ML19351G378

Forwards Addl Info Re NUREG-0737,clarification of TMI Action Plan Requirements,In Response to NRC Request

ML19351G378
Person / Time
Site:	Farley
Issue date:	02/13/1981
From:	Clayton F ALABAMA POWER CO.
To:	Schwencer A, Varga S Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.4.2 NUDOCS 8102230603
Download: ML19351G378 (20)

Text

_

Alacama Powsr Company 600 Nortn 18tn street Post Office Box 2641 Birmingnam. Alabama 35291 Teleonone 205 250-1000 F. L. CLAYToN, JR.

semor V+ce P'esident Alabama Power t% soumem e'ect?mtem February 13, 1981 Docket fios. 50-364 50-348 ,

Director, iluclear Reactor Regulation U. S. fluclear Regulatory Commission Washington, D. C. 20555 ~j

d-Attention: Mr. A. T. Schwencer -

Mr. S. A. Varga Gentlemen:

JOSEPH M. FARLEY NUCLEAR PLANT - Uti!TS 1 & 2 CLARIFICATION OF Tfil ACTION PLAft REQUIREMENTS (flVREG-0737 )

Based on discussions with NRC Staff personnel during the week of February 9,1981, additional information was requested with regard to several items in the APCo original submittal of January 14, 1981.

Alabama Power Company submits the enclosed revised responses on each of these issues. These positions have been discussed with the flRC Staff and are considered by Alabama Power Company to satisfy the Staff's questions.

If you have any questions, please advise.

Yours very truly,

' 0' h

[a ..h L.v. aM al mF. L. Clayton, Jr.

I FLCjr/BDM:de

. Enclosures -

cc: Mr. R. A. Thomas (w/ enclosures)

Mr. G. F. Trowbridge Mr. L. L. Kintner "

J Mr. W. H. Bradford "

O}'

Mr. E. A. Reeves s i //

8102280 60 3

4 II.E.4.2 CONTAINMENT ISCLATION DEPENDABILITY 40 Previous Rescanse By letters of June 30, 1980 for Unit 2, and October 24, 1973, Cecember 31, 1979, November 21, 1979 and March 14, 1980 for Unit 1, Alabama Power Company responded to this item for the Farley Nuclear Plant. -

~

Clarification Resoonse .

The containment pressure-high (CP-H) setpoint is selected to limit the maximum -

pressure inside containment follcwing a design basis accident, and to rrovide -

early isolation in the event of an accident but must be set high enougn to ,

avoid spurious activation. Technical Specification 3.6.1.4 requires that the containment internal pressure be limited to a maximum of 3.0 psig under normal operating conditions. The 3.0 psig limit is required to assure that the containment peak pressure does not exceed the design condition pressure of 54 psig during the LOCA condition.

The instrument error allowances are based on the changes that can occur in i the instrument channel as a function of time, environmental conditions, set-l point drift, inaccuracies that are inherently associated with instrumentation, and instrument error in test gear. The above errors, when combined statisti- -

cally, result in an error allowance of 2.5% of span or 1.8 psig. There is a 95% confidence level in this methodology. This implies that 5% of -

the time the various errors may combine to be greater than the 2.5% value.

• The errors during these instances could be as high as 4.25% of span or 3.0 psig.

! With the containment purge system inoperable due to preventive maintenance -

t or surveillance testing containment pressure would increase. A typical value of 0.6 psig was used for this discussion; however, a higher value is  :

possible depending upon the time the containment purge is out of service.

At this point, assuming the maximum instrument error of 3.0 psig, the rack components could read as much as 3.6 psig which is unacceptably close to the current staff position of 4.0 psig.

For Westinghouse plants, any SI signal actuates containment Phase A isolation along with diesel generator startup and ECCS component initiation. High containment pressure is one of the SI actuating signals. For a typical plant 3 whica bounds the FMP desion, SI will occur for a 2" RCS break in about 65 seconds and for a 4" RCS break in about 22 seconds. For containment high "

oressure to actuate Phase A isolation prior to its being actuated by low pressuri:er pressure for a 2" RCS break, the Phase A setooint would have to ~

be set below 3 psig. This is an unacceptably low value. For the 4" RCS break, a setpoint of approximately 5 psig would result in containment high l pressure and low cressurizer pressure initiating containment Phase A i isolaticn at roughly the same time. Since containment isolation is l

actuated in about 22 seconds frcm low pressurizer pressure, any reduction of isolation time is insignificant when considering offsite doses particularly i

1 l

h i -

l 1

i II.E.4.2 (continued) 41 l

! I f Core uncovery i since core (hence uncover)v doesn't begin until about 400 seconds.is a prerequisite for significa fuel damage j 9600-1980 provides core information about small break LCCA's.

