ML20031G622

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Forwards Draft Responses to Questions in NRC Re Auxiliary Sys.Responses Will Be Incorporated Into FSAR Amend
ML20031G622
Person / Time
Site: Perry  FirstEnergy icon.png
Issue date: 10/16/1981
From: Davidson D
CLEVELAND ELECTRIC ILLUMINATING CO.
To: Tedesco R
Office of Nuclear Reactor Regulation
References
NUDOCS 8110230386
Download: ML20031G622 (38)
v • d • e


Text

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t THE CLEVELAND ELECTRIC ILLUMIN ATING COMPANY P o BCX 5000 m CLEVELAND. oH:o 44101 e TELEPHcNE (216) 622-9800 m ILLUMINATING BLDG e 55 PUBLIC SoVARE Serving The Best Location in the Nation Dalwyn R. Davidson VtCE PRESIDENT SYSTEM ENGINEERING AND CONSTRUCTION October 16, 1981 Mr. Robert L. Tedesco Assistant Director for Licensing Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 Response to Request for Additional Information -

Auxiliary Systems

Dear Mr. Tedesco:

This letter and its attachment is submitted to provide draft responses to the concerns identified in your letter dated September 1, 1981 in regard to Auxiliary Systems. It is our y

intention to incorporate these responses in a subsequent amendment to our Final Safety Analysis Report.

Very Truly Yours,

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Dalwyn R. Davidson Vice Ilasident System Engineering and Construction DRD: mlb Attachment A

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M. D. Houston k(

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w91 NRC Resident Inspector 2

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410.5 Section 3.6.2.3.5 of the FSAR indicates that analyses of (3.4.1) flooding resulting from high or moderate line failures have been performed.

For areas contining high or moderate pipes, present the results of these analyses on a room by room basis to demonstrate that the plant will be able to achieve safe shutdown considering the height to which the water would rise assuming the failure of one of the pertinent sump pumps.

Response

The response to this question will be provided by November 30, 1981.

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410.6 Section 3.6.2.3.4 of the FSAR stated that compartment pressuritation (3. 6.1) analyses have been made for all compartments containing high-energy lines. Present the result of these analyses including the peak pressures and temperatures and blowdown duration, and state how the blowdown is terminated, for compartments outside containment.

Verify that essential equipment located within the compartments are capable of operating in the environment resulting from high-energy line failures.

Response

The response to this que: tion is contained in revised Sections 3.6.2.3.4, 3.11, and Tables 3.11-2 through 3.11-8.

Section 3.11 and Tables 3.11-2 through 3.11-8 are being revised as a result of Nureg 0588 and will be provided consistent with the Equipment Qualification Program review schedule.

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Restraints, connections, anchorages and the supporting structure are designed for the maximum reactions obtained from the dynamic analysis.

3.6.2.3.3.4 Emergency Core Cooling Piping System Pipe Whip Restraints -

Inside Containment Loading combinations and design criteria for ECCS piping system pipe whip restraints inside containment are as presented in Section 3.6.2.3.3.1.

3.6.2.3.3.5 Other 'ligh Energy Piping System Pipe Whip Restraints - Inside Containment 5

Loading combinations and design criteria for other high energy piping system pipe whip restraints inside containment are as presented in Sections 3.6.2.3.3.1 and 3.6.2.3.3.3.

3.6.2.3.3.6 High Energy Piping System Pipe Whip Restraints - Outside Containment 2

Loading combinations and design criteria for high energy piping systems outside containment are at presented in Sections 3.6.2.3.3.1 and 3.6.2.3.3.3.

3.6.2.3.4 Compartment Pressurization Analysis Based on the blowdown time histories generated using the criteri.7 of Section 3.6.2.2.1, all compartments containing high energy lines have been analyzed for the highest energy release rate pipe rupture event to determine the maximum loadings of the compartment. The results of these analyses are presented in Section 3.11, Tables 3.11-2 through 3.11-8, including peak pressures, NS6 l

temperatures, and duration. Means of terminating the blowdowns are indicated in

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notes to Tables 3.11-2 through 3.11-8.

These conditions then become the design parameters.

See Section 3.8.

