ML20081D055
| ML20081D055 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 03/15/1995 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20081D045 | List: |
| References | |
| NUDOCS 9503200108 | |
| Download: ML20081D055 (212) | |
Text
{{#Wiki_filter:! J ATTACHMENT (4) f t t t UNIT 1 TECHNICAL SPECIFICATION l MARKED-UP PAGES i l l i l l 9503200108 950315 PDR ADDCK 05000317 P PDR
TABLE OF CONTENTS
^ ADMINISTRATIVE CONTROLS --
i, SECTION pag A 6.1 RESPONSIBILITY ................... 6-1 , 6.2 0RGANIZATION 6.2.1 ONSITE & OFFSITE ORGANIZATIONS ........... 6-1 6.2.2 UNIT STAFF ..................... 6-1 6.3 FACILITY STAFF QUALIFICATIONS . . . . . . . . . . . . 6-5
- 6. TRAINING ......................
/ .
6.5 EW AND AUDIT , 6.5.1 PLANT TIONS AND SAFETY REVIEW COMITTEE (P C) Function ................. .... 6-5 Composition . .....,............. 6-5 Chaiman .............. ....... 6-6 I Alternates .... ................ 6-6 Meeting Frequency . . . ... .......... 6-6 Quorum ......... ............. 6-6 Res>onsibilities .... . ........... 6-6 Aut1ority . . . . . . . . . . . . ......... 6-8 Records . . . . . . ....... ........ 6-8 6.5.2 PROCEDURE REVI COMITTEE Function . ................ ... 6-8 Compositi ..................... 6-9 , Chairm ..................... 6-9
. Al.t ates ...................... 6-9 M ing Frequency . . . . . . . . . . . . . . . . . . 6-9 uorum ....................... -9 Authority . . . . . . . . . . . . . . . . . . . . . . 6 bit Records . . . . . . . . . . . . . . . . . . . . . . . 6-1 I
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i CALVERT CLIFFS - UNIT 1 XVI Amendment No. 169-l i l
3 - x . TABLE OF CONTENTS i ADMINISTRATIVE CONTROLS ; r SE6, TION fiE
. t 6.5.3 QLIALIFIEC REVIEWERS 6-10 i Function . . . . . . . . . . . . . . . . . . . .-..
thority . . . . . . . . . . . . . . . . . . . . .. .
.6-10
. . .- 6-11 Ce ification . . . . . . . . . . . . . . . .
6-11 Reco 6.5.4 0FF-SITE FETYREVIEWCOMMITTEE(OSSRC .
........ 11.
Fucction .. ........... 6 "
- Composition . . . ..... .. . . . . . . . . . - .
Qualifications . . . . . . - . . . . . . . . . . . -
'6-12: ,
6-12 i Consultants . . . . . . . . ............ 6-12 Meeting Frequency . . . . . . . . . . . . . . . . . . 6-12 Quorum . . . . . . . . .............
........... 6-13 i Review . . . . . . . . . . . . '
. ..... .. . . . . . ... . . 6-13 Audits . . . .
Authority . . . . . . . . . . . . . . ....... 6 ............... ..... 6-15 Records . : RTABLE EVENT ACTION . . . . . . . . . . . . ... . '6-15 6.6 J .. ............. .
. 6 SAFETY LIMIT VIOLATION I
+ -16 6.Y.
PROCEDURES . . . . . . . . . . . . . . . . . . . . . ' g . .g P Ac'6 RAM.s Aiv.O MA WvA L.S 6.E 6 REPORTING REQUIREMENTS-g f.u.),todiition Repo<f ; 6.#.1 t 9,..ge_f,m. Eqcsm
.................. 61 ,
6 St rtup Sp:rts . . . . . . . . . . . . . . . . . . . fr = ' 5 pert:- . . . . . . . . . . . . . . . . . . . 6-Monthly Operating Report .............. 6' 9 g.t.t . Annual Radiological Environmental- Operating i { git..p. 61
- Report ....................... i 4 .4. 3 = h rr.ud Radioactive Effluent Release Report ... 2 Core Operating Limits Report - . . . . . . . . . . . . -22 l i
~4 . 4. .f t.&. 6 Peusveixar POR V *~d .Sa fe' ty Va lve Aerort SPECIAL REPORTS . . . . . . . . .:. . . . . . . . . . 6-26 l 5.0.2 :
'6. 5.1 offsile he GIculation nni(opw]
6..r. 2 Po.stucide-t .Sa-pH 9
-- 4..C,3 frionaef Coolant .Sovece.s Outside cmfaten,enf ,
- 6. 5. f Teek n ita l Spe e if;c, f;,,, (7.g3 g,,,, z.,ny,,, g b
6.s~. .V Andiosef;'e E l'f/v en f Ce,n i<,./, pmym CALVERT CLIFFS - UNIT 1 XVII Amendment No. defr , 4 4
- TABLE OF CONTENTS ;
3-1 .- , ADMINISTRATIVE CONTROLS t GE MCTION l 6-27 l ; 6.10 RECORD RETENTION .................. ;
........ ... 6-28 l 6.11 TION PROTECTION PROGRAM 6.12 HIGH RADIA N AREA . . . . . . . . . . . . . . . . . .6-29 .l .
6-29 l 6.13 SYSTEM INTEGRITY ..................
.......... .... 6-29 l ;
6.14 IODINE MONITORING . .
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6-30 l 6.15 POSTACCIDENT S LING . . . . . . . . . . . . . . . . r 6-30
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6.16 PROCESS TROL PROGRAM (PCP) . . . . . . . .... l
..... 6-30 l . ., 6.17 0F TE DOSE CALCULATION MANUAL (ODCH) .
f 6.18 MAJOR CHANGES TO RADI0 ACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS . . . . . . . . . . . . 1 l ; t
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.. 1 CALVERT CLIFFS - UNIT 1 XVIII Amendment No. 186
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1.0 DEFINITIONS-CONTROLLED LEAKAGE 1.9 CONTROLLED LEAKAGE shall be the water flow from the reactor coolant pump seals. C0RE ALTERATION 1.10 CORE. ALTERATION shall be the movement or mantaulation of any_ component within the reactor pressure vessel with t1e vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION sh.all not preclude completion of movement of a component to a safe conservative position. CORE OPERATING LIMITS REPORT 1.11 The CORE OPERATING LIMITS REPORT is the unit specific document that provides cycle specific parameter limits for the current reload cycle. )/
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These cycle specific parameter limits shall be detertnined for each reload cycle in accordance with Specification 0.0.1.0. Plant operation within these limits is addressed in individual Specifications. S' S * # DOSE EQUIVALENT I-131 1.12 DOSE EQUIYALENT I-131 shall be that concentration of I-131 d' (pCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132 I-133, I-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." i - AVERAGE DISINTEGRATION ENERGY 1.13 i shall be the average (weighted in proportion to the concentration f of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 955; of the total non-iodine activity in the coolant. ENGINEERED SAFETY FEATURE RESPONSE TIME 1.14 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation g setpoint at the channel sensor until the ESF equipment is capable of perfonning its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values _etc.). Times shall include diesel generator starting and sequence loading delays ) where applicable. CALVERT CLIFFS - UNIT 1 1-3 Amendment No. 466-
- s 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS l REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and highest o>erating loop cold leg coolant temperature shall not exceed the limits stown in Figure 2.1-1. g APPLICABILITY
- N0 DES 1 and 2. >
ACTION:#'Whenever the point defined by the combination of the highest ' operating loop cold leg temperature and THERMAL POWER has exceeded the l h (} nSe'[] appropriate muuri et-p ssure line, be in HOT STANDBY w [.see P b-!LJ REACT 0RC005NTSYSTEMPRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia. - APPLICABILITY: MODES 1, 2, 3 '4 and 5. ACTION: MODES I and 2 ,
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A Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its ) W imit aithin 1 hour. (Inse<4 .1 F.s ei p. b-l(,D MDIS 3 dF ano d i A. Whenever the Reactor Coolant System pressure has exceeded 2750 psia, h reduce the Reactor Coolant System pressure to within its limit within i 5 minutes. _ MS b~ f i
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CALVERT CLIFFS - UNIT 1 2-1 Amendment No. 486-
. 3/4.3 INSTRUMENTATION ] TABLE 3.3-6 (Continued)
TABLE NOTATION Alam setpoint to be specified in a controlled document ' (e.g., setpoint control manual). ACTION STATEMENTS ACTION 14 - With the number of channels 0PERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the , ACTION requirements of Specification 3.4.6.1. ACTION 16 - With the number of channels OPERABLE less than required by , the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9. ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours, and:
. 1) either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or
- 2) prepare and submit a Special Report to the Comission pursuant to Sp; ific: tier. 0.0.2 within 30 days following F the event, ou ining the action taken, the cause of the inoperability, and the plans and schedule for restoring !
the system to )PERABLE status, l 1 ! 10 C FR 50 + l d i l l 1 CALVERT CLIFFS - UNIT 1 3/4 3-25 Amendment No. iM M
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3/4.3~INSTRMENTATION
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3/4.3.3' MONITORING INSTRUMENTATION , Meteorolonical Instrumentation - LIMITING COMITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in l Table 3.3-8 shall be OPERABLE. APPLICABILITY: At all times. i AC ON:
- a. With one or more required meteorological monitoring channels .
, inoperable for more than 7 days, prepare and submit a Special e Report'to the Commission pursuant to ! the next 10 days outlining the cause o;xift:1'in f the function 0.0.2 and within the : plans for restoring the channel (s) to OPE E status. l
- b. The provisions of Specifications 3.0.3 and 2 .0.4 are not ;
applicable. ~ It? c fR 50.+ SURVEILI.ANCE REQUIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the perfonnance of the CHANNEL : CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-5. ! I i
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M CALVERT CLIFFS - UNIT 1 3/4 3-27 Amendment No. @9-- , 9
-y~ , n g e -r , . - , - ,m.- ,,.n.~-. + . ~ , ,n-. , ..n e .- , , - -
3/4.3 INSTRUMENTATION TABLE 3.3-10 (Continued) , ACTION STATEMENTS ACTION 31 - With the number of 0PERABLE >ost-accident monitoring channels less than required >y Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days or. be in NOT $NUTDOWN within the next 12 hours. . ACTION 32 - With the number of OPERABLE post-accident monitoring channels one less than the Minimum Channels OPERABLE requirement in Table 3.3-10 operation may proceed provided the inoperable channel is restored to OPERABLE status at the next outage of sufficient duration. < ACTION 33 - With the number of OPERABLE post-accident monitoring channels two less than required by Table 3.3-10, either restore one inoperable channel to OPERABLE status within i 30 days or be in NOT $HUTDOWN within the next 12 hours. ACTION 34 - With the number of OPERABLE post-accident monitoring , channels one less than the Minimum Channels OPERABLE ' requirement in Table 3.3-10, either restore the system to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report , to the Comission pursuant to Sp::ift:: tim 0.0.2 within ' 30 days following the event, outlining th action taken, the , cause of the inoperability and the plans a d schedule. for restoring the system to 0PERAB'.E status. jp c g j4 gg,9_ r ACTION 35 - With the number of OPERARLE channels two less than required by Table 3.3-10, either restore the inoperable channel (s) to 0PERABLE status within 48 hours if repairs are feasible without shutting down or: ,
- 1. Initiate an alternate method of monitoring for core and Reactor Coolant System voiding;
@ 2. Prepare and submit a Special Report to the Comission pursuant to Sp;;ific; tic ' O.2 within 30 days following the event, outlining the ah taken, the cause of the inoperability and the plans anc e for restoring i the system to 0PERABLE status; and j g g 3 og
- 3. Restore the system to 0PERABLE status at the next scheduled refueling. ,
i CALVERT CLIFFS - UNIT 1 3/4 3-35 Amendment No. 499- M i
3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION Fire Detection Instrumentation
- LIMITING CONDITION FOR OPERATION i
3.3.3.7 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: Whenever equipment in that fire detection zone is required to be OPERABLE. ACT10N: With one or more of the fire detection instrument'(s) shown in Tab' e 3.3-11 inoperable:
- a. Within I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) .is located inside the containment, then inspect the containment at least once per 8 hours or monitor the ,
containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5; or unless the instrument (s) is located in fire detection zones equipped with - automatic wet pipe sprinkler systems alarmed and supervised to the Control Room, then within I hour and at least per 24 hours thereafter, inspect the zone (s) with inoperable instruments and verify that the Automatic Sprinkler System,-including the water flow alarm and supervisory system, is OPERABLE by CNANNEL FUNCTIONAL TEST.
- b. Restore the inoperable instrument (s) to OPERABLE status within i 14 days or prepare and submit a Special Report to the Commission pursuant to S::f ff:W;r, 5.0.2 within the next 30 days outlining the action ta cen, the i use of the inoperability and the plans '
and schedule for resto the instrument (s) to 0PERABLE status. l
/C c. FM S0. +
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
i SURVEILLANCE REQUIREMENTS I 4.3.3.7.1 At least once per 6 months, at least 25% of the above required fire detection instruments which are accessible during plant operation shall be demonstrated 0PERABLE by perfomance of a CHANNEL FUNCTION' . TEST. Detectors selected for testing shall be selected on a rotating basia such i CALVERT CLIFFS - UNIT 1 3/4 3-37 Amendment No. 199- 8
3/4.4' REACTOR C00UUli SYSTEN -
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BASES l adequate structural margins against burst during all normal operating, ! transient,'and accident conditions until the end of the fuel cycle.' ..This l evaluation would include the following elements: l
- 1. An assessment of the flaws found during the previous inspections.
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- 2. An assessment of the structural margins relative to the criteria of >
Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes," that can be expected before the end of the fuel ; cycle or 30 months, whichever comes first.
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- 3. An update of the assessment model, as appropriate, based on l comparison of the predicted results of the steam generator tube integrity assessment with actual inspection results from previous ;
inspections. !
.i The plant is expected to be operated in a manner such that the secondary 1 coolant will be maintained within those chemistry limits found to result in '
negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized 1 corrosion may likely result in stress corrosion cracking. The extent of :
, cracking during plant operation would be limited by the limitation of steam ,
generator tube leakage between the Primary Coolant System and the Secondary l Coolant System (primary-to-secondary leakage = 1 gallon per minute, total). : Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads 1 imposed dur.ng nomal operation and by postulated accidents. Operating ! plants have demonstrated that primary-to-secondary leakage of I gallon ! peminute can readily be detected by radiation monitors of steam generator ! blowdown. Leakage in excess of this limit will require plant shutdown and i an unscheduled inspection, during which the leaking tubes will be located ' and plugged. ! Wastage-type defects are unlikely with proper chemistry treatment of the ! secondary coolant. However, even if a defect should develop in service, it i will be found during scheduled inservice steam generator tube examinations. ; Plugging will be required for all tubes with imperfections exceeding the : plugging limit of 40% of the tube nominal wall thickness. Steam generator - tube inspections of operating plants have demonstrated the capability to ; reliably detect degradation that has penetrated 20% of the original tube ' wall thickness. t Whenever the results of any steam generator tubing inservice inspection l fall into Category C-3, these results will be propptly reported to the 0f Comission pe===t te Speciff =th= 5.0.S priorrthe resumption of plant , operation. Such cases will be considered by the Commission on a case-by- t case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of ; the Technical Specifications, if necessary. , i CALVERT CLIFFS - UNIT 1 B 3/4 4-4 Amendment No.108- T
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*- 3/4.4 . REACTOR COOLANT SYSTEM SURVEILLANCE REQUIRENENTS (Continued) ..
7.') 4.4.5.5 Recorts 0""9 . Y l' e " P o* f P e ' ** c of...
- a. Following each i ervice inspection of steam generator tubes, the number of tube plugged in each steam generator shall be reported ~
icr. S.a.2. to the Comis ion within 15 days pursuant to Specifi~"4 to c.t- So,+.,
- b. The comple results of the steam generator tube inservice
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'i inspection shallu be,.6f +iur.:,1ud-d-in th: .^. :::1 ^L eratin;;
4 -, -- - * < . . . ,. + .a r . . "..p;rt . . . i. , o for i
+ k. .. < a 4. ~
5U:!Niiti5r. 5 al.5.N.~Tiis' report shali'Inciudi~ ]
- 1. Number an extent of tubes inspected.
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- 2. Loca ton and percent of. wall-thickness penetration for each !
in cation of an imperfection. !
- 3. Identification of tubes plugged. l
- c. esults of steam generator. tube inspections which fall into Category C-3 require verbal notification of the NRC Regional Administrator by telephone within 24 hours prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine
-- cause of the tube degradation and corrective measures taken to prevent recurrence and shall be submitted within the next 30 days .@' 4 pursuant to Specific:ti:n 5.02. ;
10 c f M 50. + i
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.g, a b mi fled te fire co mm ,'.s3 lo ., pejo<. 10 1%<d j of each yea < puesvanf to lo C FR 50.4, CALVERT CLIFFS - UNIT 1 3/4 4-14 Amendment No. 108- #
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y . g TABLE 4.4-2 g r-M STEAN GUIERATOR TUBE INSPECTION i E t 15I 5AFFLt IN3VtcIIUN 21W) 5AFFLE llDt'tcIIUN JRD L. d IfOrtcIIUN 3D r- Sample 51ze Result Action Required Result Action Required ! y A minimum of 5 lubes per C-1 None N/A Result Action Requ1std E N/A m S. G. C-2 Plug defective tubes and N/A N/A M
, inspect additional 25 c-1 c-z Esive Plug defective N/A C-1 N/A None tubes in this S. G. tubes and inspect z c-z ring defective additional 45 tubes
- q in this S. G. tubes- g
' E-3 rerform action for C-3 result of - g l first sample " c-3 Perforfit action for ! C-3 result of first i sample N/A N/A g l c-5 Inspect ait tubes in Ain other this S. G., plug S. G.s are None N/A N/A defective tubes and C-1 R. m inspect 25 tubes in each
- a. other S. G.
5 24 hour verbal Some 3.s.s Perform action for N/A N/A notification to NRC with C-2 but no C-2 result of written followup additional second sample pursuant to S. G. are C-3 E m M d ;. 0.0.2. rg;G1tt onal Inspect all tuDes
/0 c Fg go,+ S. G. is C-3 in each S. G. and plug defective N/A N/A
;La tubes. 24 hour _
5 verbal notification to NRC with written E followup. pursuant ' g- to ;;;;;'f;;tian
= 6 9.A /acF#So,+
e li -
& S = 3 b Where N is the number of steam generators in the unit, and n'is the number of' steam generators inspected during an inspection i
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3/4.4 MACTOR COOLANT SYSTEM
, q 3/4.4.8- SPECIFIC ACTIVITY ,
LIMITING CONDITION FOR OPERATION
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r 3.4.8 The specific activity of the primary coolant shall be limited to:
- a. < 1.0 pCi/ gram DOSE EQUIVALENT I-131, and
- b. <_ 100/E pCf/ gram.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*:
- a. With the specific activity of the primary coolant > 1.0 pCi/ gram !
l DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4.8-1, operation may continue for up to 100 hours provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. The provisions of specification 3.0.4 are not applicable.
- b. With the specific activity of the primary coolant > 1.0 pCi/ gram h DOSE EQUIVALENT I-131 for more than 100 hours during one
- a9 continuous time interval or exceeding the limit line shown on Figure 3.4.8-1, be in at least HOT STANDBY with T.,, < 500'F within 6 hours.
- c. With the specific activity of the primary coolant
> 100/E pCi/ gram, be in at least HOT STANDBY with T,, < 500'F within 6 hours.
MODES 1, 2, 3, 4 and 5:
- d. With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/ gram, perform the sampling
'and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. L'htnev:r th: :p cific ectivity of the pr% ry celant e.ceed: 1.0- pCi/ gram DOSE EQUIVALENT-I-131-for-ift-excess cf 50 hour: foe-o ne-con t i n uou s-ti me-inte rv al-or-5-pe rc en t- o f-the uni +'t +atal yearly e serating-time-pursuant-to-ACTION-a-abover-a .
Special Report :h:11 m prpared-and-submitted-to-the-Connission puMuant te S; :ific tion . 9.2 within the next 30 day:. Thh
'- With T.,, > 500*F.
CALVERT CLIFFS - UNIT 1 3/4 4-24 Amendment No. 488- t ; v- cy y .
3/4.4 REACTOR C0OLANT SYSTEM -
--) LIMITEG CONDITION FOR OPERATION (Continued) report shall contain the results of the specific acti y lyses together with the following infonnation:
- 1. Re tor power history starting 48 hours f or to' the first samp in which the Itmit was excee ,
- 2. Fuel burnu y core region,
- 3. Clean-up flow h ory s ting 48 hours prior to the first sample in which the it was exceeded,
- 4. History of de- sing oper on, if any, starting 48 hours prior to the rst sample in h the limit was exceeded, and
- 5. The me duration when the specific activ of the primary lant exceeded 1.0 pCi/ gram DOSE EQUIVALEN 131.
SURVEILLANCE REQUIREMENTS
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i 4.4.8 The specific activity of the primary coolant shall be determined to
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be within the limits by performance of the sampling and analysis program of Table 4.4-4. CALVERT CLIFFS - UNIT 1 3/4 4-25 Amendment No. 400 $
3/4.4 -REACTOR C0OLANT SYSTEM
"~} LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
- a. With one PORV inoperable in MODE 3 with the RCS temperature Jg
< 365*F or in MODE 4, either resto,*e the inoperable PORY to OPERABLE status within 5 days or depressurire and vent the RCS through a t 1.3 square inch vent (s) within the next 48 hours; maintain the RCS in a vented condition until both PORVs have been restored to 0FERABLE status.
- b. With one PORY inoperable in MODES 5 or 6, either restore the inoperable PORV to OPERABLE. status within 24 hours, or
! depressurize and vent the RCS through a 11.3 square inch vent (s) /. I I within the next 48 hours; and maintain the RCS in this vented I condition until both PORVs have been restored to OPERABLE status.
- c. With both PORVs inoperable, depressurize and vent the RCS through $
a 12.6 square inch vent (s) within 48 hours; maintain the RCS in a vented condition until either one OPERABLE PORY and a vent of 3 1.3 square inches has been established or both PORVs have been restored to OPERABLE status,
- d. In the event either the PORVs or the RCS vent (s) are used to M mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to
') mcrA50.t S' ciff atter, 5.0.2 within 30 days. The report shall describe tie circumstances initiating the transient, the effect of the i PORVs or vent (s) on the transient and any corrective action I necessary to prevent recurrence.
- e. With less than two HPSI pumps' disabled, place at least two HPSI h pump handswitches in pull-to-lock within fifteen minutes and disable two HPSI pumps within the next four hours.
- f. With one or more HPSI loo) MOVs' not prevented from automatically j<
aligning a HPSI pump to tie RCS, immediately place the MOV handswitch in pull-to-override, or shut and disable the affected MOV or isolate the affected HPSI header flowpath within four hours, and implement the ACTION requirements of Specifications 3.1.2.1, 3.1.2.3, and 3.S.3, as applicable, l
- g. With HPSI flow exceeding 210 gpm while suction is aligned to the RWT and an RCS vent of < 2.6 square inches exists,
- 1. Immediately take action to reduce flow to less than or equal to 210 gpm.
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I EXCEPT when required for testing. CALVERT CLIFFS - UNIT 1 3/4 4-34 Amendment No. 169
3/4.4 REACTOR C00LANT SYSTEN
) 3/4.4.11 CORE BARREL MOVEMENT _
LIMITING CONDITION FOR OPERATION 3.4.11 Core barrel movement shall be limited to less than the ' Amplitude Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level. ; APPLICABILITY: M0DE 1.
- ACTION:
- a. With the APD and/or SA exceeding their applicable Alert Levels, POWER OPERATION may proceed provided the following actions are taken:
- 1. APD shall be measured and processed at least once per 24 hours,
- 2. SA shall be measured at least once per 24 hours and shall be processed at least once per 7 days, and
- 3. A Special Report, identifying the cause(s) for exceeding the a)plicable Alert Level, shall be prepared and submitted to tie Comission pursuant to Sp::ific: tier. S.0.2 within 30
'l days of detection. /0 C FN SE f-
- b. With the APD and/or SA exceeding their applicable Action Levels, measure and arocess APD and SA data within 24 hours to detemine if the core )arrel motion is exceeding its limits. With the core barrel motion exceeding its limits, reduce the core barrel motion to within its Action Levels within the next 24 hours or be in HOT STANDBY within the following 6 hours. ;
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
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CALVERT CLIFFS - UNIT 1 3/4 4-39 Amendment No. -148- M,
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) - r
-] 3/4.5.2 ECCS SUBSYSTEMS - MODES 1. 2 AND 3 (L> 1750 PSIA) _
LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsysterer, shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE high-pressure safety injection pump,
- b. One OPERABLE low-pressure safety injection pump, and
. c. An 0PERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and .
automatically transferring suction to the containment sump on a Recirculation Actuation Signal. APPLICABILITY: MODES 1, 2 and 3*. ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to 0PERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. In the event the ECCS is actuated and injects water into the
,1 h Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Sp :tft : tier. 5.0.2 within 90 days describing the circumstances ofI he actuation and the total accumulated actuation cycles to date.
i iO c FM 50. +-
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- i With pressurizer pressure > 1750 psia.
( j CALVERT CLIFFS - UNIT 1 3/4 5-3 Amendment No.169-i
i 3/4.5 ' EMERGENCY CORE C0OLING SYSTEMS (ECCS)
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3/4.5.3 ECCS SUBSYSTEMS - MODES 3 (< 1750 PSIA) AND 4 - [ LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERA 8LE: ,
- a. One' OPERABLE high-pressure safety injection pump, and ;
- b. An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and- '
automatically transferring suction to the containment sump on a Recirculation Actuation Signal. APPLICABILITY: MODES 3* and 4. i ACTION: .
- a. With no ECCS subsystem 0PERABLE, restore at least one ECCS i subsystem to 0PERABLE status within 1 hour or be in COLD SHUTDOWN :
within the next 20 hours. !
- b. In the event the ECCS is actuated and injects water into the !
Reactor Coolant System, a Special Report shall be prepared and ;
, submitted to the Commission pursuant to Specific:+1:r. 5.0.2 within 90 days describing the circumstances of th actuation and ;
the total accumulated actuation cycles to date. l 10 C FR 50 9- 7 SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the : applicable Surveillance Requirements of 4.5.2. ' l I Between 385 F and 365 F, a transition region exists where the OPERABLE f HPSI pump will be placed in pull-to-lock on a cooldown and restored to automatic status on a heatup. At 365 F and less, the required CPERABLE f HPSI pump shall be in pull-to-lock and will not start automatically. At 365'F and less, HPSI pump use will be conducted in accordance with Y :
! Technical Specification 3.4.9.3.
l With pressurizer pressure < 1750 psia. CALVERT CLIFFS - UNIT 1 3/4 5-7 Amendment No. MS- j
3/4.6 CONTAINMENT SYSTEMS , S SURVEILLANCE REQUIREMENTS (Continued) , adjacent tendons is found unacceptable, it shall be considered as , evidence of possible abnomal degradation of the containment : structure. In addition, more than one unacceptable tendon out of i those selected for surveillance (from all three tendon groups) shall be considered as evidence of possible abnomal degradation of the containment structure. , If the nomalized lift-off force of any single tendon lies below the lower bound individual, the occurrence should be considered as evidence of possible abnomal degradation of the containment structure. , In addition, detemining that the average of the nomalized lift- i off forces for each sample population (hoop, vertical, dome) is equal to or greater than the required average prestress level; 536 kips for hoop tendons, 622 kips for vertical tendons, and 555 kips for dome tendons (reference Figures 3.6.1-1, 3.6.1-2, , and 3.6.1-3). If the average is below the required average prestress force, it shall be considered as evidence of possible ' abnormal degradation of the containment structure.
- b. Removing one wire from each of a dome, vertical and hoop tendon ;
checked for lift-off force, and determining. over the entire : length of the wire: _,1
- 1. The extent of corrosion, cracks, or other damage. The
)resence of abnomal corrosion, cracks, or other damage shall se considered evidence of possible abnomal degradation of the containment structure. >
- 2. A minimum tensile strength value of 240 Ksi (guaranteed ultimate strength of the tendon material) for at least three wire samples (one from each end and one at mid-length) cut from each removed wire. Failure of any one of the wire samples to meet the minimum tensile strength test is evidence of possible abnomal degradation of the containment structure.
- c. Perfom a chemical analysis to detect changes in the chemical properties of the sheath filler grease. Any unu::ual changes in physical appearance or chemical properties that could adversely affect the ability of the filler grease to adhere to the tendon wires or otherwise inhibit corrosion shall be reported to the Commission pursuant to o.,.u within the next 30 days. y . . . (, . . - . . m . .
10 C F R 50. 4- . J CALVERT CLIFFS - UNIT 1 3/4 6-10 Amendment No. Mrt-
l 3/4.6 CONTAINMENT SYSTEMS _' ! 3 SURVEILUUICE REQUIREMENTS (Continued) _ . . 4.6.1.6.2 End Anchoraoes and Adjacent Concrete Surfaces. The structural integrity of the end anchorages and adjacent concrete surfaces shall be demonstrated by determining through inspection of a representative sample of tendons (reference Specification 4.6.1.6.1) that no apparent changes have occurred in the visual appearance of the end anchorages or their adjacent concrete exterior surfaces. Also, inspections of the pre-selected concrete crack patterns adjacent to end anchorages shall be performed during the Type A containment leakage rate tests (reference Specification 4.6.1.2) while the containment is at its maximum test pressure. 4.6.1.6.3 Containment Surfaces. The exposed accessible interior and exterior surfaces of the containment, including the liner plate shall be > visually inspected during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2). This inspection shall be performed prior to the Type A containment leakage rate test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak tightness. 4.6.1.6.4 . Reports. Any abnormal degradation of the containment structure detected during the above required tests and inspections shall be reported to the Commission pursuant to Spe-4ficaticr. 5.0.2 within the next 30 days. This report shall include a descr/ption of the tendon condition, the
, condition of the concrete (especih11y at tendon anchorages), the inspection '
procedures, the tolerances on cral: king, and the corrective actions taken. I IO C FR 50.+ t g/ CALVERT CLIFFS - UNIT 1 3/4 6-11 Amendment No. 44GL
3/4.6 C0'NTAINMENT SYSTEMS - 3/4.6.5 g>6USTIBLE GAS CONTROL Hydrocen Analyzers LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent containment hydrogen analyzers shall be OPERABLE. APPLICABILITY: MODES I and 2. ACTION: ;
- a. With one hydrogen analyzer inoperable, restore the. inoperable analyzer to OPERABLE status within 30 days or:
/ 0 C.F R 5 0. 4-
- 1. Verify containment a osphere grab sampling capability and prepare and submit special report to the Coranission pursuant to '+::i.f n tf = 5.0.2 within the following 30 days, outlining the ACTION taken, the cause for the inoperability, ,
and the plans and schedule for restoring the system to i 0PERABLE status, or
- 2. Be in at least H0T STANDBY within the next 6 hours.
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') b. With both hydrogen analyzers inoperable, restore at least one inoperable analyzer to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours,
- c. Specification 3.0.4 is not applicable to this requirement.
SURVEILLANCE REQUIREMENTS 4.6.5.1.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least { bi-weekly on a STAGGERED TEST BASIS by drawing a sample from the Waste Gas ; System through the hydrogen analyzer. l 4.6.5.1.2 Each hydrogen analyzer shall be demonstrated CPERABLE at least once per 92 days on a STAGGERED TEST BASIS by perfonning a CHANNEL I CALIBRATION using sample gases in accordance with manufacturers' I recommendations. , A CALVERT CLIFFS - UNIT 1 3/4 6-25 Amendment No.167- 8
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3/4.7 PLANT SYSTEMS y ^ SURVEILLANCE REQUIREMENTS (Continued)
- a. Sources in use - At least once per six months for all sealed
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sources containing radioactive material:
- 1. With a half-life greater than 30 days (excluding Hydrogen 3),and
- 2. In any form other than gas.
- b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer t.g. another - ,
licensee unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use. *
- c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source or detector.
4.7.9.1.3 Reports - A report shall be prepared and submitted to the Commission on an annual basispif sealed source or fission detector leakage tests reveal the presence of 10.005 microcuries of removable contamination. p usani to io c fa so.+ ! l l l l 1 l l CALVERT CLIFFS - UNIT 1 3/4 7-31 Amendment No. 166- 8
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3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System
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LIMITING CONDITION FOR OPERATION 3.7.11.1 The Fire Suppression Water System shall be OPERABLE with: s
- a. Two high pressure pumps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header,
- b. Two water supplies, each with a minimum contained volume of 300,000 gallons, and
- c. An 0PERABLE flow path capable of taking suction from the Pretreated Water Storage Tanks Numbers 11 and 12 and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves and the first valve ahead of the water flow alann device on each sprinkler, hose standpipe or spray system riser required to be OPERABLE per Specifications 3.7.11.2, 3.7.11.4, and 3.7.11.5.
APPLICABILITY: At all times. ! ACTION: i
- a. With one pump and/or one water supply inoperable, restore the !
inoperable equipment to OPERABLE status wittin 7 days or prepare and submit a Special Report to the Commission pursuant to /vcFXga.1Spect'inticr. S.O.2 within the next 30 days outlining the plans , and arocedures to be used to provide for the loss of redundancy < in tiis system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ,
- b. With the Fire Suppression Water System otherwise inoperable:
- 1. Establish a backup Fire Suppression Water System within 24 hours, and
- 2. Submit a Special Report in accordance with 4=ifi= tier. 0.M: / o c /~# .fo. 4 i
a) By telephone within 24 hours, 4 i CALVERT CLIFFS - UNIT 1 3/4 7-33 Amendment No. i n M l
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3/4.7 PLANT SYSTEMS 3/4.7.11
' FIRE SUPPRESSION SYSTEMS Soray and/or Sorinkler Systems LIMITING CONDITION FOR OPERATION 3.7.11.2 The spray and/or sprinkler systems shown in Table 3.7-5 shall be OPERABLE:
APPLICABILITY: Whenever equipment in the spray / sprinkler protected areas is required to be OPERABLE. ACTION:
- a. With.one or more of the required spray and/or sprinkler systems inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant safe shutdown systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to LOC.F450,9 Specifi dicr. 5.0.2 within the next 30 days outlining the action 04 taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
I b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.11.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPEPABLE:
- a. At least once per 31 days by verifying that each valve (manual.
power-operated or automatic) in the flow path, not locked, sealed or otherwise secured in position, is in its correct position.
- b. At least once per 12 months by cycling each valve in the flow path through at least one complete cycle of full travel,
- c. At least once per 18 months
- 1. By performing a system functional test which includes simulated automatic actuation of the system, and verifying that the automatic valves in the flow path actuate to their correct positions on a simulated test signal.
CALVERT CLIFFS - UNIT 1 3/4 7-37 Amendment No. -166- d
i I 3/4.7' PLANT SYSTEMS _ D 3/4.7.11 FIRE SUPPRESSION SYSTEMS -- Halon Systems ] l l LIMITING CONDITION FOR OPERATION 3.7,11.3 The following Halon Systems shall be OPERABLE with the storage l tanks having at least 95% of full charge weight (or level) and 90% of full charge pressure. ;
- a. Cable s areading room total flood system, and associated vertical ;
cable clase IC, Unit 1.
- b. 4160 volt switchgear room 27' & 45' elevation Unit 1. ;
APPLICABILITY: Whenever equipment protected by the Halon System is required to be OPERABLE. , ACTION:
- a. With both the primary and backup Halon Systems )rotecting the areas inoperable, within one hour establish an 1ourly fire watch with backup fire suppression equipment for those areas protected ,
by the ino >erable Halon System. Restore the system to OPERABLE status wit 11n 14 days or prepare and submit a Special Report to
) the Conrnission pursuant to Specift- tier. S.0.2 within the next 30 days outlining the action taken[:the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. g
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l l SURVEILLANCE REQUIREMENTS 4.7.11.3 Each of the above required Halon Systems shall be demonstrated ! OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, i power-operated or automatic) in the flow path is in its correct i position,
- b. At least once per 6 months by verifying Halon storage tank weight (level)andpressure.
CALVERT CLIFFS - UNIT 1 3/4 7-40 Amendment No. M G M. l d l
J 3/4.7 PLANT SYSTEMS - i 3/4.7.11 FIRE SUPPRESSION SYSTEMS
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Fire Hose Stations > LIMITING CONDITION FOR OPERATION -
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3.7.11.4 The fire hose stations shown in Table 3.7-6 shall be OPERABLE. --- i APPLICABILITY: -Whenever equipment in the areas protected by the fire hose ' stations is required to be OPERABLE. : ACTION: 1
- a. With one or more of the fire hose stations shown in Table 3.7-6 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within I hour. Restore the fire hose station (s) to OPERABLE status within 14 days or prepare and submit 'a Special Report to the Of Commission pursuant to Ep^:i 30 days outlining the action taken,
[fintia.the 0.0.2 within cause the next of the inoperability and the plans 'and schedule for restoring the fire hose station (s) to OPERABLE status.
- b. The provisions of Specificakions 3.0.3 and 3.0.4 are not applicable.
