ML20214J626

Forwards Response to Request for Addl Info Re Reg Guide 1.97 Compliance,Including Info on Pressurizer Heater Status,Per Item II.E.3.1 of NUREG-0737

ML20214J626
Person / Time
Site:	Farley
Issue date:	08/08/1986
From:	Mcdonald R ALABAMA POWER CO.
To:	Rubenstein L Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097, TASK-2.B.2, TASK-2.E.3.1, TASK-TM TAC-51088, TAC-51089, NUDOCS 8608150099
Download: ML20214J626 (11)

Text

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Mailing Address Alib"ma Power Compiny 600 North 18th Street Post Ofhce Box 2641 Birmingham. .Alabama 35291 Telephone 205 783-6090 R. P. Mcdonald Senior Vice President Fhntridge Building August 8, 1986 '

AlabamaPower cwsc .,, s, 3..

Docket Nos. 50-348 50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. L. S. Rubenstein Gentlemen:

Joseph M. Farley Nuclear Plant - Units 1 & 2 Regulatory Guide 1.97 Compliance On April 10, 1985 Alabama Power Company provided to the NRC information as requested by " Interim Regulatory Guide 1.97 Report, Joseph M. Farley Nuclear Plant Units 1 and 2" dated February 7,1985. In recent communications with Alabama Power Company, the NRC requested additional information related to the Regulatory Guide 1.97 Compliance Reports for Joseph M. Farley Nuclear Plant Unit 2 (March 30,1984) and Unit 1 (June 29, 1984) and to the April 10, 1985 Alabama Power Company letter. The requested information is attached.

If you have any questions, please advise.

Respectfully submi ed, l

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l R. P. Mcdonald i RPM /MGE:kpc-T.S.2 Attachment cc: Mr. L. B. Long Dr. J. N. Grace l Mr. E. A. Reeves Mr. W. H. Bradford 0

8608150099 860G18 8 1f (b j

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ATTACHMENT Response to NRC Request for Additional Information Regulatory Guide 1.97 Compliance

1. Item 3.3.1 Environmental Qualification Requirement Deviation The licensee has indicated that environmental qualification is not applicable for the following Category 1 and 2 instrumentation. However, no justification was submitted for this deviation.

Main steamline pressure Refueling water storage tank level Condensate storage tank level Plant vent stack flow Condenser SJAE radiation Plant vent effluent radiation Accessible area radiation Main steam effluent radiation TDAFW effluent radiation HVAC emergency damper position--control room Pressurizer heater breaker position Status of standby power and other energy sources important to safety

! NRC Request Alabama Power Company should provide information relating to the environmental qualification for each of the items listed above.

APCo Response The subject Category 1 and 2 instrumentation variables listed above, including all associated instrument loop components, are located outside areas that may constitute a harsh environment. Therefore, Alabama Power Company designated the environmental qualification guidelines of Regulatory Guide 1.97 for each of these variables as not applicable.

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2. Item'3.'3.15 Pressurizer Heater Status The licensee should install electric current instrumentation for the emergency pressurizer heaters in accordance with the regulatory guide recommendations.

NRC Request Alabama Power Company should plan to provide a current meter for each emergency group of pressurizer heaters.

APCo Response As stated in NRC letter dated February 7, 1985, the intent of this Regulatory Guide 1.97 recommendation is to provide pressurizer heater current instrumentation to determine the operating status of the pressurizer heaters that are required by Section II.E.3.1 of NUREG-0737.

At Farley Nuclear Plant the recommended current instrumentation is not provided because the operator determines the operational status of the pressurizer heaters during emergency and normal operations by utilizing pressurizer heater breaker position status lights when their respective electrical bus is energized and by utilizing pressurizer or reactor coolant system pressure indicators. In addition, periodic surveillance testing in accordance with Farley Nuclear Plant Technical Specifications verifies the capacity and operability of each prassurizer heater group.