3.6.2.3.5 Flooding Analysis Based on blowdown from high energy line breaks after Section 3.6.2.2.1 and l

leakage from moderate energy line cracks, flooding of safety related structures has been determined.

In no case are safe shutdown systams jeopar '2eA by the d

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410.7 Describe the means provided in the scram discharge system (4.6) design to meet the criteria enumerated in Section 4 of the Generic Safety Evaluation Report BWR Scram Discharge System, dated December 1,1980, and transmitted to you by NRC. letter J

dated December 22, 1980.

Response

The NRC staff's acceptance (as given in NUREG-0619) of the CRDRL deletion is based on completion of the following four requirements:

1.

Demonstration, by test of concurrent two CRD pump operation.

2.

Installation of equalizing valves between the cooling water header and the exhaust water header.

3.

Installation of flush ports on the exhaust header if carbon steel piping is atilized.

4.

Rerouting the flow stabilizer loop to the cooling water header with stainless steel piping.

In response to the first stated requirement, CEI has committed to testing concurrent two CRD pump operation.

The GE CRD system specifications (i.e., Design Spec., Design Spec. Data Sheets, P&ID, et-.) comply to the latter three requirements. The CRD system P&ID shows 1) the pressure equalizing valve as communicating with the cooling water header and the exhaust water headar, 2) the flow stabilizer loop routed to the cooling water header, and 3) both the exhaust header and the flow stabilizer leop as requiring the use of stainless steel piping.

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410.7 (Cont'd) (pg. 2)

CEI has also committed to install the non-cladded feedwater nozzles and triple-sleeve with duel piston ring type feedwater spargers. Perry's current design already has RWCU flow routing through the feedwater piping in the steam tunnel.

The above commitments meet the recommendations in Section 4 cf NUREG 0619 and NRC's generic letter 81-11 on the cubject dated February 20, 1981.

410.8 Describe the effects on the safety and operability of the (4.6) control rod drive hydraulic system assuming the drive / cooling water pressure control valve fails either closed or open.

Response

The drive / cooling water pressure control valve (PCV) in the failed closed or open condition can only effect the velocity of the control rod during insert or withdrawal.

Since the scram pressure source is independent of the drive water pressure, this PCV failure is not a safety concern.

The function of the F003 press re control valve (PCV) is to provide a means of adjusting the drive water header and cooling water header pressures.

The F003 PCV is a manually controlled motor operated valve which is controllable from the main control room.

Indicating lights are provided in the control room for the valve full open and full closed positions. Adjustment of the F003 PCV in concert with adjustments to the F002 flow control valve permit adjustment of the drive water header pressure to approximately 260 psi above vessel pressure while, at the same time, maintaining the drive cooling water header pressure at approximately 20 psi above vessel pressure.

If the F003 PCV were to fail to a full-open position, the cooling water pressure would increase and the drive water pressure would decrease. The resulting cooling water pressure increase could cause control rods to drift inward.

The existence of rod drifts would be alarmed to the control room operator for appropriate action.

The resulting drop in drive water pressure would make normal control rod notch movements impossible but would not affect the ability of the scram function.

Conversely, if the F003 PCV were to fail to a full-closed position, the cooling water pressure would decrease while the drive water pressure would increase.

The reduction in cooling water pressure (and flow) would eventually lead to high CRD temperatures being alarmed in the control room.

The CRD system's scram function would not be affected by the increase in drive water pressure.

In the limiting case, the resulting increase in drive water pressure would reach up to the shutoff pressure of the supply pump (2000 psig)..

The occurrence of this condition during withdrawal of a drive at zero reactor pressure will

410.8 (, Cont'd) (pg. 2) result in a drive pressure increase from 260 psig to no more than 2000 psig.

Calculations and tests indicate that the drive would accelerate from 3 inch /

see to approxi" *r'v 7 inch /sec. The rod movement would stop after the 4

driving sigt aoved or a rod block is enforced by the Rod Control and Information The peak fuel enthalpy for drive speeds of approximately 7 incl /sec it

,alow the 170 cal /gm. fuel design limit. Therefore, due to provisions in t.he system design and margin in the fuel design, this postulated scenario will not compromise the integrity of the fuel.

