ML20114B467
| ML20114B467 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/24/1992 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20114B464 | List: |
| References | |
| NUDOCS 9208280214 | |
| Download: ML20114B467 (54) | |
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Attachment 1 Proposed Changes to the Farley Units 1 and 2 Technical Specifications 9209280214 920824 8 DR ADOCK 0500
Farley Unit 1 l Proposed Changed Technical Specification Pages Remove Paae J_qsert Pace l
B 3/4 3-2 B 3/4 3-2 B 3/4 11-2 8 3/4 11-2*
6-15 6-15* l 6-15a*
6-16 6-16 6-22 6-22 6-23 6-23*
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- Supersedes proposed changed technical specification page submitted by Southern Nuclear Operating Company letter dated June 23, 1992.
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5 Unit 1 Marked Pages l
t CD6CCn'kfdion KVa! ts qqd to or IRSS Mn Ne INSTRUMENTATM eHiggnt cenceng d,on hmi73 STged jq IO C f' S 2 0 ) b pfe.rd iX 6 (to fofG 3 f'h5 RASES 2 6. ! 0 01 - 2.0.?A 04 % b\ t 2, C o I a m n 1.
REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (Continued)
The measurement of response time at the specified frequencies providas assurance that the reactor trip and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may j be demonstrated by either 1) in place, onsite or offsite test measurements or
- 2) utilizing replacement sensors with certf fied response times.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in %3 areas served by the individual channels and 2) the alarm or automatic action u initiated when the radiation level trip setpoint is exceeded.
Alarm / trip setpoints for the containment purge have been established for
" a purge rete of 5,000 scfm in all MODES and for purge rates of 25,000 scfm and 50,000 scfm in H00ES 4, 5, and 6. The containment purge setpoints are based a
on a release in which Xe-133 and Kr-85 are the predominent isotopes, on 4,he (
10 CIR 0Q 4 p jh B, T:b h 2, " O V: h :: for these isotopes)and on a X/Q of 5.6 x 10 $ sec/m at the site boundry.
Thehara/tripsetpointforthefuelstoragepoolareahasbeenestablished I based on a fjow rate of 13,000 scfm; a release in which Xe-133 and Kr-85 are ,
thepredominentisotopes,onthe10OTR20,,gpp:ndt3 for these isotopesy and on a X/Q of 5.6 x 10 sec/m at the sits boundry.
S. T:bh 2, "O V he: < '
3/4.3.3.2 MOVABLE INCORE DETE" TORS The OPERABILITY of the movable incore detectors with the specified sinimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux districution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
ForthepurposeofmeasuringF(Z),Fh,andF q xy a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be
, used in recalibration of the excore neutron flux detection system. Full ncore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
FARLEY-UNIT 1 B 3/4 3-2 AMENDMENT NO. 26
RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 LIQUID WASTE TREAT 14ENT This spec.theat1eh deWted. Refer Mhe OM1' DOMMCMD'1 NM
- The 6PHABUTY f the liquid radw6ste treatment sy tem ensures that' is' .
system will be avai able for use who er liquid afflue . require tr,sa t prior to release the enviro'.1 ment. The requirement t the appropria portions of this ystem be used specified provi s, assurance that n liquid effluen will be kept "a low as releasesofradfonctivematerials Thisspecificationimp)ementstherequi is reasonably Achievable". nts of 10 CFR Part .36a, General Oes gn criterion 60 o Appendix A to 10 FR Part 50 and the des gn objective give in Section II.D Appendix I to 1 CFR Part 50.
The specif ed limits governi the use of app riata portions o the liquid radwaste reatment system re specified as a suitable fraction f the dose designf jectives set for in Section II. A f Appendix I, 10 R Part 50, for liquipeffluents. OELE,7ED 3/4.11.1.4 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assuranca that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CP P:-t T', f.;;;dh 2. T:.b1: !!, C 17an-2, at the nearest potable water supply and the nearest surface water supp n an unrestricted area.
l O CFR hrt 26,602 tbx2MO 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE
~Th s spea haton clekiecl. ReTer40%e 0%dh Dose Ca ICA lotion M a ntm l-This specific tion is provided td ensure that the ~ se at any time at'.the site boundary f gaseous affluent from all units on he sit.e will be ylithin the annual dose inits of 10 CFR P t 20 for unrestri tad areas. The annual dose limits a the dosas associa d with the conce rations of 10 CFI Appendix B. T le II, Column 1. hese limits prov ereasonableass~/Part20 ance that radioac 1_ve material disc rged in gaseous e 1 vents will not asult in the exposur of an individual n an unrestricted area, either wit n or outside the site undary, to annual average concentra or.s exceeding th limits specifie in Appendix'B, T la II of 10 CFR P t 20 (10 CFR Par 20.106(b)).
For in viduals who may a times be within t site boundary, occupancy of the i ividual wil' be s ficiently low to ompensate for any increase in the ateo heric diffusion f ctor above that f the sita bounda . The specified rel ase rate limits trict, at all tim the correspond g gamma and beta d a rates above bac roundtoanindiv'p,latorbeyond a site bcundary dua ss than or equal 500 mrem / year t.o the total bod' / or to less than or e 1 o 3000 area / year o the skin. Thas release rate lim s also restrict, all tisaes, the co responding thyroi dose rate above ckground to an i ant via the cow-mil' infant pathway to ess than or equa to 1500 area / yea for the nearest c to the plant. OE44TED l
FARLEY-UNIT 1 P. 3/4 11-2 AMEMDMENT NO. &
l ADMINISTRATIVE CONTROLS
- b. In-Plant Radiation Monitorina A program which will ensure the capability to accurately detamine the airborne iodine concentration in certain plant areas where personnel may be present unde accident condiblons. This program shall include the following:
(1) Training of personnel, (ii) Procedures for monitoring, and I
(iii) Provisions for maintenance of sampling and analys6s equipment.
- c. Secondary Water Chemistry l A program for (Anitoring of secondary water chemistry to inhibit -
steam generator tube degradation. This program shall include:
l (1)- Identification of a sampling schedule for the critical
- variables and the control points for these-variables, (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, including monitoring-the condenser hotwells for evidence of condenser in-Ipekage.
(iv) Procedures for the recording and management of data,
, (v) Procedures defining corrective actions for off-control point-chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing.cf administrative events repired to initista corrective action.
- d. Post-accident Samplina A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant 4
gaseous affluents, and containment atmosphere samples under accident
! conditions. The program shall include the training of personnel, the procedures for sampling and analysis and the provisions for maintenance of sampling and analysis equipment.
I e.34 9- Insect 3 FARLEY-UNIT 1 6-15 AMENDMENT NO. - - . . . - - . . _ . . . _ . . . . . - - . . _ - . - , _ . . . . - - , - - . . , , -
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- e. Radioactive Effluent Controls Proaram
. A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for msintaining the doses to members of the public from radioactive effluents as low as
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l reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the f;11owing elements:
i) Limitations on the operability of radioactive liquid and gaseous monitoring !nstrumentation including surveillance tests and setpoint detern.ination in accordance with the methodology i;, the ODC6 cct d hmts ii) Limitations /on the concentrctions of radioactive material .
4 released in liquid effluents to unrefricted areas conforming to 10 Cf" Part 20, App;ndix 0, T:bi: !!, C lun 2,HIncert 3M iii) Monitoring, sampling, and analysis of radioactive liquid and gaseous affluents in accordance with 10 CFR 20.105 and with the roQ odology and parametet's in the 00CH, L . 20.13 0 2, iv) Limitations on the annual and quarterly doses or dose commitment to a .. ember of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to Appendix I to 10 CFR Part 50, v) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter i and current calendar year in accordance with the methodology and parameters in the ODCH at least every 31 days, l vi) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce ,
releases of radioactivity when the projected doses in a !