1 It is Alabana Power Ccepany's concern that arbitrarily reducing the contain-ment high pressure setpoint will result in an increased number of spurious SI's Spurious SI's result in unnecessary activation of the ECCS system and potential challenges to the PORV's. They result in a plant trip, a reduction in i remaining thermal sleeve fatigue cycles and a significant recovery tine from i a resulting SI. It is the opinion of Alabama Power Company that the minimal benefit, if any, which may be derived from reducing the containment high ,

pressure setpoint is far outweighed by the potential detrimental effects on the plant. It should be noted that reducing the remaining fatigue cycles could impact expected plant lifetime.

It is Alabate Pcwer Company's position that 1.4 psig above the postulated 3.6 psig (which is in no means the maximum possible value) for the typical ,

i plant evolution described above is mandatory to allow adequate margin for minor containment pressure increases associated with plant operation thereby i l preventing scurious safety injections. Alabana Power Company, therefore, I proposes to lower the CP-H setpoint to 5.0 psig.  ;

In accordance with NRC Staff request, Alabama Power Company will provide a ,

more definitive response to NUREG-0737 Positions 2, 3, 4, 5 and 7 on ,

1 Section II.E.4.2. This respon:t .ill be provided to the NRC by May 1, 1981.-

j I

)

i e*

e< ~, e----gn,*--g-*ewp-e**eeg w - v se -- e- = - -

i 4

r

II.K.3.17 REPORT ON OUTAGE OF ECC SYSTE.'iS - LICENSEE REPORT AND

PR,CPOSED TECHNICAL SPECIFICATICN CHANGES i

, Previous Res6cnse Alabama Power Company's letter dated June 26, 1980 responded to this  !

, item for Farley Muclear Plant Unit 1. f i ~

Clarification Resoonse

. Table 1 is a listing of ECC systems outages for Farley Nuclear Plant -

4 Unit 1 since:Cecemcer 1,1977 during modes in which the systems.are required i to be operable in accordance with technical specifications. Included in

this listing is the duration of the outage, the affected ECCS ccmponent, the

, cause of the outage, ano corrective action taken to return the component'to i operable status. ECC system outages routir.ely required for surveillance t j testing are not included in the listing, it should be noted that no accurate ,

records exist for the duraticn of outages of ECCS equipment during modes of operation in which this equipment is not required by technical specifications.

1 Unit 2 ECCS outages resulting in limiting conditions for operation j as defined by technical specifications will be reviewed and documented.

i This review will be conducted for each five-year time period beginning l Nith initial criticality. This documentation will include the duration and date of the outage, the affected ECCS component, the cause of the  ;

i outage, and torrective action taken to return the ccmponent to operating status. Surveillance testing fallina under the above criteria will be included in this review. It should be noted that no accurate records

exist for the duration of outages of ECCS equipment during modes of opera-tion in whicb this equipment is not required by technical specifications..

The documentation of this review will be reported to the NRC every five years. .

I i

1 I

+

9 YW*4

• e* y'r y w -eqy V~ryy Yguy'T'-"y9qWtv wq w-gP y wWy 9 r,- at g w g w=q% 4- ase.+

-pg-- 9 et - mm-.yyggip-i.p +- ae-11-Mayg3 =^g-l- -" y gyeaae %r

~

5 '

e III.A.I.2 UPGRADE E!1ERGENCY SUPPORT FACILITIES Previous Resoonse By letters of August 1, 1980 and June 20, 1980 for Unit 2 and October 24, 19/9, December 31, 1979, and March 14, 1980 for Unit 1, Alabama Power Company responded to this item for the Farley Nuclear Plant.

Clarification Rescanse NUREG 0737 requires no clarification for this item.

TSC Rocm Construction of the TSC which consists of the following area has been ccmpleted:

1. tionitoring and Display Area

2. Conference Area

3. Planning and Coordination Area

4. Document Room Area Construction activities are complete and area turnover is in progress.

The TSC furniture is scheduled to be in place by February 20, 1981.

Selected documents are scheduled to be available in the TSC by Februarj 20, 1981. Transfer of support function from the temporary TSC to the permanent TSC is scheduled for February 23, 1981.

Data Monitoring The CRT hookup capability is installed and will be tested during the week of February 16, 1981. The Units 1 and 2 line printers are installed -

and operational. The two-pen recorder was inadvertently cmitteo in the original design. Design to install a two-pen recorder for each unit has i been requested and is scheduled to be completed by February 18, 1981.