I In both of the cases described above, the manually operated bypass PCV (F004) in conjunction with isolation gate valves located upstream and downstream of the F003 PCV would enable the operators to take corrective action.

In conclusion, although the failure to the full-open or full-closed position of the drive / cooling water PCV would cause perturbation in the CRD system operation, it does not present a safety problem or affect the scram capability i

of the CRD system.

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410.9 Describe the means provided in the control rod drive system (4. 6) design to meet the criteria enumerated in Sections 4 and 8 of NUREG-0619, BWR Feedwater Nozzle and Control Rod Driven Return Line Nozzle Cracking and verify that this design is in full compliance with those sectiona of the document.

Response

The design of the scram discharge system can be found in 4.6 of the FSAR.

The criteria given in the Generic Safety Evaluation Report - BWR Scram Discharge System, dated December 1, 1980, are organized under different headings. These headings are 1) funcrional criteria, 2) safety criteria,

3) operational criteria, 4) design criteria and surveillance criteria.

The revised scram discharge system meets the criteria enumerated in the Generic Safety Evaluation Report - BWR Scram Discharge System. A summary of each criteria is given below along with a discussion of how the scram discharge system complies.

Functional Criteria Functional Criterion 1 The scram discharge volume shall have sufficient capacity to receive and contain water exhausted by a full reactor scram without adversely affecting control-rod-drive scram performance.

GE Compliance:

A minimum scram discharge volume of 3.34 gallons per drive is specified through the system design specifications.

This minimum scram discharge volume is based on conservative assumptions as to the performance of the scram system.

In the event of a coolant leak into the SDV an automatic scram will occur before the required SDV available volume is threatened.

Safety Criteria Safety Criterion 1 No single active failure of a component or service function shall prevent a reactor scram, under the most degraded conditions that are operationally acceptable.

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410.9 (Cont'd) (pg. 2)

GE Compliance:

!;o single active failure in the scram system design will prevent a reactor scram. The GE scram discharge system design meets the NRC acceptance criterion for Scfety Criterion 1.

Partial loss or full loss of service functions will result in either not adversely affecting the r. cram system function or a full reactor scram.

The GE system requirements state that there shall be no reduction in the pipe sire of the header piping going from the HCUs to and including the Scram Discharge Instrument Volume (SCIV).

This hydraulic coupling permits L

operability of the scram level instrumentation prior to loss of system function. The scram level instrumentation are redundant and diverse to assure no single active failure or common mode failure prevents a reactor scram.

Safety Criterion 2 No single active failure shall present uncontrolled loss of reactor coolant.

GE Compliance:

Redundant Scram Discharge Volume (SDV) vent and drain valves are provided as psrt of the SDV modifications. The redundant SDV valve configuration assures that no single failure can result in an uncontrolled loss of reactor coolant. An additional solenoid operated pilot valve controls the redundant vent and drain valve.

The vent and drain system is therefore sufficiently redundant to avoid a failure to isolate the SDV due to solenoid failure. The vent and drain valve's opening and closing sequences are controlled to minimize excessive hydro-dynamic forces.

The modifications will be installed in Perry Units 1 and 2.

Safety Criterien 3 The scram discharge system instrumentation shall be designed to provide redundancy, to operate reliably under all conditions, and shall not be adversely affected 'oy hydrodynamic forces er flow characteristics.

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410.9 (Cont'd. ) (pgo 3)

GE Compliance:

Diverse, and redundant level sensing instrumentation on the Scram Discharge Instrument Volume (SDIV) is provided for the automatic scram function.

SDIV water level is measured by utilizatiec of both float sensing and pressure sensing devices.

Instrument tapa nave been relocated from the vent and drain piping to the SDIV to protect the level sensing instrumentation from the flow dynamics in the scram discharge system.

Each SDIV has a redundant instrument loop. A one-out-of-twc twice logic is employed for the automatic scram function. This instrumentation arrangement assures the automatic scram function on high SDIV water level in the event of a single active or passive failure.

T*,ese SDV modifications will be installed in Perry Units 1 and 2.

Safety Criterion 4 System operating conditions which are required for scram shall be continuously monitored.

GE Compliance:

See GE response to Safety Criterion 3.