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l 10 cf R 50.4-SURVEILLANCE REQUIREMENTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7-6 'shall be demonstrated CPERABLE: -- 2
- a. At least once per 31 days by visual inspection of the station to assure all required equipment is at the station. Hose stations located in the containment shall be visually inspected on each scheduled reactor shutdown, but not more fre.quently than every -
31 days. bv At least once per 18 months for hose stations located outside the containment and once per REFUELING INTERVAL for hose stations ~ inside the containment by:
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- 1. Removing the hose for inspection and re-racking, and
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- 2. Replacement of all degraded gaskets in couplings. -
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CALVERT-CLIFFS - UNIT 1 3/4 7-42 Amendment No. M6 d
3/4.7 PLANT SYSTEMS 3/4.7.11
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FIRE SUPPRESSION SYSTEMS Yard Fire Hydrants and Hydrant Hose Houses LIMITING CONDITION FOR OPERATION 3.7.11.5 The following yard fire hydrants and associated hydrant hose houses shall be OPERABLE: *
- a. f6 yard hydrant and associated hydrant hose house, which provides primary protection for Unit 2 RWT blockhouse.
- b. #7 yard hydrant and associated hydrant hose house, which provides primary protection for Unit 1 RWT blockhouse.
APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE. ACTION:
- a. With one or more of the yard fire hydrants or associated hydrant ,
hose houses inoperable, within I hour have sufficient additional lengths of 2-1/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire lydrant or associated hydrant hose 9 house is the primary means of fire suppression. Restore the hydrant or hose house to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specif4cetion 5.0.2 within the next 30 days outlining the action taken,f the cause of the inoperability, and the plans and schedule for re storing the hydrant or hose house to OPERABLE status,
- b. The pr] visions of Specifications 3.0.3 and 3.0.4 are not !
applic able.
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to F12 SO . f l l CALVERT CLIFFS - UNIT 1 3/4 7-45 Amendment No. M6- d
j f 3/4.7. PLANT SYSTEMS !- 3/4.7.12 PENETRATION FIRE BARRIERS i LIMITING CONDITION FOR OPERATION 3.7.12 All' fire barrier penetrations (i.e., cable penetration barriers. fire doors and fire dampers), in fire zone boundaries, protecting safe shutdown areas shall be OPERABLE. , APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required fire barrier penetrations inoperable within one hour either establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on'at least one side of the inoperable fire barrier and establish an hourly fire watch p(atrol; includingor verify the waterthe flowoperability of Automatic alarm and supervisory system)Sprinkler on both System sides of the inoperable fire barrier. Restore the inoperable fire barrier penetration (s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Comission pursuant to !
h /OcFA50.+ Spri#inti= 5.0.2 within the next 30 days outlining the action taken, the cause of the inoperable penetration and plans and schedule for restoring the fire barrier penetration (s) to ) OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.12 Each of the above required fire barrier penetrations shall be ! verified to be OPERABLE:
- a. At least once per 18 months by a visual inspection.
- b. Prior to returning a fire barrier penetration to functional i' status following repairs or maintenance by perfomance of a visual inspection of the affected fire barrier penetration (s).
l l l CALVERT CLIFFS - UNIT 2 3/4 7-47 Amendment No. M 3 8 l l
l 6.0 ADMINISTRATIVE CONTROLS ' I 1 3 6.1 _ RESPONSIBILITY _ h 6.1.1 operation an{d shall delegate in writing the succession to thisThe Ilant responsibility during his absence. jneladiory fh e p lw$ Specific 6.2 ORGANIZATION YifI'd # 8"#"
.p,,) y;ti;nj "th5'Yi,"">
e re.spo" sWM o.o 6.2.1 ONSITE & OFFSITE ORGANIZATIONS ,f the P /*I"'" M I nes e TechmuI SteciNeo fim Onsite and offsite organizations shall be esta ished for unit operation ' and corporate management, respectively, The on te and offsite organizations shall include the positions for act vities affecting the safety of the nuclear power plant.
- a. Lines of authority, responsibility and comm ication shall be established and defined for the highest manag ent levels through intemediate levels to and including all operat g organization positions. These relationships shall be documen d and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and re ationships, and job descriptions for key personnel positions, or l g equivalent foms of documentation. These requirements j shall be !
documented in FS?", Ch:pter 12, =d upd:ted in eccerdence .ith 18 10 CFR 50.71(ef. fA e v F.5 A R. h b. The llant Generel ganager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
- c. The Vice President - Nuclear Energy shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable perfomance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.
~d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. l 6.2.2 UNIT STAFF I ye un it .s fa $ oqnet izaffe .1 s ha Il K. EL.. = duty :hift shall .: amp:=d :f at ic=t the -4a4""""
shift cr= = p=iti= shsa in T&ble 0.2-1. is e Iv.4 e the fof}vw; ; CALVERT CLIFFS - UNIT 1 6-1 Amendment No.169-
l 6.0 . ADMINISTRATIVE CONTROLS
- i ^ b. At least one licensed Operator.shall be in the control Room when !
fuel is in the reactor, . . .
- c. ' At least two licensed Operators shall.be present in the Control !
Room during reactor STARTUP, scheduled reactor shutdown, and l during recovery from reactor trips. ;
- d. An individual qualified in radiation protection procedures shall ,
be on site when fuel is in the reactor. j i
- C ll CORE A ALTERATIONS after the initial fuel loading s alredii . :r:i==d by either a licensed c or Operator .
> or Senior Reactor 0 el Handling who has no ! other ra ;....6 esponsibilities during l e .f. A site Fire Brigade of at least 5 members shall be maintained l onsite at all times. The Fire Brigade shall not include the ! minimum shift crew necessary for safe shutdown of both units l (4 members) or any personnel requir for other Assential ; functions during a fire emergency. m.scJt~2(.see p. 6 r l f Ar. The S :Yi$t YNt - _[$r0;:r: tion: shall hold or have held a [ senior reactor Sg:rti::r $;,op;erator license r 0::r:ttent, ShiftatSupervisor Calvert Cliffs. andThe Generel.supeyLwe 1 Control Room Supervisor shall iold a senior reactor operator license. !
.. The Control Room Operator shall hold a reactor operator license.. l' .:." ?) < JI t a.. A totaI of ' tYst <ee non- He en. sed ope
- f& s Si,a ({ be. a3.s n'g n ed fo the lAsifIA2.
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.sliff crews,
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Q ,[M3er Y S 3J.f-(Jee p. 6 - +j -
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Q .[MSer$ 2 O ($ ee f. b l l
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CALVERT CLIFFS - UNIT 1 6-2 Amendment No. M9-
--,p-- - __ , _
i 6.0 ADMINISTRATIVE CONTROLS - - l ^' ' TABLE 6.2-1 . MINIMUM SHIFT CREW COMPOSITION' ondition of Unit 1 - Unit 2 in MODES 1. 2. 3 o 4 l N / APPLICABLEMDES LICENSE f CATEGORY 1, 2, 3 & 4 j 5&6 SO L** \ 2 [ 2*
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0L** 3 3 Non-Licensed \ 3 / 3 Shift Technical Advisor Y 1" Condition of t 1 - Unit in MODES 5 or 6
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APPLIkBLEMODES CATEGORY 1. 2, 3 & 4
\ 5&6 S O L** / 2 \ 1*
O L** [ 3 \2 ! Non-Licens/d 3 \3 Shift Tp[hnical Advisor 1" 0\
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CALVERT CLIFFS - UNIT 1 6-3 Amendment No. M9-
6.0 ADMINISTRATIVE CONTROLS 7 T? ALE 5.:'-; (Cuntinsed) d7 3, A A # TfaiE %0 TAT!0%
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(SesP p) 8 C;;; =t in:1:d: the licerred Senter R:::ter Operater :r Scr.ior I h!bh: ff r f T Ne -b d '0fra so s+ h kg k:::E y ; P;p,n.w&....yZ...}=Msti5$iE Bl.+A unitt 1 i m icensea on ea M8L spapa /;.aad .,,g,,,,,,(
" " d'
@ / h. Shift crew composition may be ledan the minimum requirements *for a period of time not to exceed)t hours in order to accomodate 3 L. sert f, unexpected absence of on duty shift crew members provided imediate action is taken to restore the shift crew composition to-within the
(.scep,p2} minimum requirements of Tele 6.? -' 7 f- I M T TMbe qualified / serve as follows: (1) With unit in MODE S or 6, and the other unit in H0DE 1, 2, 3 4 or 4, the L holder other than the Shift Supervisor shall ser e as the STA. (2) With one unit defuele the other unit in 1, 2, 3 or 4, the STA shall be an SOL ho in addit o the one SOL required. . (3) With both units in MODE , 3 or 4, t TA shall be an SOL J f holder in addition e two SOLs require . l 4 l As an alternat to the above, the STA may be an indiv with the / )
- following mum qualifications: a bachelor degree or equiv nt in i a sci fic or engineering discipline with specific training in ,
p design and response and analysis of the plant for transients / ' nd accidents. M J Aeplacul Ah Inse f 3 (ncyt m,). 1
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CALVERT CLIFFS - UNIT 1 6-4 Amendment No. 1 &
In.se A 3 (see P 6-1) ,
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- g. One Shift Technical Advisor (STA) shall be assigned to the shift crew when either unit is' in MODE 1,2,3 or 4 and shall be filled as follows:
l l j Ih I (1) by the Shift Supervisor or an on-shift Senior Operator License (SOL holder arovided the individual meets the Commission Policy Statement on ngineering Expertise on Shift; or
/ l (2) by an individual meeting the minimum STA education and training requirement of !
) Specification 6.3.1; or l
i
) (3) by a SOL holder previously approved as an exception to the minimum STA !
education requirements of Specification 6.3.1, provided the following conditions [ are met. (i) With both units in MODE 1,2,3 or 4, the STA shall be an SOL holder in L addition to the two SOL holders required. )
- I (ii) With one unit in MODE 1,2,3 or 4 and the other unit in MODE 5 or 6, the !
STA shall be an SOL holder other than the Shift Supervisor. l I (iii) With one unit in MODE 1,2,3 or 4 and the other unit defueled, the STA shall be a SOL holder in addition to the one SOL holder required. ! N m - - I I
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l i-i i i i i
i 6.0 ADMINISTRATIVE CONTROLS - 6.3 FACILITY STAFF OUALIFICATIONS . . . 6.3.1 Each member of the facility staff shall meet or exceed the minimum ualifications of ANSI N18.1-1971 for comparable positions, except for
- 1) the Radiation Safety Engineer who shall meet or exceed the ;
. qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift Technical Advisor who shall have a Bachelor's Degree or equivalent in a ! scientific or engineering discipline with specific training in plant ! design, and response and analysis of the plant for transients and f accidents. Add;tionaf e s'cep fions to AVJ T /v13./ -117/ a re 2 con ta ined in Tabt a is -I o f the Qval& Asswa~e e Policy i g' fe < the calve.4 ct;H* Wlea< Po we< Plad, ~
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@ 6.4Mg and replacement trai shall be maintaine Training and shal r exceed e ,wwir---
4 ;;. M facility staff on of the General Supervisor - Nuclear and recommendations of t Sectio NSI N18.1-1971 and 10 CFR 55.59(c), as _- - _ _ - y .,g-HA # 5.4.2 A tr:f-in; ;r:gr= f:r th: Fire Brigad shall 6 : int:ined und:r-OM the direction Of t e ".:::;:r Sele:r S:fety :nd Plan 9ng Dep:r+==t-and i shAl meet er :::::d the requirements of NFPA 27, 1975 edition.
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" A
- 6. REVIEW AND AUDIT 1
Q 6.5.1 NT OPERATIONS AND SAFETY REVIEW COMMITTEE (POSRCA'
, (seep.6.2) l r
FUNCTION 6.5.1.1 The POSRC all function to advise the ant General Manager on . all matters related t uclear safety. COMPOSITION t 6.5.1.2 The POSRC shall be com of at least seven, but no more than ten, members, including the C irman. embers shall collectively have experience in the followin reas: Nuclear Operatio Electrical an ontrols Maintenance Chemistry Mechanica Maintenance Nuclear ngineering Radi on Safety P1 Engineering ' sign Engineering
- s. .
CALVERT CLIFFS - UNIT 1 6-5 Amendment No. If+
6.0 ADMINISTRATIVE CONTROLS - 31 ers shall be appointed in writing by the Plant General Manag . . . . i ers shall have a minimum of eight years power plant experi ce of which a mi mum of three years shall be nuclear power experience. t least one , member hall have an SRO license on Calvert Cliffs Units 1 nd 2. CHAIRMAN , 6.5.1.3 The iman and alternate Chaimen of th POSRC shall be l appointed in wr ing by the Plant General Manager. Chairmen shall have a minimum of 10 yea s power plant experience of w ch a minimum of three years shall be nucL ar power experience. ALTERNATES 6.5.1.4 All alternate memb rs shall b ap)ointed in writing by the Plant , General Manager. Alternate mbers all save a minimum of eight years > power plant experience of whi am imum of three years shall be nuclear power experience. t MEETING FREQUENCY l 6.5.1.5 The POSRC shall mee at least once per calendar month and as convened by the POSRC Chai an or one o the designated alternates. 1 OUORUM 6.5.1.6 A quorum of e POSRC shall include he Chaiman or one of the designated alternate hairmen and shall consis of a majority of the members, including 4 ternates. No more than ha f of the quorum shall be i alternates, includ 4g an alternate chaiman. . 1 RESPONSIBILITIE 6.5.I'.7 The1 SRC shall be responsible for the follow ng except for those items design ted for review by the Procedure Review Comgittee or Qualified 3 Reviewer in accordance with Specification 6.5.2 and 6.5.3 respectively:
- a. eview of 1) all procedures required by Specifica ion 6.8 and changes thereto, and 2) any other proposed proced es or changes -
thereto as determined by the Plant General Manager o affect nuclear safety. . ..] CALVERT CLIFFS - UNIT 1 6-6 Amendment No. M 9-
6.0 ADMINISTRATIVE CONTROLS _ O Cross-disciplinary reviews of these procedures are conductd in accordance with administrative procedures in addition to he reviews conducted by POSRC, the Procedure Review Comit e, or Qualified Reviewer.
- b. view of all proposed tests and experiments that fect nuclear s ety.
- c. Revi of all proposed changes to Appendix A T chnical Speci cations.
- d. Review o 11 )roposed changes or modific tons to plant systems or equipmen t1at affect nuclear safety. .
- e. Review of the ant Security Plan and mplementing procedures and shall submit rec nded changes to he Off-Site Safety Review Comittee.
- f. Review of the Emerge y Plan an implementing procedures and shall submit recomen chang to the Off-Site Safety Review Comittee.
- g. Review of changes to the P CESS CONTROL PROGRAM and the 0FFSITE DOSE CALCULATION MANUAL.
- h. Review of all 10 CFR 5 .59 Safe Evaluations that support procedures in 6.5.1.7.a and chan s or modifications in 6.5.1.7.d.
- 1. Investigation of 11 violations of t Technical Specifications including the p paration and forward g of reports covering evaluation an recomendations to preve t recurrence to the Plant General Mana r, the Vice President - Nu ear Energy and to the Chaiman of he Off-Site Safety Review Co ittee.
J. Review o all REPORTABLE EVENTS.
, k. Revie of facility operations to detect potenti 1 safety hazards.
- 1. Rev w of any accidental, unplanned or uncontrolle radioactive r ease that exceeds 25ft of the limits of Specifica ion 3.11.1.2,
.11.2.2 or 3.11.2.3, including the preparation of rbports covering evaluation, recomendations and disposition at the corrective action to prevent recurrence and for forward (ng of these reports to the Plant General Manager and the Off-Skte Safety Review Comittee.
\
1 CALVERT CLIFFS - UNIT 1 6-7 Amendment No.149-
6.0 ADMINISTRATIVE CONTROLS . i Dh m. Performance of special reviews, investigations or analyses and_. reports thereon as requested by the Chairman of the Off-Site Safety Review Committee. I AUTHORITY . 6.5.1.8 The P nt Operations and Safety Review Committee sh : ,
- a. Recome to the approval authority approval or isapproval of procedur considered under 6.5.1.7.a. ,
- b. Recomend t the Plant General Manager app al or disapproval of items conside ed under 6.5.1.7(b) through h) above. j
- c. Render determin ions in writing with gard to whether or not each item conside ed under 6.5.1.7(a through (h) above constitutes an unr fewed safety q stion.- '
- d. Evaluate root causes d rec ded a'ctions to recurrence for items co sidere under 6.5.1.7(i)preventthrough (1).
- e. Provide written notificat within 24 hours to the Vice President - Nuclear Energy d the Chaiman of the Off-Site .
Safety Review Comittee d agreement between the POSRC and the !
, responsible approval au ority n the case of item 6.5.1.7.a or between the POSRC and he Plant eneral Manager; however, the ,
Plant General Manage shall have esponsibility for resolution of - such disagreements ursuant to 6.1 above. RECORDS l 6.5.1.9 The POSRC sh maintain written minutes o each meeting and ; copies shall be prov ed to the Vice President - Nuc tar Energy, Chairman ; of the Off-Site Saf y Review Comittee, and the Plan 1. General Manager. , 6.5.2" PROCEDU REVIEW COMITTEE FUNCTION 6.5.2.1 e Procedure Review Comittee may function to review tems listed in Speci cation 6.5.1.7(a) in lieu of review by POSRC or Quali d Reviewe as directed by the Plant General Manager. l l l
*- I CALVERT CLIFFS - UNIT 1 6-8 Amendment No. 40-
r . l I _6.0 ADMINISTRATIVE CONTROLS 3 COMPOSITION - 6.5.2. The Procedure Review committee shall be composed of a Chai an and , eight dividuals who shall collectively have expertise in the ar s ! containe in Technical Specification 6.5.1.2. ; Members sha ' be appointed in writing by the Plant General Ma ger. Members shal' have a minimum of eight years power plant exp tence of which a minimum of t ee years shall be nuclear power experience At least one member shall be POSPC member or alternate. The charter for the Procedure Review Committee all include a description of members p, qualifications, , functions, and rep ts and shall be described in plan administrative procedures. The Pro edure Review. Committee may be d ssolved at the discretion of the Pla General Manager. CHAIRMAN 6.5.2.3 The Chairman and al ernate Chairme of the Procedure Review Committee shall be appointed writing b the Plant General Manager. Chairmen shall have a minimum eight y ars power plant experience of , which a minimum of three years s all b nuclear power experience. ALTERNATES 6.5.2.4 All alternate members all b ap)ointed in writing by the Plant General Manager. Alternate m bers sha lave a minimum of eight years power plant experience of w ch a minimu of three years shall be nuclear power experience. MEETING FRE0VENCY 6.5.2.5 The Proce re Review Committee shall me t at least once per calendar month an as convened by the Chairman or is designated alternates. OUORUM 6.5.2.6 A quorum for the Procedure Review Committee shal consist of the Chai n or one of the designated alternate Chairmen and th e primary or alte ate members provided at least four disciplines are re sented. CALVERT CLIFFS - UNIT 1 6-9 Amendment No. 169-
6.0 ADMINISTRATIVE CONTROLS - S AUTJiORITY _. 6.5.2. The Procedure Review Comittee shall:
- a. approval of 3 Recomend to the Approval ocedures considered Authority under 6.5.1.7 a . (a proval or d
- b. Ren r deteminations in writing with regard o whether or not each ocedure under 6.5.1.7(a) constitutes an unreviewed safety .
questio '
- c. Provide wr ten notification within 24 ours to the Vice Presi. dent - clear Energy and the C iman of the Off-Site Safety Review ittee of disagre nts between the Procedure Review Comitte and the res)onsi e approval authority. The Plant General Man er shall lave esponsibility for resolution of such disagreements rsuant to .1.1 above.
t RECORDS 6.5.2.8 The Procedure Review Com ee shall maintain written minutes of each meeting and copies shall be p o ided to the Plant General Manager. 6.5.3 OUALIFIED REVIEWERS FUNCTION 6.5.3.1 The Plant Gene 1 Manager may design e specific procedures or ) classes of procedures escribed in Specificatio 6.5.1.7.a to be reviewed by Qualified Reviewer in lieu of review by POSR or the Procedure Review Comittee. AUTHORITY j 6.5.3.2 Qua fied Reviewers shall: l
- a. ecomend to the approval authority approval or isapproval of designated procedures and changes considered unde 6.5.1.7.a. and
. Render detemination in writing with regard to whet r or not i each procedure under 6.5.1.7.a constitutes an unrevi ed safety question.
l me & , CALVERT CLIFFS - UNIT 1 6-10 Amendment No. M9
6.0 ADMINISTRATIVE CONTROLS 'N h c. Provide written notification within 24 hours to the Vice President - Nuclear Energy and the Chairman of the Off- te Safety Review Committee of disagreements between the alified Reviewer and the ap)roval authority. The Plant Gene 1 Manager hall have responsi)ility for resolution of such d . agreements p rsuant to 6.1.1 above. , CERTIFICATION 6.5.3.3 Qualifie Reviewers shall be nominated, tr ned, and certified in accordance with ad nistrative procedures. Certif cation shall be by a
. department manager.
6.5.3.4 Certification equirements of person 1 designated as Qualified Reviewers shall be in a ordance with admini rative procedures. Qualified Reviewers shall ave:
- a. A Bachelors degree n engine ing, related science, or technical discipline, and two ears o nuclear power plant experience; OR
- b. Six years of nuclear w plant experience
- c. Equivalent combi tion of edu tion and experience as approved by a Department Ma ager.
RECORDS 6.5.3.5 Review of rocedures by Qualified Rev wers shall be documented in accordance with a inistrative procedures. 6.5.4, OFF-SITE AFETY REVIEW C0FNITTEE (OSSRC) FUNCTION 6.5.4.1 T e Off-Site Safety Review Committee shall func ion to provide independ t review and audit of designated activities in e areas of:
. nuclear power plant operations
- b. nuclear engineering CALVERT CLIFFS - UNIT 1 6-11 Amendment No. F&t-
6.0 ADMINISTRATIVE CONTR0LS
- c. chemistry and radiochemistry
. metallurgy and non-destructive examination j
- e. strumentation and control
- f. rad logical safety
- g. mechan al and electrical engineering
- h. quality a urance practices l
COMPOSITION ) 1 6.5.4.2 The OSSRC shall b composed of at le t seven members, including the Chaiman. Members of t OSSRC may be f om the Nuclear Energy Division or other BG&E organization o from organiza ions external to BG&E and shall collectively have expertise in 11 of the reas of 6.5.4.1.- 00ALIFICATIONS 6.5.4.3 The Chairman, members and ternate members of the OSSRC shall be appointed in writing by the Vice P e ident - Nuclear Energy and shall have an academic u ;' ee in engineerin or physical science, or the equivalent, and in addition '11 have a mi mum o five years technical experience in one or more area; ,.ven in 6.5 4.1. No ore than two alternates shall participate as voting member in OSSRC a ivities at any one time. CONSULTANTS 6.5.4.4 Consultants s 11 be utilized as dete ined by the OSSRC Chairman to provide expert ady ce to the OSSRC. MEETING FRE0VENCY 6.5.4.5 The' O C shall meet at least once per six m ths. OUORUM 6.5.4.6 he quorum of the OSSRC necessary for the perfoma e of the OSSRC review nd audit functions of these Technical Specifications hall consist of mor than half the OSSRC membership or at least four member whichever is g ater. This quorum shall include the Chaiman or his appo ted CALVERT CLIFFS - UNIT 1 6-12 Amendment No. M9-i
i I i 6.0 ADMINISTRATIVE CONTROLS j lh alt rnate and the OSSRC members, including' appointed alternates, meeti g . the equirements of Specification 6.5.4.3. No more than a minority quo shall have line responsibility for operation of the plant. the l j REVIEW 6.5.4.7. The SSRC shall review: '
- a. The fety evaluations for 1) changes to proc ures, equipment or system and 2) tests or experiments com)1ete under the provisi s of 10 CFR 50.59, to verify t1at uch actions did not constitu an unreviewed safety question.
- b. Proposed ch ges to procedures, equi t or systems which involve an un eviewed safety questio as defined in 10 CFR 50.59. ;
- c. Proposed tests o experiments whi involve an unreviewed safety '
question as defin in 10 CFR 50 9.
- d. Proposed changes in chnical pecifications or this Operating License.
- e. Violation of codes, regu tions, orders, Technical Specifications, licens re irements, or of internal procedures or instructions havin nucle r safety significance.
- f. Significant operat g abnormal ies or deviations from normal and expected perform ce of plant eq ipment that affect nuclear 4 safety. i
- g. All REPORTAB EVPdTS.
- h. All recog zed indications of an unant ipated deficiency in some aspect o design or operation of safety elated structures, system , or components.
,i. Rep ts and meetings minutes of the POSRC.
AUDITS 6.5.4. 1 Audits of facility activities shall be performed der the cogni nce of the OSSRC. These audits shall encompass: CALVERT CLIFFS - UNIT 1 6-13 Amendment No. 169
6.0 ADMINISTRATIVE CONTROLS 31 a. The conformance of facility operation to provisions contain d - within the Technical Specifications and applicable licens ! conditions at least once per 12 months.
- b. e performance, training and qualification of the e re l fa lity staff at least once per 12 months.
- c. The r uits of actions taken to correct deficien es occurring in facilit equipment, structures, systems or met d of operation that aff t nuclear safety at least once per months.
- d. The perfom ce of activities required by e Quality Assurance .
Program to me t the criteria of 10 CFR P t 50, Appendix B at ! least once per 4 months.
- e. The Safeguards Co ingency Plan and plementing procedures at !
least once per 12 ths in accord cewith10CFR73.40(d). -
- f. Any other area of faci ty oper ion considered appropriate by the OSSRC or the Vice P siden - Nuclear Energy.
- g. The Facility Fire Protectio Program and implementing procedures at least once per 24 month .
- h. An independent fire pro ction a d loss prevention program i inspection and audit s all be per med at least once per 12
? months utilizing eit r qualified fsite licensee personnel or an outside fire pro etica firm.
- 1. An inspection an audit of the fire pro ection and loss prevention prog m shall be perfomed by qualified outside fire consultant at ast once per 36 months. l.
- j. The radiolo cal environmental monitoring pr ram and the results thereof at east once per 12 months.
- k. The OFF TE DOSE CALCULATION MANUAL and implement ng procedures at lea once per 24 months.
- 1. The OCESS CONTROL PROGRAM and implementing procedur s for ,
pro essing and packaging of radioactive wastes at leas once per i 24 m mths. l
- m. he perfomance of activities required by The Quality Assu nce Program for effluent and environmental monitoring at least o ce per 12 months.
-m e ,g CALVERT CLIFFS - UNIT 1 6-14 Amendment No. M 9-l 1
6.0 ADMINISTRATIVE CONTROLS
' 6.5. 8.2 Review of facility activities shall be performed under the, >
JJ . cogni nce of the OSSRC. These reviews shall encompass:
- a. The Facility Emergency Plan and implementing pt-ocedures at least ceper12monthsinaccordancewith10CFR50.54(t)
AUTHORITY 6.5.4.9 The OSSRC all report to and advise the Vice P sident - Nuclear i Energy on those areas f responsibility specified in Se tions 6.5.4.7 and 6.5.4.8. RfCORDS 6.5.4.10 Records of OSSRC acti ties shall' b prepared, approved and distributed as indicated below:
- a. Minutes of each OSSRC meet gs 11 be prepared, approved and forwarded to the Vice Prest - Nuclear Energy within 14 days following each meeting.
- b. Reports of reviews encomp4 sed b Section 6.5.4.7 above, shall be prepared, approved and f arded t the Vice President - Nuclear Energy within 14 days f lowing com etion of the review.
- c. Audit reports encomp sed by Section 6. 4.8 above, shall be forwarded to the Vi e President - Nuclea Energy and to the ,
management positi7..s responsible for the a eas audited within 30 days after compi tion of the audit. 6.6 REPORTABLE EVENT TION ! 6.6.1 The followin actions shall be taken for REPORTABLE ENTS: ;
.a. The Co ission shall be notified and a report submi ed pursuant to th requirements of 10 CFR 50.73, and
- b. Ea REPORTABLE EVENT shall be reviewed by the POSRC an the 3 r ults of this review shall be submitted to the OSSRC a the ;
ice President - Nuclear Energy. l i h,# I CALVERT CLIFFS - UNIT 1 6-15 Amendment No. E !
?
6.0 ADMINISTRATIVE CONTROLS Gh
~
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b b. The soon NRC as possible Operationsand in all Center cases shall within be onenotified hone as hour. by telep[The Vice l h h President - Nuclear Energy and the 96SRG shall b notified within
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@ / e.
24 hours. The re;:-t
' A.x A Safety ths!!Limit Violation by Report
- shall be " 'prepared--
- - + " - ' g '- 'e O,3 5e rnisd th: .O';Z. !
f (1) ;;11 etle c'rsst=::: pr::: din; t;: v';l; tin, (2) -ffuts ! j af +h- vi:::t': upen faci'ity :: ;; =t:, :y:tc. Or :t ::ter=, j
=d (5) =rr=tive actien tihen te pr:v=t recurr=ce.
l ' 1 a/ Jt'. TM Sif:ty '.i:!t Yiel:ti= ";,~. i. J. ell b:* submitted to the O@ Comission, the within 14 days of he violation. and the Vice President - Nuclear Energy offsite re + e w t%dion i 4 _ 6.6 PROCEDURES ;
- 4
'j 6.8.1 Written procedures shall be established, implemented and maintained ;
covering the activities referenced below: ! i
- a. The applicable procedures recommended in Appendix A of Regulatory uide 1.33. Revision 2, February 1978. i
$ JnSed S l g a 9: fueling per;ti= .
- c. Serveill=:: =d te:t activities of sefety related equip =r.t. !
,g d. Security Pl= i=p1==ttti=.
- - ser:ency m = !=p, ==t:ti=. i
@ f. Fir: ".etectier. ."re;rn impl:nnt: tic $
@ dr. T'h e amount of overtime worked by plant staff members perfoming safety related functions must be limited in accordance with the l
NRC Policy Statement on Working Hours (Generic Letter No. 82-12). ; i s d- h X POS"O is ;nly r ;; ired te revin Pre Pretect!= pr=edere and chang = th;r:t: which :ffect =10:r =fety. ! CALVERT CLIFFS - UNIT 1 6-16 Amendment No.149-
. ;N.! ,
t i i l
.l In.se<f 5 (see p. 6-u) .
i i i
- b. The emergency o lement the h
. requirements ofREbratind 737 and brocedures required to im[upplement 1 NUREG-0737, !
as stated in Generic letter 82-33; and > Q c. All programs specified in Specification 6.5. - l 1
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6.0 ADMINISTRATIVE CONTROLS j 3 h. PROCESS CONTROL PROGRAM implementation. h 1. 0FFSITE D0SE CALCULATION MANUAL implementation. h 6.8.2 The nt General Manager may designate specific procedures r classes of pro dures in writing to be reviewed by the Procedure eview Committee or by alified Reviewers in lieu of review by the P C. Review i' by the Procedure R iew Comittee shall be in accordance wit paragraph 6.5.2. Review by Qu ified Reviewers shall be in accordan with paragraph l 6.5.3. l l 6.8.3 Procedures listed i .8.1 shall be approved the Plant General ! s, Superintendents or General Supervisors l Managerorby)cognizantManathat (or Directors report dir tlyprior to a toManag implementation as ! specified by administrative requ ements. T approval authority for specific procedures or classes of ' ocedur shall be designated in writing by the Plant General Manager and sia a different individual from the ! Qualified Reviewer. l l 6.8.4 Each procedure of 6.8.1 a e and ch ges thereto shall be reviewed periodically as set forth in a nistrative p cedures. 6.8.5 Temporary changes . procedures of 6.8.1 a ve may be made provided:
- a. The intent the original procedure is not 1tered.
- b. The el ge is approved by two members of the pla t management ,
~ sta , at least one of whom holds a Senior Reacto perator's ense on the unit affected. ;
. The change is documented, reviewed by the POSRC, the Pro dure .
Review Comittee, or by a Qualified Reviewer and approved the j designat_ed agphthority witgof _implementat n. _ _ l
?3 _d6 2b'" P& 9(s ee p. N6 .29), i o + 19~ }
@ J 6.f4 REPORTING RE0VIREMENTS "O'JT!"E REPORTS h 6.9.1" In :dditien te the app!!ceble repertin; repf re ?"t: Of Titl: 10, c:d: :f r:d:ra ":;;1:tten:, the fell:=f ng r:;:rt: ths!! be ::britt:d to th: Direct:r f the Re;f; .;l Offi:c :f Inspectica 5"d Eafercement :100:
C. M 5e a^ted. The follo win g << po<fs shall be sub-tHed in l a c c o. 6-17) .
.MN _ - x
~
- c. Licensee initiated changes to the ODCM:
l ! 1
/ 1. Shall be documented and records of reviews performed shall be retained.
This documentation shallcontain: l i l { (a) Sufficient information to support the change (s) together with the i appropriate analyses or evaluationsjustifymg the change (sk f (b) A determination that the chang s maintain the levels of radioactive f efDuent control rec uired by 10 20.1302,40 CFR 190,10 CFR l 50.36a, and 10 CFR 50, A ndixI, and not adversel impact the ; accuracy or reliability of e uent, dose, or setpoint ca culations; / j
- 2. Shall become effective after review and acceptance by the onsite review .
function and the approval of the plant manager; and
\ !
l 3. Shall
/ !
entire be submitted ODCM to or as.part of the NRC in the concurrent withform of a coNoactivelete,le the Ra ble cobof ffluen elease I the Report for the period of the report in which any change in the ODCM was k
; made. Each change shall be identified by marlangs in the margin of the i i
' affected pages, clearly indicating the area of the page that was changed, and _j shall indicate the date (i.e., month and year) the change was implemented. / ! I i I i i
}
J u e d to (.Se' P. 6- t ') 6.5.4 Technical Specifications (TS) Bases Control Program I l This program provides a means for processing changes to the TS Bases.
- a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
f i. A change in the TS incorporated in the license; or k li. A change to the UFSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.
\ c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
/ i
! l 4
- d. Proposed changes to the TS incorporated in the license or proposed changes to the UFSAR or Bases that involves an unreviewed safety question shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(c).
s l l l i i l 1
l0(Seef.6,-l7) . x - 6.5.5 Radioactive Effluent Controls Program l This pr,ogram conforms to 10 CFR 50.36a for the control of radioac8ve effluents and for I [ maintammg the doses to members of the public from radioactive effluects as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shallinclude the following elements:
- a. Limitations on the functional capability of radioactive liquid and gaseous f monitoring instrumentation including surveillance tests and setpoint i determination in accordance with the methodology in the ODCM;
- b. Limitations on the concentrations of radioactive material released in li effluents to unrestricted areas, conforming to 10 CFR 20, Table Appendix B,qu.d II, Column 2;
/
- c. 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; i d.
/
Limitations on the annual and guarterly 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 10 CFR 50, Appendix I;
- e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar j year in accordance with the methodology and parameters in the ODCM at ,
f least every 31 days; {
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these j systems are used to reduce releases of radioactivity when the projected doses (
in a period of 31 days would exceed 2% of the guidelines for t be annual dose i ) or dose commitment, conforming to 10 CFR 50, Appendix I; i f g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary c0aforming to the dose associated with 10 CFR 20, Appendix B, Table II, Column 1; j 1
- h. 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 10 CFR 50, Appendix 1, l
- i. Limitations on the annual and ublic from /
iodine-131, iodine-133, tritium, quarterly and all doses radionuclides in to a member particulate orm of the with I halflives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and
- j. Limitations on the annual dose or dose commitment to any member of the i public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. j !
/
N- l 4
I t i 6.0 ADMINISTRATIVE CONTROLS - '3 STA KUP REPORT _ 6.9.1.1 g 041 shall be s sumary itted report following of p(lant
- 1) receipt STARTUP of an and operating power escalation license tes .
h (2) amendmen (3) installatio o the license involving a planned increase in f fuel that has a different design or has en er level, , I manufactured by a fferent fuel supplier, and (4) modifi ions that may have significantly a red the nuclear, thennal, or hyd ulic performance , of the plant. 6.9.1.2 The startup report all address each the tests identified in i 06/ the FSAR and shall include a c cription of- measured values of the ; operating conditions or characte tics ob ned during the test program ' and a comparison of these values w d gn predictions and ;
~
specifications. Any corrective acti that were required to obtain satisfactory operation shall also d ribed. Any additional specific details required in license con ions b ed on other comitments shall be ; included in this report. 06g 6.9.1.3 Startup report all be submitted wi in (1) 90 days following - completion of the st up test program, (2) 90 d s following resumption or commencement of co rcial power operation, or (3 months following i initial critica y, whichever is earliest. If the rtup report does not cover all th events (i.e., initial criticality, com tion of startup test progr , and resumption or commencement of commercia ower , operati , supplementary reports shall be submitted at leas every three i E.8 mont until all three events have been completed. ;
- 8. 6. I occ updional hdi,dios Exposwe Reped '
7 2I _ _
, j
/5~.0.1.1 ^nne:l*reporty ::'ferin; the acti"itier Of the unit :: d:ccribed ibelce for the previter :: lend:r ye:r shall be submitted prior to March 31- of each year. Th: intthl repert :h:11 he s"httted pri:r t: ":rch I ef the l year felbr in; i-iti:1 crith:lity, -
h S.9.1.5 k pert: requir:d ;r. :n ;;; :1 S M th:11 h : h d ; h K A tabulation on an annual basis of the number of station, utility, and otgr personnel (including contractors) receiving exposures gre:t., th;a 100 mrem /yr and their 3ssociated man rem exposure according to work and job functions X(e.g., reactor operations and surveillance, inservice inspection, routine 4k bdepe.de-f 3 Pe"Y fxI AW 9 ebSf*I5efa.s l O h
- f ceapa tiona l Jo.se feeA(ISF.s& wiII.se calvui CI4fs bepa<afel tepe,de/
y, 1/ A single submittal may be made for : multiple unit :t: tion. submittal should combine those sections that are common to aMT% UnTts, at th st:tien #
-_ ___ - f ase-f l 1;;t h M This tabulation supplements the requirements of 10 CFR 20.4@.
