that it is necessary to demonstrate operability of these components

, tore frecuently.

Thus we do not find this proposed change acceptable and recommend that the original surveillance interval for an operaoility check of this ecuipment be maintained (4 to 8 months) and a full functicnal test be performed at annual or at refuelirg intervals not to er:eed 13 montns.

b)

A new surveillance require ent for calibration at eacn refueling interval has been crocosed for local indicators for the helium curtftcation dryer inlet te?:erature, for the helium purification puted:wn line pressure and for the reactor plant cooling nater surge tank cover gas pressure.

From tne informaticn provided it is not clear that:

(1) these proposed surveillances are sufficient to assure operational readiness of these com;onents, and (2) the cc?corents to be given i

surveillance provide a complete set to assure operational readiness of the systems they

serve, j

Therefore, information needs to ce provided addressing the aDove.

RESPCNSE TO CCvvE'J n6 a) PSC concurs with the f.4 comment (s) and will implement the acove requi re?en t( s )

in applicable Surveillance Pro:edures ard associated schedules.

O RESPONSE TO COMMENT e6 (Cont'd) b.1)

A new surveillance requirement for calibration at each refueling interval for the helium purification cryer inlet temperature, the helium purification pumpdown line pressure and for the reactor plant cooling water surge tank cover gas pressure has been proposed by P5C.

These pr cosed surveillances assure operational readiress of the components.

Temperature instruments are supplied with ACM electrical cower to monitor helium purification cryer inlet temperature and provide information acout the acequacy of cooling for processing of reactor coolant by the purification system.

The pressure instruments in the helium purificaticn system pumpdown line allow ronitorieg of the progress of PCRV depressurtration and also monttors PCRV pressure subsecuently.

Each PCRV liner cooling outlet subheacer is equipped with a local temperature irdicator to indicate the potential need to increase system pressure to prevent boiling; that pressure is menitored at tne surge tanks where the cover gas pressure can be adjusted as required.

b.2)

The new proposed surveillances concerning the aforementioned instru ents provice information to assure ccerational readiress of the systems they serve, procedure SR-RE-50-X, (10-15-35), ACM INSTRUMENTS CALIER4710N, is utilized to calibrate Iccal indicators for the elium purification dryer inlet te cerature (TI-23157), pu ec;-9 line pressure (PI-23162 and PI-23163),

and syste-46 surge tank cover gas pressure (PI-4665 and PI-4666).

Inese calibrations are performed annually and are consistent with the Standard Technical Specifications.

The functions to be monitored by these instruments include the progress of PCRV depressurization, cooling for acequate operation of the helium purification system, prevention of boiling in the PCRV liner cooling water syste9, and radioactivity of gaseous effluents.

o COMMENT #7 7.

SR 5.2.24 Reactor Auxiliary Cooling Water Systems Surveillance.

The title of this surveillarce requirement has been changed from Circulating Water Makeup System Survaillance. We find this acceptable.

l a) and c) No NRC comments.

t b)

The surveillance interval for functionally testing l

each circulating water pump is proposed to be i

extended to monthly from weekly. As the monthly i

interval is in accordance with the ASME Ccde and surveillance requirements have been added to the proposed change regarding instrument calibration, pump performance capability and mechanical condition, we find the proposed change acceptable, c)

The peccosed surveillance recuirement aculd be a new requirement pertaining to the integrity of the circulating water makeup pond embankments.

The proposed aedition is consistent with LWR service water requirements and is considered acceptacle, i

e)

The proposed surveillance require *ent would be a l

rea re;uirement certaining to tne testing of each service water pump and the associated instru ents.

We nave revieaed these -e1.irements and have found the9 in partial accord with the ASWE Code. We

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aculd find the? accectable if the licensee either conforms to the detailed require ents of the ASVE i

Cede or provides an acce: table alternative.

l f)

The proposed surveillance recutre ent would be a new requirement pertaining to the testing of each l

reactor plant cooling water Du o ard the a

associated instruments. We have revieaed these i

re:vire ents and have found the9 ir cartial accord r

l att9 the ASVE Code.

'ae would find the9 accept 4Dle

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ff the licensee either conforms to the detailed l

recuirements of the ASME Code or provides an j

accerta:1e alternative.

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The croposed surveillance requirement ould be a new reautrement pertaining to the testing of eacn i

purification cooling water pump and the associated i

intstru ents, he have reviewed these r+;utrements

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and have found them in partial accord with the ASVE Code. We wculd find them acceotable if the licensee wither conforms to tne cetailed

% utrements in accordance with the ASSE Coce or

,11 vices an acceptable alternative.

h)

The procosed sur.e111ance requirement ould be a new reautrement pertaining to the testing and calibration af instruments and valves used for I

automatic isolation of portions of the reactor plant cooling water system.

We find that an interval of each refueling cycle not to exceed 13 j

months for a full stroke test of each valve is

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acceptable for those valves that cannot be tested I

during plant operation.