31-day period would exceed 2 percent of the guidelines for i
the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, at a 4 +weg conantmt(Ons of
" vii) Limitations Aon the doce r:te^reculth; #romradioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dese: :::cciated 4 th 10 CF9 Part 20, Appendix S, T ble !!, Cole 1, T i Insert 3B !
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INSERT 3A ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 -- 20.2401), Table 2, Column 2 INSERT 3B ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 - 20.2401), Table 2, Column 1, which corresponds to a dose rate of 500 mrem / year total effective dose equivalent, i
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ADMINISTRATTW Gmiiwi 5 _
6.9.1.2 The startus report shall address eacn of the' tasta identified in the i j Final safety Analysis Aspert and shall include a description of the measured !
values of the operating conditions or characteristics attained durir.g the test program and a comparison of these. values with desip predictions and specifications.
Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in.1icense canditions based on other couritments shall be included in this report.
8.9.1.3 startup reports shall be subeitted within (1)-90 days following .
completion of the startup tast program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the startup Report does not cover all three events (i.e., initial criticality, completion of startte test program.
and restauption or commencement of commercial operation) supplementary reports shall be submitted at least, every three months until all three events have been completed.
ANNUAL, rep 0RT1/
- 5. 9.1. 4 Annual Mports covering the activities of the unit as described below for the previous calendar year shall be suositted prior to March 1 of each year. Tne initial report shall be submitted orier ta Marc.1 1 of the year felicwing initial c-iticality.
6.9.1.5 Reports required on an annual basis shall inclues:
- a. A tabulation on an annual basis of the number af station, utility, and staer eersonnel (including contractars) receiving exposures greater than 100 aren/yr and their associated mances exposure according to wort and joo functions,W e.g. , reactor operations and surveillance, inservice inspection, routine maintenance, special asintenanca (describe maintenanca), wasta processing, anu refueling. The dose assignments to various duty functions any be estimate:: based on pockas dosimetar. T1.D. or film badge measurements. Sanil saposures totalling less than 20 percent af the individual total dose need not be accountad for. In the aggregata, at least 80 percent of the total whole body dose received from external sourcas should be-assiped ta specific major work functions, j i
M Asingle suosittal may be made for a' multiple unit station. The submittal l should conoine those sections that are common to all units at the- station.
MThis tabulation supplements the requirements of ;20. t07 of 10 CFR Part' 20.
paraya e h 2.0.22 %
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! Fast FY-UNTT 1 6-15 AMOWENT NO. 25 t
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4 ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA
- 20. l b ol 60 6.12.1 I " alarm signal" . required by paragrapn 20.2'J ic)p(iieu mi of 10 of the20, CFR " control each high device" radiationc area in wh4ch the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be l barricaded and conspicuously posted as a high radiation area and entrance theretc shall be controlled by requiring issuance of a Radiation Work Permit.' Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after tne dose rate level in the area has been established and personnel have been made knowledgeable of them,
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is resconsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics Superviscr.
6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnd witn radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthori:ed entry, and the keys shall be maintained under the administrative control of tne Shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods' of access by personnel under an
, approved Radiation Work Permit which shall specify the dose rate levels in the immeciate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel-with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem"' that are located within large areas, such as PWR containment, wnere i no enclosure exists for purposes of locking, and no enclosure can be reasonably l I
constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flasning light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote l (such as use cf closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.
' Healtn Pnysics personnel or personnel escorted by Health Physics personnel l shall be exemot fr:m the RWP issuance requirement curing the performance of !
their assigned radiation protection duties, provided they are following plant radiation protection procedures for entry into high radiation areas.
- Measurement made at te2 from source of raoicactivity.
30 c nn _
FARLEY-UNIT 1 6-22 AME1J4ENT NO. : /.
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- , ADMINISTRATIVE CONTROLS 4
6.13. PROCESS CONTROL PR03 RAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.
6.13.2 Licensee initiated changes to the PCP: j T ngeet 7
- 1. 1ve
! Shall Ef fluentbe stj$mitted "elease Report for tothe theperiod Co $ission in wh in the s[ch the change (
4 This sub ittal shall conta n:
- a. S ficiently detailr infomation to otally support t e rationale l r the change wit ut benefit of ad itional or suppl mental
] nformation;
)
b A determination that the change y d not reduce the overall conformance of the solidified w ite program to e ting criteria fo
, solid wastes' and i
/ c. Documentat'on of the fact th t the change has een reviewed an l
/ found acc ptable by the POP . '
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- 2. Shall becom effective upon r jview and approva in accordance w' h '
- Soecifica on 5.5.3.1. /
1 j 6.14 0FF5'TE DOSE CALCULATION MANUAL (00CM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
! 5.14.2 Licensee initiated changes to the ODCM:
4 a/Imert E l
! 1. all be submitted to the Commissio in the Monthly 0 , rating Report i ithin 90 days of the date the cha e(s) was made ef ective. This submittal shall ontain:
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- a. Sufficien+ y detailed inforjiation to totally upport the ratio ale i
for the ange without benefit of additiona or supplemental informa ion. Information submitted should consist of a pack ge of
' those ages of the CDCM o be enanged wit- each page number d and provi ed with an appro 1 and date box, ogether with appe priate anal ses or evaluatio justifying the. hange(s);
- b. A etermination tha the change will ot reduce the at uracy or eliability of dos calculations or etpoint determir tions; and
! c. Documentation of the fact that th cht.nge has been , eviewed and found acceptab' . by the PORC.
- 2. Shall become eff ctive upon -evie and approval in ccordance with 3:ecificatic" 6 5.2.1.
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FARLEY-UNIT 1 6-23 AMEN 0 MENT NO.97-4
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- 1. Shall be documented and records of reviews performed shall be 4 retained as required by Specification 6.10.2.o. This i documentation sna11 contain:
- a. Sufficient information to support the change : gather with
- the appropriate analyses or evaluations justifying the l change (s) and
- ' r- 2 0.1302_
- b. A determination that the change will maintain the/ level of a radioactive effluent control required by 10 CFR 20.100, 40
! CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part
- 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
I j 2. Shall become effective after review and acceptance by the PORC
- and the approval of the General Manager - Nuclear Plant, i j 3. Shall be submitted to the Commission in the form of a complete,
- legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the
{' pericd of the report in which any change to the ODCM was made.
Each change shall be identified by markings in the margin of the i
affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the
, change was implemented.
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MSJE@ENTATION BASES RECTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM MSTRUMENTATION (Continued)
The measurement of response time at the specified frequencies provides assurance that the reactor trip and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or
- 2) utilizing replacement sensors with certified response times.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILIT? of the radiation monitoring channels ensures that 1, the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.
Alarm / trip setpoints for the containment purge have been established for a purge rate of 5,000 scfm in all MODES and for purge rates of 25,000 scfm and 50,000 scfm in MODES 4, 5, and 6. The containment purge setpoints are based on a release in which Xe-133 and Kr-85 are the predominant isotopes, on concentration values equal to or less than the effluent cencentration limits stated in 10 CFR 20, Appendix B (to paragraphs 20.1001 - 20 2a01), Table 2, Column 1 for these isotopes, and on a X/Q of 5.6 X 10-6 sec/n? at the site boundary.
The alarm / trip setpoint for the fuel storage pool arei has been established I based on a flow rate of 13,000 scfm; a release in which Xe 133 and Kr-85 are the predominant isotopes, on concentration values equal to or less than the effluent concentration limits stated in 10 CFR 20, Appendix B (to paragraphs 20.1001 - 20.2401), Table 2, Column 1 for these isotopes, and on a X/Q of 5.6 X 10-6 sec/m3 at the site boundary.