The Emergency Response Procedures (ERPs) in use at Farley Nuclear Plant were developed based on the Westinghouse Owners Group Emergency Response Guidelines. These ERPs direct the operator to manually align the pressurizer heater emergency power source when offsite power is not available and caution the operator to check that sufficient diesel capacity is available prior to energizing the pressurizer heaters. In addition, the operator is subsequently cautioned to maintain diesel load within the continuous load rating limit. Such capacity verification and subsequent i caution to maintain diesel load ratings provide for sufficient pressurizer i

heater load verification. (Diesel generator watt meters [ load indicators]

l have been identified as a part of the control room design review for improved readability. These meters are expected to be replaced or modified for acceptable readability.) After the emergency buses and the heaters are energized, the operator verifies the operational status of the pressurizer heaters by periodically checking the emergency pressurizer heater group breaker position status lights and by monitoring the pressurizer and reactor coolant system pressure indicators.

l l Other control room indications are available to the operator during l emergency operations that could be used to indirectly determine the l operating status of the pressurizer heaters. Although not all of the following indications are included in the scope of Regulatory Guide 1.97, these indications are available to the main control room operator.

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a) Core Subcooling Monitor Caution Alarm This alarm alerts the operator that reactor coolant system pressure is nearing saturation conditions. Such conditions could be indicative of a pressurizer heater malfunction.

b) Pressurizer Liquid and Vapor Space Temperature Indicators Pressurizer liquid and vapor space temperature provide a diverse means of monitoring pressurizer heater performance and pressurizer pressure, c) Pressurizer Heater Breaker Tripped Alarm This common alarm is provided to alert the operator whenever any pressurizer heater group (s) supply breaker trips due to an overcurrent condition.

d) Diesel Generator Megawatt Indicator The operator can observe an increase in diesel generator load when its associated pressurizer heater group is energized.

e) Diesel Generator Current Indicator The operator can observe an increase in diesel generator output amperes when its associated pressurizer heater group is energized.

f) Emergency 4.16KV Bus Incoming Current Indicator If offsite power is available, the operator could observe a slight increase in the emergency bus incoming amperes when its associated pressurizer heater group is energized.

9) Emergency 600V Load Center Status Light The operator can verify that the emergency 600VAC load center for a given pressurizer heater group is energized by observing the respective load center's energized status light.

h) Emergency 600V Load Center Breaker Trip Alarm This common alarm is initiated whenever a given emergency 600VAC load center's feeder breaker is open thus alerting the operator to a possible loss of a pressurizer heater group power supply.

1) Emergency 600V Load Center Ground Fault Or Breaker On Local Control Alarm This corrmon alarm alerts the operator whenever a ground fault occurs or when one of the feeder breakers is selected for local control on a given emergency 600VAC load center associated with the pressurizer heaters.

The operational guidance. (including the verification of diesel generator load capacity) provided by emergency response procedures, the direct and indirect pressurizer heater operational status indications, and the verification of pressurizer heater capacity and operability through periodic surveillance testing obviate the need for pressurizer heater current instrumentation at the Farley Nuclear Plant.

3. Item 3.3.23 Radiation Exposure Rate The licensge should install Category 2 instrumentation with a range of 10-1 to 10 R/hr in areas where access is required to service equipment important to safety.

NRC Request Alabama Power Company should address long-term surveillance monitoring for the auxiliary building.

APCo Response Revision 3 of Regulatory Guide 1.97 recommends that a Type D, Category 3 variable with a range of 10-1 R/hr to 104 R/hr be provided for the detection of area radiation exposure rate inside buildings or areas where access is required to service equipment important to safety for the purpose of detection of significant releases, release assessment and long-term surveillance.

The instrumentation currently installed and the portable instrumentation currently available at the Farley Nuclear Plant are sufficient to provide for the detection of area radiation exposure rate inside buildings or areas where access is required to service equipment important to safety. The following summarizes Alabama Power Company's effort to ensure personnel access to post-accident vital areas and Farley Nuclear Plant's capabilities to provide for long-term surveillance monitoring in these areas.