__ _ ~ ~. - .Jp hI; CALVERT CLIFFS - UNIT 1 6-18 Amendment No. M 9-
themimin a.s ce,,f
. 6.0 ADMINISTRATIVE CONTROLS f[(nse.d jee p. 6.i I1. fd,3; meter 7h
' maintenance, special maint ance (describe mainten ce), waste processing, and refueling). The dose assignment to v ious duty '
, or functions f81= hd;:may be estimates based on pockettotalling dosimeter T p than 20%
@ r ur= nt:. Small exposures of the individua.1 total dose need not be accounted for. In the ler_
aggregate, at least 80% of the total whole body dose received from exter 1 sources shaN be assigned to speciAff' r wr Agri c If b' T @ functions.ynj;fp L Shools{ Cejo.s eme tereeleis pe<so, Ve, (,see p. w e
= ct:= g:=r:ter t'Ae in:ervice
. b. The e phe ===t c!=pu =: per
_ _ , ,t, .i..,__ , ederf f:=d e uag the r: pert p;r ei d (r:ference
~
O~ b.6. 4 Pre.sivAicA Po vs .Sdef t v alves A@*'t y(. %ocumentat#9 p( all failures and challenges to the pressurizer
- PORVs or safety valves.
JnSed uce f E 12 I7) [A repwf .she il be .submiHel P'I*' 'f" %5h I O e% i f { ON1,HLY OPERATIN < ,v 6.9.1.6 Routine reports of operating statistics and shutdown experience shal? be submitted on a monthly basis to the Dirrter, Off'= Of.I=pect'en and Enfer=xat, L'.5. N=1=r R:;;ht:ry Cc-inten, u=hin;t , 0.C. 20555, ATIN. D===t Centrol Ock, "ith : =py to th: Regien:!
!.dminhtr:t:r :nd t; th: "RC Reid=t !=p=ter, no later than the 15th of each month following the calendar month covered by the report.
..~% b.b.2
"' @ ^ ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- The Annsa l 5.0.1.7 Routine
- Radiological Environmental Op: rating Report)( covering the operation of the unit during the previous calendar year shall be submitted prior to Hay 1 of each year. juo,,; jo,;n, p,og ,,m i The frn =1 R:d h hgical Env 1 Ope- thg Leport)( shall include sumaries, interpretations, nd an analys}s of trends of the resu.lts of the 73 Radiological Environmental
.ur=ilh=e ::t444 .[ir ties ..m=t:
for the report"*/eriody , M 3 nchdMg i : =mp:rf = "ith pr=::r:ti=:1 :ted!=, with creratten:1 contr u = :ppr:prhte, =d wit; previe= =v'rennat:1 =fvei'hnee repert:, =d = nen:nnt Of th: Oberved impet: Of the punt oper:t hn en thy =dferm=t. Th r:p:rt: :h:ll ch : i = h d: the result: ef land uw ceme=: required by Specification 3.2.2. h The Arn=1 R:diele 1 1 Envirenant:1 Operating Leport)'shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and {igures in the ODCH, as well-as (The inaie iai p'c vs ole d s ha ll be t o s.s b ien f w i i Ir -fhe ohjeef:<es _ out final in the off.sile Ase ca(cv la flon Ma n'a } (OD C N } , a no/ in 10cFM sos APpendit I, .Secflon s .2 B 3, .22. 2.3, w,,( y , c O@
- A single submittal may be madey for C a l v ee d C // f /.s , Tb-
.submif tal .s hould combin e th os e .scefIen.s fhat a ee.
ccm mon to bett, uni t.s , CALVERT' CLIFFS - UNIT 1 6-19 Amendment No.149- .
, 9 g g
& 1 :.' * %i.
. N :. -
6.'O ADMINISTRATIVE CONTROLS N.' summarized and tabulated results of these analyses and measurements in the fonnat of the table in the Radiological Assessment Branch Technical , Position, Revision 1. November 1979. In the event that some individual . results are not available for inclusion with the report, ths report shall_ - be submitted noting and explaining the reasons for the missing results. i The missing data shall be submitted as soon as possible.in a. supplementary > report. h T radip eports shall also include the following: a sumary description of the al environmental monitoring program; at least two legible all sampling locations keyed to a table giving dist - i' maps cover s and directions the central point between the two contai ! buildings; the resu " licensee participation in th erlaboratory O-n. Comparison Program, requ' by Specification 3.1 , discussion of all ! deviations from the sampling dule of Ta .12-1; and dist.pssion of ' all analyses in which the LLD req able 4.12-1 was not Ichievable. The Annual Radiological E ental Ope ng Report will include the identification of th se of unavailability samples (if any), and will describe the lo ns used for the replacement s es. The report will also inclu y pennanent changes in the sample loca which could ' ap)ea the monitoring program. It will include a revis ure(s) and ta e for the ODCM reflecting the new location (s). _a h RADIO 4.CTIVE + EFF6UENT+ + REllASE REkORT" S.9.1.8 Re e Radioactive Effluent Release Reportf. covering the operation of the unit durin; the pr=S : 5 = nth: : Oper:tha sg1 be h submitted The 2:dictetiv: ig.ing d:g:jf,teg{:gg:ggry Effhat n
"cle= Seporty shall include g' 1,$e{no ms 3:Q:gr.
a sun ary fo~de< of the tkg g,,i,$
., 12 ment 4
quantities of radio c. ive liquid and gaseous effluents and solid waste l; released fro'n the 'n cutl hed in ":ge h tery Guide 1.21, ": = uring, Cv:h: ting, 2nd R:perting Radictethity in S: lid U;;t;; ;r.d "ch== cf ( R:dincth: ".:t:rSh b Li:;;id nd Cnsu: Effh =t: fr a. Light U;ter-C:: led Nach:r "=:r Phnt:," ":vbica 1, J :: 1974, with d:t: :r trized-ca : ;uerterly bath f:1hwing the femat of Apar .dh S thereef. Th e -on a te< s'a i provided .shall be consiski will, the objec tive c nilined in'ihe OPCM en o! Proc eu Ce*,foi Preyeo m a ,o/ 1., c o s fe m o or e e wfk "l 10 c f M .50, App eerdir E, .See ficos 5. B. l . 7 [I>,se<f10_CF 13 (seeAp _Sg.. 6- 3 3pg *Q h X On: np th:11 :: :r -':th= near the tralude the nr: di6+ stm+4ane. SITS 00' ==Y, ; n::nd :h:1-1 A single subm tal may be made for Calvert Cliffs, since the Radwaste Systems are ommon to both units. h N 17. lic" Of :rb-is k: with the Sci-Annual "ap;rt, S;", Sr"- - j an=14: ruu!ts =>y ha erb-ited b : rupH:- :- tery reprt chb 10^ Ey: after January 1 ead July 1 Of an year. CA RT CLIFFS - UNIT 1 6-20 Amendment No. M9-The e y e -+ sLII al.w i< c ide c harg es to pae opctM in seco<,larce. with spe c i fte a f s o ., t . s. l e ,
6.0 ADMINISTRATIVE CONTROLS h Th Radioactive Effluent Release Report to be submitted within 60 daya ... aft January 1 of each year shall include an annual summary of urly I meteo logical data collected over the previous year. This ann 1 summary i may be ither in the form of an hour-by-hour listing on magne c tape of i
-wind s d, wind direction, atmospheric stability, and prect tation (if l measured or in the fom of joint frequen wind speed, j wind direc ton, and atmospheric stability.py Thisdistributions same rep shall include an sessment of the radiation doses due to the dioactive liquid !
and gaseous fluents released from the unit or station uring the previous f calendar year. The assessment of radiation doses shal be perfomed in ; accordance wit the methodology and parameters in th 0FFSITE D0SE ! CALCULATION MAN L(0DCM). l The Radioactive E luent Release Report to be su tted 60' days after January 1 of each ar shall also include an ass ssment of radiation doses' 1 to the likely most posed MEMBER OF THE PUBLI from reactor releases and ' other nearby uranium uel cycle sources, inc1 ing doses from primary effluent pathways and irect radiation, for e previous calendar year to show confomance with CFR Part 190, Envi onmental Radiation Protection , Standards for Nuclear Po er Operation. A eptable methods for calculating , the dose contribution fro liquid and ga . ous effluents are given in Regulatory Guide 1.109, Re 1, October 977, and NUREG-0133, " Preparation of Radiological Effluent Te nical Spe ifications for Nuclear Power Plants." The Radioactive Effluent Releas R orts shall include the following l
; infomation for each class of so waste (as defined by 10 CFR Part 61) ;
shipped offsite during the repor eriod:
- a. Container volume, l
- b. Total curie quanti .y (specif whether detemined' by measurement !
or estimate). ! l
- c. Principal radi nuclides (specify hether determined by measurement o estimate),
- d. Source of ste and processing emplo d (e.g., dewatered spent resin, co acted dry waste, evaporato bottoms),
- e. Solidif cation agent or absorcent (e.g., cement).
The Radioactiv Effluent Release Reports shall inclu a list and description o unplanned releases from the site to UN STRICTED ARE/eS of radioactive aterials in gaseous and liquid effluents m e during the reporting riod. n lieu of submission with the first half year Radioacti e Effluent Release Report, this sumary of required meteorological data may be
; retained.on site in a file that shall be provided to the NRC\upon s request.
CALVERT CLIFFS - UNIT 1 6-21 Amendment No. 469
1 6.0 -ADMINISTRATIVE CONTROLS ! Te c an during thective r Effluent Release Reports shall in_ciTRULPROGRAM(PCPfe eriod to the PROCESS , and to
@ the 0FFSITE DOSE CALCU locations for dose cale U
sie
, as well as a listing of new the annual land use census h pursuant to on 3.12.2. -
set}\
;7 Y
- 4a.1.0- CORE OPERATING LIMITS REPORT (fot.
a (5"P69 N y -
) s.t.5a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and i
shall be documented in the COLR for the following: : 2.2.1 - 3.1.1.1 i 3.1.1.2 ) 3.1.1.4 3.1.3.1 3.1.3.6 3.2.1 3.2.2.1 3.2.3 3.2.5 3.9.1 %
- b. The analytical methods used to detemine the core operating limits shall be those previously reviewed and approved by the NRC; specifically, those described in the following documents: ,
(1) CENPD-199-P, Latest Approved Revision, "C-E Setpoint , Methodology: C-E Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems," January 1986 (2) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 1: C-E Calculated Local Power Density and Themal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II," December 1979
~
(3) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 2: Combination of System Parameter Uncertainties in Themal Margin Analyses for Calvert Cliffs Units 1 and 2," January 1980 (4) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 3: C-E Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for Calvert Cliffs Units 1 and 2," March 1980 (5) CEN-191(B)-P. "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2 " December 1981 CALVERT CLIFFS - UNIT 1 6-22 Amendment No. 146
- 6.0 ADMINISTRATIVE CONTROLS i s
i
'(6) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr.
(BG&E), dated June 24,1982, Unit 1 Cycle 6 License Approval (Amendment No. 71 to DPR-53 and SER) (7) CEN-348(B)-P, " Extended Statistical Combination of Uncertainties," January 1987 (8) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan l (BG&E), dated October 21, 1987, Docket Nos. 50-317 and 50-318. " Safety Evaluation of Topical Report CEN-348(B)-P, Extended Statistical Combination of. Uncertainties" (9) CENPD-161-P-A, " TORC Code, A Computer Code for Detemining , the Themal Margin of a Reactor Core," April 1986 (10) CENPD-162-P-A, Latest Approved Revision, " Critical Heat Flux ' Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 1. Uniform Axial Power Distribution" (11) CENPD-207-P-A, Latest Approved Revision, " Critical Heat Flux , Correlation of C-E Fuel Assemblies with Standard Spacer ; Grids Part 2, Non-Unifom Axial Power Distribution" /q , (12) CENPD-206-P-A, Latest Approved Revision, " TORC Code, : Verification and Simplified Modeling Methods" ;
)
(13) CENPD-225-P-A, Latest Approved Revision, " Fuel and Poison Rod Bowing" (14) CENPD-266-P-A, Latest Approved Revision, "The ROCS and DIT Computer Code for Nuclear Design" (15) CENPD-275-P-A, Latest Approved Revision, "C-E Methodology for Core Designs Containing Gadolinia - Urania Burnable Absorbers" (16) CENPD-382-P-A, Latest Approved Revision, "C-E Methodology for Core Designs Containing Erbium Burnable Absorbers" (17) CENPD-139-P-A, Latest Approved Revision, "C-E Fuel Evaluation Model Topical Report" (18) CEN-161-(B)-P-A, Latest Approved Revision, " Improvements to Fuel Evaluation Model" (19) CEN-161-(B)-P, Supplement 1-P, " Improvements to Fuel Evaluation Model," April 1989 CALVERT CLIFFS - UNIT 1 6-23 Amendment No. IM-
6.0 ADMINISTRATIVE CONTROLS (20) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated February 4, 1987, Docket Nos. 50-317 and 50-318. " Safety Evaluation of Topical Report CEN-161-{B)-P. Supplement 1-P, Improvements to Fuel Evaluation Model" (23) CEN-372-P-A, Latest Approved Revision " Fuel Rod Maximum Allowable Gas Pressure" (22) Letter from Mr. A. E. Scherer (CE) to Mr. J. R. Miller (NRC),datedDecember 15,1981, LD-81-095 Enclosure 1-P, "C-E ECCS Evaluation Model Flow Blockage Analysis" (23) CENPD-132, Supplement 3-P-A, Latest Approved Revision,
" Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and 11 Designed NSSS" (24) CENPD-133, Supplement 5 "CEFLASH-4A, a FORTRAN 77 Digital Computer Program for Reactor Blowdown Analysis," June 1985 f
(25) CENPD-134, Supplement 2, COMPERC-II, a Program for b Emergency Refill-Reflood of the Core," June 1985 (26) Letter from Mr. D. M. Crutchfield (NRC) to Mr. A. E. Scherer (CE), dated July 31,1986, " Safety Evaluation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related Licensing Topical Reports" (27) CENPD-135. Supplement 5-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 (28) Letter from Mr. R. L. Baer (NRC) to Mr. A. E. Scherer (CE), dated September 6,1978, " Evaluation of Topical Report CENPD-135, Supplement 5" (29) CENPD-137, Supplement 1-P, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977 (30)CENPD-133, Supplement 3-P,"CEFLASH-4AS,AComputerProgram for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident," January 1977 (31) Letter from Mr. K. Kniel (NRC) to Mr. A. E. Scherer (CE), ' dated September 27,1977, " Evaluation of Topical Reports CENPD-133,, Supplement 3-P and CENPD-137, Supplement 1-P" (32) CENPD-138, Supplement 2-P, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977 s .- CALVERT CLIFFS - UNIT 1 6-24 Amendmer.t No. 4%-
6.0 ADMINISTRATIVE CONTROLS - (33) Letter from Mr. C. Aniel (NRC) to Mr. A. E. Scherer, dat~ed April 10,1978, " Evaluation of Topical Report CENPD-138, Supplement 2-P" (34) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. J. R. Miller (NRC) dated February 22,1985, "Calvert Cliffs Nuclear Power Plant Unit 1; Docket No. 50-317, ; Amendment to Operating License DPR-53, Eighth Cycle License Application" (35) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr. ! (BG&E), dated May 20,1985, " Safety Evaluation Report Approving Unit 1 Cycle 8 License Application" t (36) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. R. A. Clark (NRC), dated September 22, 1980, " Amendment to Operating License No. 50-317, Fifth Cycle License Application" (37) Letter from Mr. R. A. Clark (NRC) to Mr. A. E. Lundvall, Jr. (BG&E), dated December 12, 1980, " Safety Evaluation Report Approving Unit 1, Cycle 5 License Application" )( ' (38) Letter from Mr. J. A. Tiernan (BG&E) to Mr. A. C. Thadani (NRC), dated October 1,1986, "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2, Docket Nos. 50-317 & 50-318 Request for Amendment" (39) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated July 7,1987, Docket Nos. 50-317 and 50-318, i Approval of Amendments 127 (Unit 1) and 109 (Unit 2) (40) CENPD-188-A, Latest Approved Revision, "HERMITE: A Multi- l Dimensional Space-Time Kinetics Code for PWR Transients" (41) The Full Core Power Distribution Monitoring System ; referenced in Specifications 3.1.3.1, 3.2.2.1, 3.2.3, and the BASES is described in the following documents: (a) CENPD-153-P, Latest Approved Revision, " Evaluation of ; Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed Incore Detector System" t (b) CEN-199(B)-P,"BASSS,UseoftheIncoreDetectorSystem ! to Monitor the DNB-LC0 on Calvert Cliffs Unit I and ! Unit 2," November 1979 i m/ CALVERT CLIFFS - UNIT 1 6-25 Amendment No. M6-i l
m . . _ __ __ _
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6.0' ADMINISTRATIVE CONTROLS l (c) letter from Mr. G. C. Creel (BG&E) to NRC Document - : Control Desk, dated February 7,1989, "Calvert Cliffs I Nuclear Power Plant Unit No. 2; Docket 50-318 Request for Amendment, Unit 2 Ninth Cycle License Application" (d) Letter from Mr.- S. A. McNeil, Jr. (NRC) to Mr. G. C. Creel (BG&E), dated January 10, 1990. " Safety ; Evaluation Report Approving Unit 2 Cycle 9 License s/
/
Application" ,
- c. The core limits operating (limits shall be detemined such that alle.g., !
a plicable t..emal hydraulic limits, Emergency Core Cooling Systems (ECCS) . limits, nuclear . limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis ari met.
- d. The COLR, including any mid-cycle revisions or supplements, shall '
be provided up_on__ issuance for each reload cycle to the NRC.
?f!!!jh? f I'~l4 -
SPEMAL REPORTS 6.9.2 cial reports shall be submitted to the Regional Admi istrator of the NRC Re. nal Office within the time period specified for ach report. i These reports hall be submitted covering the activities i ntified below ; pursuant to the quirements of the applicable referenc pecification: {
- a. ECCS Actua on, Specifications 3.5.2 and 3 .3.
- b. Inoperable Sei ic Monitoring Instrum ation, Specification 3. .3.
- c. Inoperable Meteorolog' al Instr entation, Specification 3.3.3.4.
- d. Seismic Event Analysis, S ification 4.3.3.3.2. !
- e. Core Barrel Movement, ecifi tion 3.4.11. l
- f. Fire Detection Ins umentation, S cification 3.3.3.7.
- g. Fire Suppressi Systems, Specificatio 3.7.11.1, 3.7.11.2, )
3.7.11.3, 3.7 1.4, and 3.7.11.5. ! i
- h. Penetrati Fire Barriers, Specification 3. 2. l
- i. Steam enerator Tube Inspection Results, Specif) tion 4.4.5.5.a and .
J. pecific Activity of Primary Coolant, Specification 3. 8. s
-h a CALVERT CLIFFS - UNIT 1 6-26 Amendment No. 446-2-
g9 4
. - - , .- . - . - . - - -nn- -- ~ -. -
6.0 ADMINISTRATIVE CONTROLS
) k. Containment Structural Integrity, Specification .4.6.1.6. -
. Radioactive Effluents - Calculated Dose and Total Dose, ;
Specifications 3.11.1.2, 3.11.2.2, 3.11.2.3, and 3.11.4. l
- m. Tadioactive Effluents - Liquid Radwaste Gaseous Radw te and ntilation Exhaust Treatment Systems Discharges. -
Sp cifications 3.11.1.3 and 3.11.2.4. ,
- n. Radi ' ogical Environmental Monitoring Program, Speci cation 3.12.1.
l
- o. Radiatto Monitoring Instrumentation, Speci cation 3.3.3.1 ;
(Table 3. 6).
- p. Overpressure rotection Systems, Specif ation 3.4.9.3. ,
- q. Hydrogen Analyz rs, Specification 3. 5.1.
- r. Post-Accident Inst umentation, Sp ification 3.3.3.6.
t 6.10 RECORD RETENTION 6.10.1 The following records shall retained for at least five years. . l
.)
- a. Records and logs of faci ity peration covering time interval at !
each power level. (
- b. Records and logs of incipal ma tenance activities, !
inspections, repair and replacemen of principal items of ( equipment related o nuclear safety.
- c. All REPORTABLE ENTS. j
- d. Records of s veillance activities, inspe ions and calibrations required by hese Technical Specifications.
- e. Records reactor tests and experiments. y
- f. Record of changes made to Operating Procedures. !
- g. Rec ds of radioactive shipments. f I
- h. cords of sealed source and fission detector leak te is and !
esults. 1 Records of annual physical inventory of all sealed source ! material of record , CALVERT CLIFFS - UNIT 1 6-27 Amendment No. 166 M j
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-- 6.0 ' ADMINISTRATIVE CONTROLS !
- 6. 0.2 The following records shall be retained for the duration of i Fac ity Operating License Records and drawing changes reflecting facility desig }
modifications made to systems and equipment describ in the inal Safety Analysis Report.
- b. Re ords of new and irradiated fuel inventory, f 1 transfers and ass bly burnup histories. l
- c. Recor of facility radiation and contamina on surveys. ;
- d. Records radiation exposure for all ind iduals entering i radiation ontrol areas. ;
i
- e. Records of g eous and liquid radioac ve material released to the environs, j
- f. Records of trans nt or operatio 1 cycles for'those facility components identi ed in Table .7-1.
- g. Records of training nd qual ication for current members _ of the !
plant staff. !
- h. Records of in-service i ections performed pursuant to these Technical Specification . j
- i. Records of Quality A uranc activities identified in the NRC j approved QA Manual lifeti records. l
- j. Records of revie performed f changes made to procedures or !
equipment or te ews of tests a experiments pursuant to ; 10 CFR 50.59. ! I
- k. Records of etings of the POSRC, e Procedure Review Consnittee, l and the OS C. t
- 1. Records f the service lives of all sa ety related snubbers !
includ g the date at which the service ife commences and assoc ated installation and maintenance cords. , t 6.11 RADIAI ON PROTECTION PROGRAM Procedur for personnel radiation protection shall be pr ared consistent I with t requirements of 10 CFR Part 20 and shall be appro d, maintained
'and ered to for all operations involving personnel radiat on exposure. j i
}
r CALVERT CLIFFS - UNIT 1 6-28 Amendment No. 486- [ l 1 l 1
T l 6.0 ADMINISTRATIVE CONTROLS 64 2 HIGH RADIATION AREA In lieu of the " control device" or " alarm signal" required b 6.12. 10 CFR 2 . 3(c)(2):
- a. A hig radiation area in which the intensity of diation is greater an 100 mrem /hr but less than 1000 em/hr shall be barricadec d conspicuously posted as a gh Radiation Area and -
entrance ther o shall be controlled issuance of a Special or Radiation Work it and any in dual or group of individuals pennitted to enter ch area , all be provided with a radiation monitoring device whi inuously indicates the radiation dose rate in the area.
- b. A high radiat area in which t tensity of radiation is greater th 000 mrem /hr shall be su et to the provisions of 6.12.1. , above, and in addition locked icades shall be
-) rov ed to prevent unauthorized entry into s areas and the c s shall be maintained by the Supervisor - Rad ion Control aerations and the Operations Shift Supervisor on du under t.1eir separate administrative control.
~ - ~ v ^
- 6. SYSTOi I"TEGRITY /W/ 6"#I"*' Y 60 "" 5 M5 M8 C""Id##'*~/~
The licensee shall implement a program # to reduce leakage from systems
- ) outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:
/
- a. Provisions establishing preventive maintenance and periodic '
) visual inspection requirements, and I
- b. Leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
- 6. IODINE MONITORING ' b#7 The license all implement a arogram* which will ensure the capabilit to accurately de e the air.)orne iodine concentration in vit eas under accident condit o This program shall include th owing:
- a. Training of personn'el,
- b. Procedures for moni , and
- c. Provi or maintenance of sampling and analysis eq nt.
It is acceptable if the licensee maintains details of the program in j plant operation manuals (e.g., chemistry procedures, training instructions, maintenance procedures. ERPIPs), s CALVERT CLIFFS - UNIT 1 6-29 Amendment No. IM E In.s< t 9 fodnd e Cse e p. s-n ) ..
~ , , ,
f' bi" '
.Tn. sed e besp. 4-n)~ '
\
.6.0 ADMINISTRATIVE CONTROLS
, n. , --
9' Op LJ / 6.H POSTAC81DENTSAMik,.ING The licensee shall estabitsh, implement and maintain a program 7 which will ensure the capability to'obtain and analyze reactor coolant, radioactive fodines and-particulates in plant gaseous effluents, and containment . {Iatmosphere samples under accident conditions. The program shall include the following:
- a. Training of personnel, e l I .b. Procedures for sampling and analysis, and-
) Provisions for maintenance of sampling and analysis equipment.
I c.
.w- _
16 PROCESS CONTROL PROGRAM (PCP) 6.16. The PCP shall be approved by the Commission prior to impleme tion. 6.16.2 Lice e initiated changes to the PCP:
- a. Shall b ubmitted to the Commission in e Semiannual Radioactiv ffluent Release Report f the period in which the change (s) was ade. This submittal all contain:
. / 1. An evaluation pporting t premise that the change did not reduce the overa confo nce of the solidified waste product to existin r eria for solid wastes; and
- 2. A reference to t date nd the POSRC meeting number in which thechange(s) s reviewe and found acceptable to the POSRC.
- b. Shall become fective upon revie y the POSRC and approval of the Plant eral Manager.
h 6.17 0FFSIT 7 SE CALCULATION MANUAL (ODCM) 6.17.1 e ODCM shall be approved by the Commission prior o l
'impi ntation.
i
~^
It is acceptable if the licensee maintains ' details of the progr'am in j l j plant operation manuals (e.g., chemistry procedures, training ' instructions, maintenance procedures, ERPIPs). _ ._ - --~_ CALVERT CLIFFS - UNIT 1 6-30 Amendment No. 4 M M \
Repla x ee/ N2awHhInsed nest pae < 7 }
%er a ipecy.s,,}
f.7 6.0 ADMINISTRATIVE CONTROLS . h S ." . 2 Licensee initiated' changes to the ODCM:
- a. all be submitted to the Comission in the ::!:n-d. .
j Ra ctive Effluent Release Report for the per in which the change was made effective. This submitta all contain:
- 1. Sufficien infomation to support rationale for the change. In ation submitted uld consist of a package of those pages of e ODCM to b hanged with each page numbered and provided with hang umber and/or change date together or evaluations justifying the with change app)ropriate (s ; anal
- 2. A detemination at the chang ill not reduce the accuracy or reliabili of dose calculatio or setpoint l determina ns; and
- 3. Docu tation of the fact that the change ha een reviewed f a ' found acceptable by the POSRC. A ,,3fje ,. ,,,, ./ b ,,g g,,,
T 4 Q i b. all become effective upon review by the pGSRC and appro of U/ the{1antCener:1ganager.
- -- ~ _
Tu T e T. e, V. h S.10.1 Licensee initiated major changes to the Radioactive Waste Systems
/(liquid, gaseous and solid) shall be reported to the Commission in the Semi:nne:1 Radioactive Effluent Release Report for the period in which the
[y modification to the waste system is completed. The discussion of each change shall contain: i i i I l
/ a. A description of the equipment, components and processes !
involved. !
- b. Documentation of the fact that the change including the safety ;
. analysis was reviewed and found acceptable by the POSRE. !
onsiieeevi<sfun14,,
- - ~
I f~ l L nw j G (se e p. 6-ac), l .j' CALVERT CLIFFS - UNIT 1 6-31 Amendment No. 106 M
6 ATTACHMENT (5) 1 6 UNIT 2 TECHNICAL SPECIFICATION l MARKED-UP PAGES l l i
TABLE OF CONTENTS
. t
- 3. ADMINISTRATIVE CONTROLS SECTION f8E 6.1 RESPONSIBILITY ...................
6-1 ; 6.2 ORGANIZATION ; 6.2.1 ONSITE & OFFSITE ORGANIZATIONS .......... 6-1 6.2.2 UNIT STAFF .................... 6-1 ! 6.3 FACILITY STAFF QUALIFICATIONS . . . . . . . . . . . 6-5
- 6. TRAINING ....-.................
((} ! 6.5 IEW AND AUDIT i 6.5.1 PLA PERATIONS AND SAFETY REVIEW COMMITTEE (P C) , Functio ..................... 6-5 1 Compositio .............. .... 6-5 : Chairman .. ............ ..... 6-6 i
) Alternates ... ........ ....... 6-6 .
Meeting Frequency . ...... ........ 6-6 i Quorum ....... .............. 6-6 Responsibilities .... ........... 6-6 i Authority . . . . . . . . ........... 6-8 ! Records.. . . . . . . . . . . .......... 6-8 ; 6.5.2 PROCEDURE REVIEW ITTEE < Function .............. ...... 6-8 Composition ............. ..... 6-9 ' Chairman .................... 6-9 Alterna ................. .. 6-9 6-9 Meeti Frequency . . . . . . . . . . . . . . . . i Qu ...................... 6-9 hority . . . . . . . . . . . . . . . . . . . . . 6-10 ; Records . . . . . . . . . . ... . . . . . . . . . . 6-10 ' l 1 s_ .I CALVERT CLIFFS - UNIT 2 XVI Amendment No. 149
' TABLE OF CONTENTS .e g h . ADMINISTRATIVE CONTROLS SECTION 6.5.3 QUALIFIED REVIEWERS 6-10 nction . . ................... 6-10 Au rity . . . . . . . . . . . ... ._. .:... . . .
6-11 Cert ation . . ... . . .-. . . . . . . ... . . 6-11 Records . . ................ .. 6.5.4 0FF-SITE SAFE EVIEW C0lHITTEE (OSSRC
....... 6-11 Function . . . . . ........
6-12. Composition . . . . .
.......... 6-12 Qualifications .... .. .
6-12 Consultants . . . . . . . ............ 6-12 Meeting Frequency . . . .. ..........
...... ............... 6-12 Quorum 6-13 Review .....
. . .. . . 6-13 Audits ... ............
6-15 Authority . .............. ....
". ................... 6-15 Records .
6-15 6.6 PORTABLE EVENT ACTION . . . . . . . . . . . . . .
.) 6 N 6 SAFETY LIMIT VIOLATION . . . . . . . . . . . . . ..
- 6. PROCEDURES . ................... . f-16 5
- 6. 5 PM06RA MS A WD pl A yvA L.S 6.i b REPORTING REQUIREMENTS 4 Qcp.qtafjom/ Mado'aHon Espo.svee Repo<f 6 7 6.f.1 no.. m nurunia ... ...............
6 8
.Otertdp R: pit ... ............... 6 6 r
g .4. + e=, n:p=ts Monthly Operating Report ............. 19
- Annual Radiological Environmental Operating (eg 4.x Report . . . ................... -] .
Srif.. ,ni Radioactive Effluent Release Report .. -2 ) 4.4 3 .......... 6-2, l
'4.65 46.4 Core Operating PeeJ3eri.uae PonLimitsv en)Rep Sn rt lefy vsIve Me po< f
= r: = = .................. 6-2 1 s.e.2 ,
< g,5. 1 oppsite %e G twl< Hon Ma nva) (oMd
, s, s. 2 Po.sts ec tole nt .Sa ~p/my
% Prima <y Co*la d .Sov'ecs Od'Ue Co" f* I"
- c" Y 4.s. 3 c . S. + Teei, n :c-a I .s;peesi.*e* fo'o~4TS) Bases Con %I P?i,j<am j g . 5. .g h 4,'o nfive Efflue~f cont <ols P oy ra m CALVERT CLIFFS - UNIT 2 XVII Amendment No. MS--
s .;g _;
TABLE OF CONTENTS
] ADMINISTRATIVE CONTROLS j SECTION AGE 6.10 RECORD RETENTION ................. 6-27 l 6.11 RADI ON PROTECTION PROGRAM ........... .
6-28 l 6.12 HIGH RADIAT AREA . . . . . . . . . ...... 6-29 l 6.13 SYSTEM INTEGRITY ................. 6-29 l 6.14 IODINE HONITORING . . . ............. 6-29 l 6.15 POSTACCIDENT SAMP NG . . . . . . ........ 6-30 l 6.16 PROCESS C0 OL PROGRAM (PCP) . . . . . . .... 6-30 l
- 1) 6.17 0FFS E DOSE CALCULATION MANUAL (ODCH) ...... 6-31 j' 6.18 MAJOR CNANGES TO RADI0 ACTIVE LIQUID, GASE0US AND SOLID WASTE TREATHENT SYSTEMS . . . . . . . . . -31 l :
1 l l 1 i
) . .. . l i
.Q.. CALVERT CLIFFS - UNIT 2 XVIII Amendment No. & 4 e e q . ' e
l l 1.0 DEFINITIONS j I CONTROLLED LEAKAGE , 1.9 CONTROLLED LEAKAGE shall be the water flow from the reactor coolant pump seals. ; C0RE ALTERATION 1.10 CORE ALTERATION shall be the movement or manipulation of any componeat within the reactor pressure vessel with the. vessel head removed and fuel in the vessel. Suspension of CORE ALTERATI6N shall not preclude
-completion of movement of a component to a safe conservative position.
CORE OPERATING LIMITS REPORT 1.11 The CORE OPERATING LIMITS REPORT is the unit specific document that M provides cycle specific parameter limits for the current reload cycle. q - W These cycle specific parameter limits shall be determined for each reload / cycle in accordance with Specification 6,.4.1.9. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT I-131 1.12 DOSE EQUIVALENT I-131 shall be that concentration of I-131 g-
;) (pCi/ gram) which alone would produce the same thyroid dose as the quantity , ' ' and isotopic mixture of I-131,1-132, I-133,1-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
E - AVERAGE DISINTEGRATION ENERGY 1.13 E shall be the average (weighted in proportion to the concentrati6n P ; of sum each radionuclide of the average betainand thegamma reactorenergies coolant per at the time of samp(ling) disintegration in MEV) forof the is6 topes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. I ENGINEERED SAFETY FEATURE RESPONSE TIME j 1.14 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time R interval from when the monitored parameter exceeds its ESF actuation i setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. _
) !
I CALVERT CLIFFS - UNIT 2 1-3 Amendment No. MS- )
3/4.3 INSTRUMENTATION 3 , TABLE 3.3-6 (Continued) TABLE NOTATION Alarm setpoint to be specified in a controlled document (e.g., setpoint control manual). , ACTION STATEMENTS ACTION 14 - With the number of channels OPERABLE less than required by the Minimum Channels CPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1. r ACTION 16 - With the number of channels OPERABLE less than required by the Minimum Channels OPER/BLE requirement, comply with the ACTION requirements of Specification 3.9.9. ; ACTION 30 - With the number of channels OPERAELE less than required by the Minimum Channels OPERABLE requirement, initiate the i preplanned alternate method of monitoring the appropriate ; parameter (s), within 72 hours, and:
- 1) either restore the inoperable channel (s) to OPERABLE !
status within 7 days of the event, or ;
- 2) prepare and submit a Special Report to the Comission pursuant to Sp:: ficetten S.9.2 within 30 days following h the event, outli ing the action taken, the cause of the inoperability, a d the plans and schedule for restoring the system to OP MBLE status.
I
. Io c.FR 50.+
) , CALVERT CLIFFS - UNIT 2 3/4 3-25 Amendment No. % 1 I l i
f 3/4.3 INSTRUMENTATIM 3/4.3.3 WNITORINg_ INSTRUMENTATION Meteorolonical Instrumentation . LIMITING CONDITION FOR OPERATION ! 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE. APPLICABILITY: At all times. t ACTION: '
- a. With one or more required meteorological monitoring channels !
inoperable for more than 7 days, prepare and submit a Special , Cs) Report to the Comission pursuant to Spectfic tf= 5.0.2 within the next 10 days outlining the cause of the i[alfunction and the plans for restoring the channel (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3 LO.4 are not applicable.
t o C FM 70. 4-- l SURVEILLANCE REQUIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation ! channels shall be demonstrated OPERABLE by the perfomance of the CNANNEL CNECK and CNANNEL CALIBRATION operations at the frequencies shown in Table 4.3-5. > t CALVERT CLIFFS - UNIT 2 3/4 3-27 Amendment No. 476- M
3/4.3 INSTRUMENTATION ' TABLE 3.3-10 (Continued) , ACTION STATEMENTS ACTION 31 - With the number of OPERABLE >ost-accident monitoring channels less than required ay Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days or be in NOT SHUTDOWN within the next 12 hours. ACTION 32 - With the number of OPERABLE post-accident monitoring channels one less than the Minimum Channels 0PERABLE requirement in Table 3.3-10, operation may proceed provided the inoperable channel is restored to 0PERABLE status at the next outage of sufficient duration. , ACTION 33 - With the number of OPERABLE post-accident monitoring channels two less than required by Table 3.3-10, either restore one inoperable channel to OPERABLE status within 30 days or be in NOT SHUTDOWN within the next 12 hours. ACTION 34 - With the number of OPERABLE post-accident monitoring channels one less than the Minimum Channels OPERABLE , requirement in Table 3.3-10, either restore the system to 0PERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report O4- to the Commission pursuant to S :tfic:tf:r. 5.0.2 within 30 days following the event, outliIn:ing the a tion taken, the cause of the inoperability and the plans d schedule for restoring the system to OPERABLE status. jo c g g go, .p. ACTION 35 - With the number of OPERABLE channels two less than required by Table 3.3-10, either restore the inoperable channel (s) to CPERABLE status within 48 hours if repairs are feasible without shutting down or:
- 1. Initiate an alternate method of i.ionitoring for core and Reactor Coolant System voiding; !