Safety Criterion 5 Repair, replacement, adjustment, or surveillance of any system component shall not require the scram function to be bypassed.

GE Compliance:

The SDIV scram level instrumentation arrangement and trip logic allows instrument adjustment or surveillance without bypassing the scram function or directly causing a scram.

Each level instrument can be individually isolated without bypassing the scram function. A one-out-of-two twice trip logic is employed. Plant procedures will insure that the scram function is not bypassed during 'urveillance, repair, or replacement of any system component.

Operational Criteria Operational Criterion 1:

Level instrumentation shall be designed to be maintained, tested, or calibrated during plant operation without causing a scram.

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410.9 (Cont'd) (pg. 4)

G.E. Compliance:

I See GE response to Safety Criterion 5.

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Operational Criterion 2:

The system shall include sufficient supervisory instrumentation and alarms to permit surveillance of system operation.

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f GE Compliance:

Supervisory instrumentation and alarms such as accumulator trouble, scram valve air supply low pressure, and scram discharge volume not drained 1

alarms, are adequate and permit surveillance of the scram system's readiness.

Operational Criterion 3:

The system shall be designed to minimize the exposure of operating personnel to radiation.

1 GE Compliance:

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consideration in equipment design and location.

Operational Criterion 4:

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Vent paths shall be provided to assure adequate drainage in preparation j

for scram reset.

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GE Compliance:

t A vent line is provided as part of the scram discharge s stem to assure proper drainage in preparation for scram reset.

G.E. -pecifications require the vent ac be provided by a dedicated vent line with a non-i submerged discharge to the atmosphere.

Furthermore, additional vent l

capability is provided by the vent line vacuum breakers. The vacuum j

breakers are required to have a differential pressure no greater than 5 inches of water.

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410.9 (Cont'd) (pg. 51 Operational Criterion 5:

Vent and drain functions shall not be adversely affected by other system interfaces. The objective of this requirement is to preclude water backup in the scram instrument volume which could cause spurious scram.

GE Compliance:

The SDV vent and drain lines are required to be dedicated lines that discharge into the Radwaste System. Vacuum breakers on the SDV vent lic.e and shut-off valves on the SDV vent and drain lines preclude water from siphoning back into the SDIV from the Radwaste System.

Design Criteria Design Criterion 1:

The scram discharge headers shall be sized in accordance with GE OER-54 and shall be hydraulically coupled to the. instrumented volume (s) in a manner to permit operability of the scram level instrumentation prior to loss of system function.

Each system shall be analyzed based on a plant-specific maximum in-leakage to ensure that the system function is not lost prior to initiation of automatic scram.

Maximum in-leakage is the maximum flow rate through the scram discharge line without control-rod motion, summed over all control rods. The analysis should show no need for vents or drains.

GE Compliance:

As discussed in response to Functional Ctiterion 1, a minimum scram discharge volume of 3.34 gallons per drive is specified through the system design specifications. Furthermore, the GE system requirements state that there shall be no reduction in the pipe size of the header piping going from the HCUs to and including the SDIV. The SDIV shall be directly connected to the scram discharge volume at the low point of the scram discharge header piping. These requirements satisfy the NRC's acceptance criteria for Design Criterion 1.

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410.9 (Cont'd) (pg. 6)

Design Criterion 2:

Level instrumentation shall be provided for automatic scram initiation while sufficient volume exists in the scram discharge volume.

i GE Compliance:

See Response to Functional Criterion 1 and Design Criterion 1.

Design Criterion 3:

Instrumentation taps shall be provided on the vertical instrument volume and not on the connected piping.

GE Compliance:

See Response to Safety Criterion 3.

Design Criterion 4:

The scram instrumentation shall be capable of detecting water accumulation in the instrumented volume (s) assuming a single active failure in the 4

instrumentation system or the plugging of an instrument line.

GE Compliance:

1 See response to Safety Criterion 3.

Design Criterion 5:

Structural and component design shall consider loads and conditions including those due to fluid dynamics, thermal expansion, internal l

pressure, seismic considerations and adverse environments.

GE Compliance:

The SUV and associated vent and drain piping is classified as important to safety and required to meet the ASME Section III Class 2 and Seismic Category I requirements.