3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accura %1y represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuring QF (Z), F3H, N and Fxv a full incore flux map is used, Quarter-core flux maps, as defined in WCAF-8648, June 1976, may be used in retalibration of the excore neutron flux detection system. Full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
FARLEY-UNIT 1 B 3/4 3-2 AMEN 0 MENT N0.
. RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 L10010 WASTE TREATMENT This specification deleted. Refer to the Offsite Dose Calculation Manual.
3/_4.11.1.4 L10VID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20.1302(b)(2)(1), at the nearest potable water supply and the nearest surface water supply in an unrestricted area.
3/4.11.2 GASE0US EFFLUE_NJji 3/4.11.2.1 DOSE RATE This specification deleted. Refer to the Offsite lose Calculation Manual. '
1 FARLEY-UNIT 1 B 3/4 11-2 AHr'10 MENT NO.
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ADMINISTRATIVE CONTROLS l b. In-Plant Radiation. Monitoring
- A program which will ensure the capability to accurately determine i the airborne iodine concentration in certain plant areas where
- personnel may be present under accident conditions. This program j
shall include the following:
l (1) Training of personnel, (ii) Procedures for monitoring, and I
(iii) Provisions for. maintenance of sampling and analyses equipment.
! c. Secondary Water Chemistry A program for monitoring of secondary water chemist y to inhibit l steam generator tube degradation. This program shall include:
- (i) Identification of a sampling schedule for the critical variables and the control points for these variables, i (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, including monitoring the condenser hotwells for evidence of condenser in-leakage, (iv) Procedures for the recording and int 'gement of data, (v) Procedures defining corrective actions for off-control-point chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
- d. Post-accident Samolim '
A program which will ensure the capability to obtain and analyze reactor coolant, rac'ioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the training of personnel, the procedures for sampling and analysis and the provisions for maintenance of sampling and analysis equipment.
- e. Radioactive Effluent Controls Proaram A program shall be provided conforming with 10 CFR 50.36a for the control c radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as FARLEY-UNIT 1 6-15 AMENDMENT NO.
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. ADMINISTRATIVE CONTROLS reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
i) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, ii) Limitations at all times on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 - 20.2401),
Table 2, Column 2, iii) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, iv) Limitations on the annual a;d quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to Appendix I to 10 CFR Part 50, y) Determination of cumulative and projected dose contributions from radioactive effluents for the current cajendar quarter and current calendar year in accordance with the methodolegy and parameters in the ODCM at least every 31 days, vi) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, vii) Limitations at all times on the concentrations of radioactive material released in gaseous effluents to areas beyond the-site boundary conforming to ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 -
20.2401), Table 2, Column 1, which corresponds to a dose rate of 500 mrem / year total effective dose equivalent,
, viii) Limitations on the annual and quarterly air doses resulting from noble gases released in gasecus effluents from each unit -
to areas beyond the site boundary conforming to Appendix I to 10 CFR-Part 50, FARLEY-UNIT 1 6-15a AMENDMENT N0.
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ADMINISTRATIVE CONTROLS 6.9.1.2 The startup report shall address each of the tests identified in the Final Safety Analysis Report and shall inc1LJe a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.
Any corrective actions that were required to obts n satisfactory operation shall 4 also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every three months until all three events have been completed.
ANNVAL REPORTl/
6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.
6.9.1.5 Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated m;nrem exposure according to work and job functions,2/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20 percent of the individual total dose need not be accounted for. In the aggregt.te, at least 80 percent of the total whole body dose received from 9xternal sources should be assigned to specific major work func. ions.
1/ A single submittal may be made for a multiple unit station The submittal should combine those sections that are common to all units at the station.
2/ This tabulation supplements the requirements of paragraph 20.2206 of 10 CFR Part 20.
FARLEY-UNIT 1 6-16 AMENDMENT NO.
4 ADMINISTRATIVE CONTROLS Q2 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.1601(a) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area,
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them,
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics Supervisor.
6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved Radiation Work Permit which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels s'ach that a major portion of the body could receive in one hour a dose in excess of 1000 mrem ** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual area:;, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.
Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are following plant radiation protection procedures for entry into high radiation areas.
- Measurement made at 30 cm from source of radioactivity.
FARLEY-UNIT 1 6-22 AMENDMENT NO.
ADMINISTRATIVE CONTROLS
- 5. 13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.
6.13.2 Licensee initiated changes to the PCP:
- 1. Shall be documented and records of ruiews performed shall be retained as required by Specification 6.10.2.o. This documentation shall contain:
- a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
- 2. Shall become effective after review and acceptance by the PORC and the approval of the General Manager - Nuclear Plant.
6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
6.14.2 Licensee initiated changes to the ODCM:
- 1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.o. This documentation shall contain:
- a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR En.36a, and Appendix 1 to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- 2. Shall become effective after review and acceptance by the PORC and the approval of the General Manager - Nuclear Plant.
- 3. Shall be su'mitted c to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by marking in the margin of the affected pages, clearly indicating the area of the page that was changed, and shal? indicate the date (e.g.,
month / year) the change was implemented.
FARLEY-UNIT 1 6-23 AMENDMENT NO. ;
Farley Unit 2 Proposed Changed Technical Specification Pages Remove Paae Insert Pace B 3/4 3-2 B 3/4 3-2 B 3/4 11-2 B 3/4 11-2*
6-15 6-15*
6 15a*
6-16 6-16 6-22 6-22 6-23 6-23*
Supe sedes proposed changed technical specification page submitted by j Southern Nuclear Operating Company letter dated June 23, 1992.
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coric.e.ntrMdtn VMw2.s egM h or less bo Tne INSTRUMENTATION q {.hent e,once dryi,on ljmd5 g dcd t q 10cFR Ro, A PPdx E Ro Pa eagraph s BASES 2 0. I o 01 - 2 o, MOD, Tic k 2j C o 6 cm 1 7
REa~. TOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM Tii$1RUMENTATION (Continued)
The measurement of response time at the specified frequencies provides assurance that the reactor trip and ESF actuation associated with each i
channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total chanc.el test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or
, 2) utilizing replacement sensors with certified response times.
4 3/4.3.3 MONITORING INSTRUMENTATION L 4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels aie continually measured in the areas served by the individual channels and 2) the alarm v, automatic action is initiated when the radiation level trip setpoint is exceeded.
Alarm / trip setpoints for the containment purge have been established for a purge rate of 5,000 scfm in all MODES and for purge rates of 25,000 scfm and 50,000 scfm in MODES 4, 5, and 6. The containment purgeq setpoints are based on a release in which Xe-133 and Kr-85 are the predominent isotopes, on-%e <
10 Cfil 29 App:qidS, Table 2, W values for these isotopes,and on a X/Q of 5.6 x 10 6 sec/m at the site boundry. '
The Alarm / trip setpoint for the fuel storage pool area has been established based on a flpw rate of 13,000 scfm; a release in which Ve-133 and Kr-85 are the predomindn't isotopes, on the 10 C " 20,,gppendig E, Tab!c 2, unc v31.,gs p for these isotopesy and on a X/Q of 5.6 x 10 sec/m at the site boundry.
3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuring F q (Z), F g, and F a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system. Full incore flux maps or symmetric incore tiiimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inopa,able.
FARLEY-UNIT 2 B 3/4 3-2
-= _ _. -
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, i RADIOACTIVE EFFLUENTS BASES 3/4.11.1.3 LIQUID WASTE TREATMENT Th5 seech-hon deleted. Refertothe 04sh Dose Galculedh %nu.o j .
The OPERABILITY f the liquid ra aste treatment system ensures that his system will be avai ble for use when 'er liquid effluents require trea ent prior to release t the environment. The requirement f. hat the appropri e portion of this stem be used whe specified provi s assurance that the releases of radi active materials n liquid effluen will be kept "a low as is reasonably a ievable". This ecification imp ments the requir ments of 10 CFR Part 50 6a, General Desi Criterion 60 o Appendix A to 10 CFR Part 50 and the desig objective given 'Section II.D o Appendix I to 10 CFR Part 50.