- 2. Prepare and submit a Special Report to the Comission pursuant to Sp :tff::tter 5.0.2 within 30 days following the event, outlining the etion taken, the cause of the inoperability and the pla s and schedule for restoring the system to OPERABLE sta s; and t o C FR S'0. f
- 3. Restore the system to OPERABLE status at the next scheduled refueling.
i CALVERT CLIFFS - UNIT 2 3/4 3-35 Amendment No. 476 7
+
l 3/4.3 INSTRUMENT 4 TION : i 3/4.3.3 MONITORING INSTRLMENTATION i Fire Detection Instr = antation *- LIMITING CONDITION FOR OPERATION I 3.3.3.7 As a minimum, the fire detection instrumentation'for each fire ; detection zone shen in Table 3.3-11'shall be OPERABLE' - Mf}ICABL .
"Y: Whenever equipment in that fire detection zone is required to be OPIMLE.
i ACTl0N: With one or more of the fire detection instrument (s) shown in Tab' e 3.3-11 inoperable: i'
- a. Within 1 hour establish a fire watch patrol to inspect the -
zone (s) with the inoperable instrument (s) at least once per hour, ! unlesstheinstrument(s)islocatedinsidethecontainment,then i inspect the containment at least once per 8 hours or monitor the i containment air temperature at least once per hour at the ! Iocations listed in Specification 4.6.1.5; or unless the ! instrument (s) is located in fire detection zones equipped with ' automatic wet pipe sprinkler systems alsmed and supervised to i the Control Room, then within I hour and at least per 24 hours l thereafter, inspect the zone (s) with inoperable instruments and l verify that the Automatic Sprinkler System. including the water . flow alam and supervisory system, is OPERABLE by CHANNP. i FUNCTIONAL TEST. L
- b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission '
pursuant to 5:::ift::t'= 5.0.2 within the next 30 days outlining ! the action tacen, the ause of the inoperability and the plans and schedule for resto ng the instrument (s) to 0PERABLE status. 10 CFM 30.+ ;
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
i . SURVEILLANCE REQUIREMENTS l h 4.3.3.7.1 At least once per 6 months, at least 25% of the above required ; fire detection instruments which are accessible during plant operation l shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST. i Detectors selected for testing shall be selected on a rotating basis such l CALVERT CLIFFS - UNIT 2 3/4 3-37 Amendment No. U 6- 8
l 3/4.4 itEACTOR COOLANT SYSTEM l ,i -SURVEILLANCE REQUIREMENTS (Continued) d " "3 Oc 'epod Pv **cd 4.4.5.5 Reoorts
- a. Following each inservice inspection of steam generator tubes, the number of tu s plugged in each steam generator shall be reported to the Comi sion within 15 days pursuant to Sp::ift::tter. 10 CFR SOA 0.0.2.
b._ The comple results of the steam generator tube inservice h inspection hall be9 t M period fa which S;;;ific:ti:r. S.9.#.thi f ec!'Md in the ir;;; tica '--"el 0; r:tir;
- ca-aleted 5.b) . T11s report shall include:
" pert to (a"~uant fer
- 1. Number and tent of tubes inspected.
- 2. Locatic nd percent of wall-thickness penetration for each indic on of an imperfection.
- 3. ntification of tubes plugged.
- c. sults of steam generator tube inspections which fall into ategory C-3 require verbal notification of the NRC Regional Administrator by telephone within 24 hours arfor to resumption of plant operation. The written followup of tiis report shall provide a description of investigations conducted t) determine cause of the tube degradation and corrective measures taken to J) prevent recurrence and shall be submitted within. the next 30 days f pursuant to Sp :ffic: tier, 5.0.2.
Io C FR 50. f p bm n' f f e d io lhe Co m m i.s.s e'o 1 pr ice fa F1a re h I of each year p w.svani to IoCFRso.Q ' 1
)
CALVERT CLIFFS - UNIT 2 3/4 4-14 Amendment No. M5- # ' i l - ~ - l
O g TABLE 4.4-2 g r-a
% STEAM EENERATOR TUOE INSPECTION
- 4 15T 5 m r d Apprt61194 IIIe .%ru d. AIDrtbl AUR Jus L . J. Asp rttIAust 35 h
5 ample size Result Result Action Required Action Required Result Action Required S 5 A minimum of 5 lubes per G-1 Rone N/A N/A N/A N/A " q Q S. G. C-Z
~
Plbg tiefective tubes and C-1 None R/A N/in $
, inspect additional 25 c-z Plug detective c-1 none tihes in this S. G. tubes and inspect n E additional 45 tubes pi,9 a,y,cti,, g y C-2 tubes C in this S. G. Perfonn action ro C-3 for C-3 result $
of first sagle !Q C-3 Perfom action for M C-3 result of first N/A N/A sample h C-3 inspect all tubes in AII other w this S. G., plug S. G.s are None N/A N/A defective tubes and C-1 A inspect 25 tubes in each a other S. G. ,' Some 5. S.s Perfom action for cn C-2 but no C-2 result of N/A N/A additional second sample 24 hour verbal S. G. are C-3 notification to NRC with written followup pursuant to 0;;;;i'ic ti 7. 5.0.2. Additional Inspect all tubes jo cgg go, f S. G. is C-3 in each S. G. and N plug defective N/A N/A tubes. 24 hour E verbal notification @- to NRC with written g followup pursuant a g to,.......... 2 O G,%3.10 c FA So.+
. S = 3 h Where N is the number of steam generators in the unit, and n is the number of steam generators Q inspected during an inspection .
4
3/4.4 REACTOR COOLANT SYSTEM
^'
i 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:
- a. $ 1.0 pCf/ gram D0SE EQUIVALENT I-131, and
- b. 5 100/E pCi/ gram.
APPLICABILITY: MODES 1, 2, 3, 4 and 5. , ACTION: MODES 1, 2 and 3*:
- a. With the specific activity of the )rimary coolant > 1.0 pC1/ gram D0SE EQUIVALENT I-131 but within t1e allowable limit (below and to the left of the line) shown on Figure 3.4.8-1, operation may 1 continue for up to 100 hours provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.
) b. With the specific activity of the primary coolant > 1.0 pCi/ gram
' D0SE EQUIVALENT I-131 for more than 100 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4.8-1, be in at least HOT STANDBY with T.,, < 500*F '
within 6 hours.
- c. With the specific activity of the primary coolant > 100/E pCi/ gram, be in at least HOT STANDBY with T.,, < 500 F within 6 hours. ,
MODES 1, 2, 3, 4 and 5: )
- d. 'With the specific activity of the primary coolant > 1.0 pCi/ gram l DOSE EQUIVALENT I-131 or > 100/E pCi/; ram, perfon, the sampling i and analysis requirements of item 4 a) of Table 4.4-4 until the j specific activity of the primary coolant is restored to within its limits. Whenever the speci'i ::tivity of the pri=:ry coel:nt eseeds 1.0 pC1/gre. ^^5E E^'_'IYALST I-131 fer ir. :::e:s ,
of 50 h:;r: fer ene can+inna m +4 - 5ttrv:1 er 5 p;rcent f the l unit'; t;tal yearly ep: rating t% purce::t te ACTI" (:) :E:ve, a Sped al Dannet shal' be prepared :nd : bmitted to the With T.,, > 500 F. CALVERT CLIFFS - UNIT 2 3/4 4-24 Amendment No. M5 t i
3/4.4 lEACTOR COOLANT SYSTEM , D LIMITING CONDITION FOR OPERATION (Continued) (] ) 3 ission pursuant to Specification 6.9.2 within the r. ext ys. This report shall contain the results of the cific activ analyses together with the following info on:
- 1. Reactor wer history starting 48 hours pr to the first i sample in ich the limit was exceeded, ,
- 2. Fuel burnup by e region. .
- 3. Clean-up flow history tart' 48 hours prior to the first sample in which the limu as exceeded. :
- 4. History of de-gass operatio if any, starting 48 hours prior to the fi sample in whic he limit was exceeded, !
and :
- 5. The ti duration when the specific activi f the primary coo t exceeded 1.0 pCi/ gram DOSE EQUIVALENT 1.
SURVEILLANCE REQUIREMENTS
.I 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of ;
Table 4.4-4. . l b s ) , CALVERT CLIFFS - UNIT 2 3/4 4-25 Amendment No. 446r 3r i
3/4.4 REACTOR COOLANT SYSTEM ? LIMITING CONDITION FOR OPERATION (Continued)
].
ACTION: With one PORY inoperable in MODE 3 with RCS temperature 5 305'F I a. or in MODE 4, either restore the inoperable PORV to OPERABLE I status within 5 days or depressurize-and vent the RCS through a t 1.3 square inch vent (s) within the next 48 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status,
- b. With one PORV inoperable in MODES 5 or 6, either restore the l inoperable PORY to OPERABLE status within 24 hours, or l depressurize and vent the RCS through a > 1.3 square inch vent (s) / ;
within the next 48 hours; anc maintain tee RCS in this vented ! condition until both PORVs have been restored to OPERABLE status. l
- c. With both PORVs inoperable, depressurize and vent the RCS through p a t 2.5 square inch vent (s) within 48 hours; maintain the RCS in i a vented condition until either one OPERABLE PORY and a vent of !
i
> 1.3 square inches has been established or both PORVs have been restored to OPERABLE status. ]
- d. In the event either the PORVs-or the RCS vent (s) are used to $
. mitigate an RCS pressure transient, a Special Report shall be ,
. prepared and submitted to the Comission pursuant to j
") McF# fo,+ S:riftstir. 0.".2 within 30 days. The report shall describe !
$v tie circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence. j
- e. With less than two HPSI umps' disabled, place at least two HPSI f pump handswitches in pul -to-lock within fifteen minutes and disable two HPSI pumps within the next four hours.
- f. With one or more HPSI loo) MOVs' not prevented from automatically S aligning a HPSI pump to t1e RCS, immediately place the MOV handswitch in pull-to-override, or shut and disable the affected MOV or isolate the affected HPSI header flowpath within four hours, and implement the action requirements of Specifications 3.1.2.1, 3.1.2.3, and 3.5.3, as applicable, j
- g. With HPSI flow exceeding 210 gpm while suction is aligned to the jk i RWT and an RCS vent of < 2.6 square inches exists, !
- 1. Immediately take action to reduce flow to less than or equal ,
to 210 ppm. j
~!
)
s. Except when required for testing. l CALVERT CLIFFS - UNIT 2 3/4 4-34 Amendment No.165- l
\
l
+!
- - ~ - .. - . - . - . - . - - __
3/4.4 REACTOR COOLANT SYSTEM
~'
i 3/4.4.11 CORE BARREL MOVEMENT , LIMITIM CMDITIM FOR OPERATIM - 3.4.11 Core barrel movement shall be limited to less than the Amplitude i Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for j the applicable THERMAL POWER 1evel. , APPLICABILITY: MODE 1.
]
ACTION:
- a. With the APD and/or SA exceeding their applicable Alert Levels, POWER OPERATION, may proceed provided the following actions are i taken: ,
1.. APD shall be measured and processed at least once per - ; 24 hours, ,
- 2. SA shall be measured at least once per 24 hours and shall be processed at least once per 7 days, and . -
- 3. A Special Report, identifying the cause(s) for exceeding the ;
applicable Alert Level, shall be prepared and submitted to : J f the Commission pursuant to !;::tfi::ti: . 5.".2 within I l') 30 days of detection. /C c.FR 50 1- )
- b. With the APD and/or SA exceeding their applicable Action Levels, ;
measure and process APD and SA data within 24 hours to detemine ! if the core barrel motion is exceeding its limits. With the core barrel motion exceeding its limits, reduce the core barrel motion-to within its Action Levels within the next 24 hours or be in NOT STANDBY within the following 6 hours..
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l CALVERT CLIFFS - UNIT 2 3/4 4-38 Amendment No. Mfr X
3/4.5 EMERGENCY CORE C0OLING SYSTEMS (ECCS) ; 3/4.5.2 ECCS SUBSYSTEMS - MODES 1. 2 AND 3 (2 1750 PSIA) l l LIMITING CONDITION FOR OPERATION , 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE high-pressure safety injection pump,
- b. One OPERABLE low-pressure safety injection pump, and
- c. An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.
APPLICABILITY: MODES 1, 2, and 3*. ACTION: ,
- a. With one ECCS se' system inoperable, restore the inoperable subsystem to 0PERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. In the event the ECCS is actuated and injects water into the
) Reactor Coolant System, a Special Report shall be prepared and i submitted to the Comission pursuant to Sp::ifi : tic , 5.%2 within 90 days describing the circumstances of he actuation and the total accumulated actuation cycles to date.
10 cFR foA-J With pressurizer pressure > 1750 psia. CALVERT CLIFFS - UNIT 2 3/4 5-3 Amendment No. 449-
i i 3/4.5 EMERGENCY C0RE C0OLING SYSTEMS (ECCS)
. 3/4.5.3 ECCS SUBSYSTEMS - MODES 3 (< 1750 PSIA) AND 4 4
LIMITING CONDITION FOR OPERATION 3.5.3' As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One' OPERABLE high-pressure safety injection pump, and l
- b. An OPERABLE flow path capable of taking suction from the !
refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal. APPLICABILITY _: MODES 3* and 4. ACTION:
- a. With no ECCS subsystem OPERABLE, restore at least one ECCS :
i subsystem to OPEPABLE status within 1 hour or be in COLD SHUTD0WN within the next 20 hours.
- b. In the event the ECCS is actuated and injects water into the i Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to S;;;;ifisticr 5.0.2 :
j
@ within 90 days describing the circumstances of phe actuation and the total accumulated actuation cycles to date. ;
10 CFR 50 f SURVEILLANCE REQUIREMENTS , 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of Specification 4.5.2. l l l i
# Between 350 F and 305"F, a transition region exists where the l OPERABLE HPSI pump will be placed in pull-to-lock on a cooldown and l restored to automatic status on a heatup. At 305 F and less, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start automatically. At 305'F and less, HPSI pump use will be i conducted in accordai.ce with Technical Specification 3.4.9.3.--- i l
s)
- With pressurizer pressure < 1750 psia.
1 CALVERT CLIFFS - UNIT 2 3/4 5-7 Amendment No. 449-l
i
. I 3/4.6 CONTAINMENT SYSTEMS 7 SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.6.3 Containment Surfaces. The exposed accessible interior and exterior surfaces of the containment, including the liner- plate shall be visually inspected during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2). This inspection shall be performed prior to the Type A containment leakage rate test to uncover any ,
evidence of structural deterioration which may affect either the . containment structural integrity or leak tightness. l 4.6.1.6.4 Reports. Any abnormal degradation of the containment structure detected during the above required tests and inspections shall be reported Oep to the Commission pursuant to Specif'- This report shall include a descripidon; of thetier, 0.0.2 tendon within the condition, the next 30 days. condition of the concrete (especial y at tendon anchorages), the inspection procedures, the tolerances on crack ing, and the corrective actions taken. i 10 CFR 50.f r F u.- l CALVERT CLIFFS - UNIT 2 3/4 6-10 Amendment No. dWHh
. . = _ -.
l f 3/4.7 PLANT SYSTEMS t SURVEILLANCE REQUIREMENTS (Continued) - -
- a. Sources in use - At least once per six months for all sealed !
sources containing radioactive material: ;
- 1. With a half-life greater than 30 days (excluding !
Hydrogen 3),and
- 2. In any fonn other than gas.
- b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another l licensee unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use. ;
- c. Startuo sources and fission detectors - Each sealed startup ,
source and fission detector sha' I be tested within 31 days prior ! to being subjected to core flux or installed in the core and ' following repair or maintenance to the source or detector. 4.7.9.1.3 Reports - A report shall be prepared and submitted.to the , Commission on an annual basi if sealed source or fission detector leakage tests reveal the presence of > .005 microcuries of removable contamination. . i pu<s n nt to to cfg so,+ l CALVERT CLIFFS - UNIT 2 3/4 7-31 Amendment No. l e - M
i f
' 3/4.7' PLANT SYSTDtS l N 3/4.7.11 FIRE SUPPRESSION SYSTEMS Fire Suporession Water System l
LIMITING CONDITION FOR OPERATION !
~
3.7.11.1 The Fire Suppression Water System shall be OPERABLE with:
- a. Two high pressure pumps, each with a capacity of 2500 gpm, with ;
their discharge aligned to the fire suppression header,
- b. Two water supplies, each with a minimum contained volume of 300,000 gallons, and
- c. An 0PERABLE flow path capable of taking suction from the ;
Pretreated Water Storage Tanks Numbers 11 and 12 and transferring ! the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant ^ curb valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser recuired to be OPERABLE per Specifications 3.7.11.2, 3.7.11.4, anc 3.7.11.5. APPLICABILITY: At all times.
- ACTION:
i
- a. With one pump and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status within 7 days or prepare and submit a Special Re) ort to the Comission pursuant to 1
/OcF# 50.+ Spr!'icet' . '.".2 wit 11n the next 30 days outlining the plans O4- and )rocedures to be used to provide for the loss of redundancy in t1is system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable,
- b. With the Fire Suppression Water System otherwise inoperable:
- 1. Establish a backup Fire Suppression Water System within 24 hours, and i^
- 2. Submit a Special Report in accordance with em.,4<4,.+,-- e a e. ,
*r - ' *#*6* !
{ a) By telephone within 24 hours, i 1 CALVERT CLIFFS - UNIT 2 3/4 7-33 Amendment No. -161 M l l 1
3/4.7 PLANT SYSTEMS U 3/4.7.11 FIRE SUPPRESSION SYSTEMS
~"
r Fire Hose Stations LIMITING CONDITION FOR OPERATION j 3.7.11.4 The fire hose stations shown in Table 3.7-6 shall be OPERABLE. , APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. - ACTION: i
- a. With one or more of the fire hose stations shown in Table 3.7-6 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within I hour. Restore the fire hose station (s) to OPERABLE status withir.14 days or prepare and submit a Special Report to the Of Commission 30 days outlining pursuant the actioto Sp=[i'intia.
taken, 0.S.2 the cause of thewithin the next inoperability and the plans and schedule for restoring the fire hose station (s) to OPERABLE status,
- b. The provisions of Specificaftons 3.0.3 and 3.0.4 are not applicable.
}..
/O cFR 50.4-SURVEILLANCE REQUIREMENTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7-6 shall be demonstrated OPERABLE:
- a. At least once per 31 days by visual inspection of the station to assure all required equipment is at the station. Hose stations located in the containment shall be visually inspected on each scheduled reactor shutdown, but not more frequently than every 31 days.
- b. At least once per 18 months for hose stations located outside the containment and once per REFUELING INTERVAL for hose stations inside the containment by:
- 1. Removing the hose,for inspection and re-racking, and
- 2. Replacement of all degraded gaskets in couplings.
./
CALVERT CLIFFS - UNIT 1 3/4 7-42 Amendment No. le6 M
p, .; 3/4.7' PLANT SYSTEMS 1
- h. 3/4.7.11 FIRE SUPPRESSION SYSTEMS Yard Fire Hydrants and Hydrant Hose Houses !
i LIMITING CONDITION FOR OPERATION ! f 3.7.11.5 The following yard fire hydrants and associated hydrant hose ! houses shall be OPERABLE: ' i
- a. f6 yard hydrant and associated hydrant hose house, which provides !'
primary protection for Unit 2 RWT blockhouse.
- b. #7 yard hydrant and associated hydrant hose house, which provides primary protection for Unit 1 RWT blockhouse. ,{
APPLICABILITY: Whenever equipment in the areas protected by the yard fire ! hydrants is required to be OPERABLE. i i ACTION: : l
- a. With one or more of the yard fire hydrants or associated hydrant ,
hose houses inoperable, within I hour have sufficient additional i lengths of 2-1/2 inch diameter hose located in an adjacent '! OPERABLE hydrant' hose house to provide service to the unprotected ! area (s) if the inoperable fire hydrant or associated hydrant hose
.j house is the primary means of fire suppression. Restore the j hydrant or hose house to OPERABLE status within 14 days or i prepare and. submit a Special Report to the Commission pursuant to ;
O4-
- cr'^ = tie- 5.".* within the next 30 days outlining the action
- taken, $he cause of the inoperability, and the plans and . schedule ,
for rettoring the hydrant or hose house to OPERABLE status. l
- b. The prcvisions of Specifications 3.0.3 and'3.0.4 are not !
applicable.
)
10 cFB 506 ! c .Y i f I i CALVERT CLIFFS - UNIT 2 3/4 7-45 Amendment No.163- I !
I L 3/4.7. PLANT SYSTEMS- [ -4. '
) 3/4.7.12 PENETRATION FIRE BARRIERS LINITING CONDITION FOR OPERATION 3.7.12 All fire barrier penetrations (i.e., cable penetration barriers. l
.firedoors and fire dampers), in fire zone boundaries, protecting safe .
shutdown areas shall be OPERABLE. , APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required fire barrier penetrations inoperable within one hour either establish a continuous fire ,
watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the ; inoperable fire barrier and establish an hourly fire watch ! p(atrol; or the including verify water the flow operability of Automatic alarm and Sprinkler supervisory system)Systems on both sides of the inoperable fire barrier. Restore the inoperable ' fire barrier penetration (s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission pursuant to , h locfA 50.t Sp :if t:"'er 5.9.2 within the next 30 days outlining the action taken, theforcause of the la
.s schedule restoring theinoperable fire barrier penetration penetration and to p(s)ns and - .,) OPERABLE status. .
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. -
i i SURVEILLANCE REQUIREMENTS 4.7.12 Each of the above required fire barrier penetrations shall be i verified to be OPERABLE:
- a. At least once per 18 months by a visual inspection.
- b. Prior to returning a fire barrier penetration to functional [
status following repairs or maintenance by perfonnance of a ! visual inspection of the affected fire barrier penetration (s). , i
).
CALVERT CLIFFS - UNIT 1 3/4 7-47 Amendment No. IM M l i
9
)
6.0 ADMINISTRATIVE CONTROLS 7 6.1 RESPONSIBILITY Oq 6.1.1 operation and shall delega}te in writing the succession to thisThe gia ' responsibility during his absence. l J spee,L l
, ine ttHes Idi.,y 'Ih eo senf npele<.p a 6.2 ORGANIZATION g. g. g , .pj, ,
6.2.1 ONSITE & OFFSITE ORGANIZATIONS a[n f48 thesef #Teet,meaf 8 / //*45 spea;p;uf,.,, r/8//ard/e/ Onsite and offsite organizations shall be establi ed for unit operation - and corporate management, res sectively. The onsit and offsite organizations shall include t1e positions for activi ies affecting the safety of the nuclear power plant.
- a. Lines of authority, responsibility and comuni tion shall be ,
established and defined for the highest manageme levels through intemediate levels to and including all operatin organization positions. These relationships shall be documente and updated, as appropriate, in the form of organization charts, nctional descriptions of departmental responsibilities and rel tionships, and job descriptions for key personnel positions, or i equivalent forms of documentation. These requirements hall be documented in FSP CMpter M. and get:d ' 20:rt,:: .dth 10 CFR 50.M (:) . f A e uF.S A A q b. The giant Ceneni ganager shall be responsible for overall unit ' safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant. I
- c. The Vice President - Nuclear Energy shall have corporate I responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable perfomance of the staff in operating, maintaining, and providing technical support to the j plant to ensure nuclear safety. !
I 'd . The individuals who train the operating staff and those who carry l out health physics and quality assurance functions may report to l the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. 6.2.2 UNIT STAFF 23 ' 2"i?+ A Y TA*"d%?NL th*!)u.a +u _,_,_._
@ M -
sM 4 2. .. . . . n. ...: L, Q.:
- r. r.. ' .: o..1... r. c.q m. . a ; ,
in. lvde +he fo llo wt.,y : ) CALVERT CLIFFS - UNIT 2 6-1 Amendment No. 149-
~--" -
6.0 ADMINISTRATIVE CONTROLS '. [ b. At least one licensed Operator shall be in the Control Room when fuel is in the reactor. ,
- c. At least two licensed Operators shall be present in the control Room during reactor STARTUP, scheduled reactor shutdown,_ and ;
during recovery from reactor trips.
- d. An individual qualified in radiation protection procedures shall
- be on site when fuel is in the reactor. .
=
-t Ali *I ALTERATIONS after the in' ti=1 ft: 1 .m ng shall be !
directly sup niawe : c L a l' censed Senior Reactor Operator ! or Sent perator Limited to rs=; :::M14na who has no ! concurrent responsibilities during this operRTB1r. ---
- e. .P. A site Fire Brigade of at least 5 members shall be maintained
@ onsite at all times. The Fire Brigade shall not include the minimum shift crew necessary for safe shutdown of both units i
j (4 members) or any personnel requireJ_for other estantial ; functions during a fire emergency. [J+tsed.2[see-p.~ 6-S] ; o ma.<*9ee - - ~ - The :;.print;; a+;oint a .M 1;;r Opereti;a; shall hold or have he d a
/g f.g. j i
senior reactor operator license at Calvert Cliffs. The General.s.y,,.,/,, ; S:; = i;;r " :10:r 0::retiea;, Shift Supervisor and control ; Room Supervisor shall iold a senior reactor operator license. -t The Control Room Operator shall hold a reactor operator license. } of leas t
- a. A tota) oh iI,<e e n on- licensed opea h , i
$ L si,a ll be a.ss,9 nad to %e unit i + 1.
_ shif t c r e w.s . l t i is Inseds 3a+(see p.6-+),;
-- ~ _
)
@ 10<< 1 2o(sea p. s- + )
l i i l l _ . . . t f l CALVERT CLIFFS - UNIT 2 6-2 Amendment No. M9-l t
f 6.0 ADMINISTRATIVE CONTROLS 3 TABLE 6.2-1 . O t MINIMUMSHIFTCREWCOMPOSITIOJf ndition of Unit 2 - Unit 1 in MODES 1. 2. or 4 , N / ! APPLICABk MODES LICENSE e CATEGORY 1. 2, 3 & 4 / 5&6 ; SOL" \ 2 [ 2* j OL" \ 3 / 3 Non-Licensed \ [ 3 Shift Technical Advisor Y1" 1" Condition of it 2 - Uni 1 in MODES 5 or 6
/ \
., APPLkBLEMODES ,
LICENSE CATEGORY 1, 2, 3 & 4 \ S&6 ! SOL" / 2 \ l' , OL" [ 3 \2 - Non-Licens/ 3 \3 ' Shift Te/nical Advisor 1" Oh
/
N l l CALVERT CLIFFS - UNIT 2 6-3 Amendment No.-H9-i I
1 6.0 ADMINISTRATIVE CONTROLS TO' E 5.2-1 (0;ntinued)
- N
( .inse<f ;to I h 00:: n t include the licented Senter Pe :t:r Operater Or Serier i Re;;t:r Operater Limited to Fuel ":nd!'ng. : p:rvising 005 ( ALTEP_^.TJ0Mi duri= f'Ja2 releadiac. _ ,f fo egg f,,f9 h
' g, ts (,,,)(>)(j) ,,,)
(i'*Li'.::t set.se2
- ::.. oj]Wq:i,illbe. divi _dj.1 1: 1icensed gg,ggcoon f f,, ea,s O23 5 4.2,24 y A. Shift crew composition may be lesyT'[n9heh minimum requirement for JH5ed t '
a period of time not to exceed Thours in order to accommodate 9,3 g, (,-;-) unexpected absence of on duty shift crew members provided imediate action is taken to restore the shift crew composition to within the "I i 8nts,e 2.1. f" - M The STA shall be qualified [ serve as follows: - M h (1) W or 4, one unit in MODE 5 or 6, and the other unit in MOD e SOL holder other than the Shift Supervisor all serve
, 2, 3 /
l as the S
/
eled and the other unit i (2) With one unit the STA shall be a OL holder in addi M0DE 1, 2, 3 or 4 n to the one SOL / required, t
! (3) With both units in MODE 1 , r 4, the STA shall be an SOL holder in addition to quired.
\ two SOL i
/
the above, the STA may be individual with the fAsanalternativ following mi m qualifications: a Bachelor's Deg or equivalent I in a scie fic or engineering discipline with specific ining in plant sign and response and analysis of the plant for transients and accidents. q v-Repla e e d wit h 2nsed 3(nv1 pp) J CALVERT CLIFFS - UNIT 2 6-4 Amendment No.149-
.[Hje83(Seef.6.2.)
x _. - h g. One Shift Technical Advisor (STA) shall be assigned to the shift crew when either unit is in MODE 1,2,3 or 4 and shall be filled as follows: i ;
@l (1) by the Shift Supervisor or an on-shift Senior Operator Ucense (SOL) holder arovided the individual meets the Commission Policy Statement on Engineering
- Expertise on Shift; or 1
/ 7 h I
~ (2) by an individual meeting the minimum STA education and training requirement of Specification 6.3.1; or I
(3) by a SOL holder previous) approved as an exception to the minimum STA , Q [) education requirements o Specification 6.3.1, provided the following conditions are met. , I (i) With both units in MODE 1,2,3 or 4, the STA shall be an SOL holder in addition to the two SOL %olders required. I (ii) With one ue.*t in MODE 1,2,3 or 4 and the other unit in MODE 5 or 6, the STA shall be an SOL holder other than the Shift Supervisor. (iii) With one unit in MODE 1,2,3 or 4 and the other unit defueled, the STA shall be a SOL holder in addition to the one SOL holder required. s m _ 1 i i
6.0 ADMINISTRATIVE CONTROLS
^
6.3 FACILITY STAFF OVALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Safety Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift Technical Advisor who shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents. ANitional eu ep tiens to Alv.GT W 19. /- 19 7/ w e O26 contain ad in Ta l>le IB-1 of the Qvalif A.ss m nc e. Po is'cy
& 6 G
' U' "' #" " ' ?# "' ' '"Y'
@ and replacement training program for the 6hetcaining~iidHErthe-dirretion shall be maintainec e factMty st'aff of the Canaral-30porvisor - Nuclear Training and shall ==et :;r cu.een the7equirementtand recomendations of SWSI N18.1-1971 and 10 CFR 55.59(c), as applicabk. $
h ininh :r:gr:m for the Fire Brigade #IsNa~11 be nint:ined under L the direction Of t:: " n:ger-Nucle:r S:fety and Picaning Dep:rtment ad >c shall meet er ex:::d the requirements of NFPA 27, 1975 edition. Inse<+ .T Csee p. s. .y g . REVIEW AND AUDIT 6.5.1 LANT OPERATIONS AND SAFETY REVIEW COMMITTEE (POSR FUNCTION 6.5.1.1 The RC shall function to advise the ant General Manager on all matters rel ed to nuclear safety. COMPOSITION 6.5.1.2 The POSRC shall co sed of at least seven, but no more than ten, members, including the aiman. Members shall collectively have exper:1~ence in the followi a as: Nuclear Operati s Electrical a Controls Maint nce Chemistry Hechanic Maintenance
. Nuclea Engineering Radi ion Safety P1 t Engineering sign Engineering d
CALVERT CLIFFS - UNIT 2 6-5 Amendment No. M
6.0 ADMINISTRATIVE CONTROLS
^ 03' r hn6 aa 'a* d '" r'a' 6' '6 "* a " ra' " " a r-rs shall have a minimum of eight years power plant experience o which' a mi mum of three years shall be nuclear power experience. ,At le t one membe hall have an SRO license on Calvert Cliffs Units 1 and 2.
1 CHAIRMAN l 6.5.1.3 The airman and alternate Chaimen of the POSR shall be appointed in ting by the Plant' General Manager. Cha n shall have a : minimum of 10 ye s power plant experience of which a inimum of three ! years shall be nuc ar power experience. ; I ALTERNATES j 6.5.1.4 All alternate ers shall be apso ted in writing by the Plant i General Manager. Alternate ers shall ye a minimum of eight years ! power plant experience of wh a minimu f three years s. hall be nuclear i power experience. ; MEETING FRE00ENCY } 6.5.1.5 The POSRC shall meet at eas once per calendar month and as ; convened by the POSRC Chaiman one o the designated alternates. ; t _D.. ; OVORUM 'l 6.5.1.6 A quorum of the OSRC shall include e Chaiman or one. of the designated alternate ch men and shall consist of a majority of the ! members, including alt nates. No more than hal of the quorum shall be ! alternates, includin an alternate chaiman. i t RESPONSIBILITIES l t
- 6. 5. f.' 7 Th'e P RC shall be responsible for the followin except for those i items design ed for review by the Procedure Review Commi tee or Qualified '
Reviewer in ccordance with Specification 6.5.2 and 6.5.3, espectively:= i 1
- a. eview of 1) all procedures required by Specificat n 6.8 and j changes thereto, and 2) any other proposed procedure or changes '
. thereto as determined by the Plant General Manager to. ffect nuclear safety.
. -1
& l' CALVERT CLIFFS - UNIT 2 6-6 Amendment No. 449-y= _. 5:.39. 1. .
6.0 ADMINISTRATIVE CONTROLS ^' Cross-disciplinary reviews of these procedures are con cted in , acc'ordance with administrative procedures in addition o the reviews conducted by POSRC, the Procedure Review C ittee, or ualified Reviewer.
- b. Re ew of all proposed tests and experiments th affect nuclear safe .
- c. Revi'w e f all proposed changes to Appendix Technical Specific tons.
- d. Review of a 1 proposed changes or modif cations to plant systems or equipmtnt hat affect nuclear safe .
- e. Review of the ant Security Plan d implementing procedures and ,
shall suSmit rec ended changes the Off-Site Safety Review Comittee.
- f. Review of the Emerge y Plan d implementing procedures and shall submit recomen d cha es to the Off-Site Safety Review Comittee.
- g. Review of changes to the CESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL
- h. Review of all 10 CFR 0.59 Sa ty Evaluations that support procedures in 6.5.1 .a and cha es or modifications in 6.5.1.7.d.
- 1. Investigation o all violations of e Technical Specifications including the eparation and forwar ng of reports covering evaluation an recomendations to pre nt recurrence to the Plant General Man er, the Vice President - clear Energy and to the Chaiman of the Off-Site Safety Review C ittee.
J. Review o all REPORTABLE EVENTS.
. .k . Revie of facility operations to detect poten tal safety hazards.
- 1. Rev ew of any accidental, unplanned or uncontro ed radioactive r ease that exceeds 25% of the limits of Specif ation 3.11.1.2,
.11.2.2 or 3.11.2.3, including the preparation o reports overing evaluation, recomendations and dispositio of the corrective action to prevent recurrence and for forw ding of these reports to the Plant General Manager and the Of Site Safety Review Comittee.
- ~ ^
j CALVERT CLIFFS - UNIT 2 6-7 Amendment No. H 9
r 6.0' ADMINISTRATIVE CONTROLS
'h m. Perfomance of special reviews, investigations or ' analyses and.
reports thereon as requested by the Chairman of the Off- te : Safety Review Comittee. AUTHORITY 6.5.1.8 The P nt Operations and Safety Review Comittee hall: ,
- a. Recomme to the approval authority approval r disapproval or_
procedure considered under 6.5.1.7.a.
- b. Recomend to he Plant General Manager a royal or disapproval of items conside d under 6.5.1.7(b) throu i (h) above.
- c. Render detemina ons in writing wit regard to whether or not each item conside d under 6.5.1.7 through(h)above constitutes an unre iewed safety estion.
- d. Evaluate root causes 'd reconn nded actions to prevent recurrence for items c sider under 6.5.1.7(1) through (1). -
- e. Provide written notificat within 24 hours to the Vice President - Nuclear Energ nd the Chaiman of the Off-Site Safety Review Comittee sagreement between the POSRC and the i responsible approval au orit in the case of item 6.5.1.7.a or ,
between the POSRC and he Plan General Manager; however. the
~~)
Plant General Manage shall hay responsibility for resolution of ; such disagreements rsuant to 6. 1 above. REC 0kDS 6.5.1.9 The POSRC sha maintain written minutes f each meeting and copies shall be provi ed to the Vice President - Nu lear Energy, Chaiman of the Off-Site Saf y Review Comittee, and the Pla General Mar,sger. , i 6.5.27 PROC 5DUR REVIEW COMMITTEE FUNCTION i l 6.5.2.1 e Procedure Review Comittee may function to revie items listed in Speci cation 6.5.1.7(a) in lieu of review by POSRC or Qual fied Review as directed by the Plant General Manager. U CALVERT CLIFFS - UNIT 2 6-8 Amendment No. H9-1 ~ ^ *
"~ ' a , _- _ _ . _ . ,__ ((
-6.0 ADMINISTRATIVE CONTROLS-Oh COMPOSITION 6.5. 2 The Procedure Review Comittee shall be composed of a airman and eight dividuals who shall collectively have expertise in th areas contain in Technical Specification 6.5.1.2. l Members sh 'l be appointed in writing by the Plant Gener Manager.