However, the interval for I

a functional check of those valves and instruments i

capable of being tested by a partial stroke should l

be performed semi-annually in accordance with the l

prececent of SR 5.2.21 or cuarterly in accardance l

l with the guidance of the ASVE Code.

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RES?ONSE TO CCvvi'.T C I

SR 5.2.24 e) e d c) in, above requirements will be incorporated L

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into existing Surveillance Procedures and their associated j

senedules.

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f) and g):

The major put: cara etees, as y:ve revien l

l indicates, are in agreement with the ASME Code Subsection IGP and are pertinent for inclusion into the Technical $cecifications.

However, there are many other detailed requirtaents listed in the i

ASWE Code which will be included in esisting clant surveillance i

procedures, but are too cetailed for inclusion in the Technical i

Soecificattoas.

It is DSC's practice to use Section Al Code

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requirements in t*e approorlate plant surveillance procedure.

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h)

PSC concurs with the NRC comment to limit the time between full functional tests to

,10 more than 13 months, partial stroking and full stroking of the valves used for automatic

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isolation of portions of the reactor plant cooling water system and the purification cooling water system will be performed twice annually as recer erded by the NRC.

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CCMMENT #8 8.

SR 5.3.4 - Safe Shutdown Cooling valves Surveillance The licensee proposes to test valves used for safe shutdoon cooling on an annual basis or folic.ing scheduled plant shutco.n.

Tnis is not acceptable except for cases obere it is not chysically possible to perform a more frequent surveillance.

For tnose valves tnat can de tested curing reactor operation, are required to initiate and function during safe shutcc.n cooling, and which are not called upon to operate during cornal plant c eration, the licensee should provide testing require ents and intervals in conformance witn the ASME Code.

R E 5 00 N S E TO CCVV EN_T__* B.

P5C concurs with the NRC comment which Itmits time between tests to 13 montns and reautres partial stroking in accordance with Paragraphs l

IGV-3411 and IGV-3412 of Section XI, Division 2 of the ASME Pressure vessel Code. Tne aeove reautrement will te imolemented in 4:pitcable Surveillance Procedures a9d their asscciated schedules.

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1 CCMMENT #9 i

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SR 5.3.9 - Safety Valves Surveillance a)

This proposed surveillance would require verification of safety valve setooints at five year intervals for the steam generator su;erneater, reheater and steam / water daic tank.

The requirement is satisfactory provided 1) that a senedule for additional testing is developed for any valve in a system that fails to function on a regular test and 2) that an acceptable test procedure is ceveloped or referenced.

Conformance to the ASME Code (Subsection IGV) for Class C valve testing would meet the above requirements and simplify i

I the development of an acceptable technical specification j

requirement.

I b)

The licensee proposed that all other Class I safety valves not covered by other surveillance requirements shall be t

i setpoint tested at 10 year intervals.

This is unacceptable.

I The licensee snould conform with the ASME Code in this i

matter.

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BESPCNSE TO C0YMENT #9 t

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Ceneral - Please note that Amendment #39, ni:h was approved 1/25/84 l

Superseced the text of SR 5.3.9 wnich the NRC reviewed.

viponses are based on the

urrent tent.

l a) A schedJle for additional testing reflected in IGV-3513 of ASME Code Section XI, Division 2 will ee incorporated in appropriate pro:ecures and schedules.

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Folloaing are D$C Surve111ar:e Prc:edures devol ped for Safety L

Valve surveillances:

1.

Steam Generator $uperheater Safety.alve Test (Main Steam) i SR 5.3.9.1-2.5Y (Issue 3) Ef fective Date 1-4-85.

2.

Steam-Generator Reneater Safety Valve Test. SR 5.3.9.2-2.SY i

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(!ssue 4) Effe:tive Cate 12-14-94

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Outo Tank $afety Valve Test SR 5.3.9.3-2.5Y (Issue 4)

Effective Date 1-11-35.

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RESPCNSE TO CCMMENT 89 (CONTINUE 0J b) All other Class I safety valves not covered by Technical Soecification Surveillance Requirements are testec at 5 year l

intervals per the ASME Code.

Therefore, this paragraon has ceen l

deleted frem the 12/30/83 amendment application.

Furthermore, these other s a fe ty valves are not relied upon to prevent or l

mitigate any accident that is analyzed in the FSAR.

They are l

provided for pressure boundary integrity only.

All Class I safety valves are presently tested according to applicable Code requirements and will be included in the separate ISIT Program for FSV.

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l COMMENT #10 10.

SR 5.4.4 - PCRV Cooling Water System Temperature Instrument Surveillance l

a)

The proposed surveillance requirement clarifies the

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monthly monitoring of the PCRV cooling system water I

inlet temperature, individual tube water outlet I

temperatures, and the associated outlet temperature alarms.

We find the proposed clarification acceptable, b)

The prooosed surveillance requirement clarifies requirements for annual calibration of the temperature renitoring

scanner, the inlet and outlet header temperature indicators, and the outlet subheader temperature indicators.