The specifie limits governing the use of appro inte portions of the liquid radwaste tr atment system wer ,specified as a itable fraction f the dose design obj ctives set forth 'n Section II.A o Appendix I, 10 R Part 50, for liquid ef luents. pgt e gc>
3/4.11.1.4 LIQUID HOLDUP TAdKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled releasa of the tanks' contents, the resulting concentrations would be less than the limits of 10 CIR rart 20, Appr. dix B. T 21: II, Cel m 2, at the nearest potable water supply and the nearest surface water suppl n an unrestricted area.
l 0 C FR Part 2 0.1302.(W2(il 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specWicotoo delded RefeMo-fhe OMsh Dose 6)euhhog r%mpl.
This specifi ation is providedj to ensure that the dose at any time at the site boundary f m gaseous e.ffluen 's from all units o the site will be'within the annual dos limits of 10 GFR art 20 for enrestr cted areas. The nr ual dose limits a e the doses as,soci ted with the conc trations of 10 CF Part 20, Appendix B, able II, Column 1. These limits pro de reasonable as rance l that radio tive material dfs arged in gaseous ffluents will-not result in i the expos e of an individua) in an unrestricte area, either wit in or outside l the site oundary, to annuaY averbge concentr ions exceeding t limits j specifi d in Appendix B, T/blo f II of 10 CFR rt 20 (10 CFR Pa 20.106(b)).
l For i ividuals who may t times be within +ie site boundary, he occupancy of the dividual will be fficiently 16w to compensate for an increase in the atm pheric diffusion actor above that f r the site bound: y. The specifie re ease rate limits r. strict, at all tir s, the correspon ng gamma and bet se rates above ba ground to an indiy! dual at or beyon the site boundar to ess than or equal o 500 mrem /ye&r ts the total body o to itm, than or qual to 3000 mrem / year o the s' Kin. Thes release rate lim ts also restrict, at all times, the c rrespanding thyroi dose rate above ackground to an i fant via the cow-mil -infant pathway t less than or equ- to 1500 mrem /ye for the nearest c to the plant. DELETED FARLEY-UNIT 2 B 3/4 11-2 A MEMDrnEUT hJ0.
ADMINISTRATIVE CONTROLS 1
- b. In-PIsv. Radiation Monitorino A progree which will ensure the capability to L.surataly determine the airborne iodine concentration in certain plant areas when personnel may be present onder accident conditions. Tht) program shall include the following:
4 (i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for saintenance of sampling and analyses equipment.
j c. Secondary Water Chemistry A program for monitoring e. secondary water chemistry to inhibit
- taan generator tube degr dation. This program shall include:
(1) Identification of a sampling schedule for the critical
- variables and the control points for these variables, (ii) Identification of the procedures used to sensure the values of the critical variables, (iii) Identification of process sampling points, including monitoring the condenser hotwells for evidence of condenser in-leakage.
(iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for off-control point-chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
- d. Post-accident Sampline
- A program which will ensure the capability to obtain and analyze i
reactor coolant, radioactive iodines and particulates in plant
, gaseous effluents, and containment atmosphere samples under accident-conditions. The program shall include the training of personnel, the procedures for sar Sling and analysis and the provisions for maintenance of sampling and analysis equipment.
e.A. q. Insect 3 FARLEY-UNIT 2 6-15 knendment No. U
1HSERT 3
- e. Radioactive Effluent Controls Proaram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. _ The program (1) shall be contained in the ODCM, (2) shall be implemented by c erating procedures, and (3) shall include remedial actions to be taken wicuever the program limits are exceeded. The program shall include the following elements:
- 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including survelliance tests and setpoint determination in accordance with the methodology in the ODCM, cd cdl bee.5 ii) LimitationsAon the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to:10--CFP, Part 20, Apper.d b S, T:ble !!, Cel m ?. y gg ili) Monitoring, sampling, and analysis of radioactive liquid and gaseous ef fluents in accordance with 10 CFR 20.105 and with -
the methodology and parameters in the ODCM, t.-20. l S 02.
iv) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to Appendix ! to 10 CFR Part 50, v) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM st least every 31 days, vi) Limitations on the operability and use of the liquid and gaseous effluent treatment syttems to ensure that the appropriate portions of ^52se systems are used to reduce releases of radioactivity when the projected doses in a 31-day period eould axceed 2 percent of the guidelines for the annual dose or <tse commitment conforming to Appendix 1 to 10 CFR Part 50, at _J l b mesthec.o vii) LimitationsAon ne dere -r.cr; teat at.^reru a g (ps r4a gk radioactive material released ir. gaseous effluents to areas beyond the site boundary conforming to:the dc::: a;;cciated .;ith 10 CFP, Part 20, Appe!Jh D, Tabir !!, Crl= 1, V Lcret 2B_
m______.________.__________ .
INSERT 3A ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 - 20.2401), Table 2, Column 2, INSERT 30 (a) ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 - 20.2401), Table 2, Column 1, which corresponds to a dose rate of 500 mrem / year total effective dose equivalent, 1
ACMINI m A C/E ** m Ct.J 5.9.1.2 Na sur==. scar saali accress eacn of the tasu identified in ta Final Safety Analysis Aecort ana tr.all inclues' a cascription of us measurec values of us acersting concisions or :..aracuristica sotainea, curing us tast oregram and a c:scarison of these values with design precinions ,ang s:ecifications.
Any corr $ctive actions tr.at were recuired to catain satisfactar/. coeration .snall also :e oescribec. Any additional scocific ceuils required in licanse concitions Dased on other esamitments shall be included in tais report.
5.9.1.3 Starwn reports shall be submit *.ed within (1) 90 days following ccmolation of us startue test progras (2) 90 days following rssumotion or esamencement of commercial power coerstien, or (3) 9 sonths following initial criticality, wnicnever is earliest. If the Startup Aeoort coes not cover all tarse events (i.e. , initial criticality, completion of startuo test program, and resumotion or ccamencement of commercial operation) succlementary re9eru shall te suosittad at least every three acntas until all tarse events have teen c:moleted.
ANNUAL RESCRU 6.9.1.4 Annual recorts covering the anivities of the unit as described belcw for ce revious calendar year shall be sucaittad, neier to Maren 1 of esca year. No initial recort shall to summit.ac prior to Merca 1 of the year following initial criticality.
4 6.9.1.5 tecoru recuired on an annual basis shall include:
- a. A taculation on an annual tasis a the nunner of station, utility, and cus- :ersonnel (inclucing c:ntrsetore) receiving ex:osurts greatar uan nort 100 and erse/yr and tayr associated santes ex osure accorcing to joo functions,- e.g. , rescur operations and surveil .ance, inservice inscoction, reutine saintananca, special saintenance
(:escrite maintanance), wasta precassing, and refueling. The cose assignments to various duty functio.1s say Le estimated basec on poexet cosimeter, TLO, or fils cacge seasurements. Small exacsures tstalling less taan 20 percent of Oe individual total cose need not be accountac for. In the aggregata, at least 80 percent of the tstal =nole bocy dose received from extarnal sour:ss snould te assigned to specific sajor work functions.
M A single suosittal may be asce for a sulti:le unit station.
~he sutaittal snculo c:moine those sections that are common to all units at the station.
2/
- This taculation sucolemants taa recuire.sents of HGdO7 of 10 0?R Part 20.
pa eao g rogh 20.120b t
raatrt-uwt 2 4g
ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATICH AREA d o. t uo \ Uh 6.12.1 Indieu of the " control device" or "alann signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 m em/hr but less than 1000 mrem /hr shall be barricided and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.* Any individual or group of individuals parmitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation rnonitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made af ter the dose rate level in the area has been established and personnel have been made knowledgeable of them.