Members sha' have a minimum of eight years power plant perience of which , a minimum of ree years shall be nuclear power expert ca. At least one member shall b a POSRC member or alternate. The ch ter for the Procedure Review Comitte shall include a description of rship, qualifications, : functions, and re rts and shall be described in ant administrative procedures. The P cedure Review Comittee may dissolved at the discretion of the P nt General Manager. CHAIRMAN 6.5.2.3 The Chairman and 1 ternate Chat n of the Procedure Review l Comittee shall be appointe in writing y the Plant General manager. Chaimen shall have a minimu of eight ears power plant experience of which a minimum of three year shall e nuclear power experience. i I ALTERNATES ^
) 6.5.2.4 All alternate members hall be apaointed in writing by the Plant ,
General Manager. Alternate m bers s 11.1 ave a minimum of eight years l power plant experience of w ch a mini m of three years shall be nuclear power experience. MEETING FREQUENCY 6.5.2.5 The Procedu e Review Comittee shall et at least once per i calendar month and s convened by the Chairman his designated alternates. OUOROM 6.5.2.6 A uorum for the Procedure Review Comittee sh 11 consist of the Chairman one of the designated alternate Chaimen and hree primary or alterna members provided at least four disciplines are presented. 1 d CALVERT CLIFFS - UNIT 2 6-9 Amendment No. H9-
~I __
__ _ _ _ _ _ _ _ _ __iE____
i 6.0 ADMINISTRATIVE CONTROL 5 .[ Wh AUTHDRITY , 6.5.2. -The Procedure Review Comittee shall:
- a. !
Recomend cedures considered to the approval authority under 6.5.1.7 a). (approval or di pproval of !
- b. Re er deteminations in writing with regard whether or not-each rocedure under 6.5.1.7(a) constitutes u'nreviewed safety t quest n. l
- c. Provide itten notification within 24 urs to the Vice :
President Nuclear Energy and the Cha n of the off-Site l Safety Revi Comittee of disagrc s between the Procedure. t Review Comi ee and the responsib1 approval authority. The ! Plant General nager shall have spnsibility for resolution of - such disagreeme s pursuant to 6 .I above. i RECORDS i 6.5.2.8 The Procedure Review C ee shall maintain written minutes of each meeting and copies shall be p vided to the Plant General Manager. t
,. 6.5.3 OUALIFIED REVIEWERS l 0 l FUNCTION 6.5.3.1. The Plant Gene 1 Manager may design te specific procedures or i classes of procedures escribed in Specificati 6.5.1.7.a to be reviewed ;
by Qualified Reviewer in lieu of review by POS or the Procedure Review l Comittee. 'i l i AUTHORITY ! 6.5..T.2 . Qusi f ed Reviewers shall: l
- a. R comend to the approval authority approval or isapproval of ;
esignated procedures and changes considered und 6.5.1.7.a. and -l b Render detemination in writing with regard to whet r or not i each procedure under 6.5.1.7.a constitutes an unrevi ed safety ; question. ; D i CALVERT CLIFFS - UNIT 2 6-10 Amendment No.149-l
~~. ,ww rg-, -g , . , - , - - , - , - - - - - , ..,,-.m--e- w- . --
_ _ ~ . _ . - _ _ . _ _ t 4 t 6.0 ADNINISTRATIVE CONTROLS ; r h- l. c. Provide written notification within 24 hours to' the Vice
"' President - Nuclear Energy and the Chairinan of the Off- ite !
Safety Review Committee of disagreements between the lified -i Reviewer and the approval authority, The Plant Gener Manager i shall have responsibility for resolution of such di greements j ursuant to 6.1.1 above. l l CERTIFICATI0 , j Reviewers shall be nominated, 'tra ed, and certified in ! 6.5.3.3 Qualif l accordance with a inistrative procedures. Certiff tion shall be by a department manager. . 6.5.3.4 Certifuatio requirements c' personn- designated as Qualified ! Reviewers shall be in cordance with adminis ative procedures. j ave: '! Qualified Reviewers shall !
-a. A Bachelors degre in engineer g, related science, or technical l discipline, and N:p ars of uclear power plant experience; j R l
\
- b. Six years of nuclear po r plant experience j
.- OR l
- c. Equivalent combinat on of e cation and experience as approved by j a Department Hang r. !
/ !
RECORDS i 6.5.3.5 Review of pro edures by Qualified R iewers shall be documented in I accordance with admi strative procedures. 6.5.4 0FF-SITE S TY REVIEW COH41TTEE (OSSRC)
- i FUNCTION 6.5.4.1 The ff-Site Safety Review Committee shall nction to provide independent eview and audit of designated activities n the areas of: 1 1
- a. nuclear power plant operations-
. nuclear engineering U
- CALVERT CLIFFS - UNIT 2 6-11 Amendment No. 1 4 i
- =- .- - - -_ _ _ .
6.0 ADMINISTRATIVE CONTRCLS 7 c. chemistry and radiochemistry metallurgy and non-destructive examination
- e. instrumentation and control
- f. ra ological safety
- g. mecha ical and electrical engineering
- h. quality ssurance practices COMPOSITION 6.5.4.2 The OSSRC shall e composed of at least s en members, including the Chairman. Members of he OSSRC may be from e Nuclear Energy Division or other BG&E organization r from organizatio external to BG&E, and shall collectively have expe ise in all are of 6.5.4.1. l OUALIFICATIONS !
6.5.4.3 The Chaiman, members and ernate members of the OSSRC shall be appointed in writing by the Vice Pr ident - Nuclear Energy and shall have an academic degree in engineering r physical science, or the equivalent, and in addition shall have a min um o five years technical experience in one or more areas given in 6.5 .1. No re than two alternates shall , participate as voting members n OSSRC ac vities at any one time. t CONSULTANTS 6.5.4.4 Consultants s 11 be utilized as determ ed by the OSSRC Chaiman to provide expert ady ce to the OSSRC. MEETING FREQUENCY 6.5.4"5 The 0 RC shall meet at least once per six mont s. OVORUM 6.5.4.6 he quorum of the OSSRC necessary for the perfomance of the OSSRC : review nd audit functions of thesc Technical Specifications sh 11 consist of mor than half the OSSRC membership or at least four members, whichever > is g ater. This quorum shall include the Chaiman or his appoin ed CALVERT CLIFFS - UNIT 2 6-12 Amendment No. -149-1
.,p...
6.0 ADMINISTRATIVE CONTROLS T ' al rnate and the OSSRC members, including appointed alternates, mee ing the equirements of Specification 6.5.4.3. No more than a minorit of the ' quo shall have line responsibility for operation of the plant. REVIEW. , 6.5.4.7 The SRC shall review: l
- a. The sa ety evaluations for 1) changes to pr edures, equipment or systems nd 2) tests or experiments compl ed under the provisio of 10 CFR 50.59, to verify t such actions did not constitute n unreviewed safety questi .
- b. Proposed cha es to procedures, eq pment or systems which involve an unr viewed safety ques on as defined in 10 CFR 50.59.
- c. Proposed tests o experiments ich involve an unreviewed safety
- question as define in 10 CF 50.59.
- d. Proposed changes in , chn cal Specifications or this Operating License.
- e. Violation of codes, gu tions, orders, Technical Specifications,11 nse re uirements, or of internal procedures
., or instructions h ing nuc1 r safety significance.
)
- f. Significant op ating abnormal ties or deviations from nonnal and expected perf ance of plant e ipment that affect nuclear safety. t
- g. All REPO ABLE EVENTS.
- h. All re gnized indications of an unant ipated deficiency in some :
aspec of design or operation of safety elated structures, syst.ms, or components.
- i. R orts and meetings minutes of the POSRC.
AUDITS ( 6.5.4. 1 Audits of facility activities shall be perfonned nder the ! cogni ance of the OSSRC. These audits shall encompass: [ f
.)
CALVERT CLIFFS - UNIT 2 6-13 Amendment No. 449- ! r - - . - , . , 7
i 6.0 ADMINISTRATIVE CONTROLS i O 31 .. The conformance of facility. operation to provisions contained within the Technical Specifications and applicable licens t conditions at least once per 12 months.
- b. e perfonnance, training and qualification of the en re fa 111ty staff at least once per 12 months.
f
- c. The suits of actions taken to correct deficien es occurring in facili equipment, structures, systems or meth of operation '
that a ct nuclear safety at least once per months. ;
- d. The perfo ance of activities required by e Quality Assurance :
Program to et the criteria of 10 CFR Pa 50 Appendix B at least once p 24 months.
- e. The Safeguards ntingency Plan and plementing procedures at :
least once per 1 months in accorda e with 10 CFR 73.40(d).
- f. Any other area of 111ty oper on considered appropriate by the OSSRC or the Vic Presiden - Nuclear Energy.
- g. The Facility Fire Prote to Program and implementing procedures at least once per 24 mont .
I
- h. An independent fire pr ect n and loss prevention program
- inspection and audit all be erformed at least once per I ._)
12 months utilizing ither qua fied offsite licensee personnel ' or an outside fire rotection f . t
- i. An inspection a audit of the fir protection and loss :
prevention pr ram shall be perform by a qualified outside fire i consultant a least once per 36 month > l
- j. The radio gical environmental monitorin program and the results i
thereof t least once per 12 months. ;
- k. The FSITE DOSE CALCULATION MANUAL and impi enting procedures at ast once per 24 months. ,
"1. e PROCESS CONTROL PROGRAM and implementing proc res for-processing and packaging of radioactive wastes at i st once per '
24 months. 3 , CALVERT CLIFFS - UNIT 2 6-14 Amendment No.149-l . . . . , .. m.v .
6.0 ADMINISTRATIVE CONTROLS
^'
- m. The perfonnance of activities required by the Quality Assu ance '
. Program for effluent and environmental monitoring;at leas once per 12 months.
6.5.4.8.2 Review of-. facility activities shall be perfonned un r the cognizance f the OSSRC. These reviews shall encompass: ,
- a. The acility Emergency Plan and implementing pro dures at least <
once er 12 months in accordance with 10 CFR 5 4(t). AUTHORITY 6.5.4.9 The OSSRC sh 11 report to and advise the ice President - Nuclear Energy on those areas f responsibility specift in Sections 6.5.4.7 and 6.5.4.8. RECORDS 6.5.2.10 Records of OSSRC ac vities s all be prepared, approved and distributed as indicated below:
- a. Minutes of each OSSRC m ing shall be prepared, approved and '
forwarded to the Vice P ident - Nuclear Energy within 14 days following each meetin . ;
)
- b. Reports of reviews neompass by Section 6.5.4.7 above, shall be :
prepared, approv and forwar d to the Vice President - Nuclear ! Energy within 1 days following completion of the review. ,
- c. Audit report encompassed by Sect n 6.5.2.8 above, shall be fonearded t the Vice President - clear Energy and to the managemen positions responsible for he areas audited within 30 days aft completion of the audit. i i
6.6 REPORTABL EVENT ACTION 6.6.1 The f 11owing actions shall be taken for REPORT LE EVENTS:
- a. he Comission shall be notified and a report s mitted pursuant l to the requirements of 10 CFR 50.73, and Each REPORTABLE EVENT shall be reviewed by the POSR and the results of this review shall be submitted to the OSS and the Vice President - Nuclear Energy.
CALVERT CLIFFS - UNIT 2 6-15 Amendment No. -149-l l
6.0 ADMINISTRATIVE CONTROLS' O@ .4. 7 SAFETY L "!T VIOLATIO"
.7.1 sh f ll:Tini ::I$0n3 N: b IN On #"
$b0 v:nI : b:f I) biOiI it
,u t. a.l. .m e.n. A.
pg g - - Th; facility :h:ll b: ple :d in :t-1;;;t ".0T ST."."O7 with!; :n: (see , h / 2-I
@ b. The NRC Operations Center shall be notified by tele hone as soon as possible and in all cases within one hour. The Vice
@ President - Nuclear Energy and the OSSAE shall be notified within 24 hours. ofhite re d o feetion
@ c. :
d, m A Safety Limit Violation Report shall be preparedy The r:p rt
@ N:ll b; reviered by th: POSRC. This 7:::rt th:1! de:Cribe (1) :pplicab!e circu~tance: prc: ding t; viol: tion (2) effec 4s of the vie!ation uper facility : mponents, :ystems-er-structures, and (3) correctiv: :: tion taken to preventwecur+ence.
*N
- g. The Safety Li-it Yiclet4:n R; pert-shall be) submitted to the ?
{o Comission, the GF6AC and the Vice President - Nuclear Energy within 14 days of!the violation.
%{y.s;fe eeview /1mch*cn ~
4-O4I 6.4- PROCEDURES
- . 4- ;) 6.E.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
- a. The applicable procedures recomended in Appendix A of Regulatory 4 (, Guide 1.33, Revision 2, February 1978.
, inse. f 5
.. P,efueling Oper:tient.
tA
- . Surveillance and test activitie: Of cafety related-equipment. ,
@ dAeuri4y-Plen-finpl+ ment +t4on,,
.' aAergency-Plan-implementatien-g f. Fire Prot;; tier. Progr:m implementat4 ens
@ /. c The amount of overtime worked by plant staff members performing safety related functions must be limited in accordance with the I
NRC Policy Statement on Working Hours (Generic Letter No. 82-12). [ L X h PCSRC is Only required to review fir: Protection procedures-and ch:nges theret: which affest-nucle:r ::fety CALVERT CLIFFS - UNIT 2 6-16 Amendment No. 149
0 InSeet 5(see p. 6-u) i I h b. The requirements emergency of bratindbarocedures RE lement the rekuired to * [upplement 1 737 and NU EG-0737
- as stated in Generic Letter 82-33; and
?
- c. All programs specified in Specification 6.5. ,
( 4 I f
}
I i t r e - - - - - -, -, , - - - - ,,
1 l 6.0 ADMINISTRATIVE CONTROLS ! Nh h. PROCESS CONTROL PROGRAM implementation. h 1. OFF5ITE DOSE CALCULATION MANUAL implementation. h 6.8.2 classes of PlantGeneralManagermaydesignates$ytheProceduecific cedures in writing to be reviewed or Reviewprocedur ! Committee or Qualified Reviewers in lieu of review by the SRC. Review i by the Procedu Review Comittee shall be in accordance h ; paragraph 6.5.2. eview by Qualified Reviewers shall b n accordance with ! paragraph 6.5.3. , l 6.8.3 Procedures list in 6.8.1 shall be approv by the Plant General ! agers, Superintende , or General Supervisors i Managerorby) (orDirectors that cognizant report rectly to a Man r prior to implementation as , specified by administrative ufrements. e approval authority for l specific procedures or classes f >roced s shall be designated in writing i by the Plant General Manager and iall e a different individual from the Qualified Reviewer. i I 6.8.4 Each procedure of 6.8.1 ve a changes thereto shall be reviewed' periodi'c ally as set forth in inistrat procedures. > t 6.8.5 Temporary changes procedures of 6. . above may be made provided: i
- a. The intent the original procedure is altered. !
'.) b. The c ge is approved by two members of the nt management sta , at least one of whom holds a Senior React Operator's !'
ense on the unit affected.
. The change is documented, reviewed by the POSRC, the Pro dure l Review Comittee, or by a Qualified Reviewer and approved the :
5 o f/rv __ __ designated approval authorityg4 days of implementatio _ { 5 j
!" Sed
- b 1 bec P 6-30, 2 bee p. s-3o), 9 (.see p. g.Q , o ; , q
~
6.J6 REPORTING REQUIREMENTS - R W NF-AEPS E j
@ ,oi r . aa m . . . . u. . . u ,. . u . _ _ _ _ _. . _
cm cr.xs L.L.,;.;<rs m ,rinz w mim :lJ;nM' _. . , _ - _ . . . , m , - .. DahibN55 b5ibE'b[ N5-bb5Ibb' 5[Ibfbh:5 55t bb5:$5 th....". .".
. . ...'....N f The fo llo winy repo.4.s .s h ll be .s ub m /H e,/ in a c-c o < h n c e wi t h 10 C Fit 50. + , 1 l
i j S4 . m-J
.gs (Inse;fis9 fooinole (see p. 6-3o ; g 29)
CALVERT CLIFFS - UNIT 2 6-17 Amendment No.149-
i l i l 1
)
f MSe rl 6 (.S e e p. 6-17) l 1 6.5 PROGRAMS AND MANUALS i The following programs shad be established, implemented, and maintained. 6.5.1 Offsite Dose Calculation Manual (ODCM) l \ - 1 t a. The ODCM shall contain the methodology and parameters used in the calculation l 1 of offsite doses resulting from radioactive gaseous and liquid effluents, in the ' / calculation of gaseous and liquid effluent monitoring alarm and trip set in the conduct of the Radiological Environmental Monitoring andProgram; points,
- b. The ODCM shall also contain the radioactive effluent controls and radiological k environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Re Specification 6.63, respectively. ports required by Specification 6.6.2 and i
-i 7 in.se<1 7 (see p. 6-17)
N -
- c. Ucensee initiated changes to the ODCM: !
- 1. Shall be documented and records of reviews performed shall be retained. :
This documentation shall contain. 1 i I (a) Sufficient information to support the change (s) together with the appropriate analyses or evaluationsjustifymg the change (sk : f (b) A determination that the change (s) maintain the levels of radioactive ' I effluent control rec uired by 10 CFR 20.1302,40 CFR 190,10 CFR i 50.36a, and 10 CFR 50, A ndixI, and not adverse , accuracy or reliability of e uent, dose, or setpoint cagculations; impa
- 2. Shall become effective after review and acceptance by the onsite review
./
1 function and the approval of the plant manager; and
- 3. lete,le ible ofthe /
Shall be submitted entire ODCM to the NRC as part of or concurrent in Rad with the the form of a comfonctivekffluIn i ! Report for the period of the report in which any change in the ODCM was j ; 3 made. Each change shall be identified by markmgs in the margin of the ; affected pages, clearly indicating the area of the page that was changed, and ; shall indicate the date (i.e., month and year) the change was implemented. / I
-- e
l l jn.Se d te (.see f.I,-l7) 6.5.4 Technical Specifications (TS) Bases Control Program h This program provides a means for processing changes to the TS Bases. 057 L a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b. Licensees may make changes to Bases without changes do not involve either of the following: prior NRC approval provided th i
- i. A change in the TS incorporated in the license; or i i ii. A change to the UFSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.
- c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR
- d. Proposed changes to the TS incorporated in the license orproposed changes to the UFSAR or Bases that involves an unreviewed safety question shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented j without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
h d !9 (.see p.g m l x _ h 6.5.5 Radioactive Effluent Controls nogram l This conforms to 10 CFR 50.36a for the control of radioactive effluents and for ! t [ main reasonab the doses to members of the public from radioactive effluents as low as ievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the , j program limits are exceeded. The program shall include the following elements: ;
- a. !
f Umitations on the functional monitoringinstrumentationincl ca$ ding surveillance tests and setyomta 1 determination in accordance with the methodology in the ODCM;. !
- b. Umitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 CFR 20, Appendix B, Table ,
II, Column 2; f
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in sccordance with 10 CFR 20.1302 and with the methodology and !
parameters in the ODCM; j
- d. Limitations on the annual and guarterly doses or dose commitment to a !
/ member of the public from radioactive materials in liquid effluents released L
from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;
- e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar ;
j year in accordance with the methodology and parameters in the ODCM at i T least every 31 days; ;
]
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these j i systems are used to reduce releases of radioactivity when the projected doses [ j in a period of 31 days would exceed 2% of the guidelines for 11e annual dose 1 or dose commitment, conforming to 10 CFR 50, Appendix I; j f g. Limitations on the dose rate resulting from radioactive material released in !'
gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR 20, Appendix B, Table II, Column 1, :.
- h. Limitations on the annual and quarterly air doses resulting from noble gases I released in gaseous effluents from each unit to areas beyond the site -
boundary, conforming to 10 CFR 50, Appendix I; f
- i. Limitations on the annual and quarterly doses to a member of the ublic from / l iodine-131, iodine-133, tritium, and all radionuclides in particulate orm with l halflives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and
- j. Limitations on the annual dose or dose commitment to any member of the /
public due to releases of radioactivi y and to radiation from uranium fuel cycle sources, conforming to 40 CF 190. ! I
i 6.0 ADMINISTRATIVE CONTROLS j O MARTUP REPORT i l 6.9. 1 A sunnaryfollowing report of p(lant startup and power escalation testi shall submitted 1) receipt of an operating license, l nt to the license involving a planned increase in p level, ) (2) amen (3) instal tion of fuel that has a different design or has ! manufacture y a different fuel supplier, and (4) modifici ons that may i have signific ly altered the nuclear, thenna1, or hydr ic perfonnance of the plant. l 6.9.1.2 The startup port shall address each o he tests identified in Ogf the FSAR and shall inc e a description of t measured values of the i operating conditions or c racteristics obt ed during the test program I r and a comparison of these y es with de n predictions and ! specifications. Any correcti actiorcescribed. hat were required to obtain i satisfactory operation shall als Any additional specific i details required in license cond based on other connitments shall be included in this report. j all be submitted hin (1) 90 days following h 6.9.1.3 Startu reports completion of t e sta p test program (2) 90 "s following resumption or rcial . power operation, or (3c months following i commencement of co ' initial critic e events y, whichever is earliest. If the TUP Report does not cover all t (i.e., initial criticality, comp ufon of STARTUP test pro m, and resumption or connencement of connerciaL ower oper on), supplementary reports shall be submitted at leas very three m) manfis until all three events have been completed. l C.b } Occ v pdiom l Ruliafton Expo.swe Repod ' ANN'J^.L "E"^"TS il p. 6-iq{
)
v:rin;; th: ;;tivitic; cf th: unit :: deteribed [ eporty::::had:r y::r shall be submitted prior to March 31 of bel:" f:r the prt'ti::: The initt:1 r:p:rt : hell b: ::b:!tted prier t: M:rch 1 Of th: l ('eachyear.
.... <.iiom4o, 4.m.i..,-u..
----_n
- m_- .
h 5.".1.5 R: pert; 7:q;;r:d :: :n :n ,;;; beei; shall L JJe. )
@ f.p A tabulation on an annual basis of the number of station.
utility, and ot4er personnel (including contractors) receiving l l t exposures exposure according ;r::t., th:n to work 100 andmrem job /yr and their functions E, e.g.,Js(sociated reactor man rem j he Jdepe.eled spe.,+feeI.sfua,elsMk& f ' rePode/ o f u.opaticeI of0.se Fo ?(E.SES1)wt/I s epmfe ty,be c.alved Cliff.s 1/ A single submitti l may be made for : titiple unit :t:tica. The submittal should combine those sections that are common to g units, l zn.5e4 iz at tt: :t:tir h _ _ _ _ _ M [ This tabulation supplements the requirements of 10 CFR 20.407.
~- _--_ __
.u o CALVERT CLIFFS - UNIT 2 6-18 Amendment No. 449-1
,s F **'*g *'
. t fh*I*"I"*scus l
. 6.0 ADMINISTRATIVE CONTROLS
( t(see
& <tia_ p. g-l__g dosime fee .
Wh operations and surveillance , inservice inspection, maintenance, special maintabance (describe maintenan e), waste utine l The dose assignment to v ious duty processing, functions may and refueling)* based on pocket dosimeter, be estimatis , or - O47 ff 5 had:- - rr:c: :.tr. Small exposures totalling 20% . of the individual total dose need not be accounted for. In the , aggregate, at least 80% of the total whole body dose rec !
@g from external sourcet shah be assioned to specif or work i functions. b,.see t ]) S oolof . e.leefr ic pe<jona) jn#ef$ $) '**Y
*g ag)
.. b. The : il:t O n ltf ef it r ;= =t r teh f r :=f:: i (Se* ,i=l::ti=: pr,f:.=d
___,n-a,_ , dr' ; the E -:-+ pried (--f:= = e 5,g - C. 4.1, Pee.ssunz efPoR V.a svY,V'N'Jt P' S
% {peumenta44Wpf all failures and challenges to the pressurizer ,
AN yed C" PORVs reprefor safety
.shall bevalves.
su bml++ed prie< io Ma<< 4 I of each yeae (,see P' y 6
= - - --
~- i h a.fi. MONTHLY w =
OPE < dM ~
~
6.9.1.6 Routine reports of operating statistics and shutdown experience i j shall be submitted on a monthly basis t th: "in:t:r, Offf = :f !=;=ti= 1 nd Enf;=: ::t, ".0.1:1.7 %;;ht:ry 01 !='=, "=hin;t=, ".0. l (I 20555 ^. " : 0 r =t 0;r.t=1 "=h, with : r;" te the ".:;i =? i ( 9fetet=t= =d t: the ""O "=td=t =p=3r, no later than the 15th of )
.~ \ each month following the calendar month covered by the~ report. i 3 - -
l 6.6. A
# A ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- The AnnoI d
S.0.1.7 "=tfee Radiological
. Environmental Operating ReportY covering the operation of the unit during the previous calendar year shalT be submitted prior to May 1 of each year. g,, ; f,,. f.,, g, ,,,m The " =1 ":dich;i=1 E=tr- __:te! 0- atf e; pporty'shall include summaries, interpretations, l
@ +gdlological .. ! edit; 2.ce- Environmental
!== t'th ;-
re:illence ;;th'tf=f
==tf:=1 r fordthe anreport"$eriody ted!=, rf th :;;=tt=:1 analyds ,of tren O74 j r t-th = :;;;=;r = :t:, =d rit; ;=;' = e- 4=7. =t: 1 2 ---ai! r--
s I
= ; thrt:,=';' = d rn- = n :=.t == :t :f =th:; ;in Ofth:
th:- ;1=t e;re+4a=
-- . Th: = 2r;;d
- h: i =t:11 :!=' 3:hd: r!te ef 1=ad ==
' e r = = = ;f = d by Sp =ifi et k ,0.12.2.
h7 The ".==1 %dSI ;i=1 Ociin:=:t:1 Oa^-etfe; hport)( shall include the results of analysh of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the lable and gigures in the ODCM, as well as sumarized and tabulated results of these analyses and measurements in the { ryne ma le,-tal p<ovodel .shall be e,ons ste ,f wi+h the oh
~
out lined I., th e off.syc Dese Calevlef **o*r Ma nual Co.0ca t%jeefim n,(
, in 10 t FM So, Appe.vfW 4 Sectfo*t.> 1E'.B.2, ye'. 3 3, ,,q y ( ,
d@
- A single submittal may be madey for &fged Cli/[>, The.
.s ubmit ta l .s ho uld c o m bin e common to bot k us. ;f.s .
the.se seeiton s + k ,,f a v e. CALVERT CLIFFS - UNIT 2 6-19 Amendment No. 449-
. % , w ,n, - - - , - -
, , .. x = . - , =
6.0 ADMINISTRATIVE CONTROLS l fonnat of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. : The missing data shall be submitted as soon as possible in a supplementary report. T orts shall also include the following: a sumary description ofps radiolo environmental monitoring program; at least two legib ing locations keyed to a table giving di'st s and covering all ent buildings; directions from the ral point between the two cont cipation in the Inte oratory Comparison the results of licensee p Program, required by Specifica n 3.12.3; ; and discussion ssion of all deviations from the sampling schedule of Tab of all analyses in which the LLD required by Tabi .
- as not achievable.
The Annual Radiological ronmental 0)erating Re will include the cause of unavaila)ility of sample fany),andwill identification of t ons used for the replacement samples. T eport will describe the 1 uld also inclu any pemanent changesItinwill the sample locations whic include 'a revised figure and ap)ea
~
the monitoring program. ta e for the ODCM reflecting the new location (s). c.6.3 ' O77 ^ SEP!?="AL RADIOA.CTIVE. + o EFFLUENT a REU:ASE REdORT** w e
) 5.".1.3 9:uPine Radioactive Effluent Release Resorty covering the operation of the unit during " ^
the' prevf eu:"'"'"'d-" 5 mentu ef :p:r:tten shall be
@ '"bmitted;","' MGs"i',2WWUWEb..'Q.e'.]Vida
, }
The n;dic::tive Effluent Rele::e ReportX shall inciu e a**o me t sum ry
/
ledn
~ ef&ie'd of t 12 **r*r thJ '
quantities of radioac ;ve liquid and gaseous effluents and solid waste
* ':: eutifned ia Regu! tery Cufde 1.21. "M:::uring, released from the Ev:!ucting, and Rep rting Radioactivity 4" (alid M :te and Releste: ef !
R: die::tive Matert:1: 1" Liquid and C::::u Ef#1 ent: frem Light M:ter C cled Meclear Dewer ."l:nt:," " vi:i n 1, June 1974, -f th d:ta curart::d en : qu:rter!y b::f: fe!!=/ing the f=:t :f App;ndix S ther::f. ohjeeft m
-l he .ma lerial peo n de .shall he con.sidewf w:th th e out lin e / in -t h e O D c M anoi Peaceu confol Proy a m and
;n con fo, mnca mY4 je cyg sqq3nd to WSs Appe~hI'
@ .seet,'on 1e . s.t. In. sea 13Q p.i.si)
+
N Om nw shall : ver :t ti;ns neer the !!!: "0"=2Y, e ::::nd :h;11 include the mora distan+ ctatf:nt. A single submittal may be made for Calsert Cliffs, since the Radwaste Systems are comon to both units. N In 11:e f s"hm4"f er wi+h +he ?:-f Annu; Rep;rt:, cr, Sr" ' analy::: re: ult: may be submit +ad <- a cup)!emen+ery report wf tM-L 120 days after knu:ry 1 and July-1 ef ::: year, fH Th e, po<f n.sha naccorda ll a Iso in elude e.h w9es % H,, c e wi+k Specifica tien 6, 5. l c . p p ( ,4 on i CALVERT CLIFFS - UNIT 2 6-20 Amendment,No. 149
)
i
l 6.0 ADMINISTRATIVE CONTROLS Th Th Radioactive Effluent Release Report to be submitted within 60 days
- aft January 1 of each year shall include an annual summary of hourly ry met logical data collected over the previous year. This annual s i may be ither in the fom of an hour-by-hour listing on magnetic ta of I wind sp d, wind direction, atmospheric stability, and precipitatto (if !
measured or in the form of joint frequepcy distributions of win speed. l wind direc ion, and atmospheric stability . This same report sh j include an sessment of the radiation doses due to the radioac ve liquid and gaseous fluents released from the unit or station durin the previous calendar year. The assessment of radiation doses shall be p formed in- . accordance with he methodology and parameters in the OFFSI DOSE l CALCULATION MANU (0DCM). ! The radioactive Ef ent Release Report to be submitted 0 days after ! Ja'nuary 1 of each ye shall also include an assessmen of radiation doses j to the likely most exp sed MEMBER 0F THE PUBLIC from eactor releases and other nearby uranium fu 1 cycle sources, including ses from primary l effluent pathways and di ct radiation, for the pr tous calendar year to t show conformance with 40 R Part 190, Environme al Radiation Protection I Standards for Nuclear Powe Operation. Accepta e methods for calculating the dose contribution from 1 uid and gaseous fluents are given in Regulatory Guide 1.109, Rev. October 1977 and NUREG-0133, " Preparation ! of Radiological Effluent Techn al Specific ions for Nuclear Power l Plants." ; The Radioactive Effluent Release R ort shall include the following infomation for each class of solid te (as defined by 10 CFR Part 61) : lj d: l shipped offsite during the report pe
- a. Container volume,
- b. Tnal curie quantity specify wh ther detemined by measurement !
orestimate), i
- c. Principal radion lides (specify wh her detemined by measurement or stimate), i I
- d. Source of w te and processing employe (e.g.,dewateredspent Il resin, com cteddrywaste,evaporatorbttoms),
- e. Solidif ation agent or absorbent (e.g., c ent). !
( The Radioactiv Effluent Release Reports shall include list and { description o unplanned releases from the site to UNR RICTED AREAS of l' radioactive terials in gaseous and liquid effluents ma during the reporting p tod. n lieu of submission with the first half year Radioacti Effluent Release Report, this summary of required meteorological da i may be retained on site in a file that shall be provided to the NRCpon d request. 3 CALVERT CLIFFS - UNIT 2 6-21 Amendment No.149-- l
,- -,_,-.-w,,,_r ,.r ,-. m ,, - - - - - , - -_m -- -
6.0 ADMINISTRATIVE CONTROLS h ctive Effluent Release Reports shall include an during the re od to the PROCESS CON neer 1mide PCP), and to
@ the 0FFSITE DOSE CALCULATI N
~
locations for dose a (C% as well as a listing of new ons identificii ensus pu p cation 3.12.2. 17 h ' S.0.1.0- CORE OPERATING LIMITS REPORT (c 0z.R) (Seep.6-I9) b.65 a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: 2.2.1 3.1.1.1 3.1.1.2 , 3.1.1.4 3.1.3.1 3.1.3.6 ' 3.2.1 3.2.2.1 3.2.3 3(
/
3.2.5 3.9.1
- b. The analytical methods used to detennine the core operating limits shall be those previously reviewed and approved by the
- NRC; specifically, those described in the following documents:
(1) CENPD-199-P, Latest Approved Revision, "C-E Setpoint Methodology: C-E Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems," January 1986 (2) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 1: C-E Calculated Local Power Density and. Thennal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II," December 1979 (3) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 2: Combination of System Parameter Uncertainties in Thennal Margin Analyses for Calvert Cliffs Units 1 and 2," January 1980 (4) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 3: C-E Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for Calvert Cliffs Units 1 and 2," March 1980 (5) CEN-191(B)-P, "CETOP-D Code Structure and Modeling' Methods for Calvert Cliffs Units 1 and 2," December 1981
.~: '
u ) CALVERT CLIFFS - UNIT 2 6-22 Amendment No. 43
l 6.0 ADMINISTRATIVE CONTROLS (6) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr. , (BG&E), dated June 24,1982, Unit 1 Cycle 6 License Approval l (AmendmentNo.71toDPR-53andSER) .
-(7) CEN-348(B)-P, " Extended Statistical Combination of l Uncertainties," January 1987 i i
(8) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated October 21, 1987 Docket Nos. 50-317 and 50-318, " Safety Evaluation of Topical Report CEN-348(B)-P, Extended Statistical Combination of Uncertainties" . (9) CENPD-161-P-A, " TORC Code, A Computer Code for Detemining the Themal Margin of a Reactor Core," April 1986 l (10) CENPD-162-P-A, Latest Approved Revision, " Critical Heat Flux ' Correlation of C-E fuel Assemblies with Standard Spacer Grids Part 1 Unifom Axial Power Distribution" , (11) CENPD-207-P-A, Latest Approved Revision, " Critical Heat Flux ' Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 2, Non-Unifom Axial Power Distribution" (12)CENPD-206-P-A,LatestApprovedRevision,"TORCCode, E Verification and Simplified Modeling Methods" (13) CENPD-225-P-A, Latest Approved Revision " Fuel and Poison
~) Rod Bowing" (14) CENPD-266-P-A, Latest Approved Revision, "The ROCS and DIT Computer Code for Nuclear Design" (15) CENPD-275-P-A, Latest Approved Revision, "C-E Methodology I for Core Designs Containing Gadolinia - Urania Burnable Absorbers" . .
(16) CENPD-382-P-A, Latest Approved Revision, "C-E Methodology for Core Designs Containing Erbium Burnable Absorbers" (17) CENPD-139-P-A, Latest Approved Revision, "C-E Fuel Evaluation Model Topical Report" (18)CEN-161-(B)-P-A,LatestApprovedRevision,"Improvementsto ! Fuel Evaluation Model" (19) CEN-161-(B)-P, Supplement 1-P, " Improvements to fuel Evaluation Model," April 1989 CALVERT CLIFFS - UNIT 2 6-23 Amendment No.1&- F= *,e,= +K ,*** 4 - --
i
- 6.0 ADMINISTRATIVE CONTROLS w
) (20) Letter from Mr. S.. A. McNeil, Jr. (NRC) to.Mr. J. A. Tiernan I (BG&E), dated February 4,1987. Docket Nos. 50-317 and - !
50-318,'" Safety Evaluation of Topical- Report CEN-161-(B)-P, Supplement 1-P, Improvements to Fuel Evaluation Model" (21) CEN-372-P-A,- Latest Approved Revision " Fuel Rod Maximum , I Allowable Gas Pressure" r ('22)~ Letter from Mr. A. E. Scherer (CE) to Mr. J. R. Miller (NRC), dated December 15,1981, LD-81-095 Enclosure 1-P. !
"C-E ECCS Evaluation Model Flow Blockage Analysis" (23) CENPD-132. Supplement 3-P-A, Latest Approved Revision, .