Shielding Design Review In response to NUREG-0737, Item II.B.2, Alabama Power Company conducted a plant shielding design review to ensure post-accident access to vital areas. The shielding design review identified areas in the auxiliary building required for post-accident access, required the implementation i of plant shielding modifications, and resulted in the development of post-accident radiation zone maps and in the revision of some plant procedures associated with the Emergency Plan. (NRC letter dated December 15, 1983 stated that the requirements of NUREG-0737, Item II.B.2.2 [ Plant Modifications] have been met for Farley Nuclear Plant.)

The following table identifies areas in the Farley auxiliary buildings that are post-accident access areas and their respective occupancy periods.

Area Occupancy Period Control room, primary access 24 h/ day point, technical support center Health physics access control area 24 h/ day Hallways 316/2316 (connect I h/ day Units 1 and 2)

Hallways 409/2409 (RCA entrance, 1 h/ day el. 155 ft)

Hallways 322/2322 (outside sample 1 h/ day room)

Cable spreading rooms 1/2 h*

Filter rooms 2 h/ day

• Switchgear rooms (el. 121 ft) 1/2 h*

Hot shutdown panels 24 h/ day

• Component cooling water pump 1/2 h*

rooms Corridors 161/2161 1/2 h*

Stairway 1 Transit to elevations at west side of each auxiliary building Stairway 2 Transit to el. 77 ft from el. 83 ft in each auxiliary building Stairway 8 Transit to elevations at north and east sides of each auxiliary building

• NOTE: Access to these areas may be required post-accident, but no l specific activity is anticipated.

Plant shielding modifications were required to ensure that first-priority safety actions can be achieved in the post-accident postulated radiation fields in the auxiliary building. These shielding modifications included the following:

a) Shielding was added to the exposed portion of 3-inch line GCC-12 in the area of the seal injection filter valve station to reduce the dose rate in this area, b) Temporary shielding was placed around the containment atmosphere radiation monitor (RE-11 and RE-12) to reduce the dose rate in the corridor.

c) The reactor coolant system (RCS) sample discharge line was re-routed so that spent samples are returned to the volume control tank (VCT) by a more direct route without entering the letdown line.

d) Additional shielding was added outside the containment auxiliary personnel hatch to minimize the potential exposure rates for the access control area and Technical Support Center at the 155-foot elevation.

e) Electrical disconnects were installed outside the electrical penetration room for the normally locked-out, motor-operated valves in the emergency core cooling system (ECCS) flow path to ensure personnel access to a vital area within one hour after an accident. In addition, a shield door was added outside the Unit 2 electrical penetration room door and in Unit 1 the door was relocated to support access of personnel obtaining post-accident RCS samples.

A series of post-accident radiation zone maps was developed for each level of the Farley Nuclear Plant auxiliary buildings. These zone naps establish eight different possible zone boundaries based on given zone dose rate limits and represent the correlation of the maximum anticipated dose rates, insnediately following a worst-case accident, with the required operator actions and resultant necessary accessibility to vital areas. Also, these zone maps, along with specific plant emergency procedures, provide a means of determining the maximum anticipated radiation levels for each area of the auxiliary building requiring re-entry immediately after a worst-case accident and are used to establish dose rate reduction versus time after the accident.