Design Criterion 6:

The power-operated vent and drain valves shall close under loss of air and/or electric power. Valve position indication shall be provided in the control room.

GE Compliance:

The present vent and drain valve design operation meets'this criterion.

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410.9 (Cont'd) (pg. 7)

Design Criterion 7:

Any reductions in the system piping flow path shall be analyzed to assure system reliability and operability under all modes of operation.

GE Compliance:

See response to Design Criterion 1 Design Criterion 8:

System piping geometry (i.e., pitch, line size, orientation) shall be such that the system drains continuously during normal plant operation.

GE Compliance:

All SDV piping is required to be continuously sloped from its high point to its low point.

Design Critesion 9:

Instrumentation shall be provided to aid the coerator in the detection of water accumulation in the instrumented volume (s) prior to scram initiation.

GE Compliance:

The present alarm and rod block instrumentation meets this criterion.

Design Criterion 10:

Vent and drain line valves shall be provided to contain the scram discharge water, with a single active failure and to minimize operational exposure.

GE Compliance:

See Reponse to Safety Criterion 2.

Surveillance Criteria Implementation of surveillance prc:edures to comply with the following surveillance criteria will be included in the plant surveillance program.

(The GE recommended Standard Technital Specification complies with the intent of the Safety Evaluati'on Report's Surveillance Criteria.)

410.9 (Cont'd) (pg. 8) l Surveillance Criterion 1:

i Vent and drain valves shall be periodically tested.

Surveillance Criterion 2:

Verifying level detection instrumentation shall be periodically tested in place.

Surveillance Criterion 3:

The operability of the entire system as an integrated whole shall be demonstrated periodically and during each operating cycle, by demonstrating scram instrument response and valve function at pressure and temperature at approximately 50% control-rod density.

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410.10 With regard to the new fuel storage, section 9.1.1.3.1 of the (9.1.1)

FSAR indicated that the new fuel storage arrangement will not exceed a keff of 0.95 assuming the new fuel storage area was dry or flooded with unborated water. Verify that a keff equal to or less than 0.98 will be maintained with new fuel of the highest anticipated reactivity assuming optimum moderation, for example, foam, spray, small droplets or mist.

Response

Under normal fuel handling operations the new fuel vault is always dry. The optimum moderator condition is considered to be the same as the abnormal condition of v alt flooding because of the very low probability of any other moderator density condition.

Administrative controls are used to prohibit the introduction of low density hydrogenous material into the fuel storage region.

For example,

" Fuel shall not be stored in the new fuel vault when there are construction activities on the refueling floor or construction debris in the vicinity of the new fuel vault unless a solid cover is placed over the vault which would preclude criticality due to inundation by low density such as water mist or spray from a fire hose."

Since the new fuel vault is designed to have keff ;E.95, including

'.1 calculational biases and uncertainties, with the abnormal full flooded condition then keff will also be;[.98.

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410.11 Regarding the seismic design of the spent fuel storage areas (9.1.2) in containment and the interacdiate building, i

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The FSAR does not indicate that the spent fuel pool n

liner plate was designed to seismic Category I requirements. Discuss why a failure of the liner plate resulting from an SSE will not result in radio-active release from ona of the following: mechanical damage to the spent fuel, significant loss of water from the pool which could uncover the fuel, loss of ability to cool the fuel due to flow blockage caused by portions of the liner plate falling on top of the spent fuel, and damage to safety-related equipment as a result of the pool leakage.

(2)

The FSAR does not discuss whether the gates used to separate the cask pit, the spent fuel storage pool, the fuel transfer pool, and the fuel storag-and j

preparation pool and the gates used to separate the steam dryer storage pool, the fuel storage pool, and the fuel transfer pool were designed to seismic Category I requirements. The seismic category of the 1

gates should be documented.

If the design does not meet seismic Category I requirements, discuss how a failure of tt.e gates as a reault of an SSE will not result in similar conditions as stated for the pool liner in part (1) of this question.

Response

The response to this question is provided in revised Section 9.1.3.3.

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pressure indicators and flow indicators monitor the condition of the filter-demineralizers.