" Calculative Methods for the C-E Large Break LOCA Evaluation l Model for the Analysis of C-E and )( Designed NSSS" )
.I (24) CENPD-133, Supplement 5. "CEFLASH-4A, a FORTRAN 77 Digital .j Computer Program for Reactor Blowdown Analysis," June 1985 :
(25) CENPD-134, Supplement 2 "COMPERC-II, a Program for p . Emergency Refill-Reflood of the Core," June 1985 l i (26) Letter from Mr. D. M. Crutchfield (NRC) to Mr. A. E. Scherer (CE), dated July 31,1986, " Safety Evaluation of Combustion - l Engineering ECCS Large Break Evaluation Model and Acceptance- l m for Referencing of Related Licensing Topical Reports" .i O . (27) CENPD-135, Supplement 5-P, "STRIKIN-II, A Cylindrical j Geometry Fuel Rod Heat Transfer Program," April 1977 i i (28) Letter from Mr. R. L. Baer (NRC) to Mr. A. E. Scherer (CE), . dated September 6,1978, " Evaluation of Topical Report CENPD-135, Supplement 5" . f (29) CENPD-137, Supplement 1-P " Calculative Methods for the C-E l Small Break LOCA Evaluation Model," January 1977 l i (30) CENPD-133, Supplement 3-P, "CEFLASH-4AS, A Computer Program j for the Reactor Blowdown Analysis of the Small Break Loss of ; Coolant Accident " January 1977 (31) Letter from Mr. K. Kniel' (NRC) to Mr. A.' E. Scherer (CE), : dated September 27,1977, " Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P" , i (32) CENPD-138, Supplement 2-P " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977 ! l CALVERT CLIFFS - UNIT 2 6-24 Amendment No. IG l l~ a
~
t 6.0 ' ADMINISTRATIVE CONTROLS 3 i (33) Letter from Mr. C. Kniel (NRC) to Mr. A. E. Scherer, dated April 10,1978, " Evaluation of Topical Report CENPD-138, Supplement 2-P" (34) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. J. R. Miller (NRC) dated February 22,1985, "Calvert ' Cliffs Nuclear Power Plant Unit 1; Docket No. 50-317, Amendment to Operating License DPR-53. Eighth Cycle License l Application" , (35) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr. (BG&E), dated May 20, 1985 " Safety Evaluation Report ! Approving Unit 1 Cycle 8 License Application" l (36) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. R. A. Clark (NRC), dated September 22, 1980, " Amendment , to Operating License No. 50-317, Fifth Cycle License ! Application" g ! (37) Letter from Mr. R. A. Clark (NRC) to Mr. A. E. Lundvall, Jr. ! (BG&E), dated December 12,1980, " Safety Evaluation Report : Approving Unit 1, Cycle 5 License Application" ! t (38) Letter from Mr. J. A. Tiernan (BG&E) to Mr. A. C. Thadant (NRC), dated October 1,1986, "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2 Docket Nos. 50-317 & 50-318 Request
-) for Amendment"
( ' i (39) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BGLE), dated July 7,1987, Docket Nos. 50-317 and 50-318, Approval of Amendments 127 (Unit 1) and 109 (Unit 2) (40) CENPD-188-A, Latest Approved Revision "HERMITE: A Multi- : Dimensional Space-Time Kinetics Code for PWR Transients" l [ (41) The Full Core Power Distribution Monitoring System referenced in Specifications 3.1.3.1, 3.2.2.1, 3.2.3, and the BASES is described in the following documents: (a) CENPD-153-P, Latest Approved Revision, " Evaluation of Uncertainty in the Nuclear Power Peaking Measured by i the Self-Powered, Fixed Incore Detector System" (b) CEN-199(B)-P,"BASSS,UseoftheIncoreDetectorSystem to Monitor the DNB-LCO on Calvert Cliffs Unit 1 and ' Unit 2," November 1979 on e . . w) CALVERT CLIFFS - UNIT 2 6-25 Amendment No. 18
1 6.0 kDMINISTRATIVE CONTROLS I 1 D (c) Letter from Mr. G. C. Creel (BG&E) to NRC Document l Control Desk, dated February 7,1989, "Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket 50-318. Request for Amendment, Unit 2 Ninth Cycle License Application" (d) Letter from Mr. S. A. McNeil, Jr. (NRC) to - Mr. G. C. Creel (BG&E), dated January 10, 1990, " Safety Evaluation Report Approving Unit 2 Cycle 9 License Application" , y
- c. The core operating (limits shall be detemined such that r a aplicable limits tiemal hydraulic limits, Emergency Core Cooling Systems (ECCS) '
limits, nuclear limits such as SDM, transient analysis limits - and accident analysis limits) of the safety analysis are met. i
- d. The COLR, including any mid-cycle revisions or supplements, shall
__ be provided upon issuance for each reload cycle to the NRC. I"M d R SS'c P.A ~ g ShCTA[R'EP0RTS_ cial reports shall be submitted to the Regional Administra ) ort. r of l 6.9.2 ., the NRC Re onal Office within the time period specified for each - These report shall be submitted covering the activities identiication: ed aelow pursuant to th equirements of the applicable reference spec in i "V
- a. ECCS Actu on, Specifications 3.5.2 and 3.5.3.
- b. Inoperable Sei ic Monitoring Instrumentati ,
Specification 3. . 3. Inoperable Meteorolog al Instrument on, Specification 3.3.3.4. c.
- d. Seismic Event Analysis, 5 cific on 4.3.3.3.2. .
- e. Core Barrel Movement, Spect tion 3.4.11.
- f.': Fire Detection Instrum ation, S cification 3.3.3.7. j tems, Specificati 3.7.11.1, 3.7.11.2,
- g. Fire Suppression 3.7.11.3, 3.7.1 . , and 3.7.11.5. l i
- h. Penetratio tre Barriers, Specification 3.7. 2. t
- i. Steam nerator Tube Inspection Results, Speciff tion 4.4.5.5.a and .
J. pecific Activity of Primary coolant, Specification 3. . Containment Structural Integrity, Specification 4.6.1.6 _ , 6-26 Amendment No. 143 CALVERT CLIFFS - UNIT 2 X , y es:.2QW:'.Cp;f. .a
i 6.0 ADMINISTRATIVE CONTROLS 3 1. Radioactive Effluents - Calculated Dose and Total Dose, Specifications 3.11.1.2, 3.11.2.2, 3.11.2.3, and 3.11.4. ; m Radioactive Effluents - Liquid Radwaste, Gaseous Radwa e and Ventilation Exhaust Treatment Systems Discharges, ecifications 3.11.1.3 and 3.11.2.4.
- n. Ra ological Environmental Monitoring Program S cification 3.1 .
- o. Radiat n Monitoring Instrumentation, Speci cation 3.3.3.1 (Table 3 -6).
- p. Overpressu Protection Systems, Specif ation 3.4.9.3. I
. q. Hydrogen Anal ers, Specification 3. 5.1. .
- r. Post-Accident In trumentation Spe ification 3.3.3.6.
@ 6.10 RECORD RETENTION 6.10.1 The following records sh 11 b retained for at least five years: ,
- a. Records and logs of faci ty operation covering time interval at each power level.
, b. Records and logs of incipa maintenance activities, inspections, repair and repla ment of principal items of ,
- equipment related o nuclear s fety.
4 I
- c. All REPORTABLE ENTS.
- d. Records of s veillance activities, inspections and calibrations l required by these Technical Specific tons. ;
i
- e. Records reactor tests and experiment j
,, f.. Recor of changes made to Operating Proce ures.
1
- g. Re rds of radioactive shipments.
l
- h. cords of sealed source and fission detector 1 k tests and I esults.
. Records of annual physical inventory of all sealed s rce material of record. i N/ .
CALVERT CLIFFS - UNIT 2 6-27 Amendment No. 143- M
.. *" m n
t 6.0 ADMINISTRATIVE CONTROLS-
- 6. 0.2 The following records shall be retained for the duration of the .l
]' ' ' Fa lity Operating License - ,
Records and drawing changes reflecting facility design modifications made to systems and equipment described in' e ! Final Safety Analysis Report. !
- b. R ords of new and irradiated fuel inventory, fue1 ~ nsfers'and i ass ly burnup histories, j
- c. Recor of facility radiation and centaminatio urveys. .j
- d. Records o radiation exposure for all indivi uals entering :
radiation ntrol areas. {
- e. - Records of. ga ous and liquid radioacti e m&terial released to .
the environs.- 'l
- f. Records of transi t or operationa cycles for those facility .;
components identif d in Table 5 .1. j I
- g. Records of training a d quali cation for current members of the plant staff.
1
- h. Records of in-service ins ctions perfonned pursuant to these Technical Specification . ;
- 'T l ic7
- 1. Records of Quality A urance tivities identified in the NRC l
~
approved QA Manual s lifetime ecords.
- j. Records of revie performed for anges made to procedures or equipment or r iews of tests and e eriments pursuant to 1 10 CFR 50.59. .i
't
- k. Records of eetings of the POSRC, the P cedure Review Comittee, -
and the O RC.
- 1. Recordt of the service lives of all safety lated snubbers
.. . inclu ng the date at which the service life . nces and asso iated installation and maintenance record 6.11 RADI ION PROTECTION PROGRAM Procedu s for personnel radiation protection shall be prepared onsistent with e requirements of 10 CFR Part 20 and shall be approved, ma tained and hered to for all operations involving personnel radiation ex sure.
- CALVERT CLIFFS - UNIT 2 6-28 Amendment No. & M OO8 g g9 . g T4 #
l
t 6.0 ADMINISTRATIVE CONTR6LS 3g 2 HIGH RADIATION AREA , j
.6.12.1 lieu of the " control device" or 'alam signal by !
10 CFR Pa 0.203(c)(2): - l
- a. A hig diation area in which the intens of radiation is v greater 100 mren/hr but less than ares /hr shall be - -
.i barricaded a conspicuously posted a High Radiation Area and ;
entrance thereto all be contro d by issuance of a Special or !
. Radiation Work Pem and any ividual or groep of individuals !
pemitted to enter suc shall be provided with a radiation ! monitoring device which nuously indicates the radiation dose rate in the area. !
- b. A high radiatto rea in which the in ity of radiation is :
greater tha arem/hr shall be subjec the provisions of i 6.12.1.a ove, and in addition locked barr es shall be l provi to prevent unauthorized entry into such as ana the '! ke shall be maintained by the Supervisor - Radiati ontrol !
>erations and the Operations Shift Supervisor on duty u r- :
t1eir separate administrative control. ! h 6.lYS4H9HM6GRf4Y Pei muy ca./o.t Some.s N.s.'Je canfai~med b l
.The licensee shall im>1ement a progra/to reduce leakage from systems (
.)
/ outside containment t sat would or could contain highly radioactive fluids-
>during a serious transient or accident to as low as practical levels. This program shall include the following:
- a. Provisions establishing preventive maintenance and periodic visual inspection requirements, and
- b. Leak test requirements for each system at a frequency not to exceed refueling cycle intervals. i
- q,, f 9 h h MONITORING (see P. 4-/ 7)
The' licer$see s ram whic sure the e ability , to accurately detemine the air >o concentration in vital areas-4 under accident condition program sha the following: d It is acceptable if the licensee maintains details of the program in
@ plant operation manuals (e.g., chemistry procedures, traininf
instructions,maintenanceprocedures.ERPIPs).
=
CALVERT CLIFFS - UNIT 2 6-29 Amendment No. MS ,.Y
.in.s e<f 9 fednof e ls'e P 6-I 7) . . ,.
l I 6.0 ADMINISTRATIVE CONTROLS i a. of personnel, I
- b. Procedures for monito. fp g,F-s y)
- c. Prov maintenance of sampling an equipment. i ENT SAMRLING f The licensee shall establish, implement and maintain a program
- which will ensure the capability to obtain and analyze reactor coolant, radioactive i ,
iodines and particulates in plant gaseous effluents, and containment . I katmospheresamplesunderaccidentconditions. The program shall include ' the following:
- a. Training of personnel, ;
- b. Procedures for sampling and analysis, and i i
I c. Provisions for maintenance of sampling and analysis equipment. j i -- _ -- _ - ~ m - _
- 6. 6 PROCESS CONTROL PROGRAM (PCP) 6.16.1 e PCP shall be approved by the comission prior 5 implementa n. ,
.')
6.16.2 Licensee
- itiated changes to the PCP:
- a. Shall be su ted to the Comiss in the Semiannual l Radioactive Ef nt Release R rt for the period in which the i change (s) was made. This su ttal shall contain- !
- 1. An evaluation su the premise that the change did not !
reduce the overa confo nce of the solidified waste
- i product to ex ing criteri or solid wastes; and l l
- 2. A refer e to the date and the P C meeting number in which ;
the c ges(s) was reviewed and foun cceptable to the i P0 .
- b. S 1 become effective upon review by the POSRC a approval of !
he Plant General Manager. 7 e JnSed 2 (see p. 4-11} 3
]'
h It is acceptable if the licensee maintains details of the pff#am in plant operation manuals (e.g., chemistry procedures, training instructions,maintenanceprocedures.ERPIPs). CALVERT CLIFFS - UNIT 2 6-30 Amendment No. MB d I l i
t 6.0- ADMINISTRATIVE CONTROLS hh .0.17 OTTOITC 000E CA:.00;.AT:0% :"J.:.JAL l0=:0 WP keff,'[,b,f 7- j h 5.17.1 Tim Z:;; et,;11 M ;ppr:nd by th: C __i::ier, prf:r t :
'-P-- ;t tM .
j h'* [fh'Pf e 43 . . Licensee initiated changes to the ODCN:
- a. Shall be submitted to the Connission in the 4em4emwe+ . l ioactive Effluent Release Report for the period which the !
cha e(s) was made effective. This submittal-s contain:- , ;
- . 1 i
ionale for the ent-information to support the
- 1. Suff .
change. nfonnation submitted sho consist of a package of : those page f the 00CM to be c ged with each page numbered ! and provided th a change n er and/or change date together ! nalyses evaluations justifying the r withap(propriat changes); .
) ,
- 2. A detennination at the ange will not reduce the accuracy i tions or setpoint or reliabilit of dose calc determinat s; and-
- 3. Docu tation of the. fact that the c ge has been reviewed !
,. a found acceptable by the POSRC. ,. f , ,,.y gg,.
all become effective' upon review by the POGRC an royal of b.
' Od the~ giant 0:au;lganager. -
9} n . (, S.10 "*J0" 0"'MCES TO "AO n 0*CT:'!: LIGUIG. 0A0COUS AM-30tt1HlASTE T" :AT".:ST SYSTE".S , p _ _
= _
O17 (Griert Licensee initiated major changes to the Radioactive Waste Systems (liquid, gaseous and solid) shall be reported to the Comission in the !
':=.i;;;=1 Radioactive Effluent Release Report for the period in which the
- j '
modification to the waste system is completed. The discussion of each 6 change shall contain: f i
.. a.. A description of the equipment, components and processes-i involved. !
i
- b. Documentation' of the fact that the change. including the safety i analysis was reviewed and found acceptable by the P06AG.
on. site revo'ew lbndien h - l
-[*13 e d G(Jee p,4,-20) j CALVERT CLIFFS - UNIT 2 6-31 Amendment No. 443- I l e .
9 #C [ e*? ' t?
- r 4 -e -- ---m - , - , , - ,
W . 5 7 1 l ATTACHMENT (6) . i f t l s
'\ .
UNIT 1 PROPOSED TECHNICAL SPECIFICATION : PAGES i e t
?
k r l l i I l i
~- . - . . _--,- . . , _ - _ .
. 6.0 ABMINISTRATIVE CONTROL 5 :
i 6.1 RESPONSIBILITY
]
6.1.1 The plant manager shall be responsible for overall facility l I operation and shall delegate in writing the succession to this i responsibility during his absence. j 6.2 ORGANIZATION i 6.2.1 Onsite & Offsite Orcanizations Onsite and offsite organizations shell be established for unit operation > and corporate management, res actively. Thc onsite and offsite 1 organizations shall include tie positions for activities affecting the i safety of the nuclear power plant. j
- a. Lines of authority, responsibility and comunication shall be j established and defined for the highest management levels through ;
intemediate levels to and including all operating organization ; positions. These relationships shall be documented and updated, i as appropriate, in the fom of organization charts, functional : descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in ' , ecuivalent foms of documentation. These requirements, including the plant-specific titles of personnel fulfilling the ; responsibilities of the positions delineated in these Technical ; Specifications, shall be documented in the Updated Final Safety j Analysis Report (UFSAR). i
- b. The plant manager shall be responsib10 for overall unit safe l ;
operation and shall have control over those onsite activities-
- c. Th ce P siden - Nuc1 Energ s all ve c rpo respor.sibility for overall plant nuclear safety and shall take j any measures needed to ensure acceptable perfomance of the staff !
in operating, maintaining, and providing technical support to the i plant to ensure nuclear safety. l 4 d. The individuals who train the operating staff and those who carry ? out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have . sufficient organizational freedom to ensure their independence from operating pressures. i 6.2.2 Unit Staff : i The unit staff organization shall include the following:
- a. A total of at least three non-licensed operators shall be i assigned to the Units 1 and 2 shift crews. ;
I CALVERT CLIFFS - UNIT 1 6-1 Amendment No. !
~ . . .
6.0 ADMINISTRATIVE CONTROLS i
- b. At least one licensed Operator shall be in the Control Room when ,
fuel is in the reactor. , i
- c. At least two licensed Operators shall be present in the Control i Room during reactor STARTUP, scheduled reactor shutdown, and during recovery from reactor trips. ,
i
- d. An individual qualified in radiation protection procedures shall ,
be on site when fuel is in the reactor. j
- e. A site Fire Brigade of at least five niembers shall be maintained l onsite at all times. The Fire Brigade shall not include the i minimum shift crew necessary for safe shutdown of both units j (four members) or any per.ionnel required for other essential l functions during a fire emergency. Fire Brigade training shall i meet the requirements of NFPA 27, 1975 edition.
li
- f. The operations manager shall hold or have held a senior reactor .
operator license at Calvert Cliffs. The supervisor, Shift i Supervisor and Control Room Supervisor shall hold a senior ; reactor operator license. The Control Room Operator shall hold a i reactor operator license. j
- g. One Shift Technical Advisor (STA) shall be assigned to the shift l crew when either unit is in MODE 1, 2, 3 or 4, a id shall be i fille) as follows: !
- 1. By the Shift Supervisor or an on-shift Senior Operator !
License (SOL) holder, provided the individual meets the l Commission Policy Statement on Engineering Expertise on ! Shift; or j I
- 2. By an individual meeting the minimum STA education and l training requirement of Specification 6.3.1; or i
- 3. By an SOL holder previously approved as an exception to the !
minimum STA education requirements of Specification 6.3.1, I provided the following conditions are met: l (a) With both units in MODE 1, 2, 3 or 4, the STA shall be an ! SOL holder in addition to the two SOL holders required; ; i (b) With one unit in MODE 1, 2, 3 or 4 and the other unit in ! MODE 5 or 6, the STA shall be an SOL holder other than t the Shift Supervisor; and (c) With one unit in MODE 1, 2, 3 or 4 and the other unit defueled, the STA shall be an SOL holder in addition to ' the one SOL holder required. { l CALVERT CLIFFS - UNIT 1 6-2 Amendment No.
6.0 ABMINISTRATIVE CONTROLS
- h. Shift crew composition may be less than the minimum recuirements of 10 CFR 50.54(m)(2)(1) and Specifications 6.2.2.a anc 6.2.2.g for a period of time not to exceed two. hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
- 1. Licensed Operators shall be licensed for both units.
6.3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for ; (1) the Radiation Safety Engineer who shall meet or exceed the : qualifications of Regulatory Guide 1.8. Septembe.e 1975, and (2) the Shift l Technical Advisor who shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant l design, and response and analysis of the plant for transients and i accidents. Additional exceptions to ANSI N18.1-1971 are contained in ! Table IB-1 of the Quality Assurance Policy for the Calvert Cliffs Nuclear - Power Plant. i 6.4 PROCEDURES l l 6.4.1 Written procedures shall be established, implemented and maintained l covering the activities referenced below:
- a. The applicable procedures recommended in Appendix A of Regulatory l Guide 1.33. Revision 2. February 1978; l i
- b. The emergency operating procedures required to implement the .
requirements of NUREG-0737 and NUREG-0737 Supplement 1, as ! stated in Generic Letter 82-33; and
- c. All programs specified in Specification 6.5
(
- d. The amount of overtime worked by plant staff members perfoming l safety-related functions must be limited in accordance with the ;
NRC Policy Statement on Working Hours (Generic Letter 82-12). l, t 1 i l CALVERT CLIFFS - UNIT 1 6-3 Amendment No. j
6.0 ADMINISTRATIVE CONTROLS ! 6.5 PROGRAMS AND MANUALS I The following programs shall be established, implemented and maintained: ), 6.5.1 Offsite Dose Calculation Manual (0DCM)
- a. The 0D01 shall contain the methodology and parameters used in the -
calculation of offsite doses resulting from radioactive gaseous l and liquid effluents, in the calculation of gaseous and liquid : effluent monitoring alam and trip setpoints, and in the conduct l of the Radiological Environmental Monitoring Program; and l
- b. The ODCM shall also contain the radioactive effluent controls and ;
radiological environmental monitoring activities and descriptions j of the infomation that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent ! Release Reports, required by Specifications 6.6.2 and 6.6.3, i respectively. J
- c. Licensee initiated changes to the ODCM: l l
\
- 1. Shall be documented and records of reviews perfomed shall be ,
, retained. This documentation shall contain: i i (a) Sufficient information to support the change (s) together i with the app)ropriate the change (s - analyses or evaluations justifying l (b) A detemination that the change (s) maintain the levels of ! radioactive effluent control required by 10 CFR 20.1302, ! 40 CFR Part 190, 10 CFR 50.36a, and 10 CFR Part 50, . Appendix I, and not adversely impact the accuracy or i reliability of effluent dose, or setpoint calculations; I
- 2. Shall become effective upon review by the onsite review function and approval of the plant manager; and
- 3. Shall be submitted to the NRC in the fom of a complete, legible copy of the entire ODCM as part of or concurrent with 4
the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change ' shall be identified by markings in the margin of the affected i pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) l the change was implemented. lt l l CALVERT CLIFFS - UNIT 1 6-4 Amendment No. i 1
I l 0 6.0 ADMINISTRATIVE CONTROLS j 1 6.5.2 Post-Accident Samolina l l The licensee shall establish, implement and maintain a program
- which will l ertsure the capability to obtain and analy7e reactor coolant, radioactive l iodines and particulates in plant gaseous effluents, and containment i atmosphere samples under accident conditions. The program shall include' l the following: !
^
- a. Training of personnel, j
- b. Procedures for sampling and analysis, and l
- c. Provisions for maintenance of sampling and analysis equipment. l l
6.5.3 Primary Coolant Sources Outside Containment The licensee shall im>1ement a program
- to reduce leakage from systems outside containment tsat would or could contain highly radioactive fluids j during a serious transient or accident to as low as practical levels. This- '
program shall include the following: I
- a. Provisions establishing preventive maintenance'and periodic l visual inspection requirements, and
{.
- b. Leak test requirements for each system at a frequency not to j exceed refueling cycle intervals. ;
-l 6.5.4 Technical Specification Bases Control Procram l This program provides a means for processing changes to the Technical !
Specification Bases, j
- a. Changes to the Bases of the Technical Specificattor.s shall be ,
made under appropriate administrative controls and reviews. , t
- b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
- 1. A change in the Technical Specifications incorporated in the license; or
- 2. A change to the UFSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59. l l
It is acceptable if the licensee maintains details of the program in I i plant operation manuals (e.g., chemistry procedures, training ! instructions, maintenance procedures ERPIPs). ! CALVERT CLIFFS - UNIT 1 6-5 Amendment No. l l
l t 6.0 ABNINISTRATIVE CONTROLS
- c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
- d. Proposed changes to the Technical Specifications incorporated in the license or proposed changes to the UFSAR or Bases that !
involve an unreviewed safety question shall be reviewed and ' approved by the NRC prior to implementation. Changes.to the Bases implemented without prior NRC approval shall be provided to theNRConafrequencyconsistentwith10CFR50.71(e) 6.5.5 RADIOACTIVE EFFLUENT CONTROLS PROGRAM This program confoms to 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 shall j be contained in the ODCM, shall be implemented by procedures, and shall ; include remedial actions to be taken whenever the program limits are l exceeded. The program shall include the following elements: l
- a. Limitations on the functional capability of radioactive liquid I and gaseous monitoring instrumentation, including surveillance :
tests and setpoint detemination, in accordance with the ! methodology in the ODCM;
- b. Limitations on the concentrations of radioactive material i released in liquid effluents to uvestricted areas, confoming to 1 10 CFR Part 20. Appendix B. Table II, Column 2; i l
- c. Monitoring, sampling, and analysis of radioactive liquid and !
gaseous affluents, in accordance with 10 CFR 20.1302, and with ! the methodology and parameters in the ODCM; l
- d. Limitations on the annual and quarterly doses or dose comitment to a member of the public from radioactive materials in liquid i effluents released from each unit to unrestricted areas, i confoming to 10 CFR Part 50 Appendix I; l
- e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and i current calendar year, in accordance with the methodology and i parameters in the ODCM, at least every 31 days; j
- f. Limitations on the functional capability and use of the liquid !
and gaseous effluent treatment systems to ensure that appropriate i portions of these systems are used to reduce releases of l radioactivity when the projected doses in a period of 31 days i , would exceed 2% of the guidelines for the annual dose or dose j commitment, confoming to 10 CFR Part 50, Appendix I;
- g. Limitations on the dose rate resulting from radioactive material !
released in gaseous effluents to areas beyond the site boundary I confoming to the dose associated with 10 CFR Part 20, l Appendix B. Table II, Column 1; ; CALVERT CLIFFS - UNIT 1 6-6 Amendment No. 4 l
r 6.0 ADMINI511ATIVE CONTROL 5 I
- h. 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, confoming to 10 CFR Part 50, Appendix I; j
- 1. Limitations on the annual and quarterly doses to a member of the !
public from iodine-131, iodine-133, tritium, and all i radionuclides in particulate fom with half lives > 8 days in l gaseous effluents released from each unit to areas beyond the j site boundary, confoming to 10 CFR Part 50,' Appendix I; and i
- j. Limitations on the annual dose or dose commitment to any member !
of the public due to releases to radioactivity and to radiation i from uranium fuel cycle sources, confoming to 40 CFR Part 190. l 6.6 REPORTING RE0UIREMENTS l The following reports shall be submitted in accordance with 10 CFR 50.4. ) 6.6.1 Occupational Radiation Exposure Report
- 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 man rem exposure according to work and job i functions (e.g., reactor operations and surveillance, inservice inspection, i routine maintenance, special maintenance [ describe maintenance), waste ! processing, and refueling). This tabulation supplements the requirements i of 10 CFR 20.2206. The dose assignment to various duty functions may be ! estimates based on pocket dostmeter, electronic personal dosimeter or i themoluminescent dosimeter. Small exposures totalling < 20% of the- ! individual total dose need not be accounted for. In the aggregate, at i least 80% of the total whole body dose received from external sources ' should be assigned to specific major work functions. The report shall be i submitted prior to March 31 of each year. 6.6.2 Anpual Radioloofcal Environmental Operatina Report The Annual Radiological Environmental Operating Report covering the i operation of the unit during the previcus calendar year shall be submitted prior to May 1 of each year. t The report shall include summaries, interpretations, and analyses of trends i of the results of the Radiological Environmental Monitoring Program for the ; reporting period. The material provided shall be consistent with the objectives outlined in the ODCM, and in 10 CFR Part 50, Appendix I, ! Sections IV.B.2, IV.B.3 and IV.C. : i i A single submittt.1 may be made for Calvert Cliffs. The submittal i should combine those sections that are common to both units. : Occupational dose from Independent Spent Fuel Storage Installation ! will be reported separately. ' CALVERT CLIFFS - UNIT 1 6-7 Amendment No. i
l 6.0 ADMINISTRATIVE CONTROL 5 l The report shall include the results of analyses of all radiological . ! environmental samples and of all environmental radiation measurements taken ! during the period pursuant to the locations specified in the table and i figures in the CDCN, as well as summarized and tabulated results of these i
. analyses and measurements in the fomat of the table in the Radiological i Assessment Branch Technical Position, Revision 1. November 1979. In the !
event that some individual results are not available for inclusion with the ! report, the report shall be submitted noting and explaining the reasons for i the missing results. The missing data shall be submitted as soon as j possible in a supplementary report. l ; i 6.6.3 Radioactive Effluent Release Report
- I l
The Radioactive Effluent Release Report covering o ration of the unit '! shall be submitted in accordance with 10 CFR 50.3 .e., time between i submittal of the reports must be no longer than 12 nths). The reports shall include a sumary of the quantities of radioactive liquid i and gaseous effluents and solid waste released from units. The material .i provided shall be consistent with the objectives outlined in the ODCN and i Process Control Program and in confomance with 10 CFR 50.36% and i- l 10 CFR Part 50. Appendix I, Section IV.B.1 g g Licensee initiated major changes to the Radioactive Waste Systems (liquid, ( kj ' gaseous and solid) shall be reported to the Commission in the Radioactive ! Effluent Release Report for the period in which the modification to the .l waste system is completed. The discussion of each change shall contain: l
- a. A description of the equipment, components and processes l involved; and i i
- b. Documentation of the fact that the change including the safety I analysis was reviewed and found acceptable by the onsite review function.
The report shall also include changes to the ODCM, in accordance with ! Specification 6.5.1.c l 6.6.4 Monthly Doeratina ReDort Routine reports of operating statistics and shu' tdown experience shall be l submitted on a monthly basis, no later than the 15th of each month j following the calendar month covered by the report. , l A single submittal may be made for Calvert Cliffs, since the Radwaste - Systems are common to both units. ' I CALVERT CLIFFS - UNIT 1 6-8 Amendment No.
l 6.0 ADMINISTRATIVE CONTROLS l 1 6.6.5 Core Doeratino Limits Report (COLR) l l
- a. Core operating limits shall be established prior to each reload l cycle, or prior to any- remaining portion of a reload cycle, and !
shall be documented in the COLR for the following: l 2.F.1 3.1.1.1 3.1.1.2 3.1.1.4 l 3.1.3.1- ; 3.1.3.6 : 3.2.1 l 3.2.2.1 3.2.3 l 3.2.5 i 3.9.1 !
- b. The analytical methods used to determine the core operating i limits shall be those previously reviewed and approved by the !
NRC; specifically, those described in the following documents: (1) CENPD-199-P, Latest Approved Revision, "C-E Setpoint i Methodology: C-E Local Power Density and DNB LSSS and LCO l Setpoint Methodology for Analog Protection Systems," ! January 1986 l t (2) CEN-124(B)-P, " Statistical Combination of Uncertainties j Methodology Part 1: C-E Calculated Local Power Density and i Therwal Margin / Low Pressure LSSS for Calvert Cliffs Units I t and II," December 1979 j (3) CEN-124(B)-P, " Statistical Combination of Uncertainties 4 Methodology Part 2: Combination of System Parameter l' Uncertainties in Thernal Margin Analyses for Calvert Cliffs Units 1 and 2," January 1980 (4) CEN-124(8) P, " Statistical Combination of Uncertainties ' Methodology Part 3: C-E Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for ! Operation for Calvert Cliffs Units 1 and 2 " March 1980 , i (5) CEN-191(B)-P,"CETOP-DCodeStructureandModelingMethods ! for Calvert Cliffs Units 1 and 2." December 1981 i (6) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr. ! (BG&E), dated June 24,1982, Unit 1 Cycle 6 License Approval i (Amendment No. 71 to OPR-53 and SER) j i I l CALVERT CLIFFS - UNIT 1 6-9 Amendment No. j - , , , - , , , . -.- - - - ,.-v -
. --m-*---w--ar +--r = -~ ~ * *r w
ll 6.0 ADMINISTRATIVE CONTR0LS (7) CEN-348(B)-P, " Extended Statistical Combination of Uncertainties," January 1987 , (8) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan , (BG&E), dated October 21, 1987 Docket Nos. 50-317 and 1 50-318. " Safety Evaluation of Topical Report CEN-348(B)-P. ' Extended Statistical Combination of Uncertainties" (9) CENPD-161-P-A, " TORC Code A Computer Code for Detemining l the Themal Margin of a Reactor Core," April 1986 , (10) CENPD-162-P-A, Latest Approved Revision, " Critical Heat Flux Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 1. Uniform Axial Power Distribution" l (11) CENPD-207-P-A, Latest Approved Revision, " Critical Heat Flux Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 2, Non-Uniform Axial Power Distribution" (12) CENPD-206-P-A, Latest Approved Revision " TORC Code, Verification and Simplified Modeling Methods" (13) CENPD-225-P-A, Latest Approved Revision, " Fuel and Poison Rod Bowing" (14) CENPD-266-P-A, Latest Approved Revision, "The ROCS and DIT Computer Code for Nuclear Design" (15) CENPD-275-P-A, Latest Approved Revision "C-E Methodology - for Core Designs Containing Gadolinta - Urania Burnable Absorbers" (16) CENPD-382-P-A, Latest Approved Revision, "C-E Methodology for Core Designs Containing Erbium Burnable Absorbers" ! (17) CENPD-139-P-A, Latest Approved Revision, "C-E Fuel Evaluation Model Topical Report" (18) CEN-161-(B)-P-A, Latest Approved Revision, " Improvements to Fuel Evaluation Model" (19) CEN-161-(B)-P, Supplement 1-P, " Improvements to Fuel Evaluation Model," April 1989 ; (20) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated February 4,1987. Docket Nos. 50-317 and I 50-318 " Safety Evaluation of Topical Report CEN-161-(B)-P. i Supplement 1-P, Improvements to Fuel Evaluation Model" (21) CEN-372-P-A, Latest Approved Revision, " Fuel Rod Maximum ! Allowable Gas Pressure" ! CALVERT CLIFFS - UNIT 1 6-10 Amendment No. l
I 6.0 ADMINISTRATIVE CONTROLS ! (22) Letter from Mr. A. E. Scherer (CE) to Mr. J. R. Miller l (NRC),datedDecember 15,1981 LD-81-095. Enclosure 1-P, "C-E ECCS Evaluation Model Flow Blockage Analysis" (23) CENPD-132, Supplement 3-P-A, Latest Approved Revision,
" Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS" (24)' CENPD-133 Supplement 5. "CEFLASH-4A, a FORTRAN 77 Digital i Computer Program for Reactor Blowdown Analysis," June 1985 (25) CENPD-134, Supplement 2. "COMPERC-II, a Program for Emergency Refill-Reflood of the Core," June 1985 (26) Letter from Mr. D. M. Crutchfield (NRC) to Mr. A. E. Scherer :
(CE), dated July 31, 1986, " Safety Evaluation of Combustion i Engineering ECCS Large Break Evaluation Model and Acceptance - for Referencing of Related Licensing Topical Reports" ! (27) CENPD-135, Supplement 5-P, "STRIKIN-II A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 + (28) Letter from Mr. R. L. Baer (NRC) to Mr. A. E. Scherer (CE), I dated September 6,1978, " Evaluation of Topical Report ' CENPD-135, Supplement 5" * (29) CENPD-137, Supplement 1-P, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977 , (30) CENPD-133, Supplement 3-P, "CEFLASH-4AS, A Computer Program ! for the Reactor Blowdown Analysis of the Small Break Loss of J Coolant Accident," January 1977 (31) Letter from Mr. K. Kniel (NRC) to Mr. A. E. Scherer (CE), dated September 27,1977, " Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P" (32) CENPD-138, Supplement 2-P, " PARCH. A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977 (33) Letter from Mr. C. Aniel (NRC) to Mr. A. E. Scherer, dated April 10,1978, " Evaluation of Topical Report CENPD-138, Supplement 2-P" (34) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. J. R. Miller (NRC) dated February 22, 1985, "Calvert ; Cliffs Nuclear Power Plant Unit 1; Docket No. 50-317, 1 Amendment to Operating License DPR-53, Eighth Cycle License l Application" CALVERT CLIFFS - UNIT 1 6-11 Amendment No. l
6.0 ADMINISTRATIVE CONTROLS (35) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr. (BG&E),datedMay 20, 1985, " Safety Evaluation Report Approving Unit 1 Cycle 8 License Application" (36) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to . Mr. R. A. Clark (NRC), dated September 22, 1980, " Amendment to Operating License No. 50-317, Fifth Cycle License Application" (37) Letter from Mr. R. A. Clark (NRC) to Mr. A. E. Lundvall, Jr. > (BG&E),datedDecember 12, 1980, " Safety Evaluation Report Approving Unit 1, Cycle 5 License Application" (38) Letter from Mr. J. A. Tiernan (b5aE) to Mr. A. C. Thadani (NRC), dated October 1,1986, "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2. Docket Nos. 50-317 & 50-318 Request for Amendment" (39) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated July 7,1987. Docket Nos. 50-317 and 50-318, Approval of Amendments 127 (Unit 1) and 109 (Unit 2) (40) CENPD-188-A, Latest Approved Revision, "HERMITE: A Multi-Dimensional Space-Time Kinetics Code for PWR Transients" (41) The Full Core Power Distribution Monitoring System referenced in Specifications 3.1.3.1, 3.2.2.1, 3.2.3, and the BASES is described in the following documents: (a) CENPD-153-P, Latest Approved Revision, " Evaluation of ' Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed Incore Detector System" (b) CEN-199(B)-P, "BASSS, Use of the Incore Detector System to Monitor the DNB-LCO on Calvert Cliffs Unit I and Unit 2," November 1979 (c) Letter from Mr. G. C. Creel (BG&E) to NRC Document Control Desk, dated February 7,1989 "Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket 50-318. Request for Amendment, Unit 2 Ninth Cycle License Application" - (d) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. G. C. Creel (BG&E), dated January 10, 1990, " Safety Evaluation Report Approving Unit 2 Cycle 9 License Application" l I l l l CALVERT CLIFFS - UNIT 1 6-12 Amendment No. l
6.0 ADMINISTRATIVE CONTROL 5
- c. li i The~ core operating applicable limits (e.mits g., fuel shall themalbe mechanical detemined such limits,that all core thermal hydraulic limits, Emergency Core Cooling Systems limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are .
met. I
- d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. ,
F 6.6.6 Pressurizer PORY and Safety Valve Report A report shall be submitted prior to March 1 of each year documenting all failures and challenges to the pressurizer PORVs or safety valves. , l
)
I i l l l l l I i CALVERT CLIFFS - UNIT 1 6-13 Amendment No.