Installation of the two-pen recorder'is scheduled for April 1,1981. The j line printers presently have block trending capability and will be utilized

for trending until the two-pen recorder is installed. The TV monitoring system has been installed and is being tested. With the exception of the two-pen recorder the data monitoring is scheduled to be operational by i February 23, 1981.

l l

e w , as e - -

-+.--.m--r v-e _a-,.y am9, .-w,y = yqy,y.. v. -

e,-y7 y & ,-.

• t grw 1-

Communications The TSC communications equipment is scheduled to be installed during the week of February 16, 1981.

Habitabili ty The habitability system has been installed and is in final testing.

Completion of the habitability system is scheduled for February 23, 1981.

Fire Protection The TSC is enclosed by three-hour fire rated walls with Class A air tight fire doors. A water hose cabinet and C0g reel are presently located within the TSC area. One C02 and one dry chemical fire extinguisher is scheduled to be installed by February 20, 1981. Emergency breathing apparatus and spare air bottles are new provided for control room personnel.

Electrical Power fiormal and emergency provisions are as follows:

a. Closed Circuit TV and-Computer Hardware - Regulated voltage distri-bution panels 18 and 2B, channel 2.

b. HVAC - 600-V MCCs-lF and lG. These are shared MCCs and can be powered from the diesel generators upon LOSP.

c. Wall Receptacles - 120/208-V distribution panel LL. This shared panel can be powered frcm the diesel generator upon LOSP.

d. Lighting - 26 flourescent fixtures, four 40-W lamps each; 277-Vac from MCC-lF and 1G (Unit 1) and MCC-2CC and 200 (Unit 2); 25-W emergency lighting; 8-hcur battery packs, two 2-head tied to ,

MCC-2CD (Unit 2), train B.

The Operational Support Center is scheduled to be completed by February 23, 1981.

The permanent Emergency Operations Facility is scheduled to be completed by May, 1982. The interim Emergency Operations Facility has been established in the startup office complex. The interim EOF contains all necessary support ccmmunications and facilities.

3 zo

,.y II.3.2 DESI'G ?1 REVIEW OF PLAtti SHIELDIitG Ali0 EllVIR0:4MEIITAL QUALIFICATI0fi 0F EQUIPMEllT FOR SPACES / SYSTEMS WHICH t%Y BE USED Ill POST-ACCIDE?ti OPERATI0 tis C{ Previous Rescense -

By letters of June 20, 1980 and August 1, ^1980, for Unit 2 and October 24, 1979,

?!ovember 21, 1979, December 31, 1979, March 14, 1980, and May 5,1980 for Unit 1

- Alabama Power Ccmpany dccumented ccmmitments and actions taken for the Farley f;uclear Plant related to this item.

x .

.:_ Clarification Rescanse .

f;. .

2 -

E,;4 '

~

?- A design review for the Farley Plant - Units 1 and 2 was conducted by '

Bechtel Power Corporation, using the TID source terms and the 10 CFR 20 and GDC19, 60-64 of Appendix A to 10 CFR 50, dose criteria.

This shielding design reviev considered several classifications ~ of systems which included recirculation systems,. systems which are extensions of the containment atmosphere, portions of the liquid sampling system, and portions of the letdcwn system.

The liquid and gaseous radwaste systems were not included in these analyses.

The gasecus system was eliminated since the reactor ' vessel head vent would be

__ used for degassing operations rather than the VCT. The leak reduction program instituted at Farley fluclear Plant, and venting of the reactor by the reactor vessel head vent and/or PORVs rather than the letdown system and VCT, minimized

'the need for the liquid waste processing system and therefore it was not considered. The high activity radioactive lab and counting roca for the aff.ected

. unit was not included among those' areas where access is considered vital.after an accident since for~two unit operation'these areas in'th'e:unsffected unit Will be '

U-

. . utilized' for post-accident analyses. - '

--[ *

' Access areas with their corresponding post-accident occupancy time for Units .

,, .1 and 2 are. listed belew: , .

Onit 1

(;_ -

~~'

. Area Occuca'ncy Period 6 Control Rocm 24 hr/ day -

Health Physics Area 24 hr/ day -

Primary Access Point 24 hr/ day -

Passageway to Unit 2 1 hr/ day .

Hallway 409 1 hr/ day Electrical Penetration Recms ** 1 hr (approximately 1 hr.

afteraccident)

• • Design change to eliminate occupancy requirement is being considered, Zone maps will be updated as necessary.