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics Supervisor.
6.12.2 in addition to the reouirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the Dody could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthori:ed entry, and the keys shall be maintained under the administrative control of the Shif t Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved Radiation Work Permit which shall specify the dose rate levels in the lamediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to persennel with radiation levels such that a major porcion of the bojy could receive in one hour a dose in excess of 1000 mrem ** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably A
constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning t device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.
4 k
- Healtn Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWF issuance requirement during the performance of their assigned radiation protection duties, provided they are following plant radiation prottction procedures for entry into high radiation areas.
"Measyrement made at W from source of radioactivity.
FARLEY-UNIT 2 6-22 AMENTENT NO. 49
ADMINISTRATIVE CONTROLS 6.13. PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.
6.13.2 Licensee initiated changes to the PCP:
he et ~7 ,
1.
Shall be subn)(tted to the Commi sion in the semi-apnual Rcdioactive /
Ef fluent Re base Report for th period in which 'e change (s) was made, This submi tal shall contain: y '.
- a. Suf ciently detailed nfomation to tota) y support the rat fo the change witho benefit of additi4nal or supplemental) nale i formation;
- b. A detemination at the change did et reduce the overa 1 conformance of e solidified waste program to exiti criteria for solid wastes; d
- c. Documentati i of the f act that he change has bee eviewed and found acce able by the PORC DELETEb
- 2. Shall become/ effective upon review and approval i accordance with Specificat}tn 6.5.3.1. /
6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior'to implenentation.
6.14.2 Licensee initiated changes to the ODCM:
j Lget g
- 1. #all h be submitted t'o the Cocmission i ^ the Monthly Operaging Report within 90 days of/he date the chang ) was mada effec (ve. This submittal shall ontain:
- a. Sufficien)1y detailed infom oion to totally su ort the rationale for the phange without oene t of additional o supplemental infom (ion. Infomation ubmitted should co sist of a package those ages of the ODCM be changed with ch page numbered a d prov ded with an approv and date box, to ether with appropri te ana yses or evaluation justifying the ch nge(s);
- b. & detemination tha the change will n reduce the accura y or reliability of dos calculations or s tpoint deteritinati s; and
. Documentation o he f act that the change has been rev ewed and found accepta by the PORC.
DELETE D S. Shall become effective upon reviewf nd approval in ac rdance with
/ Specification 6'.5 3.1.
/ /
e FARLEY-UNIT 2 6-23 AMENDMENT N0d9 t
INSERT 8
?
?; 1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.o. This
] i documentation shall contain:
F
$ a. Sufficient information to support the change together with r the appropriate analyses or evaluations justifying the
/ change (s) and K r- 20. l 3 0 2 f'
- b. A determination that the change will maintain the/ level of radioactive effluent control required by 10 CFR 20.105, 40 CFR Part 190, 10 CFR 50.36a, and Appendix 1 to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- 2. Shall become effective after review and acceptance by the PORC and the approval of the General Manager - Nuclear Plant.
- 3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCH as a part of or concurrent with
- the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.
Each change shall be identified by markings in Se margin of the affected pages, clearly indicating the area c ) page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
Unit 2 Typed Pages e
m.......,.,, . . - . _ . . . . . . . _ . . . . . . . . . .
, . . - , i '
INSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (Continued)
The measurement of response time at the specified frequencies provides assurance that the reactor trip and ESF actuation associated with each channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or
- 2) utilizing replacement sensors with certified response times.
344.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.
Alarm / trip setpoints for the containment purge have been established for a purge rate of 5,000 scfm in all MODES and for purge rates of 25,000 scfm and 50,000 scfm in MODES 4, 5, and 6. The containment purge setpoints are based on a release in which Xe-133 and Kr-85 are the predominant isotopes, on concentration values equal to or less than the effluent concentration limits stated in 10 CFR 20, Appendix B (to paragraphs 20.1001 - 20.2401), Table 2, Column 1 for these isotopes, and on a X/Q of 5.6 X 10-6 sec/m3 at the site boundary.
The alarm / trip setpoint for the fuel storage pool area has been established l based on a flow rate of 13,000 scfm; a release in which Xe-133 and Kr-85 are the predominant isotopes, on concentration values equal to or less than the effluent concentration limits stated in 10 CFR 20, Appendix B (to paragraphs 20.1001-20.240g)a,tthesiteboundary.
5.6 X 10-6 sec/m Table 2, Column I for these isotopes, and on a X/Q 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuringQF (Z), F$H, trd Fxy a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system. Full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
FARLEY-UNIT 2 B 3/4 3-2 AMENDMENT NO.
RAHL0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 L10V10 WASTE TREATMEN'i This specification deleted. Refer to the Offsite Dose Calculation Manual.
3/4.11.1.4 LIOMID HOLDUP TANKS Restricting the quantity of radioactive material contair.eri in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20.1302(b)(2)(i), at the nearest potable water supply and the nearest surface water supply in an unrestricted area.
3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSLRATr This specification deleted. Refer to the Offsite Dose Calculation Manual.
5 FARLEY-UNIT 2 B 3/4 11-2 AMENDMENT NO.
, ADMINISTRATIVE CONTROLS
- b. In-plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in certain plant areas where personnel may be present under accident conditions. This program shall include the following:
(i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analyses equipment.
- c. Secondary Water :heh11stry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:
(i) Identification of a sampling schedule for the critical variables and the control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, including monitoring the condenser hotwells for evidence of condenser in-leakage, (iv) Procedures for the recording t.nd management of data, (v) Procedures defining corrective actions for off-control-point chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
- d. Eost-accident Samoling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the training of personnel, the procedures for sampling and analysis and the provisiens for maintenance of sampling and analysis equipment,
- e. Radioactive Effluent Controls Proaram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as FARLEY-UNIT 2 6-15 AMENDMENT N0.
i l
ADMINISTRAYlVE CONTROLS reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
i) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, ii) Limitations at all times on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 - 20.2401),
Table 2, Column 2, iii) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with
, the methodology and parameters in the ODCM, iv) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to Appendix 1 to 10 CFR Part 50, v) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameter
- in the ODCM at least every 31 days, vi) Limitations on the operability and use of the liquid and gaseous effluent treat :ent systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix 1 to 10 CFR Part 50, vii) Limitations at all times on the concentrations of radioactive material released in gaseous effluents to areas beyond the site boundary conforming to ten times the concentrations stated in 10 CFR Part 20, Appendix B (to paragraphs 20.1001 -
20.2401), Table 2, Column 1, which corresponds to a dose rate of 500 mrem / year total effective dose equivalent, viii) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to Appendix ! to 10 CFR Part 50, FARLEY-UNIT 2 6-15a AMENDMENT NO.
4 ADMINISTRATIVE CONTROLS
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6.9.1.2 The startup report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.
Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every three months until all tiree events have been completed.
ANNVAL REPORTV 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the yea-following initial criticality.
6.9.1.5 Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated manrem exposure according to work and job functions,U e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions.
V A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.
U This tabulation supplements the requirements of paragraph 20.2206 of 10 CFR Part 20.
FARLEY-UNIT 2 6-16 AMENDMENT NO.
1
. ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.1601(a) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto Gall be controlled by requiring issuance of a Radiation Work Permit.* Any n dividual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation d9se rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics Supervisor.
6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnP under an approved Radiation Work Permit which shall specify the dose ram levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem ** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.
Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are following plant radia ion protection procedures for entry into high radiation areas.
- Measurement made at 30 cm from source of radioactivity.
FARLEY-UNIT 2 6-22 AMENDMENT N0.
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ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM 2C P) 6.13.1 The PCP shall be approved by the Commission prior to implementation.