Plant Procedures The Farley Nuclear Plant Emergency Plan meets the requirements of Appendix E to 10CFR50. Detailed plant procedures concerning the implementation of the Emergency Plan describe post-accident assessment actions, in-plant surveys, exposure control measures and re-entry. ,

l The permanently installed process, area and effluent monitors at Farley Nuclear Plant may be used for radiological assessment. During emergency conditions in-plant surveys are performed as appropriate by radiation l monitoring teams using the same portable monitors and sampling equipment that are used during normal plant operations. Dose rate limits for l emergency response are established by the Emergency Director ard the guidelines of emergency procedures. Re-entry is made only upon the direction of the Emergency Director and all re-entries require the continuous monitoring of the dose rate along the re-entry route and in the desired access area. Whenever a re-entry into the plant is required (e.g., performance of post-accident sampling, emergency repairs, positioning electrical breakers), the re-entry is also used to assess radiological and/or hazardous conditions in order to establish more definite personnel exclusion area boundaries and to de : ermine the extent of damage.

Auxiliary Building Area Radiation Monitors The following table summarizes the auxiliary building permanently installed area radiation monitors that are currently available for radiological assessment.

Channel Description Range R-1 C6atrol Room 1 x 10-4 to 1 x 101 R/hr R-1B Technical Support 1 x 10-4 to 1 x 101 R/hr Center R-3 Radiochemistry Lab 1 x 10-4 to 1 x 101 R/hr R-4 Charging Pump Room 1 x 10-4 to 1 x 101 R/hr R-5 Spent Fuel Building 1 x 10-4 to 1 x 101 R/hr R-6 Sampling Room 1 x 10-4 to 1 x 101 R/hr R-8 Nuclear Laundry (Unit 1) 1 x 10-4 to 1 x 101 R/hr Compacting Station (Unit 2)

R-9 Sample Panel 1 x 10-4 to 1 x 101 R/hr Room (Unit 2) b30 Radwaste Area Ventilation 10 to 106 cpm return from 100 foot elevation and below R-31 Radwaste Area Ventilation 10 to 106 cpm return from 121 foot elevation R-32 Radwaste Area Ventilation 10 to 106 cpm return from 139' foot elevation R-33 Radwaste Area Ventilation 10 to 106 cpm i

from 155 foot elevation R-34 Access Control Area Venti- 10 to 106 cpm lation Return (Unit 1) l R-35A,B Con.puter Room Air Handling 10 to 106 cpm l Unit Air Intake (common l for both units) l In additinn to the area monitors, the following permanently installed auxiliary building process radiation monitors are also available for

!. radiological assessment.

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Channel Description Range .

R-10 Penetration Room Filtration 1 x 10-9 to 1 x 10 6 uCi/cc Discharge Monitoring System R-14 r, Plant Vent Gas Monitoring 5 x 10-7 to 1 x' 10-4 uC1/cc System J

R-21 Vent Stack Monitoring 1 x 10-9:to 1 x 10-6 -uC1/cc

. System R-22 Vent Gas Monitoring System 5 x 10-7 to;1.x 10-4 uC1/cc R-25A,8 Fuel Handling Monitoring 1 x 10-6 to 1 x 10-3 uCi/cc System

. Portable Radiation Monitoring Instruments The portable radiation monitoring instruments currently available for post-accident surveys in the auxiliary building are listed on the following table.

Radiation Instrument Detected Range Remarks GM survey meter Beta, 0-1000 R/h Fishing pole type gamma with extendible-probe; 20 in. to 13 f t .

Neutron rem Thermal 0-5000 mR/h A BF tube inside a counter through moderator for fast detection of neutrons neutrons up to 10 MeV Proportional Alpha 0-500,000 cpm Gas alpha counter proportional

, chamber Juno-type Beta, 0-5 R/h Air ionization chamber survey gamma chamber Juno-type Beta, 0-50 R/h Air ionization chamber survey gamma chamber Scintillation Alpha 0-2M cpm An alpha counter scintillation crystal covered with a thin mylar window Ionization Beta, 0-20,000 R/h Probe is survey meter gamma detachable with extension cable

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

Post-Accident Vital Areas Of the plant areas that are accessible post-accident, only the control room has a permanently installed radiation monitor considered to be a Regulatory Guide 1.97 accessible area radiation instrument variable for Farley Nuclear Plant because this area is continuously manned post-accident. For other areas of the plant where occasional or infrequent access is required or may be required, the Farley portable Regulatory Guide 1.97 plant / environs radiation instrument variables are considered to be acceptable instrument variables to meet the Regulatory Guide 1.97 accessible area raiiation instrument variable reconaendation for Farley Nuclear Plant.