II .3.2. (Centinued) 27 Unit 1

,.J Area Cccucancy Pericd Hallway 322 (Outside Sample Recm) *** 1 hr/ day

((. Gas Analysis Rcca *** 1 hr/ day Cable Spreading Reca 1/2 hr.*

Filter Roems 2 hr/ day

• Switchgear Rocms (Elev.121') 1/2 hr.* -

Hot Shutdcwn Panel 24 hr/ day * .

~

.d. CCW Pump Rcca 'l/2 hr.* -

W._.

f;,- Corridor 161 1/2 hr.*

RHR Heat Exchanger Rcca ~1/2 hr.*

Stainvay No.1 -

Transit to elevaticns at west side of aux. b1dg.

Stairway No. 2 *** , Transit to elevations 77'/83' .

Stairway ~No. 8 Transit to elevations at

. F north and east sides of 4 aux. b1dg.

• Infrequent-brief access to these areas may be required post-accident. .

• • • Ccmplete evaluation of these areas is based on resolution of sh'i elding

~

associated with the electrical penetration rocms.

• 3,. .

~.

i - ,-

,: ~ -

~

Unit 2 - - -

, _:  ;. . ~

. .s . . . .

5~ '

Area -

~

Occuoancy Period I.

{;' Control Room 24 hr/ day T~ i Technical Support Center ~24 hr/ day .-

Health Physics Area 24 hr/ day Primary Access Point -

24 hr/ day .

I hr/ day Passageway to Unit 1 (2402) -

Hallway 2409 1 hr/ day ,

Electrical Perietration Rocms ** l'hr (apprcximately~

i 1 hr. after accident)'

l .

l ** Design change to eliminate occupancy requirement is being considered.

l Zone maps will be updated as necessary.

I

u.a.c tuonanuec) 23

.- Unit 2 Area Occucancy Period

' Hallway 2322 (Outside Sample Room for Liquid Sample) *** 1 hr/ day Spectro Photemeter *** I hr/ day Cable Spreading Roem 1/2 hr * ~

Filter Rooms 2 hr/ day * -

Switchgear Rooms (Elev.121')

. l'/2 hr* .

.g-

• . Hot Shutdown Panel 24 hr/ day *

.9 . .

S. CCW Pump Room 1/2 hr

• Corridor 2161 1/2 hr *

, RHR Heat Exchanger Rcom 1/2 hr *

~ Stairway No. 1 Transit to elevation at west side of aux. bldg.

Stairway No. 2. *** Transit to elevations g

77'/83'.

Stairway No. 8 Transit to elevations at north and east sides of aux. b1dg.

T

• Infrequent-brief access to these areas may be required post-accident.

.*** Ccmplete evaluation of these areas 'is based on resolution of. shielding

_1 ' t .

. associated with the electrical penetration rooms. ,-; ;

' . - Each of these areas has been analyzed to determine the dose rates following ./ ,

an accident. - - -

The Primary Access Point was'not initially considered but was later designated a I-A area for direct radiation. The security center will be included in the radiation zone maps. The main control recm and the technical support center were '

considered as areas requiring continuous occupancy. .

As a result of these studies the following shielding modifications are listed for Units 1 and 2: -

( l. Add shielding to the portion of line 3" GCC-12 which is exposed in' the area of the Seal Injection Fil.ter valve station to reduce the dose rate in this area. -

e

, + . - - - - , , - .-  % . .- - -

11.3.2 (Continued) 29

2. Place temporary shielding at the contair. ment radiation monitor to reduce dose rate in the corridon (RE-011, 012).

3. Re-route the RCS sample discharge line so that spent samples are returned by a more direct route to the VCT without entering the letdown line.

l. . Add additional shielding outside the auxiliary personnel hatch to minimize potential effects at the Elevation 155' for the access control area and Technical Support Center.

5. The recuirement to add shielding at hydrogen analyzers; around lines to hydrogen analyzers; and around reactor coolant and containment air sample lines was due -to the fact that cersonnel access was required within one hour after the accident. Further analysis has shown that design modifications which remove the necessity of operator action frcm. this area was more practical than the shielding modifications. These modifications include either the relocation of of eight breakers or the addition of disconnect devices outside the electrical penetration room for the normally locked out valves in the ECCS flow path. Cne of these modifications will be comoleted at the first outage of sufficient duration but no later than the first refueling on Unit 2 and third refueling on Unit 1.

The following is a discussion of the comcuter programs used for these analysis.