9.1.3.3 Safety Evaluation 6

The maximum abnormal heat load is 46.8 x 10 Btu /hr. This value is the sum of the decay heat from 3,402 bundles discharged over an eight year period, plus a sequential full core off-load which fills the fuel handling pools in the area of the intermediate building (4,020 bundles) and stores 130 bundles in the containment pool (see Table 9.1-2).

With both FFCC system pumps and heat exchangers operating, the pool temperature will rise to 154.*F.

This value is derived from a conservative analysis which overestimates the heat loading by approximately 13%. Under realistic conditions, the pool temperature will not exceed 150*F.

To prevent this condition, supplemental cooling capacity is available through a permanent cross tie to the RHR system which is no longer required to shut down the reactor. The supplemental cooling capacity, in conjunction with the fuel pool cooling capacity, will reduce the fuel pool water temperature to 106*F.

Any time the RHR system is used to supplement the spent fuel cooling system to maintain pool water temperature below 150*F, the reactor of the unit whose RHR system is being used, will be placed and maintained in a cold shutdown condition and the refueling mode as long as the RHR system is needed to supplement the FPCC system. Except for this emergency condition, the fuel pool cooling system is capable of maintaining the fuel pool below 127*F without any assistance for all other heat loads that could conceivably occur under normal operating conditions. The 150*F temperature limit is set to assure that the fuel handling area of the intermediate building environment does not exceed equipment environmental limits. Fuel pool cooling pump motors are designed to operate in a 150*F environment. The temperature limit for the fuel pool demineralizer resin is also 150*F.

The fuel storage pool is designed to ensure that no single failure of structures or equipment will cause inability to maintain irradiated fuel submerged in water, to re-establish normal fuel pool water level, or to remove s

decay heat from the pool. The spent fuel pool walls with liner plates are 79 designed to seismic Category I requirements as discussed in Section 3.8.4.

To 9.1-28

limit the possibility of pool leakage around pool penetrations, the pool is lined with stainless steel to provide a high degree of integrity. No outlets or drains are provided in the fuel pool that might permit the pool to be drained below a safe shielding level.

Inlet lines 9.1-28a

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i extending below this level are equipped with siphon breakers to prevent inadvertent pool drainage.

Interconnected drainage paths are provided behind the liner welds. These paths are designed to prevent pressure buildup behind the liner plate, to prevent the uncontrolled loss of contaminated pool water to other relatively cleaner locations within the fuel handling area of the intermediate building, and to provide expedient liner leak detection and measurement. The paths are formed by welding channels behind the liner weld joints and are designed to permit gravity drainage to the radwaste system.

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draining the cask pool will not uncover the fuel in the fuel pool. Alarms will alert the operator if the water level in the fuel pool is low.

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addition, a concrete weir is provided between the pools to prevent uncovering of the fuel. Whenever a gate is moved between its sealing position and storage position, precautions are taken to prevent the gate from falling into the spent fuel pool.

Lifting and moving is performed by the fuel handling j

platform. The auxiliary hook on the platform attaches to the gate lifting i

lugs by a hook and sling arrangement. Also, a full capacity safety line is l

secured to the gate during all periods of movement and seating. This redundant arrangement precludes the possibility of the gate ever coming into contact with any portion of the spent fuel assemblies. The gates are D

classified as Safety Class 2 equipment and are designed to Seismic Category I g

requirements.

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Makeup water from the condensate storage tank is provided to the pool to replace evaporative and leakage losses.

If the failure of the normal makeup I

water system occurs, a permanent connected makeun supply is available from the Seismic Category I emergency service water syscem.

Two manually operated valves in the intermediate building mast be opened to allow makeup water to enter the system. The manually operated valves are locked closed to prevent accidental discharge of water f rom the emergency service water system into the f

fuel pool system.

Details of the emergency service water system are shown in

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Figure 9.2-1 and discussed in Section 9.2.1.

j The fuel pool cooling and cleanup system is designed as Safety Class 3 and

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Seismic Category I except for the non-safety class filter-demineralizer i

system, cask pit drain subsystem and fuel transfer tube subsystem.