)
i ATTACHMENT (7) l i l i PROPOSED CHANGES TO QA POLICY
1 i j De proposed QA Policy changes shown on the following pages reflect: 1) the relocation of certam sections from the Technical Spannc=tian Adirunistrative Controls; 2) the reduction of the audit fmp=acim found in the Tect.nical VArada= Administrative Controls and in Regulatory Guide 1.33-1978' and 3) the
-e-*iaa to the plant p.de biennial review requirement of ANSI N18.7-1976. De latter two are considered a radnetian to the QA Policy commitment previously =~~w, and are hereby subnutted in accordance with 10 CFR 50.54(a)(3).
l Aucht Frequencies ANSI N18.7-1976 requires audits of selected aspects of operatonal phase activities to be ! performed with a frequency commensurate with their safety significance and in such a manner as to assure that an audit of all safety-related functions is completed within a period of two years. Technical Specificaten 6.5.4.8.1 and Regulatory Guide 1.33-1978 specified more frequent ; intervals for some audits.
]
Baltimore Gas and Electric Company proposes that the audit frequency for all safety-related functions be at least once every two years (except as otherwise required in regulations). He more i frequent audit intervals do not allow Baltimore Gas and Electnc Company management the ! ficxibility to devote auditing resources consistent with the strength of performance and safety j significance of an activity. Experience has shown that some audits are performed more frequently than deemed appropriate for the function Audit subjects that were included in Technical Specification 6.5.4.8.1, and are not already required by regulations or specifically mentioned in ANSI N18.7-1976, are now included in QA Pohey . section IB.18, " Audits." { l 1 Biennial Review of Praceres Plant procedures are subject to pres ri . ..r. tic controls which continually identify procedure revisions. De pros si ..r tic controls meet the intent of the biennial review process from both a I
*Maie=1 and practical standpomt because they constitute dynanuc, rather than static, procedure review methodology. Hus, the biennial review process is redundant to the established progr . r tic controls and is no longer considered necessary.
The programs and processes described below achieve the same desired outcome as periodic procedure resiews:
- 1. He plant modification process requires potentially affected procedures to be reviewed for changes. He necessary procedure changes or revisions are then tracked to ensure proper modification close-out.
- 2. An administrative procedure is in place to ensure that changes made to the Calvert Cliffs i Technical Specifications are reflected in appropriate procedures. !
4
- 3. Control processes are in place that contam mechanisms for meing procedure j improvements. Additionally, control procedures for technical procedures require periodic i review questions to be answered during every resision. l l
1
.i f
- 4. An insre reporting system and corrective action program is in place which allows f procedure users to identify problems and recommand solutens on issue reports. Issue l reports ere reviewed by cognizant supervision and assigned to owner orgariintians for corrective action Corrective action often includes procedure changes. :
5
- 5. Issue reports involving procedures are tracked and trended 'Ihis process will idantdy procedure trend issues so their root causes can be corrected j
- 6. - An add.i.h.tive procedure is in place which clearly states --aar+=+iaan regardag procedure use and adherence This includes stoppmg the job when procedure -
dise=-ies are identified. An imendiate change process is in place to support quick l procedure changes while still meeting necessary review and approval requirements. ,
- 7. An admmistrative procedure is in place which estabhshes requiremerds for -:='=>:-t ,
event investigations and root cause analysis. . Event investigations include investigations following an unusual incident, such as an accident, an ==-vaar*ad transient, significant , operator error, or equipment malf"=e'iaa Event investigations include reviewing , applicable procedures. Corrective actions specified by the event investigations may ! include procedure changes. ,
- 8. A Nuclear Safety Program has been established. The program includes procedural requirements to evaluate industry information issued by the NRC, INPO, and the reactor !
vendor, as well as other equipment vendors, for applicability to Calvert Cliffs. Actions , taken as a result of the evaluation may include procedure changes. :
- 9. An =dmi=Mrative procedure is in place to ensure vendor manual revisions and new vendor !
manuals are reviewed for impact on plant procedures.
- 10. The Nuclear Quality Assurance orgamzation reviews procedures as part of their routine J auditing activities. Actions taken as a result of an audit may include procedure ch==.aes. >
l The justification above addresses the generic guidance provided by NRR as follows: As stated above, applicable plant procedures will be reviewed following an unusual incident and i following any modification to a system. l Non-routine plant procedures (including Emergency Operatmg Procedures, Abnormal 0;-. ting l Procedures, and Emergency Response Plan Implementing Procedures) will continue to receive : biennial reviews according to admmistrative procedures. l
*Ihe bienmal internal audit of the Procedures Program will include a representative sample of plant j procedures. h audit will help ensure the acceptability of the procedures and verify that the Procedures Program is being implemented effectively.
Routine plant procedures that have not been used for two years are reviewed consistent with pre-evolution briefing requirements. Additionally, procedure use and adherence requirements state clear procedure compliance eq74ons and the need to stop the job if a procedure cannot be performed as written.
QUALITY ASSURANCE POLICY Revision 4231 \- L
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BALTIMORE GAS AND ELECTRIC COMPANY. s Y s Quality Assurance Policy for the Calvert Cliffs Nuclear Power Plant (Appendix 1B of the Calvert Cliffs Updated Final Safety Analysis Report) Approved Date (1) R. E. Denton Vice President Nuclear Energy Division
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- Co nol s&n
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/1ee. ,
SG,x d by '/"" # ' Q/P( ? WMl j i Page1of69
I QUALITY ASSURANCE POLICY Revision-42 4.1 J
, TABLE OF CONTENTS APPENDIX 1B OUALITY ASSURANCE PROGRAM FOR THE OPERATIONS PHASE' i
Section Page 1B.1 ORGANIZATION AND RESPONSIBILITIES . .. . . . . . . ......................5' , 18.2 . . . .. . . . . . 13 QUALITY ASSURANCE PROGRAM... . . . . . . . . . . . ... .. . 1 B.3 DESIGN CONTROL . .. . .. . . . .. . . . . . . . . . .. ......................................19 18.4 PROCUREMENT DOCUMENT CONTROL (5) . . . .... . . . ...................21 1 B.5 INSTRUCTIONS, PROCEDURES, AND DRAWINGS.. . . . . .. . . . .. . .. . .. . . .. 24 IB.6 - DOCUMENT CONTROL . . . ... . . . . . . . . . . . . . . . ... . . . . . . . .....................'25 i B.7 CONTROL OF PURCHASED MATERIAL, EQUIPMENT, AND SERVICES (5).. .. . 27 i B.8 IDENTIFICATION AND CONTROL OF MATERIALS, PARTS, AND. COMPONENTS (5)... . . ..... . . . . . . . . . . .. . . . . . . . . . . ....................30 l 1 B.9 CONTROL OF SPECIAL PROCESSES . .. . . . . . . .. . . . . . . . . . . . . .. . . 3 1 18.10 INSPECTION. ... ... .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . 3 3 1B.II TEST CONTROL.... ... . . . . . . . . . . . . . .. . . . . . .. .... .. . . . . . . 35 1 B.12 CONTROL OF MEASURING AND TEST EQUIPMENT. .. .... .. ... . . ... . 3 6 1B.13 HANDLING, STORAGE, AND SHIPPING ... . . . . . . . . . . . . . . . . . . . . . . . .. 37 IB.14 INSPECTION, TEST, AND OPERATING STATUS . . . ... . ... . 3 7 IB.15 NONCONFORMING MATERIALS, PARTS, OR COMPONENTS (6).. . ..... .....38 L 18.16 CORRECTIVE ACTION (6) . . . ... . . ... . . . .. .. . . . . . . . . . . . . . .38 18.17 QUALITY ASSURANCE RECORDS. .. ... . . . . . . . . . . . . . . . . . . . . .... 39 18.18 AUDITS . . . . . . . ... .. . . . .. ..40 l-1 Page 2 of 69
i p; !
- QUALITl ASSURANCE POLICY l g Revision 42.B l .j r-i LIST OF TABLES !
i 1, i l I; . Table No. . Page ~ t 1B-1 BALTIMORE GAS AND ELECTRIC COMPANY'S POSITION ON GUIDANCE - ! CONTAINED IN ANSF INDUSTRY STANDARDS AND REGULATORY GUIDES . . . 41 I i I LIST OF A'ITACHMENTS
' Attaciunent 1 Letter l l
A BASES FOR QA POLICY REVISIONS (l) . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . 60 : i
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i i LIST OF FIGURES .! Figure No. l I IB BALTIMORE GAS AND ELECTRIC COMPANY l CORPORATE ORGANIZATION... . ... . .... .. . ., ... . ... ..... . . . . . . . . . . . . . . . . . . 62 l
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I LIST OF ADDENDA Addendum No. I
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I B-1 REVIEW FUNCTIONS OF Ti1E POSR
C. PROCEDURE
REVIEW COMMITTFE l' OUAlIFIED REVIEWERS. ANDJSSRC. . 63 Page 3 of 69
QUALITY ASSURANCE POLICY Revision 42_43 l LIST OF EFFECTIVE PAGES Latest revision number is listed for pages revised after revision 38. Page Last Revision Page Last Revision Page Last Revision 1 24 M 47 41 2 25 41M 48 41 3 M 26 44M 49 41 4 4243 27 41 50 41 5 28 41 51 41 6 42M 29 41 52 41 7 42 30 41 53 41 8 42 31 41 54 41 9 42 32 41 55 41 10 42 33 41 56 41 11 42 34 41 57 12 42 35 41 58 13 41 36 59 42 14 41 37 41 60 42 15 41M 38 41 M M 16 41M 39 41 62 M i7 44M 40 44M 63 M i8 41 44M 64 E 19 41M 42 42M 61 M 20 44M 43 44M 66 M 21 41 44 41 62 M 22 41 45 41 63 M 23 41 46 41 62 M Page 4 of 69
QUALITY ASSURANCE POLICY Resision42_43 l 18.1 ORGANIZATION AND RESPONSIBILITIES All levels of organization have defmite and unique responsibilities in assuring safe, economical, and reliable i operation of Calvert Cliffs Nuclear Power Plant (CCNPP). Top level management is responsible for ensuring that policies are established, resources are authorized, management philosophy and commitments are communicated to lower levels of the organization, independent verification of management controls are performed, results are retiewed, and appropriate actions taken when necessary. ! Middle level management is responsible for translating management policies, philosophy, commitments, : and goals; applicable federal, state, and local rules and regulations; Operating Licenses, Technical l Specifications (TS), and the Updated Final Safety Analysis Report (UFSAR) into control ;;rograms for f activities such as design, procurement, constmetion, testing, operation, refueling, maintenance, repair, modification, training, plant security, fire protection, records, independent verification, and corrective action. Middle level management is also responsible for defming, measuring, and modifying the overall ! cffectiveness of control programs; taking appropriate action on the results; and keeping top management i informed of the status, adequacy, and effectiveness of control programs, and matters which could have an , impact on nuclect safety. i. First line craft and non-craft supenisors are individually responsible for ensuring that appropriate procedures are understood and used to implement each activity described in the control programs; identifying problems, seeking solutions, verifying implementation of solutions; investigating root causes of problems and taking preventive actions; ensuring that conditions adverse to plant and personnel safety are l promptly identified, reported, and corrected; detecting trends which may not be apparent to a day-to-day
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observer, recommending generic solutions for adverse trends to management, and taking appropriate actions, to achieve desired results; ensuring that employees assigned to do a job are properly qualified through appropriate training and experience; have properly qualified procedures, tools, equipment, and parts to do the job, and, ensuring that independent inspections of work are conducted in accordance with l preestablished requirements. First line non-craft supenisors are responsible to ensure that procedures are written, reviewed, and approved; first line craft supenisors may not have this responsibility. Non- ! supenisory personnel acting as job directors are responsible for ensuring that properly qualified procedures are understood and used; and ensuring that tools, equipment, and parts are on hand to do the job. Adherence to procedures is vital to the safe and reliable operation of the Calvert Cliffs Nuclear Power Plant. Personnel are responsible for adhering to established procedures, interpreting them conservatively in case of doubt, and recommending changes when necessary. Procedures with the potential to affect nuclear or personnel safety shall be strictly adhered to. When an activity controlled by such procedures cannot be acomplished as described or accomplishment of such activity would result in an undesirable situation, the work shall be stopped and the phmt placed in a safe condition. Work shall not resume until the procedure is changed to reflect correct work practices. (1) Procedures may be deviated from during emergencies to prevent or minimize injury to personnel or damage to plant equipment. Any such deviations should be thoroughly documented. (1) i Page 5 of 69
QUALITY ASSURANCE POLICY ! Revision 42_41 l l Corporate Organization and Specific Responsibilitigs f The Corporate Organization Chart of the Baltimore Gas and Electric Company (BGE) is shown in Figure ! IB-1. Persons responsible for the principal elements of the Company's Quality Assurance (QA) Program .l are as follows: (1) ! k I Chairman of the Board i President and Chief Operating Officer I Senior Vice President-Generation i Vice President Nuclear Energy Division (NED) Plant General Manager-Calvert Cliffs Nuclear Power Plant Department (CCNPPD) Manager-Nuclear Engineering Department (NED) Manager-Nuclear Quality Assurance Department (NQAD) ! Manager-Nuclear Support Senices Department (NSSD) l Manager-Nuclear Outage & Project Management Department (NOPMD) In addition to these individuals, the Vice Presidents of Corporate Affairs (CA), Fossil Energy Division (FED), Electric Interconnection and Transmission Division (EITD), General Senices Division (GSD), and ! the Management Senices Division (MSD), as well as the Managers of the System Operation and .l Maintenance Department (SOMD), Purchasing & Materials Management Department (PMMD), j Infonnation Systems Department (ISD), Facilities and Fleet Senices Department (FFSD), Fossil l Engmeering & Maintenance Department (FEMD), Fossil Support Senices Department (FSSD), and the ! Safety and Medical Services Department (SMSD) are assigned support responsibilities. (1) The above ! Managers constitute the Nuclear Program Managers who are assigned responsibilities within the QA l Program. Other departments performing any maintenance / modification activities at CCNPP are ! responsible for performing these activities in accordance with applicable QA Program requirements. This : can be accomplished by either developing their own QA Program procedures or by working to the QA ! Program through appropriate Nuclear Energy Division personnel using CCNPP procedures. (15) l Also, two advisory groups perform quality-related functions for plant operations. These are the Plant i Operations and Safety Review Committee (POSRC) and the Off-Site Safety Review Conunittee (OSSRC) ! whose makeup and responsibilities are described in4he-T6sfer-CCNPP in Addendnm 1B-1. n2) ! t Qairman of the Brasi. PresidgnLandDief Operatinn Officer. and Serior Vice President-Generation l BGE's QA Program for nuclear power plants is established under the authority of the Chairman of the Board, President and Chief Operating Officer, and Senior Vice President-Generation, who are responsible : for establishing the overall QA Policy. They assign project responsibilities to the organizations shown in i heavy-lined boxes in Figure iB-1. (1) J f i l Page 6 of 69
QUALITY ASSURANCE POLICY Revision 4M1 l The Chairman of the Board assigns authority through the President and Chief Operating Officer and Senior Vice President-Generation to the Vice President-Nuclear Energy Division. Primary responsibilities for developing, implementing, and maintaining the QA Program are assigned to Department Managers by the Vice President-Nuclear Energy Division. Managers delegate their authority as required to implement their responsibilities. (I) Quality assurance matters that cannot be resolved by the Managers or Vice Presidents are brought to the , attention of the Senior Vice President-Generation, President and Chief Operating Officer, or the Chairman of the Board for resolution. Vic. e President-Nuclear Enern Division The Vice President-Nuclear Energy Division, is responsible to the Senior Vice President-Generation for ensuring that the QA Program is developed and implemented. The authority to develop QA Program Documents is assigned to designated Nuclear Program Managers. The Vice President-Nuclear Energy Division, is also responsible for ensuring that the requirements of the QA Program that relate to the design, operation, and maintenance of the plant are implemented. His responsibility is carried out through Nuclear Program Managers. Manager-Nuclear Ouality Assurance Denartment The Manager-NQAD, is responsible for assuring an appropriate QA Program is established and effectively executed for CCNPP. lie is responsible for auditing, quality verification, vendor evaluation, and independent safety evaluation functions for CCNPP. These responsibilities include:
- 1. Developing, and revising the QA Policy.
- 2. Ensuring that QA Compliance reviews are completed for program acceptability of Control Programs (QAPs and Directives) and their revisions before they are approved. (9) l
- 3. Taking necessary corrective action, which can include the stoppage of work when manufacturing, maintenance, or modification activities fail to comply with approved specifications, plans, or procedures. Such corrective action is arranged through appropriate channels and is delegated when necessary. When a unit is operating, the Manager-NQAD, may recommend to the Plant General Manager that the plant be shut down. The Plant General Manager has the final responsibility for the overall evaluation of all aspects and implications of shutting down an operating unit.
NQAD personnel who report to the Manager-NQAD, are independent of departments, sections, and employees responsible for performing specific activities, and have sufficient authority and organizational freedom to identify quality problems; to initiate, recommend, or provide solutions through designated channels; and to verify implementation of solutions. Non-NQAD personnel who are authorized to perform activities under NQAD programs are matrixed to NQAD for the performance of such activities, and possess similar organizational freedom and independence from the activities. BGE has established that the Manager-NQAD, should have at least six years of responsible experience in engineering, design, manufacturing, construction, quality assurance, or power plant operation. as well as a knowledge of regulations and standards related to nuclear power plants. Page 7 of 69
QUALITY ASSURANCE POLICY Revision 42A3 l The organization of NQAD is shown in Chapter 12 of the UFSAR. The Manager-NQAD, delegates the following responsibilities for accomplishing the following activities: Planning and scheduling evaluations of vendor quality assurance programs. Reviewing proposed changes to QA Program documents for compliance with regulations and licensing documents. Planning, scheduling, and performing internal audits /surveillances and evaluations of on-site and off-site functions performed under the nuclear QA Program. Supporting maintenance and operations activities by performing inspections and surveillances, or by providing oversight of other department personnel as they perform inspections and surveillances. (11) Directing investigations of significant events to determine root cause, recommending corrective action, and generating appropriate reports to document the investigation results; directing a program for identifying trends within the corrective action systems. Directing reviews of the operating experience of other plants of similar design to determine the applicability of significant events with respect to CCNPP. (14) Plant General Manacer-Calvert Cliffs Nuclear Power Plant Department The Plant General Manager is responsible for operations, chemistry, radiation safety, maintenance, industrial safety and fire protection, and systems and perfonnance engineering actisities at CCNPP. He must ensure that these activities are conducted in accordance with the plant operating license and TSs, the UFSAR, the QA Program, and procedures. The Plant General Manager fulfills the position and requirements of the Plant Manager, as defined in ANSI N18.1 (1971). He, or one of his designated principal alternates, shall have acquired the experience and training normally required for examination for a senior reactor operator's license. The organization of CCNPPD is shown in Chapter 12 of the UFSAR. The Plant General Manager, delegates responsibilities for accomplishing required activities as follows:
- 1. The Superintendent-Nuclear Operations (S-NO) is responsible to the Plant General Manager, for the operation of the plant, including the general supenision of all shift operating personnel and prioritization of maintenance activities to support operations. This responsibility covers the safety of plant personnel and equipment, all fuel-handling and refueling activities, and adherence to applicable license and regulatory requirements. The Superintendent-Nuclear Operations fulfills the position and requirements of the Operations Manager as defined in ANSI N I 8.1 (1971) with the exception taken in Table 1B.I.
The Superintendent-Nuclear Operations delegates primary management responsibility to the Shift Supenisor on duty, via the General Supenisor-Nuclear Plant Operations (GS-NPO) to ensure the safe operation of the plant under all conditions. The Shift Supenisor maintains the broadest possible perspective on operational conditions that affect the safety of the plant. As the senior member of plant management on each shift, he exercises the command authority of his position to take whatever steps he deems necessary during emergency situations to place and maintain in a safe configuration any unit that may be affected.
- 2. The Superintendent-Nuclear Maintenance (S-NM) is responsible to the Plant General Manager for managing and directing activities of the Nuclear Maintenance Section to provide high quality maintenance programs, plans and schedules, and qualified personnel to perform Page 8 of 69 i
QUALITY ASSURANCE POLICY Revision 42R l maintenance functions necessary to assure the safe, reliable, and economic operation of the plant to generate power within applicable laws, standards, codes, and regulatory requirements. 4
- a. The General Supenisor-Electrical and Controls (GS-E&C), is responsible to the Superintendent-Nuclear Maintenance, for the conduct of electrical and instrument maintenance, repair, and modifications needed to keep the plant and its facilities, !
systems, and equipment in safe and eflicient working condition. He is responsible for planning and supervising or controlling the electrical and instrument maintenance activities conducted by plant maintenance personnel, and for ensuring that work is - performed in accordance with applicable Codes and Standards and that required maintenance records are developed and kept. He is responsible for controlling tools and equipment used for electrical and instrument maintenance, repair and modifications activities.
- b. The General Supenisor-Mechanical Maintenance (GS-MM), is responsible to the Superintendent-Nuclear Maintenance, for the conduct of mechanical maintenance, repair, and modifications needed to keep the plant and its facilities, systems, and equipment in safe and efficient working condition. He is responsible for planning and supenising or controlling the mechanical maintenance activities conducted by plant maintenance personnel, and for ensuring that work is performed in accordance with applicable Codes and Standards and that required maintenance records are developed and kept. He is responsible for controlling tools and equipment used for meclumical maintenance, repair, and modifications activities.
- 3. The Superintendent-Technical Support (S-TS) is responsible to the Plant General Manager for managing and directing the activities of the Technical Support Section to provide systems and performance engineering, surveillance testing administration, chemistry and radiation -
safety support, insenice inspection, and industrial safety and fire protection support to assure the safe, reliable, economic operation of CCNPP. The S-TS is also responsible for overall direction and coordination of activities to ensure compliance with the Radiological Efiluent Technical Specifications. This responsibility is carried out through the General Supervisor-Chemistry with support from the Director-Emironmental Programs and the Manager-FSSD.
- a. The General Supenisor-Chemistry (GS-C) is responsible to the Superintendent-Technical Support for the chemistry and radio- chemistry of the primary and secondary systems and for maintaining radioactive effluents within specified limits.
Additionally, the GS-C provides program management oversight of the Radiological Environmental Monitoring Program to ensure compliance with the Radiological Effluent Technical Specifications.
- b. The General Supenisor-Radiation Safety (GS-RS) under the nuclear QA Program, is responsible to the Superintendent-Technical Support for:
Ensuring the radiation protection of personnel at CCNPP. Complying with radioactive material transport regulations. l l l l i Page 9 of 69 i i
QUALITY ASSURANCE POLICY ! Revision-42E l l
- c. The Supenisor - Safety and Fire Protection is responsible to the Superintendent -
Technical Support for development, implementation and coordination of the industrial j safety program; implementing the fire prevention and fire fighting programs for the l CCNPP; and planning, scheduling and monitoring activities directiv related to safety, i fire protection, and prevention.
- d. - The General Supenisor-Plant Engineering (GS-PE), is responsible- to the Superintendent - Technical Support for providing field engineering and technical evaluation of plant systems and to evaluate and coordinate resolution of system and component problems with operations, maintenance, and engineering personnel for the Calvert Cliffs Nuclear Power Plant. Additionally, the GS-PE is responsible to the S-TS for providing plant reliability / availability testing and evaluation, plant performance improvement, and administration of the Surveillance Test Program.
- c. The Principal Engineer-Nuclear Inspection Senices is responsible to the Superintendent - Technical Support for providing inservice inspection senices for the Calvert Cliffs Nuclear Power Plant.
Manacer-Nuclear Encineerina Denartment The Manager-NED, is responsible for directing the efforts of personnel and providing resources necessary to support design, modification and engineering activities covered by the QA Program for CCNPP. These activities include nuclear, mechanical, civil, reliability, instrument and controls, and electrical engineering; nuclear fuel management; configuration management; life cycle management; plant design support; fire protection program development; engineering planning; licensing, and design and drawing. The organization of NED is shown in Chapter 12 of the UFSAR. The Manager-NED delegates responsibilities for accomplishing the following activities: Providing conceptual and detailed engineering, design and drawing, fire protection program development, data base configuration control, and documentation and maintenance of plant design bases for the power and control systems for the Calvert Cliffs Nuclear Power Plant. Directing and performing safety evaluations, preparation and review of nuclear safety accident and transient analysis, fuel management, nuclear engineering related to core physics, reactor engineering, external fuel cycin management, reliability engineering and development and integration of programs necessary to operate up to and beyond the current licensed lifetime for the Calvert Cliffs Nuclear Power Plant. Providing plant design support, and engineering planning and scheduling for the power and control systems for the Calvert Cliffs Nuclear Power Plant. Providing licensing senices; coordination and operation of various industry infonnation exchange systems; evaluation of plant events and conditions adverse to quality for reportability to the NRC and other agencies, assisting in the investigation and evaluation of events, and preparation of the reports; coordination of tracking and resolution of company commitments to the NRC; research and preparation of responses to NRC letters, bulletins, circulars and information notices; UFSAR research and revision control; maintenance and revision of the current licensing basis for nuclear power plants; coordination of all compliance-related communications with external agencies including assistance in ensuring their consistency with existing licensing basis commitments; and coordination of regulatory inspections and visits and company presentations to the NRC. Page 10 of 69
QUALITY ASSURANCE POLICY ! Revision 42A3 l i Manaaer-Nuclear Outane and Proiect Management Department The Manager-NOPMD is responsible for directing the efforts of personnel and providing resources necessary to support site integrated scheduling, outage management and project management for ; assigned projects. The organization of NOPMD is shown in Chapter 12 of the UFSAR. The { Manager-NOPMD delegates responsibilities for accomplishing the following activities: ; i Developing, implementing and maintaining a site integrated schedule which schedules all significant plant related activities at CCNPP. Managing the planning, scheduling and performing of all outages at CCNPP.' Providing overall project management for engineering, procurement, construction and testing of nuclear power plant modification for CCNPP. Manager-Nuclear Support Services Denartment The Manager-NSSD is responsible for emergency planning, training, nuclear security, onsite procurement coordination, procurement engineering, receipt inspection and storage / issue ofitems, : procedures upgrade, state regulatory matters, strategic planning, and staff services functions for CCNPP. r The organization of NSSD is shown in Chapter 12 of the UFSAR. The Manager-NSSD delegates responsibilities for accomplishing the following activities: Providing support to Managers in the Nuclear Energy Division to ensure their personnel are - properly trained and qualified to perform their assigned duties, including those duties which l implement the nuclear QA Program. Training required by special work forces and - ! contractors would be performed by the appropriate BGE Department, and/or Host Company 1 (vendor). Distributing, and coordinating the preparation of revisions to the QA Program documents; collecting, storing, maintaining, and retrieving QA records for nuclear power plants; maintaining, controlling, and distributing drawings and technical manuals related to equipment, materials, and services for nuclear power plants; coordinating investigations concerning state regulatory matters; coordinating the efforts of Nuclear Energy Division , personnel involved in the procurement of structures, systems, components, parts, and services related to the design, construction, fueling, maintenance, and modifications of CCNPP. , Establishing procedures to assure that SR and DNSR procurement documents identify technical and quality requirements; procurement (SR and DNSR) documents receive ; independent review and approval for the proper inclusion of technical and quality requirements; ensuring spare and replacement parts are suitable for their' intended ; application (s); specification of critical characteristics and acceptance criteria for dedication of commercial grade items; specification of special storage requirements for age sensitive items. , i Page 11 of 69
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QUALITY ASSURANCE POLICY Raision42_43 \
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Performing receipt inspection functions including special receipt inspections and coordinating testing performed to accept commercist grade items, designated NSR items or upgrade NSR items for use in SR applications. (S) Ensuring that the operational, maintenance, licensing, and training activities associated with plant security are effectively implemented, and that nuclear security provisions provide protection for personnel, equipment, and facilities at CCNPP against potential security threats. Directing the efforts of personnel responsible for the storage and issuance of items for CCNPP.
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Development of the annual Strategic plan for the Nuclear Energy Division including the Nuclear Program Plan.
- Directing the efforts of BGE personnel involved in emergency planning activities.
Individuals supporting the CCNPP QA Program are designated as follows: Vice President-Fossil Enerev Division The Vice President-FED, is responsible to the Senior Vice President-Generation for ensuring that the activities of FED personnel involved in CCNPP maintenance and modifications; Materials Engineering and Analysis; and radiological environmental monitoring, meet the requirements of the QA Program. This responsibility is carried out through the Manager-FEMD, and the Manager-FSSD. Manarter-Fossil Engineerine & Maintenance Denartment The Manager-FEMD, is responsible for directing the efforts of FEMD personnel involved in maintenance and modification activities at CCNPP. Manager-Fossil Support Services Denartment The Manager-FSSD, is responsible for directing the efforts of FSSD personnel involved in: (1) maintaining and operating Radiological Environmental Monitoring equipment and performing sample collection and analysis, (2) ensuring materials engineering and analysis relating to SR structures, systems, and components are completed in accordance with Company and regulatory requirements. Vice President-Electric Interconnection and Transmission Division The Vice President-EITD, is responsible for ensuring that the requirements of the QA Program that relate to the calibration of test equipment and the testing of protective relaying, and metering controls for SR clectrical power equipment are implemented. This responsibility is carried out through the Manager-SOMD. Manager-System One.rption and Maintenance Department The Manager-SOMD, is responsible for directing the efforts of personnel involved in the testing of electrical power equipment, the calibration of test equipment and the testing of protective relaying and metering controls for the electrical power equipment of CCNPP. 1 Page 12 of 69 I l
QUALITY ASSURANCE POLICY Revision 42R { i Vice President Generpl Senices Disision The Vice President-GSD is responsible for ensuring implementation of the QA Program requirements that relateto: the procurement of SR or designated NSR structures, systems, components, and senices; the constructior;. maintenance, and operation of facilities; and support senices for computer software and hardware. These responsibilities are carried out through the Manager-PMMD, Manager-ISD, and Mt. nager-FFSD. Manaccr-Purchasine and Materials Management Department The Manager-PMMD, is responsible for directing the efforts of personnel involved in the purchasing ofitems and senices for CCNPP and for the issuance of Contracts for Fitness for Duty Ac+ivities. hi_anaggr-Information Systems Denanment The Manager-ISD, is responsible for directing the efforts of ISD personnel involved in acquiring and supporting computer software and hardware. Manay;.c-Facilitigs and Fleet Senices Denanment The Manager-FFSD, is responsible for directing the efforts of FFSD personnel involved in the , planning, design, construction, maintenance. and operation of facilities and related systems directly supporting or impacting power plant operations. , Vice President-Management Senices Division The Vice President-MSD, is responsible for ensuring that the activities of MSD personnel involved with medical examination; for CCNPP operators, Nuclear Security Officers, and respirator users, meet the requirements of the regulations. The responsibility is carried out through the Manager-SMSD. hianager-Safety and Medical Senices Denanment Tlie Manager-SMSD, is responsible for directing the efforts of SMSD personnel involved with medical examinations for CCNPP operators (10CFR55), Nuclear Security Officers (10CFR73), respirator users (10CFR20), and with the Fitrass for Duty rule (10CFR26). Vice President-Corporate Affairs The Vice President-CA is responsible for ensuring that QA Program requirements related to the Radiological Effluent Technical Specifications are implemented. This responribility is carried out through ' the Director - Enviroiunental Programs. 18.2 OUALITV .1SSURANCE PROGRAM ; General Control.s
'Ihe QA Program consists of the Updated Final Safety Analysis Repon (UFSAR) Appendix IB, QA Policy, Quality Assurance Procedures, certain Nuclear Program Directives and their implementing procedures. The UFSAR Appendix IB and QA Policy are the same document except for the way changes are incorporated The QA Policy is updated when each change is approved. Revisions to the QA Policy are controlled by QA Program documents which are written to ensure compliance with 10 CFR Page 13 of 69 i
QUALITY ASSURANCE POLICY Revision 42 41 l 50.54(a)(3). He UFSAR Appendix IB is updated annually. All QA Policy changes approved during the previous year are incorporated during that update. The QA Policy identifies NRC regulatory requirements, industry standards, and specific codes applicable to the eighteen criteria contained in 10 CFR 50, Appendix B. The QA Policy also indicates action that will be taken by BGE in response to these documents and to commitments made in the UFSAR and TSs for CCNPP. Quality Assurance Procedures (QAPs) describe controls for the actions identified in the QA Policy. QAPs cover major activities related to operating a nuclear power plant, such as plant operation, plant maintenance, training, purchase ofitems and senices, calibrations, etc. Nuclear Program Directives address actions identified in UFSAR Appendix 1B. Directives identify regulatory commitments, management requirements, and assign responsibilities for business activities (i.e., design, maintenance, operations, etc.) within the BGE Nuclear Program. As directives are written and implemented, they will systematically replace QAPs. BGE's QA Program for CCNPP is applied to structures, systems, components, and activities that have been designated SR because they prevent accidents or mitigate the consequences of postulated accidents that could cause undue risk to the health or safety of the public. The QA Program is also applicable to designated NSR structures, systems, components, activities, and senices as required by in regulations. Designated NSR program requirements are based on a graded approach to Quality Assurance required to meet applicable regulatory designated requlements and guidance. The level of QA Program controls placed on designated NSR items are defir,d in QA Program documents and/or implementing procedures. The controls from other sections of tLs QA Policy are selected as necessary to meet the particular i regulations being implemented. i Controls have been established for specifying on a Quality List (Q-List) all SR structures, systems, components, and activities that are subject to the requirements of the QA Program. The Statement of Authority, in the Quality Assurance Manual for Nuclear Power Plants, signed by the Chairman of the Board, establishes the overall QA Pol;cy of BGE. This Statement sets the goal of safe and ' reliable operation of CCNPP; commits the Company to a QA Program designed to ensure the plant's ; compliance with regulatory requirements, BGE conunitments, and established practices for reliable plant operation; and requires every person involved in QA Program activities to comply with the provisions of the Program. The Policy is approved by the Vice President-NED and implemented by Nuclear Program Managers. (1) i The QA Program has established controls for BGE and its contractors as required to ensure that the criteria of 10 CFR 50, Appendix B, will be met throughout the operations phase of the plant; i.e., during activities of testing, operation, maintenance, repair, modification, and refueling. > The QA Program has also established controls to ensure that the construction, operational, and decommissioning phases for the Independent Spent Fuel Storage Installation (ISFSI) are conducted in compliance with 10 CFR 72. Activities associated with the operational and decommissioning phase shall be controlled under the CCNPP 10 CFR 50 Appendix B QA Program; existing policies, programs, i directives, and procedures stated as applicable for CCNPP are also applicable for the ISFSI. (16) Changes to the QA Program documents are issued with a transmittal notice, which is completed by the recipient and returned to indicate that the documents listed on the transmittal have been received and incorporated into the recipient's Manual. Nuclear Program Managers ensure QA Program documents are . revised as regulations, standards, results, or experience dictate. (1) The Manager-NQAD evaluates the degree of compliance with the requirements of QA Program documents and procedures. Audits are Page 14 of 69
QUALITY ASSURANCE POLICY , conducted regularly to ensure compliance with established requirements, and the results of these audits are reported to responsible management personnel. The Vice President-NED, ensures that actisities of the NQAD are aadited regularly by personnel independent of the Department. These auditors assess the effectiveness of the Department's implementation of appropriate portions of BGE's QA Program. The Vice President-Nuclear Energy Disision, evaluates the report of the independent audit to determine if changes are required to the QA Program. He is responsible for negotiating such changes with the appropriate level of management and for sending to the Chairman of the Board a copy of the audit report and an account of the corrective action taken. If a difference of opinion arises between NQAD personrel and those of other Sections or Departments, the dispute is resolved as follows: The Supenisor/ General Supenisor of the QA Unit /Section involved first tries to resolve the matter with the organization responsible for conducting the activity. If a resolution cannot be obtained, the matter is referred up through the following management personnel until it is resolved: (3)
- 1. The Manager-NQAD, and the Manager responsible for performing the activity.