6.13.2 Licensee initiated changes to the PCP:
- 1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.0. This documentation shall contain:
- a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
- 2. Shall become effective after review and acceptance by the PORC and the approval of the General Manager - Nuclear Plant.
6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
6.14.2 Licensee initiated changes to the ODCM:
- 1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.0. This documentation shall ,
contain:
- a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) 6,:.d
- b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix 1 to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- 2. Shall become effective after review and acceptance by the PORC and the approval of the General Manager - Nuclear Plant.
- 3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as e part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g.,
month / year) the change was implemented.
FARLEY-UNIT 2 6-23 AMENDMENT NO.
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I Attachment 2 Safety Assessment l' 4
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Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Changes Associated With implementation of the New 10 CFR 20 Requirements Safety Assessment Proposed Chance 1 Revise f arley Technical Specification Bases 3/4.3.3.1 regarding the acceptance criteria used to determine the fuel storage pool area gaseous activity and containment gaseous activity alarm / trip monitor setpoints.
Safety Assessment The discussion of setpoints in Bases 3/4.3.3.1 is modified to state that the fuel storage pool area gaseous activity alarm / trip monitor setpoint and the containment gaseous activity alarm / trip monitor setpoints, items 2.a and 2.b, respectively, of technical specification Table 3.3-6, are based on concentrations which are less than or equal to the effluent concentration limits (ECL) listed in the new 10 CFR 20, Appendix B, Table 2, Column 1, for Xe-133 and Kr-85 in lieu of the maximum permissible concentration (MPC) values which are listed in the old 10 CFR 20.
The fuel storage pool area gaseous activity alarm / trip monitor setpoint and the containment gaseous 1ctivity alarm / trip monitor setprints were calculated based on the old 10 CFR 20 MPC values of 3 X 10-7 uCi/ml for Kr-85 and Xe-133. The ECL values listed in the new 10 CFR 20, Appendix B.
Table 2 Column 1, for these two radionuclides are 7 X 10-7 uCi/ml and 5 A 10-7 uCi/ml, respectively. Since the ECL values are higher than the MPC values upon which the existing technical specification setpoints are based, the setpoint values are conservative and do not need to be recalculated to accommodate the new 10 CFR 20 requirements. Therefore, it is acceptable to retain the existing setpoints based on MPC values and to provide the flexibility to establish new setpoints based on ECL values contained in the new 10 CFR 20 by revising the bases.
Proposed Chanae 2 Revise Farley Technical Specification Bases 3/4.11.1.4 regarding the acceptance criteria used to determine the activity limit for the radioactive effluent liquid holdup tanks.
Safety assessment , ,
The discussion in Bases 3/4.11.1.4 is modified to state that in the event of an uncontrolled release of the outside temporary liquid holdup tanks, the resulting concentration would be less than tne limits (ECL) of the new 10 CFR 20.1302(b)(2)(i) in lieu of the limits specified .n the old 10 CFR 20, Appendix B, Table II, Column 2, which are based on MPC.
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Safety Assessment j Page 2 ;
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An evaluation was performed to determine the activity that could be released from a tank rupture based on ECL values as compared to the current technical
~l specification (Specification 3.11.1.4) limit of 10 curies which is based on HPC values contained in the old 10 CFR 20. The evaluation provided a larger allowable tank activity based on the ECL values. Since a higher activity 1 limit can be (atermined based on the ECL values, it is conservative to retain the ctrrent activity limit of 10 curies. Maintaining the activity limit at 10 curies is also consistent with the guidance contained in NUREG 0133, which states that the curie limit for a temporary tank should be limited to less than or equal to 10 curies, excluding tritium and dissolved j
4 or entrained gases which is consistent with Specification 3.11.1.4.
Proposed Chance 3 Revise proposed Farley Technical Specification 6.8.3.e(ii) submitted by Southern Nuclear Operating Company letter dated June 23, 1992, in response to Generic Letter 89-01, in order to accommodate needed operational flexibility.
Safety Assessment Proposed Technical Specification 6.8.3.e(ii) submitted by Southern Nuclear Operating Company letter dated June 23, 1992, states that liquid effluent releases to unrestricted areas must conform to 10 CFR 20, Appendix B, Table j II, Column 2. In accordance with the old 10 CFR 20, the annual dose to a member of the public upon which these concentrations are based is 500 mrem.
Although the old 10 CFR 20.106 allows effluent concentrations to be averaged over a year, the technb al specifications require that liquid effluent releases be lie ted to these concentrations at all times (i.e., for instantaneous releases). More restrictive limits were incorporated into the technical specifications by the NRC to assure that the dose limits of Appendix ! of 10 CFR 50 or the dose limits of 40 CFR 190 are not exceeded.
The basic requirements for technical specifications on effluents from nuclear power reactors are stated in 10 CFR 50.36a. These requirements indicate that compliance with effluent technical specifications will keep average annual releases of radioactive material in effluents at small percentages of the limits specified in the old 10 CFR 20.106. These
- requirements further indicate that operational flexibility is allowed, i
compatible with considerations of health and safety, which may temporarily result in releases higher than such small percentages, but still within the
, limits specified in the old 10 CFR 20.106 which references Appendix 8. Table 11 concentrations. -These referenceo concentrations are specific values which relate to an annual dose-of 500 mrem. It is further indicated in 4 10 CFR 50.36a that when using operational flexibility, best efforts shall be exerted to keep levels of radioactive materials -in effluents as low as is reasonably achievable as set forth in 10 CFR 50, Appendix 1.
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Safety Assessment Page 3 In accordance with the Introduction to Appendix B of the new 10 CFR 20, the
} liquid effluent concentration limits stated in Appendix B. Table 2, Column l 2, are based on a dose of 50 mrem in a year. Therefore, the previous NRC position that effluent releases must be limited by technical specifications to the concentrations stated in the old 10 CFR 20, Appendix B, Table II, Column 2, at .11 times, does not appear to be warranted for the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Cclumn 2, because the requirements of 10 CFR 50.36a are presented in terms of the old 10 CFR 20.106, which relates to an annual dose of 500 mrem, not 50 mrem. }
} Since a release concentration, which corresponds to a limiting value of 500 '
2 mrem in a year, has been acceptable as a technical specification limit for i
liquid effluents, which applies at all times as an assurance that the limits of 10 rFR 50, Appendix I are not likely to be exceeded, it should not be
! necessary to reduce this limit b' a factor of ten.
3 In Subpart D,Section VI, of the Supplementary Information which accompanied the new 10 CFR 20, it is stated that for power reactor licensees, compliance with the limits of 10 CFR 50, Appendix I and with the limits of 40 CFR 190 will demonstrate compliance with the limits of the new 10 CFR 20.1301, :n I which dose limits for members of the public are specified. The limits in
- 10 CFR 50, Appendix I and 40 CFR 190 are specified as annual dose limits, e
therefore, dose determinations to show compliance with these requirements are in terms of cumulative doses (doses in quarter or year for Appendix !
and doses in 4 year for 40 CFR 190). If a dose limit of 50 mrem in a year were included in a technical specification as a limit which applies at all times (i.e., a dose rate of 50 mrem / year), operational flexibility would not be available because the dose rate limit would already be very close to the dose limits specified in 10 CFR 50, Appendix I and 40 CFR 190.
Operational history at Farley has demonstrated that the use cf the concentration values associated with the old 10 CFR 20.106 as te:hnical specification limits which apply at all times has resulted in calculated doses to a member of the public that are small percentages of the limits of 10 CFR 50, Appendix 1. Therefore, the use of concentration values which correspond to annual doses of 500 mrem (ten times the concentration values ;
stated in the new 10 CFR 20, Appendix B, Table 2, Column 2) should not have '
a negative impet on the ability to continue to operate within the limits of 10 CFR 50, Appendix I and 40 CFR 190.