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410.12 Section 9.1.2.3.1 of the FSAR indicates that the geometric (9.1.2) configuration of the fuel stored in the GE racks assures that k,gg does not exceed 0.95 under all normal and abnormal storage conditions; however, the list of conditions analyzed does not include a dropped assembly lying across the top of the rack array. Verify that k gg does not exceed 0.95 for e

the condition of a dropped assembly lying across the top of the rack array.

Response

The keff of the fuel storage rack does not increase due to a dropped fuel assembly lying across the top of the rack. The handles, plenums, etc.

of the fuel assembly give a volume weigt.ted distance of water between the active fuel in the rack and the dropped fuel assembly in excess of 14 inches. Ten to twelve inches of water will sufficiently isolate one fuel assembly from another to prevent neutron interaction. Therefore, the dropped fuel assembly is isolated from the active fuel in the rack and the keff remains less than 0.95.

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410.13 With regard to the generic concern relating to the handling of i

(9.1.4) heavy loads near spent fuel, Enclosure 2 to the December 22, 1980 generic letter identified interim measures. We will require a commitment to implement these interim measures prior to the final implementation of NUREG-0612 guidelines and prior to the receipt of an operating license.

Response

A review of the controls for the handling of heavy loads at the Perry Nuclear Power Plant, as requested in the December 22, 1980 generic letter, has been completed. Three copies of the Control of Heavy Loads study, and a full-size set of drawings, were sent on June 19, 1981 to:

Mr. Darrell G. Eisenhut, Director l

Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.

20555 This review indicated that, with the exceptient of specific administrative procedures, there are no changes or modifications required to fully satisfy the NUREG-0612 guidelines. Perry Plant Department personnel will, using the Control of Heavy Loads study, identify and develop those administrative procedures necessary to implement the interim measures prior to the receipt of an operating license.

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410.14 In regard to the handling of loads over spent fuel in (9.1.4) containment and the intermediate building, provide verification that the maximum potential kinetic energy capable of being developed by all objects handled above the spent fuel racks, if dropped from the height at which it is normally handled above the storage rack, does not exceed the kinetic energy of one fuel assembly and its associated handling tool.

Response

Administrative procedures prohibit handling loads weighing greater than one fuel assembly and its associated handling tool above the spent fuel racks. Administrative procedures also restrict the height above the spent fuel racks that loads can be handled.

Therefore, the maximum kinetic energy capable of being developed over the spent fuel racks is the kinetic energy of one fuel assembly and its associated handling tool.

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410.15 Provide process and instrumentation diagrams for the potable (9.2.4) and sanitary water system which demonstrate that there are no connections to systems having a potential for containing radioactive material.

Response

The response to this question is provided in revised Section 9.2.4.2.

9.2.4 POTABLE WATER SYSTEM The potable water system supplies and distributes both hot and cold water throughout the plant for potable and sanitary purposes.

L 9.2.4.1 Design Bases Design bases for the potable water system are:

a.

The system is not safety related, i

b.

The system supplies hot and cold water in sufficient quantities for potable and sanitary purposes to the service building, control complex, turbine buildings and guard house.

c.

The system supplies cold water for safety (emergency personnel) showers and eye washes in various plant locations as required.

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d.

Malfunction or failure of any system component or piping does not adversely affect any safety related system or equipment or the ability to achieve and maintain safe shutdown.

9.2.4.2

System Description

The supply of potable and sanitary water is obtained from the Ohio Water Service Company water main, which is extended onto the site.

Potable water is 1

distributed to the plumbing fixtures located in the plant.

i Hot water is generated by two large electric hot water storage heaters located in the service building and the control complex.

Small electric water heaters are located in the turbine buildings. Hot water recirculating systems are used where excessive lengths of hot water piping warrant inclusion of such systems to maintain water temperature.

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System diagrams for the potable water system are shown on Figures 9.2-16 and b

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9.3 PROCESS AUXILIARIES 9.3.1 COMPRESSED AIR SYSTEMS The compressed air systems include the instrument air system and the service air system. The safety related instrument air system is discussed separately in Section 6.8.

9.3.1.1 Design Bases Except for that portion between the containment isolation valves, the instrument air and service air systems are non-safety related. All safety related components using compressed air are either designed to fail to a condition that corresponds with the safe shutdown of the reactor plant or are equipped with accummulators to satist'y their required air demands.