We find the proposed clarification acceptable.

c)

The pecoosed survelliance requirement would extend calibration of the inlet header and tube outlet thermaccuples from an annual interval to a five year icterval. We do not find this surveillance interval extension acceptable since no justification has been provided.

RESPCNSE TO CCvvEst d10 SR 5.4.4 4) and b) - The above require eats will be incorporated into evistic; Surveillance Procedures and their associated schedules.

SR 5.4.4 c)

Lack of direct accessibility to certain trermoccuoles during operation makes it impractical to calibrate on a yearly basis the percentage of instru ents currently specified in SR 5.4.4 Because all instruments are of similar design, and because operating conditions are not severe, it is not anticipated, as confirmed by emperience to date, that any su:stantial differences in instrument behavior will cccur over time between the various subheaders.

SR 5.4.4 was therefore changed to provide more flexibility in the selection of thermoccuoles to be caltbrated in any carticular year, while specifying a

five year calibration cycle for all the thermocouples which crovide input to the scanner.

Currently, the thermoccuoles are calibrated on a sta year cycle.

The remaining requirements of $R 5.4.4 (monthly scanner temperature alarn check and annual calibration of

$ubheader outlet temperature indicators) were found to be adequate.

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RESPCNSE TO CCM.uENT *10 (CONTINUED)

Technical Specification SR 5.4.5 currently incluces requirements for a monthly read ut of flew, a monthly check of the alarms, and an annual calibration of the flow scanner and alares.

It also reautres an annual calibration for six of the thirty-six subheader flow transmitters.

Since the reactor ray have to be shut down (to comply with LCO 4.2.14) to remove and replace the flow meters, SR 5.4.5 has been changed to allcw the calibration interval to extenc to the next seneduled shutdoon.

These existing surveillance requirements, as mocified above, are sufficient to assure that heat loads can be accurately determined and that abnormal flow and temperature conditions are alarmec.

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O CO* MENT #11 11.

SR 5.4.5 PCRV Cooling Water System Flew Instruments Surveillance The procesed surveillance would extend annual calibration of the flow scanner instruments and alarms anc :ne six subheader ficwmeters to tne next scneduled plant shutcown if they were not calibrated during tne previous year. We find this extension acceptable up to a surveillance interval not exceeding 18 months since the potentials for additional 4

plant transients are reduced and since the LWR-STS, in general, sce:ifies surveillance intervals not to exceed 18 months when utill:ing intervals of shutdown or refueling.

RESP 0'JSE TO CCMvENT all SR 5.4.5 PSC concurs with the NRC ccm. ment and will implement the above requirement in the applicable Surveillance Procecure and associated schecule, i

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o CCMMENT #12

12. SR 5.5.3 - Reactor Building Exhaust Surveillance a)-d)

No NRC com.ments.

e)

The proposed surveillance requirement would verify at weekly intervals that tne total pressure d-:p across the HEpA filter and charccal absorcer ban (s to ce less than six inches of water at filter design flow 10

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percent.

Subsequent to this proposal the Draft Upgrade Technical Specifications have been issued and it is our opinion that appropriate portions of Draft Item 4.6.5.2.c.3, " Reactor Building Exhaust System," which generally specifies the testing requirements given in tne Standard Technical Specifications, snould be utill:ed in lieu of PSC's proposal.

f)

The proposed surveillance requirement would verify annually the performance capability and mechanical condition of each exhaust fan or at the next scheduled shutdown if such verification was not performed during the crevious year.

We find this surveillance requirement acceptable provided that the surveillance interval does not exceed 15 months, since the L'aR-STS, in general, specifies survel' lance intervals not to esceed 13 montns when utili:ing intervals of shutdown or refueling.

g)

The croposed surveillance would require calibration of the instrumentation associated with the filters and fans at annual intervals or at tre next scheduled shutdown if calibration was not performed during the previous year.

We find tnese surveillance requirements are a::eptable up to a surveillance interval not exceeding 13 months since the potentials for additional plant transients are reduced and sin:e the LWR-STS, in general, specifies surveillance intervals not to exceed 13 months when utill:1ng intervals of shutdown or refueling.

RESPCNSE TO CCvvENT el:

SR S.5.3 e), f) and g) - PSC agrees with the NRC comment (s) and will imple ent the above require ent(s) in apolicable Surveillance Procedures and associated schedules.

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CCMvENT #13 13.

SR 5.7.2a - Fuel Storage Facility Surveillance The preposed surveillance would reouire an annual functional test of tne emergency ventilation system.

This surveillance, together with Darts a and b of $R 5.7.2c, have been substantially revised in Draft item 4.9.3,

" Fuel Storage Well.

We believe action on this item should be deferred until discussions on com. ment 13U telow are completed.

13U. Draft Item 4.9.3 - Fuel Storage Well The material in this Draft Item appears to represent an improved and better developed surveillance than that described in SR 5.7.2 and should be considered for inclusion in SR 5.7.2a.

RESPONSE TO CCYYENT W13 and #130 PSC concurs with the NRC comment (s) and will implement the above requirement (s) in applicable Surveillance Proce: ares and associated schedules.

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