1. Source term concentrations in pCi/cc for each isotope along with the volumetric source strengths in Mev/cc/sec. were calculated using the NUCLYD computer program. This program calculates values of the specific activity and volumetric source strength at any given decay time for a given mixture or isotopes. The program also provides an integral energy release for that given mixture of isotopes from t = 0 to any specified time. This computer program is analogous to ORIGEN.

2. Dose rates were calculated with the CYLSO computer program.

This program uses the Rockwell Point Kernel theory for one dimensional cylindrical volumetric sources. Self attenuation in the source as well as the shielding effects of various construction materials such as stell, lead, concrete and

! water are considered in the code. The code output is in l

terms of dose rate vs. distance, for various piping diameters i

and shielding configurations.

3. This computer program is similar to the SDC code.

In addition to the above study, the effects of radiation on eouipment are being considered as part of the IE Bulletin 79-01B and NUREG-0588 review. Source

' terms for LOCf. events in which the primary system may not depressurize will be addressed during the above review.

The sh.elding design evaluation is a complex iterative process. All

modifications listed above and modifications recuired to resolve the outstanding

! design issues will be completed for Units 1 and 2 by January 1,1982 except the

electrical modification described above. r

II . F.1 ADDITIONAL ACCICENT MONITORING INSTRUMENTATION C

( previcus Resconse By letters dated August 1, 1980, August 19, 1980, June 20, 1980 and July 24, 1980, for Unit 2 and October 24, 1979, November 21, 1979, December 31, 1979 and March 14, 1980, Alabama Power Company described cemitments and actions

~

taken for the Farley Nuclear plant.

T Clarification Resconse Noble Gas Effluent Monitor -

6? -

Q A. Vent Stack Monitor ,

E -

~

Alabama Power Ccmpany will install for both units an Eberline Sping 4 sampler to monitor noble gases in the plant vent stack. This sampler has a range of 10-7 to 105uci/cc using multiple detectors. . The monitor draws a sample frem the vent stack to a monitor unit located in the mechanical equipment rocm at elevation 175 of the auxiliary building. The readcut for this unit is located in the main control rocm. An auxiliary readout is located in the icw activity counting laboratory.

The noble gas measurement is perfor=ed by several detectors viewir;g a sample

'C volume. The low and medium range detectors view the same sample volume located in the SA-13 sampler assembly. The high range detector views the sample volume located in the SA-9 sampler assembly.

.- (1)_ LOW RANGE NOBLE GAS: The gas chamber is monitored by a BETA .

C. '

scintillation detector (Eberline Model RDA-3A). Background ~ ,

1 '

correction for this channel is derived frcm the gama background -

. detector, an energy-compensated GM detector (Eberline Model . ,

'10450-B28). Since the external. (ambient) gama radiation has a

.. measurable effect on the BETA measurement (particulate and gas),

Y

~~

the gamma background channel is used as a source of subtraction for both the gas measurement and the particulate measurement.

(2) MEDIUM RANGE NOBLE GAS: An energy-ccmpensated GM detector monitors the gas volume for the medium range noble gas measurement, with its .

output proportional to the gamma content of the sample. An additional identical detector is provided in the sampler shield as a measure of the external background at the sampler; this is the background detector. Thus the effects of a fluctuating external background on the medium range gas channel are nullified by measuring and subtracting the background.

l 4

9

II.F.1 (Continued) 43 j Noble Gas Effluent' Monitor (Continued)

(3) HIGH RANGE NOBLE GAS: An energy-compensated GM detector monitors the gas volume of a section of 1" stainless steel tubing for the high range ncble gas measurement. Its output is proportional to the gama content of the sample.

- An area monitor radiation detector assembly (Eberline Model DAl-1-CC)

• is mounted on the Sping 4 and provides a measure of the gama field at ,

the instrument. This detector is an energy-compensated GM tube and is calibrated in radiation dose rate. Calibration is by use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage. The Eberline Sping 4 monitor is ' -

.q.

~~-

capable of functioning both during and following an accident. Frequent.

filter replacement will ensure operability of the monitor's electronics .~

gj - after an accident. The monitor's accuracy is 2% of span . - ,

Q.

The following list sumarized by channel number and type which calibration sources are provided. -

CHANNEL CHECK SOURCE Number Tyce Content Isotcce .

1 Beta Particulate 30 microcuries 137Cs 2 Alpha Particulate .

3 Iodine (Gama) 0.5 micureurie 133Ba

. 4 Iodine-Subtraction (Gama) -

5 ,

- BetaGas-(LowRangeNobleGas) 30 microcuries 137C s 6 . Gama Area

. 0.5 microcurie 903 '90y

~

~

.7 -.