Having the oper ational flexibility discussed above is especially important in establishing a basis for effiuent monitor setpoint calculations. As discussed above, the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 2, relate to a dose of 50 mrem in a year. When applieri on an instantaneous basis, this corresponds to a dose rate of 50 mrem / year .
This is an impractical low value upon which to base effluent monitor setpoint calculations for many liquid effluent release situations when monitor background, monitor sensitivity, and monitor performance must be taken into account.
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Safety Assessment Page 4 Therefore, to :ccommodate operational flexibility needed for effluent releases, proposed technical specification 6.8.3.e(ii) submitted by Southern Nuclear Operating Company letter dated June 23, 1992, is being revised by restating the limit as ten times the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 2, to apply at all times. The multiplier of ten is proposed because the ann Jal dose of 500 mrem, upon which the concentrations in the old 10 CFR 20, Table II, Column 2, are based, is a factor of ten highor than the annual dose of 50 mrem, upon which the concentrations in the new 10 CFR 20, Appendix B, Table 2, Column 2, are based. Compliance with the limits of the new 10 CFR 20.1301 will be demonstrated by operating within the limits of 10 CFR 50, Appendix I and 40 CFR 190.
Proocsed Ch2 nae 4 Revise proposed Farley Technical Specification 6.8.3.e(iii) submitted by Southern Nuclear Operating Company letter dated June 23, 1992, in response to Generic Letter 89-01, to incorporate the new 10 CFR 20 reference regarding dose limits for individual members of the public.
Safety Assessnent Proposed Technical Specification 6.8.3.e(iii) submitted by Southern Nuclear Operating Company letter dated June 23, 1992, contained the Generic Letter 89-01 reference to the old 10 CFR 20.106 regarding radioactivity in 0
effluents to unrestricted areas. This reference is being revised to incorporate the new 10 CFR 20 reference to paragraph 10 CFR 20.1302 which 3
supersedes the old 10 CFR 20 reference to paragraph 10 CFR 20.106.
Droposed Chanj;ttji Revise proposed Farley Technical Specification 6.8.3.e(vii) submitted by Southern Nuclear Operating Company letter dated June 23, 1992, in response to teneric letter 89-01, in order to accommodate needed operational a flexibility.
S_afety Assessmeul Proposed Technical Specification 6.8.3.e(vii) submitted by Southern Nuclear Onerating Company letter dated June 23, 1992, states that dose rates due to gescous effluent releases to areas beyond the site boundary must conform to l the doses associated with 10 CFR 20, Appendix B, Table II, Column 1. In accordance with the old 10 CFR 20, the annual dose to a member of the public upon which these concentrations are based is 500 mrem. Although the old 10 CFR 20.106 allows effluent concentrations to be averaged over a year, the
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Safety Assessment Page 5 i technical specifications require that gaseous effluent releases be limited to a dose rate of 500 mrem / year (total body) which corresponds to these concentrations at all times (i.e., for instantaneous releases). More restrictive limits were incorporated into the technical specifications by the NRC to assure that the dose limits of Appendix 1 of 10 CFR 50 or the dose limits of 40 CFR 190 are not exceeded.
The basic requirements for technical specifications on effluents from n9 clear pewer reactors are stated in 10 CFR 50.36a. These requirements indicate that compliance with effluent technical specifications will keep average annual releases of radioactive material in effluents at small percentages of the limits specified in the old 10 CFR 20.106. These requirements further indicate that operational flexibility is allowed, compatible with considerations of health and safety, which may temporarily result in releases nigher than such small percentages, but still within the limits specified in the old 10 CFR 20.106 which references Appendix B, Table 11 concentrt,tions. These referenced concentrations are specific values which relate to an annual dose of 500 mrem. It is further indicated in 10 CFR D.36a tnat when using operational flexibility, best efforts shall be exerted to bep levels of radioactive materials in effluents as low as is reasonably achievable as set forth in 10 CFR S0, Appendix 1. _
In accordance with the Introduction to Appendix 8 of the new 10 CFR 20, the gaseous effluent concentration limits stated in Appendix B, Table 2. Column 1, are based on a dose of 50 mrem in a year. Therefore, the prev %us NRC position that effluent releases must be limited by technical spuifications
. to the dose rates corresponding to the concentrations stated la the old 10 CFR 20, Appendix B. Table 11, Column 1, at all times, doe', not appear to be warranted for the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 1, because the requirements of 10 CFR 50.36a are presented in terms of the old 10 CFR 20.106, which relates to an annual dose of 500 mrem, not 50 mrem. Since a limiting value of 500 mrem in a year (as a total body dose rate of 500 mrem / year) has been acceptable as a technical s)ecification limit for effluents to apply at all times as an assurance that tie limits of 10 CFR 50, Appendix 1, are not likely to be exceeded, it should not be necessary to reduce tnis limit by a factor of ten.
In Subpart D,Section VI, of the Supplementary Information which accompanied the new 10 CFR 20, it is stated that for power reactor licensees, compliance with the limits of 10 CFR 50, Appendix ! and with the limits of 40 CFR 190 will demonstrate compliance with the limits of the new 10 CFR 20.1301, in which dose limits for members of the public are specified. The limits in 10 CFR 50, Appendix 1 and 40 CFR 190 are specified as annual dose limits, therefore, dose determinations to show compliance with these requirements are in terms of cumulative doses (doses in a quarter or year for Appendix 1 and doses in a year for 40 CFR 190). If a dose limit of 50 mrem in a year were included in a technical specification as a limit which applies at all times (i.e., a dose rate of 50 mrem / year), operational flexibility would not be available because the dose rate limit would already be very close to the dose limits specified in 10 CFR 50, Appendix 1 and 40 CFR 190.
Safety Assessment Page 6 Operational history at Farley has demonstrated that the use of the dose rate corresponding to the concentration values associated with the old 10 CFR 20.106 as a technical specification limit which applies at all times has resulted in calculated doses to a member of the public that are small percentages of the limits of 10 CFR 50, Appendix 1. Therefore, the use of concentration values which correspond to annual doses of 500 mrem (ten times the concentration values stated in the new 10 CFR 20, Appendix B, Table 2, Column 1) should not have a negative impact on the ability of Farley to continue to operate within the limits of 10 CFR 50, Appendix 1 or 40 CFR 3 190.
Having the operational flexibility discussed above is especially important in establishing a basis for effluent monitor setpoint calculations. As discussed above, the concantrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 1, relate to a dose of 50 mrem in a year. When applied on an instantaneous basis, this corresponds to a dose rate of 50 mrem / year.
This is an impractical low value upon which to base effluent monitor setpoint calculations for many gaseous effluent release situations when monitor background, monitor sensitivity, and monitor performance must be
, tden into account.
Thcrefore, to accommodate operational flexibility needed for effluent releases, prop'> sed technical specificatior. 6.8.3.e(vii) submitted by Southern Nuclear vperating Company letter dated June 23, 1992, is being revised to recast the limitations in terms of concentrations instead of dose rate and to restate the limit as ten times the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 1, at all times. The multiplier of ten is proposed because the annual dose of 500 mrem, upon which the concentrations in the old 10 CFR 20 Table II, Column 1, are based, is a factor of ten higher than the annual dose of 50 mrem, upon which the concentrations in the new 10 CFR 20, Appendix B, Table 2, Column 1, are based. Compliance with the limits of the new 10 CFR 20.1301 will be '
demonstrated by operating within the limits of 10 CFR 50, Appendix I and 40 CFR 190.
Proposed Chanae 6 Revise Farley Technical Specification 6.12.1 by updating footnote ** to incorporate the new 10 CFR 20 reference regarding the control of access to high radiation areas.