The results of the Farley Nuclear Plant shielding design review for post-accident operation verified that the direct dose rate for vital areas requiring continuous occupancy such as the control room was less than 15 mR/hr. For vital areas requiring infrequent access or the corridors to these vital areas, the integrated exposure was verified to be less than the criteria of 10CFR50, Appendix A, General Design Criterion 19.

The upper range of 104 R/hr specified by Regulatory Guide 1.97 is not reasonable for vital areas requiring continuous post-accident occupancy at Farley Nuclear Plant since a person would not be expected to function in an environment that has a radiation exposure rate above the control room or technical support center area monitor's upper range of 10 R/hr.

For other areas that require or may require continuous post-accioent occupancy and are not equipped with permanently installed area monitors, such as the primary access point, the performance of periodic radiation surveys during post-accident conditions provides for adequate long-term surveillance.

In other vital areas where occasional or infrequent access may be required, the addition of the area radiation instrument variables as recommended by Regulatory Guide 1.97 is not necessary. At Farley .

Nuclear Plant specific procedures associated with the Emergency Plan are i used in the control of access and reduction of personnel exposures i

during the course of an accident and to quantitatively assess the plant

! conditions following any abnormal occurrences. In addition, when access is required to vital areas the associated re-entry procedures require surveillance of the radiation exposure rate in these areas and enroute to these areas using portable radiation detection instrumentation to ensure that dose rate limits are not exceeded and to assess the existing radiological conditions within the given access areas. These portable instruments include those Regulatory Guide 1.97 portable plant / environs radiation instrument variables identified for Farley Nuclear plant as

, follows:

I a) GM Survey Meter equipped with an extendible probe (20-inches to 13-feet) with a range of 0 to 10C0 R/hr for beta and gamma l radiation; b) Juno-type Chamber Survey Meter with a range of 0 to 50 R/hr for beta and gamma radiation; and l c) Ionization Survey Meter equipped with a detachable probe with a range of 0 to 20,000 R/hr for beta and gamma radiation.

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The existing ranges of some of these portable instruments do not satisfy all the recommended range requirements of Regulatory Guide 1,97; however, their ranges are more than sufficient to monitor the maximum dose rate for vital areas requiring infrequent access at Farley Nuclear Plant.

Summary The Farley Nuclear Plant shielding design review identified areas required for post-accider.t access, required the implementation of plant shielding modifications to ensure post-accident vital area access, and established post-accident radiation zone maps that provide for a means of determining the maximum anticipated radiation levels for each area of the auxiliary building required for post-accident access. Plant procedures associated with the farley Nuclecr Plant Emergency Plan describe post-accident assessment actions, in-plant surveys, exposure control measures and re-entry. Permanently installed auxiliary building area and process radiation monitors are available for radiological assessment. Portable radiation monitoring instruments are available for post-accident surveys.

The maximum post-accident dose rate for vital areas requiring continuous occupancy is less than 15 mR/hr. In vital areas requiring infrequent access, the integrated exposure is less than the criteria of General Design Criterion 19. The instruments currently installed and the portable instruments currently available are sufficient to provide for the detection of area radiation exposure rate inside buildings or areas where access is required to service equipment important to safety. Even if high range area radiation monitors were installed in the Farley Nuclear Plant auxiliary buildings, the actual post-accident dose rate in vital areas requiring occasional or inf requent access would be established using portable radiation monitoring instruments. Therefore, the installation of high range area radiation monitors in the auxiliary buildings of the Farley Nuclear Plant is not considered necessary.

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