Gama Gas (Medium 2ange Noble Gas) *

7. . ... . .-

.. 8 Gama Background

.~ ,

9 Gama Gas (High Range Noble Gas) .05 microcurie 903 90y The plant vent noble gas concentration in pCi/ml is detemined by sampling and/or by obtaining a value frem the plant vent stack high range monitor. -

'The plant vent flow rate is determined by the' number of operating auxiliary building exhaust fans. The release rate in curies per second is determined by the following equation:

Release rate (Ci/sec) = Concentration (pCi/ml)X ficw rate (chm) X conversion factor

{

• e e *

, , .r yc--

_,m7 - - - - ,

,y.- --,-..-,,r ..m. r ,, w .r- -e-- ,,------e- + , < - = - - + ~ - +- -

N II.F.1 (Continued) -

Noble Gas Effluent Monitor (Continued)

The above method to determine noble gas release rate is described in emergency implementating procedures. During emergencies the release rate is calculated periodically as directed by the Emergency Director to determine if the accident classification should be upgraded.

The monitors have been environmentally qualified by the vendor for the environment in which it is located.

B. Main Condenser Air Removal Monitor (SJAE) ,

The main condenser air removal exhause systems for Units 1 and 2 are .

_. monitored using the existing monitor (described in the FSAR) on the steam

';.5 - ,

jet air ejector exhaust for the normal range.of radioactivity. The accident i range of radioactivity will be monitored for Units 1 and 2 by intermediate -

m'

~

and high range detectors with overlapping ranges and located at the common vent duct for the turbine building. The accident monitor consist of 2 Eberline detectors and readouts. The intermediate range detector will be .

model Dal-lCS with an EDl-1 readout module with a range of indication of 0.1 to 100 mR/hr. The high range detector is a model Oal-4CS with an ECl-20 readout module with a. range of 10 mR/hr. to 1,000 R/hr. The relationship between mR/hr. and uCi/cc will be established for the noble ~

gas isotopes present during an accident. The range of the accident monitors in uCi/cc is from 10-5 to 103 with the normal range ~ monitor measuring concentrations down to 10-6, uCi/cc. This is the required range for the case

,r where the SJAE exhaust is combined with turbine building ventilation exhaust.

PC The readout modules will be located in the control room and will provide continuous indication. The accident detectors will be shielded from back-ground radiation with 6 inches of lead. Calibration is by use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage.

1 . .

.{ C. Steam Generator Atma' spheric Relief and Safety Valve Monitors, G

n The discharge from steam generator safety relief valves and atmospheric -:

.~

dump valves for Units 1 and 2 will be monitored by mea::uring the radiation .

levels from these steam plumes. There will be four Eberline model DAl-4CS .

g detectors per unit mounted on the main steam roof with a range of 10 mR/hr.

.: J to 1,000 R/hr. The relationship between mR/hr. and uCi/cc has been ~

.T;I. established for the noble gas isotopes present during an accident. The

~.- range of the monitors in uCi/cc will more than cover the required range from 10-1 to 103 for cases with just the PORC open to cases with the PORV and all safties open. Each detector will be connected to an -

Eberline ECl-20 readout module in the control room, providing continuous indication. Since the safety relief valve and atmospheric dump valve discharges are grouped together for each of the three steam generators,

• one detector will be used to monitor the combined effluent steam plume from each steam generator. The fourth detector is used to monitor the plume frca the steam driven auxiliary feedwater pump turbine exhaust. Each detector is collimated and background shielded with 7.5 inches of lead. Calibration is by use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage.

b c-- - -

II.F.1 (Continued)

(Noble Gas Effluent Monitor (Continued)

D. Desion and Installation Schedule for Nobles Gas Effluent Monitors The noble gas effluent monitors will be powered from a vital instrument bus. Procedures will be developed for use, calibration of the system, and dissemination of release rate infomation. The Sping-4 for both

- units is onsite hardware to support installation of the main condenser air removal monitors and the steam generator atmospheric relief and safety valve monitors are onsite or in the process of being shipped.

The installation of the vent stack monitor for Unit 2 is scheduled for March 10,1981, or prior to exceeding 5 percent power. This instrument.

.. is currently scheduled for installation in Unit 1 for prior to the end 7'u of the current refueling outage but not later than January 1,1982.

d.) The original Alabama Power Company position was to monitor the main

-i-N' condenser air removal exhaust and the discharge from the steam generator safety relief valves and atmospheric relief valves with a portable gamma survey instrument. Alabama Power Company, however, finaTized the above position based on NRC questions during the latter part of 1980. Based on the current material availability and status of the complex shieldino .

design required, installation for both units is scheduled for completion by January 1,1982. Alabama Power Company purchased the best available monitors upon finalization of this position. In order to ensure accurate reading of each of these monitors, a ccmplex shielding design is required' '

to discriminate actual readings from background including containment shine. -

~

D @

,g . .