Safety Assessment Footnote ** to Administrative Controls Section 6.12.1 currently contains the old 10 CFR 20 reference to paragraph 20.203(c)(2) regarding caution signs, labels, signals and controls. This reference is being revised to incorporate the new 10 CFR 20 reference to paragraph 20.1601(a) which 1
Safety Assessment Page 7 supersedes the old 10 CFR 20 reference to paragraph 20.203(c)(2). This change will not decrease the effectiveness of radiation protection programs at Farley to provide control of exposure from external sources in restricted areas. This change is simply administrative in nature in order to achieve implementation of the new 10 CFR 20 requirements at Farley.
Proposed Chanae 7 Revise Farley Technical Specification 6.12.2 to incorporate the new 10 CFR 20 recuirements regarding the distance used to make measurements of a source of racioactivity to determine if the major portion of a body receivas i in one hour a dose in excess of 1000 mrem. ;
Safety Assessment Administrative Controls Section 6.12.2 currently contains a requirement that measurements be made at 18 inches from a source of radioactivity to determine if the major portion of a body receives in one hour a dose in i
excess of 1000 mrem. This distance is being changed-to a value of 30 cm, ,
consistent with the requirements of the new 10 CFR 20.1601(a)(1). This '
represents a conservative change since 30 cm is a shorter distance that will result in higher dose measurements.-
Proposed Chanae 8 Revise Farley Technical Specification 6.9 l.5.a by updating footnote 2 to incorporate the new 10 CFR 20 reference regarding reports of individual monitoring.
l Safety Assessment l
I Footnote 2 to Farley Technical Specification 6.9.1.5.a currently contains the old 10 CFR 20 reference to paragraph 20.407 regarding personnel monitoring ieports. This reference is being revised to incorporate the new 10 CFR 20 reference to paragraph 20.2208 which supersedes the old 10 CFR 20 reference to paragraph 20.407. Thi: chao9 does not reduce the reporting requirements contained in 6.9.1.5.a. Tr.as cha;ge is simply administrative in nature to achieve implementation of the new 10 CFR 20 requirements.
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Safety Assessment Page B Proposed Chanae 9 Revise proposed Farley Technical Specification 6.14.2.1.b submitted by Southern Nuclear Operating Company letter dated June 23, 1992, in response to Generic Letter 89-01, to incorporate the new 10 CFR 20 reference regarding dose limits for individual members of the public.
Safety Assessment Proposed Technical Specification 6.14.2.1.b submitted by Southern Nuclear Operating Company letter dated June 23, 1992, contained the Generic Letter 89-01 reference to the old 10 CFR 20.106 regarding radioactivity in effluents to unrestricted areas. This reference is being revised to incorporate the new 10 CFR 20 reference to paragraph 10 CFR 20.1302 wMch supersedes the old 10 CFR 20 reference to paragraph 10 CFR 20.106.
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I Attachment 3 Significant Hazards Evaluation Pursuant to 10 CFR 50.92
Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Changes Associated With implementation of the New 10 CFR 20 Requirements 10 CFR 50.92 Evaluation Proposed Chanaqi The pror ..ed changes to the Farley Unit I and Unit 2 Technical Specifications are required in order to implement the new 10 CFR 20 requirements at Farley Nuclear Plant. The proposed technical specification changes involve (1) revisions to the bases and Administrative Controls Section to appropriately incorporate the new 10 CFR 20 references, (2) revisions to the Administrative Controls Section changes submitted by Southern Nuclear Operating Company letter dated June 23, 1992, in response to Generic Letter 89 01 to provide operational flexibility needed for liquid and gaseous releases, and (3) revisions to the Administrative Controls Section regarding the distance used to make measurements of radioactivity to determine if the major portion of a body can receive an excessive exposure.
Backaround By letter dated June 23, 1992, Southern Nuclear Operating Company submitted proposed changes to the technical specifications in response to Generic Letter 89-01 which allows the procedural details contained in the Radiological Effluent Technical Specifications (RETS) to be relocated to the Offsite Dose Calculation Manual (0DCM) and the Process Control Program (PCP) with appropriate programmatic controls being incorporated into the Administrative Controls Section of the technical specifications.
Following approval by the NRC, the programmatic controls will be used to revise the procedural details of the RETS, which will be incorporated into the ODCM and PCP, to reflect the new 10 CFR 20 requirements. However, the scope of Generic Letter 89-01 does not encompass all of the technical specification requirements that are impacted by the new 10 CFR 20.
Additional technical specification changes have been identified which are needed in conjunction with the Generic Letter 89-01 response to facilitate Southern Nuclear Operating Company's goal of implementing the new 10 CFR 20 requirements.
Analysis The proposed changes to the technical specifications will allow for the implementation of the new 10 CFR 20 requirements. The level of radiological control will not be reduced by the proposed changes since compliance with applicable regulatory requirements governing radioactive effluents and radiological environmental monitoring, including 10 CFR 50.3Ea, Appendix I to 10 CFR 50, and 40 CFR 190, will continue to be maintained.
s 10 CFR 50.92 Evaluation Page 2 Southern Nuclear Operating Company has reviewed the recuirements of 10 CFR 50.92 as they relate to the proposed changes and has made the following determination:
- 1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes will facilitate the implementation of the new 10 CFR 20 requirements. Compliance with othar applicable regulatory requirements will continue to be maintained. In addition, tSe existing alarm / trip setpoints for instruments which monitor the :4e > ruel pool area and containment for gaseous activity, and the curie 1:. nit for the outside temporary liquid holdup tanks are being retained. These values are based on the old 10 CFR 20 requirements which were evaluated and shown to be be conservative relative to the values that would be obtained based on the new 10 CFR 20 requirements. Also, the proposed changes do not alter the conditions or assumptions in any of the FSAR accident analyses. Since the FSAR accident analyses remain bounding, the radiological consequences previously evaluated are not adversely affected by the proposed changes. Therefore, it can be concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not involve any change to the configuration or method of operation of any plant equipment. Accordingly, no new failure modes have been defined for any plant system or component important to safety nor has any new limiting single failure been identified as a result of the proposed changes. Also, there will be no change in types or increase in the amount of effluents released offsite. Therefore, it can be concluded that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed changes do not involve a significant reduction in a margin of safety. The proposed changes do not involve any actual change in the methodology used in the control of solid radioactive wastes or radiological environmental monitoring. The methodology that will be used in the control of radioactive effluents will result in the same effluent release rate as the current methodology now being used. The operational flexibility needed for effluent releases allows the use of concentration values ten times the values given in the new 10 CFR 20.
However, this is acceptable since annual doses will be limited to the doses specified in 10 CFR 50, Appendix I and 40 CFR 190. Therefore, it can be concluded that the proposed charp do not involve a significant reduction in a margin of safety.
L.
6 10 CFR 50.92 Evaluation Page 3 CONCLUSION Based on the preceding analysis, Southern Nuclear Operating Company has determined that the proposed changes to the technical specifications will not significantly increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety. Therefore, Southern Nuclear Operating Company has determined that the proposed changes meet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.
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Attachment 4 Environmental Evaluation
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Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Changes Associated With Implementatior, of the New 10 CFR 20 Requirements Environmental E aluation Pursuant to 10 CFR 51.22(c)(9), the proposed license amendment can be categorically excluded from the requirement to perform an cnvironmental assessment or an environmental impact statement based on the following evaluation:
Southern Nuclear Operating Company has determined that the proposed changes to the Farley Unit 1 and Unit 2 Technical Specifications which will facilitate implementation of 4a new 10 CFR 20 requirements, do not affect the types or amounts of any radiological or non-radiological effluents that may be released offsite. No increase in individual or cumulative occupational radiation exposure will result from these changes.
Additionally, these changes do not involve the use of any resources not previously considered in the Final Environmental Statement related to the operation of Farley Nuclear Plant.
Based upon this evaluation it can be concluded pursuant to 10 CFR 51.22(b) that it is not necessary to perform an environmental assessment or an environmental imnact statement.
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