9.3.1.2

System Description

k The instrument and service air systems are shown in Figure 9.3-1, Figure 9.3-29 and Figure 9.3-30.

j, The service air system for each unit consists of one motor driven compressor with an integral aftercooler, an air intake filter silencer, a receiver tank and a piping network for distribution throughout the plant. A cross tie header between units is included in which distribution connections to the various plant areas are provided. An isolation valve is provided between each unit and the cross tie header.

The service air compressor for each unit is sized to provide the load capacity for the entire plant. During normal operation, the service air systems for the two units are cross connected with one compressor running and the other in the automatic standby mode.

If the service air system pressure drops below 100 psig the standby service air compressor starts automatically.

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Separate instrument air systems are provided for each unit to supply clean, dry, oil free air for control purposes throughout the plant. The instrument I

air quality is in accordance with ANSI Standard MC-11-1 (ISA-S7.3), except that the maximum particle size is 5 micrometres instead of 3 micrometres.

The normal supply of air to the instrument air system is from the respective service air system for the unit and the instrument air compressor for each unit is used as a backup. The service air compressor is cycled on and off at service air pressures of 120 psig and 125 psig respectively. The system for each unit also includes an after cooler (integral with the compressor), a receiver tank, a pre-filter, an air dryer, an after-filter, and a piping network for distribution throughout the plant. All instrument air leaving the receiver tank passes through the filters and the air dryer.

The Unit I and 2 instrument air distribution systems are cross-tied so that ait omponents of either unit can be supplied by either instrument air system gg (see Figure 9.3-1, Figure 9.3-31 and Figure 9.3-32).

k4 If the instrument air system pressure drops below 90 psig the instrument air compressor starts automatically and maintains system pressure in the 90 psig to 100 psig range. A diaphragm operated isolation valve is provided in the air supply line from the service air system. This valve closes automatically when the instrument air system pressure drops below 90 psig and is opened by a switch in the control room when the system pressure rises above 90 psig.

9.3.1.3 Safety Evaluation The instrument and service air systems have no safety-related functions as defined in Section 3.2.

Failure of these systems will not compromise any safety-related system or component and will not prevent safe reactor shutdown.

Safety related devices supplied with compressed air from this system are designed for the fail-safe mode and do not require continuous air supply under emergency or abnormal conditions. Table 9.3-1 gives a list of pneumatically operated valves required for safe shutdown and prevention or mitigation of accidents and shows that these valves assume the safe position in the event of a loss of instrument air pressure.

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410.17 In order to assure continuous reliable functioning of (9.3.1) compressed air operated valves, provide a technical specification or procedure to require testing of the instrument air quality. Describe the procedures to be followed to detect and correct degradation of the ins ~ rument air quality and the limits on degradation from the ANSI standard MC-11-1 which will be imposed on the air quality.

Response

A procedure will be developed which requires testing (" the instrument air quality, specifically dew point, oil content, and panbiol,e size.

The tests will be qualitative rather than quantitative since the design of the instrument air system is sufficient to ensure that the limits on air quality from ANSI Standard MC.ll-1 are met.

The design limit on particle size is 5 microns vice the ANSI Standard MC.ll-1 limit of 3 microns.

Sampling of the instrument air for other contaminants is not necessary due to system design but will be performed if system performance or tests indicate the possibility of contamination.

(

410.18 Regarding the Standby iquid Control (SLC) System, (9.3.5)

(1)

Your FSAR states that Figure 9.3-19 is a P&ID for the SLC system.

From our review, we conclude that figure 9.3-19 is not for the SLC system. Provide a complete P&ID for the SLC system.

(2)

Your discussion of the time that a radundant component of the SLC system may be out of operation, indicates that considerable time is available for restoring the SLC system. Verify that your proposed technical specification for the SLC system comply with the standard GE-BWR technical specification which requires operability of redundant train with 7 days.

Response

(1)

Figure 9.3-19 has been revised to provide P&lD for the SLC system.

(2)

CEI will comply with the Standard GE BWR Technical Specification with respect to the SLC System.

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