3 '- * *; ,

v. .- *,. , , ,

.-*i.** ,

.7

~

l l

II.F.1 (Continued) ,

45 Samolino and Analysis of Plant Effluents I

(. '

Alabama Pcwer Company has the capability to: provide continuous sampling of plant gaseous effluent for post accident releases of radioactive iodine and particulates at the plant vent and the condenser air removal system. The sampling method involves passing the effluent gases through a filter assembly and transporting the filter to a counting room for analysis. The sampling system has the following capabilities:

(1) Effective iodine absorption of greater'than 90% for all forms of gaseous iodine.

a. . - -

I .~ " (2) Greater than 90%. retention of particulates for 0.3 micron diameter particulates. - -

" - (3) Design intent meets sampling requirements of ANSI N 13.1-1969.I (4) Continuous collection whenever exhaust flow occurs.

(5) Analytical facilities and procedures can'sidered the design basis sample.

(6) Shielding factors were considered in the design.

• [ '

On-site laboratory capability exists to ar.alfze or measure these samples.- The sampling system design is such that plant personnel can remove samples,

, replace sampling media, and transport the samples to the on-site analysis -

S '- facility with radiation exposures that are not in excess of the GDC 19 criteria of 5 rem whole body and 75 rem to the extremities during the duration.of.the

~ '

@,- , accident assuming the design basis shielding envelope.of NUREG-0737.- ,.,,*

n

?

The Eberline Sping 4, which samples vent stack effluents, uses an isokinetic .,

no'zzle in the stack to draw its sample into its filter systam and the flow rate . -

can be adjusted at the pumping unit to attain a sample velocity that will match

6. stack flow rates. There are presently two exhaust fans that determine -

effluent velocities. In addition, there will,.be a Victoreen vacuum pump with charcoal filters that will allow the Chemistry and Health Physics Group to draw 15 minute iodine and particulate samples to b.e analyzed in the laboratory.

This pump has bypass lines that allow drawing an isokinetic sample  ;

by passing portions of the sample back to the stack.- -

l The steam jet air ejector sample point is located on the vertical section .

of the turbine building exhaust ventilation duct. Locating the sample point l on the vertical section of the exhaust duct ensures that the absorber material l is not degraded with entrapped water. '

l e

N

47 II.F.1 (Continued)

C Sampling and Analysis of Plant Effluents (Continued)

The pri=ary sampling systen for the vent stack (Sping-4) is scheduled to be installed by March 10, 1981, but prior to exceeding 5 percent power for Unit 2 and prior to return to power in Unit 1 following the current

=-

refueling outage but no later than January 1,1982, to provide indication (uci/ml) in the main control room and the counting rocm.

The sampling system for the condenser air removal systen is scheduled to be installed by March 10, 1981, but prior to exceeding 5 percent power for

-t ,

~ Unit 2 and prior to return to power in Unit I following the current refueling outage, but no later than January 1,1982.

i%.

.e.-

e l

(

1 II.F.1 (Continued) 48 Containment pressure Monitor

! [

' The present containment pressure indication provides continuous redundant indication in the main control room and has an indication range of -5 psig i to 60 psig. Additional monitoring capability with control rocm indication i

hrving a range of 0 to 210 psig is scheduled to se installed for Unit 1 by ret'arn to power af ter the current refueling outage but no later than January 1, 1982, and for Unit 2 by March 10, 1981, but no later than exceeding 5 percent

- i power. Continuous display and recording of the containment pressure is '

provided in the control roca. The indication accuracy of both the wide and narrow range instruments is 2:3.5% with a response time of less than 180 milliseconds

. ~

1 for a 10% to 90% step function change in pressure.

The environmental qualfication for these items are being addressed as a part of Alabama Power Ccmpany's response to I.E.Bulletin 79-013 and I;UREG-0588.

Alabama power Company will respond to or provide a justification for NUREG-0737 Appendix 8 requirements by June 1, 1981. Any modifications recuired~as a result of the above review will be completed by January 1,1982 or as soon as equipment meeting the requirements of NUREG-0737, Appendix B is available.

J e*

M 1

4 e

4 e

=

1 i

?

i

! -s i

. t I

+%,w.e,,,-,,wy,,-%,,-ure ,,,--.,,--,,,,,v.,,7-y-y.,,,,,,,,,,i-,,--,y,-- ,,#-r ww w - *we~ m er y ---w-,*-r, ww r e-+- *r +- +-'-rwem e -t + f 9 ' v +t'en%** + w--+ -

40~

II.F.1 (Continued)

( Containment Water Level Monitor v

The Farley Nuclear Plant present design has two wide range containment (ECCS sump) water level detectors. These detectors provide indication in the main control room that meets the wide range requirements as specified in the various clarification letters. These level transmitters and associated readout are safety grade and measure volumes up to and above 600,000 gallons.

In addition, a narrew range containment (reactor vessel cavity sump) level system meeting the various clarification letters is scheduled to be installed for Unit 1 by return to power after completion of the current refueling outage but no later than January 1,1982. The schedule for Unit 2 installation is March 10,1981, or no later than exceeding 5 percent power. The accuracy of the narrcw range level instrumentation is + 1/2 inch with an instantaneous response time. Qualification will be addressed in Alabama Power. Company's response to IE Bulletin 79-01B and NUREG-0588.

Alabama Pcwer Company will provide a description of the train separation for the containment water level monitors by June 1,1981. Any required modifications will be completed by January 1,1982, based on equipment availability.

f s

m O

( -

O e

_ _ _ ~

EO

(*

II.F.1 (Continued) v Containment Hydrocen Monitor Two indepe.adent, redundant systems for containment hydrogen monitoring are

. provided for Units 1 and 2. The design of these systems meets the requirements t

for safety-related protective systems as defined by IEEE 279-1971. The output signal of the analyzers are indicated at the analyzer panel location and are alarmed and recorded in the main control room. Each system is supplied electrical power from an independent and redundant Class lE Power Supply. .

The system meets the single failure criteria and remains operable under the postulated accident. Any single failure in one hydrogen monitoring system does not affect its redundant and independent counterpart. The accuracy of the hydrogen monitor is +E% of span with a response time of 0.45 minutes. The

.nge of indication is 0-10%. Qualificaton requirements are being addressed in Alabama Power Company's response to I.E.Bulletin 79-018 and NUREG-0588.

The indication and recording of hydrogen concentration will be initiated as required by emergency procedures in less than one hour after a safety injection initiation.

Alabama Power Company will respond to or provide justification for NUREG-0737 Item II.F.1 Attachment 6 clarifications 2 and 3 including the hydrogen recombiners

/" functioning within 30 minutes. This response or. justification will be submitted

,_, by June 1, 1981. Any required modifications will be completed by January 1,1982, based on eouipment availability meeting the NUREG-0737 requirements.

e l

l l

l l

l

( _

[.

51 II.F.1 (Continued)

Containment Hich-Rance Radiation Monitor Alabama Power Company has ordered redundant Victoreen Model 875 Radiation Detection Systems to meet the requirements for a high containment radiation '

monitor. Each system consists of an ion chamber detector, readout panel, and interconnecting cables. The monito: s will be located inside containment about

_ six feet above the operating deck and approximately S00 apart. These locations ensure the monitors are not protected by massive shielding and that they will provide a reasonable assessment of area radiation conditions inside the containment during and following an accident.

(a) Each detector is designed to measure gamma radiation.

(b) The range of each detector is 1 R/hr. to 107 R/hr. for photon radiation. -

(c) The energy response is -15% to 80 key and 8% frca 100 key to 3 Mev. -

.(d) The calibration frequency will be at a maximum interval of 18 months. Presently it will be necessary to return the monitors .

to the vendor for calibration.

pe (e) The containment high ' radiation monitors are being installed and should be operational for Unit 2 by March 10, 1981, or prior to exceeding -

5 percent power. Such monitors are being installed and should be operational prior to return to power following the current refueling outage for Unit I but no later than January 1,1982.

Victoreen has completed the preliminary qualification review and .is near ~

-completion for. the final qualification program. The radiation monitors satisfy the requirements of the vendor qualification program. The only -

. remaining component to be qualified is the electrical connection between the . :

power cable and the radiation monitor. Victoreen is currently in the process ~

~

-of qualifying this component. Alabama Power Company will update the flRC on the. qualification program as infomation becomes available. -

Capability exists for on-site calibration of the radiation monitor to 10R/hr. ~

Calibration above 10R/hr will be completed by utilizing an electronic signal.

As part of the vendor testing program, Victoreen has stated that at least one point per decade of the range between 1 R/hr. and 103 R/hr. had the calibration certified. -

w- , , , , , - - . - - -