NOTE: If the dispute is with another Unit /Section in NQAD, the issue will be settled by the Vice President-Nuclear Energy Division. (3)
- 2. The Vice President-Nuclear Energy Division. (1)
- 3. The Senior Vice President-Generation, President and Chief Operating Officer, or the Chairman of the Board.
To ensure that important activities are performed correctly, BGE conducts formal training programs for Company personnel with significant responsibilities. These programs include both initial and continuing training and are conducted in accordance with written procedures or instructions. Department Managers are responsible for ensuring that the training needs of personnel in their Departments are identified, formal training programs to satisfy those needs are developed, and the training programs are implemented in accordance with the requirements of the QA Program documents. The QA Program was developed to meet the requirements of the Regulations; and Regulatory Guides of the Nuclear Regulatory Commission (NRC), and Industry Standards of4he Nadear-Regulatory-C-ommission (NRC-)-listed below. Exceptions taken to guidance contained in these documents and equivalent BGE alternatives are stated in Table 1B-1. REGULATION 5 , 10 CFR 50.55a - Codes and Standards. 10 CFR 50.59 - Changes, Tests, and Experiments. 10 CFR 55 - Operators' Licenses. 10 CFR 50, Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. 10 CFR 72, Subpart G - Quality Assurance (ISFSI) Page 15 of 69
1 QUALITY ASSURANCE POLICY Revision 42R \ REGULATORY GUIDES , 1.8 - Personnel Selection and Training (September 1975)** 'Ilis endorses ANSI N18.1 (03/08/71)*** 1.16 - Reporting of Operating Information (as specified in Calvert Cliffs Technical Specifications). I.30 - QA Requirements for Installation, inspection, and Testing of Instmmentation and Electric l Equipment (08/l1/72)* This endorses ANSI N45.2.4 (03/01/72). ! I.33 - QA Program Requirements (Operation, Rev. 2,02/78)"#. This endorses ANSI N18.7-1976/ANS l 3.2 (02/19/76)* ** I.37 - QA Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nucicar Power Plants (03/16/73)" This endorses ANSI N45.2.1 (02/26/73)"* 1.38 - QA Requirements fbr Packaging, Shipping. Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (Rev. 2,05/77)" This endorses ANSI N45.2.2 (12/20/72)"* , 1.39 - Housekeeping Requirements for Water-Cooled Nuclear Power Plants (03/16/73)* This endorses ANSI N45.2.3 (03/15/73)"* 1.54 - QA Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants (06/73)** This endorses ANSI N101.4 (11/28/72)*" 1.58 - Qualification of Nuclear Power Plant Inspection, Examination, and Testing Persoiusel (09/80)** This endorses ANSI N45.2.6 (1978)"* 1.64 - QA Requirements for the Design of Nuclear Power Plants (10/73)* This endorses ANSI N45.2.11, Draft 3, Rev.1 (07/73). 1.68 - Preoperational and initial Startup Test Programs for Water-Cooled Power Reactors (11/73)** 1.144 - Auditing of Quality Assurance Programs for Nuclear Power Plants, Rev.1 (09/80)** This endorses ANSI N45.2.12 (1977). 1.146 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Aug. 1980).* This endorses ANSI N45.2.23 (1978)"* INDUSTRY STANDARDS ; ANSI N45.2.5 - Supplementary QA Requirements for Installation, inspection, and Testing of Structural r Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants: Draft 3, Rev.1 (l1/73). ANSI N45.2.8 - Supplementary QA Requirements for Installation, Inspection. and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Power Plants; Draft 3, Rev. 2 (09/73). j r Page 16 of 69
! - QUALITY ASSURANCE POLICY Revision 42.43 } ANSI N45.2.9 - Requirements for Collection, Storage, and Maintenance of Quality Assurance Records for Nuclear Power Plants; Draft (10/76)*" ANSI N45.2.13 - QA Requirements for Control of Procurement of Equipment, Materiais, and Senices for Nuclear Power Plants; Draft 2, Rev. 2, (10n3)*" NOTATIONS FOR REGULATORY GUIDES AND INDUSTRY STANDARDS
- NRC cndorses an Industry Standard or draft without resenation.
" NRC takes exception to or provides additional guidance in a regulatory position statement.
'" BGE takes exception to guidance offered in inancerv Standard and states alternatives.
# BGE takes excention to ouidance offered in Reoulatory Guide and ctatec alternatives.
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QUALITY ASSURANCE POLICY Revision 4233 l Procedural Controls The QA Policy and revisions thereto are reviewed by Nuclear Program Managers. QA Policy revisions are reviewed by NQAD personnel to determine if they constitute a reduction in commitments previously made to the NRC. If so, the reiisions are sent to NRC for approval prior to implementation. The Manager-NQAD reviews revisions to the QA Policy and recommends approval to the Vice President-Nuclear Energy Division. The Vice President-NED approves the QA Policy and revisions thereto. Each Quality Assurance Procedure (QAP) is prepared by one or more of the Departments responsible for conducting the activity. The QAP and revisions thereto are reviewed by NED Managers and affected Department Managers. The Manager-NQAD ensures QAP revisions are reviewed by the Quality Assurance organization and recommends approval to the Vice President-NED. The Vice President-NED approves all QAPs and revisions thereto. The Manager-NSSD ensures issuance of all QAPs and revisions thereto. (1) QA Program documents control the distribution and revision of the QA Policy and other QAPs. i Nuclear Program Directives are prepared under the direction of the Department Manager assigned by the Vice President-NED as the Program Sponsor. Each directive and revisions thereto are reviewed by affected Department Managers. The Manager-NQAD ensures directive revisions are reviewed by the Quality Assurance organization and approval recommended to the Program Sponsor. The Program Sponsor approves the directive and revisions thereto. The Manager-NSSD ensures issuance of all directives and revisions thereto. QA Program documents ensure that:
- 1. The need for special controls, processes, test equipment, tools, and skills is specified when necessary to ensure that required quality is attained in performance of the activity.
- 2. Quality is verified by inspections and tests.
- 3. Personnel who perform activities affecting quality achieve and maintain suitable proficiency through appropriate training and experience.
Administrative or Technical Procedures are prepared as needed. They establish the processes used to implement directive or QAP requirements. The controls for review and issue of procedures are discussed in Sections 1B.5 and 1B.6. Review of Operations Procedures require that CCNPP shall be operated and maintained in accordance with the plant TSs and operating license. The followmg orgamzations review plant operations to ensure that these procedures are followed: )
- 1. The Manager-NQAD provides independent verification that the requirements contained in the Plant's .
operating license, UFSAR, TSs, and plant procedures are met. This is accomplished through l quality assurance audits.
- 2. The OSSRC provides independcat verification by review that CCNPP is operated in accordance with established requirements. The OSSRC, which functions under a written Charter approved by the Vice President-Nuclear Energy Division, is composed of on-site and off-site personnel knowledgeable of in-plant operations, nuclear engineering, chemistry and radiochemistry, I metallurgy, radiological safety, instrumentation and control systems, mechanical and electrical systems, quality assurance, and environmental factors. The Page 18 of 69 i
QUALITY ASSURANCE POLICY Revision 42_41 l proceedings of all meetings are documented and sent to the Vice President-Nuclear Energy Disision, Committee members, and others designated by the Committee Chairman.
- 3. The on-site POSRC reviews matters pertaining to nuclear plant safety. This Committee screens subjects of potential concern to the OSSRC and performs preliminary investigations under the direction of the Plant General Manager. POSRC membership and functions are governed by Technical-Fpeifteations Addendum IB-1 and written procedures. [12) The results of all meetings are documented and sent to the members of the OSSRC, and others designated by the Committee Chairman.
The maintenance and repair of systems, stmetures, and components subject to the QA Program are performed by personnel under the direction of the General Supervisors of Electrical and Controls, Radiation Safety, and Mechanical Maintenance, according to written procedures and instructions as prepared by the maintenance force and approved as stated in QA Program documents. These Procedures: i
- 1. Ensure that quality-related activities, such as inspections and tests, are performed with appropriate equipment and under suitable emironmental conditions.
- 2. Indicate inspections and checks that must be made and records and data that must be kept.
- 3. Show where independent verifications of inspections or checks should be performed by specified personne' other than those performing the work. ,
When necessary, non-plant Company personnel or outside contractors are brought in to supplement the , plant work force. In such instances, the approval of work procedures and the tagging of equipment are coordinated by a member of the BGE organization responsible for the performance of the work. , Controls are established in QA Program documents to ensure that materials and parts used in the repair, maintenance, and modification of SR and designated NSR portions of the plant are appropriate for the senice intended. Written pracedures are prepared for the storage and identification of materials and parts
- to ensure that they do not deteriorate in storage and can be correctly identified before installation or use.
Equipment manufacturers and contractors used for the repair, maintenance, and modification of SR and designated NSR structures, systems, and components are required to have quality assurance programs consistent with the importance of the end-product to safety. 18.3 p_ESIGN CONTROL , Control > Plant changes which affect the design, function, or method of performing the function of a structure, l system, or component described in the UFSAR and-are c trolled by QA Program documents which are l . written to ensure compliance with Regulatory Guide 1.64 and 10 CFR 50.59. i Controls for changes, tests, and experiments conducted at CCNPP vary according to the following:
- 1. As the item or activity affected is or is not described in the UFSAR.
- 2. As the item or activity affected has been classified SR or NSR.
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QUALITY ASSURANCE POLICY Revision 42.M l
- 3. As the item or activity affects or does not affect nuclear safety.
- 4. As the proposed change, test, or experiment does or does not constitute an Unresiewed Safety Question or require a change to the TSs.
To ensure compliance with 10 CFR 50.59, the process for controlling changes, tests, or experiments has l been disided into classifications. Procedures required by QA Program documents describe and control the ! method for determining the appropriate process classification. The process classifications control the preparation and reporting of safety evaluations. Three methods of treatment are allowable:
- 1. Implementing the change, test, or experiment in accordance with Company practice for operating power plants, or in accordance with Procedures required by QA Program documents.
- 2. Implementing the change, test, or experiment in accordance with Procedures required by QA Program documents but controlling the change, test, or experiment with a process classification.
- 3. Controlling the change, test, or experiment with a process classification and not allowing the implementing actisity to begin until the review requirements of 10 CFR 50.59 and 10 CFR 50, Appendix B, have been met.
Changes, tests, or experiments which require approval by the NRC are-approved reviewed by the POSRC l i and by the OSSRC. Controls have been established to ensure that design changes to SR stmetures, systems, and components are reviewed either by the organization that made the original design or by a Responsible Design
- Organization (RDO) that meets requirements specified in ANSI N45.2.11, Section 8.0.
Responsible Desian Orcanizations RDOs, either on contract or within BGE, ensure that: ;
- 1. Applicable regulatory requirements and design bases requirements are correctly translated into j specifications, drawings, written procedures, and instreiions.
j 1
- 2. Appropriate standards for quality are specified in design documents, and deviations and changes !
from such standards are controlled. l
- 3. Suitable design controls are used in applying principles of reactor physics; making seismic, stress, thermal, hydraulic, radiation, and accident analyses; ensuring compatibility of materials; and providing accessibility for in-senice inspection.
- 4. Designs are resiewed to ensure that design characteristics can be controlled, inspected, and tested, and that inspection and test criteria are identified.
- 5. Interfaces, both external and internal, are controlled for the activities of all participating organizations.
- 6. Methods for verifying or checking, such as design reviews, alternative calculations, and qualification testing are properly chosen and followed; the most adverse design conditions are specified for test programs used to verify the adequacy of designs.
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l i QUALITY ASSURANCE POLICY l Raision-42.43 l
- 7. Individuals or groups responsible for design verification are other than the original designer and the i designer's inanediate supenisor. ;
- 8. Design and specification changes are subject to design controls and approvals applicable to the original design. j
- 9. Design documents and resisions thereto are distributed to responsible individuals and controlled to ;
prevent inadvertent use of superseded material. ;
~
- 10. Design errors and deficiencies that adversely affect SR structures, systems, and components are f documented, and appropriate corrective action is taken. - l t
i1. Design documents and reviews, records, and changes thereto are collected, stored, maintained, and [ controlled systematically. ! t
- 12. Standard off-the-shelf commercial or previously approved materials, parts, and equipment essential to the SR functions of structures, systems, and components are reviewed for suitability- of i application before they are selected.
- 13. He persons or groups responsible for design reviews and other design verification activities and I their authority and responsibilities are identified.
- 14. Design changes to NSR items initiated and approved at the plant are controlled to ensure compliance l with 10 CFR 50.59. l 15, Processes used to select suitable materials, parts, equipment, and processes for SR structures, j systems, and components includes the application of pertinent industry standards and specifications, l material and prototype hardware testing programs, and design reviews. :
- 16. Computer programs used in design are subject to design controls and program verification.
1B.4 PROCUREMENT DOCUMENT CONTROL (5) Controls have been established to specify the requirements and sequence of actions for: requesting items or [ senices; review of the requested item or senice to establish the necessary technical and quality i requirements; preparation, review and control of procurement documents; evaluation and selection of ! vendors and; control of deviations from the procurement document requirements. i The degree to which these controls are imposed on the purchase ofitems and senices by BGE for CCNPP l depends on: !
- 1. The functional (safety) classification of each item or senice as SR or NSR according to controls !
established by the RDO and l 1
- 2. The Procurement Category of the item within it's functional classification as a basic component, i commercial grade item, designated non-safety related item (DNSR) or NSR stem:
l r
- a. Commercial Grade - An item satisfying all three of the following criteria:
l
- 1. Not subject to design or specification requirements that are unique to nuclear facilities, and t
- 2. Used in appbcations other than nuclear facilities; and '!
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QUALITY ASSURANCE POLICY Revision 4233. }
- 3. Is to be ordered from the mam iacturer/ vendor on the basis of specifications set forth in the vendor's published preact description (for example, a catalog).
- b. Basic Component - An item either procured as a safety related item or as a commercial grade '
item which has been accepted and dedicated for safety related application. This term is synonymous with " Safety-Related Component".
- c. Designated Non-Safety Related - A NSR item which BGE has made a regulatory or design basis commitment; or, for plant availability reasons, BGE has implemented special controls to assure reliability. These NSR items are included within the quality assurance process.
- d. Non-Safety Related - An item that does not perform a safety related function. 1
- 3. The procurement method to be used for the item or senice:
Purchase Orders placed by BGE personnel for items or senices intended for safety related applications and DNSR items and senices fall into two categories, Nuclear Grade Method procurement and Commercial Grade Method procurement.
- a. Nuclear Grade - Purchases that are designated to be placed with vendors that maintain a 10CFR50 Appendix B quality program and supply items that meet the defmition of Basic Component. The requirements of 10CFR21 will be invoked on the vendor under this method. ,
- b. Commercial Grade - Purchases that are designated to be placed with commercial grade vendors that supply items or senices that meet the definition of Commercial Grade. These items must be dedicated for SR use by BGE.
Items and/or senices classified as DNSR will be purchased using the Commercial Grade Method with technical requirements established by :m RDO. Qualified NSSD/NED personnel trained in quality assurance program requirements with RDO authori:y review safety-related and designated non-safety related procurement documents for proper inclusion of technical and quality requirements. Personnel in NSSD/NED review safety-related and designated non-safety related procurement documents to ensure that the requirements stated therein are correct, inspectable, controllable, contain adequate acceptance and rejection criteria, and comply with the requirements of the procurement program. These reviews and approvals are documented prior to ' placement of the purchase order. All changes made to procurement documents, including specifications and other technical attachments, are ) subject to the same levels of review, approval and control that were applied in preparing and processing the original documents. Bids submitted to supply safety-related items or senices receive the same review and approval cycle as used for safety-related procurement requisitions. Vendor Selection Personnel in NQAD evaluate vendors who provide SR and designated NSR items and senices to verify they can provide acceptable items and senices. i i I l Page 22 of 69 i i __________________________..________._____________J
l l QUALITY ASSURANCE POLICY Resision4233 l Controls for Nuclear Grade Purchases i Controls have been established to ensure that, before placement of a purchase order under the Nuclear Grade method of purchase, there is esidence of the following:
- 1. The vendor has been evaluated as stated in Section IB.7 of this policy and found to have a ;
satisfactory QA program.
- 2. The item to be purchased is manufactured under the requirements of the evaluated and approved program.
Controls for Commercial Grade Purchases Controls have been established to ensure that items or services available to general industry will be sufficiently controlled to perform their SR and designated NSR function. NSSD/NED personnel will specify the acceptance methods to be used to verify the critical characteristics identified in the procurement document (s). Emeurement Document Reauirements Procedures require that procurement documents shall:
- 1. Reference part numbers or descriptions, and additional requirements to ensure that items ordered can be identified and verification can be made that each item received is the item ordered.
- 2. Contain/ reference technical requirements for the basis of design, by including the applicable regulatory requirements, component and material identification, RDO approved drawing and specification, codes, industrial standards, test and inspection requirements, and special process l
instructions such as welding, heat treating, nondestructive testing, and cleaning.
- 3. Identify the requirements of 10 CFR 50, Appendix B, wl.ivh must be complied with and described in !
the vendor's QA program, for Nuclear Grade Purchases.
- 4. Require that major contractors designated as BGE agents to purchase SR and designated NSR items i or senices must have procurement controls to ensure they purchase or acquire these items or senices in compliance with the necessary sections of ANSI N45.2.13.
- 5. Identify required documentation (i.e., drawings, specifications, procedures, inspection and ;
fabrication plans, inspection and test records, personnel and procedure qualifications, and material chemical and physical test results) to be prepared, maintained, and submitted to BGE or the ; purchaser for review and approval. !
- 6. Identify records which must be retained, controlled, maintained, or delivered to BGE or the r purchaser before use or installation of hardware. #
- 7. Specify BGE or its agent's right of access to vendor facilities and secords for source inspection, surveillance, verification and audits.
- 8. Identify requirements of the vendor's quality control process which must be implemented when providing a commercial grade item. i
- 9. Reference or specify the critical characteristics that a commercial grade item must possess to ensure [
that the item received is the item specified. ! l i Page 23 of 69 !
QUALITY ASSURANCE POLICY Revision 42_G l
- 10. Incorporate the requirements of 10 CFR 21 for Nuclear Grade procurements.
I 1. Include requirements for QA program elements to be passed on to sub-vendors. 18.5 INSTRUCTION
S. PROCEDURE
S. AND DRAWINGS Controls delineate the sequence of actions to be performed in the preparation, review, approval, and control ofinstructions, procedures, and drawings. Controls require that:
- 1. Methods for complying with each of the applicable criteria of 10 CFR 50, Appendix B, must be specified in instructions, procedures, and drawings.
- 2. Instructions, procedures, and drawings must specify appropriate quantitative (such as dimensions, tolerances, and operating limits) and qualitative (such as workmanship samples) acceptance criteria for verifying that important activines have been satisfactorily accomplished.
Controls ensure that:
- 1. The QA Policy is appaved by the Vice President-Nuclear Energy Division. (1)
- 2. QAPs are developed by Departments responsible for conducting panicular activities, reviewed by the managers of the responsible department (s) for that particular activity, and approved by the Vice President-Nuclear Energy Division. (1)
- 3. Nuclear Program Directives are prepared under the direction of the Department Manager assigned as the Program Sponsor. AfTected Department Managers review directives and their revisions. The Manager-NQAD ensures directives are reviewed by the Quality Assurance organization and approval recommended to the Program Sponsor. The responsible Program Sponsor approves directives and their revisions. Directives are prepared, reviewed, and_ approved-and-periodically reviewed according to an appendix to the Nuclear Program Directives Manual. [12)
- 4. Procedures are prepared, approved, and controlled according to the Control Procedures. Control Procedures establish review, approval, revision, and. change, and periodic review requirements for applicable procedures. [12) If format and content requirements are not contained in Control Procedures, they shall specify the document to be used to determine format and content requirements. Control Procedures are reviewed by the Quality Assurance Organization. Other procedures are resiewed by Quality Assurance on a requested basis.
- 5. Basis items added u -ing procedure revisions or changes will be recorded. (1)
Ihe Plant General Manager mav_ designate specific procedures or classes of crocedures in writing to be reliewed by the Procedure Review Committee or by Oualified Reviewers in lieu of review by the POSRC, Review by the Procedure Review Committee and by Oualifed i Reviewers shall be in accordance with Addendum IB-l. fl9) EIQcedures listed in Technical Specification 6.4 shall be approved by the Plant General Manager or by cogninnt Managers. SuocIintendents. or General Supervisors (or Directors) that reoort directly to a Manager prior to imnlementation as specified by administrative reauirements. The approval authonty for specific procedures or classes _of. procedures shall be designated in writing by the Plant Ggneral Manager and shall be a different individual from the Oualified Reviewer. H 9) Temporarr changes to procedures of Technical Specification 6.4 may br made orovided; Page 24 of 69 ,
QUALITY ASSURANCE POLICY Revision 42_O \ ;
- a. The intent of the original orocedum is not altered.
b1 The chnnpe is anoroved by two members of the oinnt management staff at least one of whom holds l n Senior Renetor Ooerator's License on the unit affected. .!
- c. The change is documented reviewed by the POSRC. the Procednre Review Committee or by n Qunlified Reviewer and anproved by the designnted anoroval authority within 14 davs of l imnlementntion. (19) i 1B.6 DOCUMENT CONTROL l Requirements have been established to control the documentation of activities controlled by the QA - ,
Program. QA Program controlled documents include the UFSAR; Operating-License, including the i Technical Specifications; EmerEency Response Plan; Security Plan; QA Policy; the ISFSI updated Safety Analysis Report (SAR) and Materials License, including Technical Specifications; procedures; . specifications; and drawings. Revisions to the QA Policy are controlled by QA Program documents which are written to ensure compliance with 10 CFR 50.54(a)(3). Alterations to the UFSAR are controlled by QA Program documents which are written to ensure compliance with 10 CFR 50.71. Alterations to the ISFSI updated SAR are controlled by QA Program documents which are written to ensure compliance with 10 CFR 72.70. Alterations to the Operating License, including the Technical Specifications, are controlled by QA Program documents which are written to ensure compliance with 10 CFR 50.59(c),10 CFR 50.90 and 10 CFR 50.92. Alterations to the ISFSI Materials License, including the technical specifications, are controlled by QA , Program documents which are written to ensure compliance with 10 CFR 72.48(c),10 CFR 72.56, and 10 : CFR 72.58. Alterations to the Emergency Response Plan are controlled by QA Program documents which are written to ensure compliance with 10 CFR 50.54(q), and with 10 CFR 72.44(f) for the ISFSI.
- Alterations to the Security Plan are controlled by QA Program documents which are written to ensure compliance with 10 CFR 50.54 (p), and with 10 CFR 72.44(c) for the ISFSI.
QAPs are required to:
- 1. Establish controls to ensure that iegulatory requirements and BGE commitments will be implemented.
- 2. Describe interdepartmental interfaces and establish controls for interdepartmental actisities.
- 3. Specify how important activities, such as plant maintenance or in-sersice inspection, are to be performed, and give sufficient detail to control the performance of the activity or to ensure that i requirements for lower-level procedures are clearly specified. j
- 4. Be prepared and controlled in accordance with QA Program documents that describe the format, sequence of topics, contents, review and approval, issue and distribution, and requirements for revision and record retention.
Page 25 of 69
QUALITY ASSURANCE POLICY Resision42_O l During the review of each QAP, compliance uith applicable criteria speciihi in 10 CFR 50, Appendix B, is verified and documented. The Manager-NSSD, is responsible for issuing, revising, and controlling QAPs. QAPs are developed by one of the departments responsible for the subject activities. Each procedure is given a compliance review by a member of the Quality Assurance Organization, and technical review by a member of one of the responsible departments. Each QAP is reviewed by department manager (s) who have responsibilities for activities governed by that QAP, and the Managers of the Nuclear Energy Division. (1) Each QAP is approved by the Vice President-Nuclear Energy Division and issued by the Manager-NSSD. (1) Directives are required to:
- 1. Establish controls to ensure that regulatory requirements and BGE commitments will be implemented.
- 2. Establish controls to ensure that management requirements will be implemented.
- 3. Assign responsibilities and interfaces within the program.
- 4. Be prepared and controlled in accordance with an appendix to the Nuclear Program Directives Manual that describes the format, contents, review and approval, and. revision.srand-periodie-rewew requirements.. (12]
Nuclear Program Directives are prepared and technically reviewed under the direction of the Department Manager assigned as the Program Sponsor. Each directive is reviewed by affected Department Managers. Each directive is given a compliance review by a member of the Quality Assurance organization. Nuclear Program Directives are approved by the sponsoring Manager after ensuring resolution and incorporation of , QA compliance review comments. (9) The Manager-NSSD ensures irsuance of each directive. Administrative and Technical Procedures are prepared when needed to implement QA Program document requirements according to a Control Procedure. Individual organizations are responsible for preparing, i revising, issuing, and controlling procedures. Each procedure is given a technical review under the l direction of the sponsoring organization. The Quality Assurance organization performs compliance reviews on Control Procedures. Other procedures are reviewed by Quality Assurance on a requested basis. Organizations that issue instructions, procedures, specifications, or drawings are required to establish controls that ensure the following:
- 1. Changes to a document are reviewed and approved by the organization that performed the original review and approval unless the control procedure designates another qualified responsible organization.
- 2. Approved changes are promptly incorporated into instructions, procedures, drawings, and other documents associated with the change.
- 3. Obsolete or superseded documents are controlled to reduce the possibility of inadvertent use.
Superseded documents retained for reference are marked and stored in separate files. Other superseded documents are removed from the files. When changes to drawings or specifications are required, change requests are prepared by the organization that desires the change. Requests are reviewed and approved by BGE RDOs. Page 26 of 69
QUALITY ASSURANCE POLICY Revision 42.M \ 1 B.7 .CPNTROL OF PURCHASED MATERIAL. EOUIPMENT. AND SERVICES (5) NQAD, NSSD, NED, and PMMD personnel are responsible for the control of purchased items and services for SR and designated NSR applications at CCNPP. The controls include: Accepting items or senices only from vendors who have been evaluated and selected in accordance with this policy. Procurement documents for spare or replacement parts of structures, systems, and components as designated under the QA Program subject to controls at least equivalent to those applied to the original equipment, or an evaluation / justification shall be documented when less stringent controls are involved. Vendor surveillance, verification and audit activities, and receipt verification are conducted to ensure the vendors comply with specified technical and quality requirements, and ensure items are identified, stored, handled and shipped in accordance with procurement document requirements. Vendor Evaluation The vendor evaluation is conducted to detennine acceptability of a vendor to provide the requested item or senice, to determine what vendor programs, procedures and documents need to be invoked by the procurement document, determining the vendor's performance history for supplying items to CCNPP and assessing the need to impose source surveillances and/or verifications during the manufacture ofitems or perfonnance of services for BGE. Vendor evaluations depend on the procurement classification of the item (s) being supplied. The National Institute of Standards and Technology (NIST), by virtue ofits being the nationally recognized standard, is an acceptable provider of calibration masters, standards or senices. Utilities holding an NRC Construction Permit or Operating License are acceptable suppliers of all items except for those items to be ! used in an ASME Boiler and Pressure Vessel Code Section til application. Neither of the above are l required to be listed on the Approved Vendors List (AVL). Nuclear Grade NQAD performs evaluations and audits to verify that the vendor has developed and implemented an acceptable quality assurance program that complies with the requirements specified in the procurement specification or proposed procurement specification. These evaluations and audits Le conducted and dm umented using written procedures or checklists that identify the QA requirements applicable to the items supplied. Commercial Grade Since BGE accepts the responsibility of verifying the conformance of commercial grade items and/or senice, they may be procured from vendors with no formal quality assurance program. In this instance, BGE dedicates the commercial grade item and/or senice for SR use. A survey may be performed of commercial vendors to assess what, if any documented controls are implemented in the manufacture ofitems or performance of senices for BGE. Vendor controls evaluated to be satisfactory may be invoked as requirements within the purchase order and may be used as part of the basis for acceptance of the item. Page 27 of 69
I l QUALITY ASSURANCE POLICY l Revision 42_43 1 1 The depth of vendor evaluation varies according to the complexity and function of the item ! involved and to the role of the vendor in acceptance of the item. Vendor Approval Upon completion of the evaluation, satisfactory vendors are added to BGE's AVL. The vendors on this list are evaluated on an annual basis and subject to re-audit or commercial grade survey on a triennial basis to verify continued compliance with BGE's requirements. An auditing organization such as NUPIC, another utility, a contractor to BGE, etc., may be used to verify that the vendor has developed and implemented a QA program that complies with 10 CFR 50, Appendix B or a commercial grade program that complies with the requirements of BGE's procurement requirements or similar requirements. Page 28 of 69
QUALITY ASSURANCE POLICY Revision 42.41 \ When required by operational considerations, an order may be placed with a vendor prior to completion of the evaluation and approval process only after obtaining the Manager-NSSD's approval. BGE's acceptance of basic component items or senices provided by an unapproved vendor is contingent on the subsequent NQAD evaluation and approval of the vendor as stated above. Ve_rification of Vendor Activities Vendor surveillance, and source verification activities are conducted by qualified NQAD personnel in accordance with written procedures or checklists. These procedures or checklists, along with the procurement documents, specify the characteristics or processes to be witnessed, inspected or verified. Personnel performing these activities are qualified to establish whether or not a vendor is capable of providing products of acceptable quality. The depth and frequency of vendor sun >cillances, verifications and audits is commensurate with the complexity and ftmetion of the item or service and the ability of the vendor to provide the necessary assurance of acceptability. When a vendor's certificates of conformance are used as part of the acceptance of an item or senice, the validity of these documents is periodically evaluated and documented by the above mentioned processes. Receip.1 NSSD is responsible for receiving and storing materials, parts, and components. Additionally, NSSD is responsible for performing standard and special receipt inspections and coordinating testing necessary to accept SR items, designated NSR items and commercial grade items for SR use. Standard receiving inspection ofitems is performed to assure the following:
- 1. The item is properly identified and that this identification corresponds with the documentation received.
- 2. Stated packaging, shipping and handling requirements have been maintained.
- 3. Items have not been damaged, workmanship is of adequate quality, and the items are adequately clean in accordance with procurement document requirements.
- 4. Documentation required by the Purchase Order has been received and is reviewed to assure that the item conforms to the purchase order requirements.
Special receiving inspection may be required if the item was not inspected at the source; when requested by the RDO or; as part of the acceptance basis for commercial grade items. A written record of the results of the NSSD receipt inspection and the disposition of received items is maintained as part of pennanent plant records. All SR and designated NSR items accepted and released for issue to a controlled storage area or released for installation or further work bear an acceptance tag and have documentation to support their acceptability. If traccability is lost or the documentation review is unsatisfactory, an item becomes subject to the controls established for non-conf-ming items. Page 29 of 69 1
QUALITY ASSURANCE POLICY - Revisiort-42E l Non-conforming items are identified and handled in accordance with Section 18.15 of this policy and, when practicable, are placed in a segregated area to prevent inadvertent installation or use until proper disposition is made. ' Documentation BGE procurement documents require vendors to provide documentation identifying the purchased item and the specific procurement requirements that are met by the item. Vendor inspection records or certificates of conformance attesting to acceptance must be in the possession of BGE before the item may be released for lnstallation or use. However, an unacceptable item may be given a " Conditional Release" if there is reasonable assurance that it can be made acceptable after insta!!ation but before the system that contains it is considered operational. Items released under
" Conditional Release" must be controlled under the Non Conformance Report (NCR) system.
Vendor requested deviations from procurement document requirements, including nonconfomtances dispwitioned "use-as-is" or " repair" must be submitted to BGE for evaluation and approval of the deviation or a recommended disposition prior to shipment. 1B.8 IDENTIFICATION AND CONTROL OF MATERIALS. PARTS. AND COMPONENTS (5) NSSD/NED personnel ensure that procurement documents require that SR and designated NSR items, , including partially fabricated sub-assemblics, are identified and controlled to prevent the use ofincorrect or defective material. Requirements for identification by use of heat number, part number, or serial number, or by other means, are referenced or stated in procurement documents. These documents require the identification to be placed on the item or in records traccable to the item so that the function and quality of the item are not affected. This identification is required to be maintained throughout fabrication, storage, crection, installation, and use. NSSD personnel ensure traceability information is correctly transferred to subdivided materials stored in the Warehouse. User organizations ensure traceability information is correctly transferred to subdivided materials after issuance from the Warehouse. NQAD is responsible for performing periodic inspections or surveillances to verify program adherence. Assigned NSSD personnel purchase identify, store, and issue items as specified by procurement controls and provide for maintaining the integrity of items and their traceability to associated documents during storage and issue. BGE contractors and their sub-contractors (who are approved to work on-site under their own QA program) are responsible for establishing and implementing programs in accordance with specified 5mments for identifying and controlling materials, parts, and components under theirjurisdiction. Identification of items important to the function of SR and designated NSR structures, systems, and components can be traced to appropriate documentation such as drawings, specifications, purchase orders, manufacturing and inspection documents, deviation reports, and physical and chemical mill-test reports. 1 I 1 l Page 30 of 69
QUALITY ASSURANCE POLICY t Revision 42_43 l Receipt SR and designated NSR items received at CCNPP are receipt inspected to verify that all requirements of j the procurement documents have been met. If a discrepancy is observed, such as damage or missing documentation, information to the effect is recorded on the receiving inspection report, and the discrepant item is identified as such and placed in a separate " hold" area when practicable. If the item is acceptable, it is identified to indicate acceptance and that it is approved for storage or installation and use. When groups ofitems in storage are subdivided, each subgroup is separately identified. If an item is found to be or is made discrepant during processing, it is identified as such and placed in a separate area when practicable. Acceptance documentation is required to be traceable to a purchase order, drawing, specification, requisition number, or assembly. As individual items are assembled, installed, and inspected, their acceptance-tag numbers are recorded in plant maintenance or operation records. After completion of tests and inspections, records that document test results and traceability are kept as part of the plant records. 1 B.9 CONTROL OF SPECIAL PROCESSES Controls , i Controls have been established for writing, qualifying, approving, and issuing procedures to control such special processes as welding, heat treating, and nondestructive testing used during the operation of CCNPP. Special Process Procedures:
- 1. Are prepared in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
- 2. Ensure that special processes are performed by qualified personnel according to qualified procedures that comply with applicable regulatory requirements.
- 3. Specify requirements for control, parameters to be considered, acceptable methods of documentation, and the codes, standards, specifications, or criteria which govern the qualification.
- 4. Define the necessary qualification of personnel, procedures, or equipment when special processes are not covered by existing codes or standards or when quality requirements for an item exceed the requirements of established codes or standards.
BGE contractors and their sub-contractors are responsible for controlling special processes used by them , and for maintaining records to verify that special processes are perfonned in accordance with requirements ! established by the portions of their QA programs that apply to special processes. Qualification of Methods Procedures, equipment, and personnel connected with special processes are qualified in accordance with applicable codes, standards, specifications, or supplementary requirements as follows:
- 1. Welding activities conducted by BGE are performed according to welding procedure specifications qualified in accordance with applicable welding requirements of the ASME Code. Each welding procedure specification is written, qualified, and approved in accordance Page 31 of 69 l
QUALITY ASSURANCE POLICY Revision 42 H \ with a controlling documented procedure. Copies of welding procedure specifications are made available to welders and, when required, to Authorized Inspectors. Before contracting for welding, the Principal Metallurgist resiews and approves non-BGE welding procedure specifications and procedure qualification records in accordance with a written procedure.
- 2. Heat-treating requirements included in welding procedure specifications are established in conformance with heat-treating requirements of the applicable ASME Code.
- 3. Nondestructive Examinations are performed to written procedures proved by actual demonstration, when practicable, to the satisfaction of the Principal Engineer - Nondestructive Examination -
Nuclear and, when required, the A Nrized Inspector. These procedures are prepared according to appropriate sections of the ASME Code for particular examination methods. Procedures. personnel qualifications, and the records that verify the Perfonnance of - Nondestmetive Examinations are kept as nuclear plant records. Nondestructive Examination Procedures describing methods not described in the ASME Code and/or SNT-TC-1 A and its Supplements are at least , equivalent to those recognized by the American Society of Mechanical Engineers and the American Society for Non-destructive Testing. Training programs acceptable to the Principal Engineer - Nondestructive Examination - Nuclear are developed to compleent these alternative methods and to establish the capability of personnel to perform the required examination according to BGE procedures and to the level of performance to which the individual will be certified. Methods of Nondestructive Examination include, but are not restricted to, radiographic, ultrasonic, liquid-penetrant, magnetic-particle, eddy-current, visual, and leak-testing examinations. Procedures are prepared to cover these examinations in accordance with a QA Program document that details the specific examination, requirements for approval, and content of the procedure, such as certification level, accept / reject criteria, examination coverage and sequence, surface preparation, test equipment, records required, permissible marking, cleanup requirements, and reference to applicable sections of the ASME Code. Qualification of Personnel Special processes are performed by certified personnel using written process sheets, shop procedures, checklists, and travelers (or equivalent), with recorded evidence of verification as follows:
- 1. BGE welders, and welders under contract to BGE, are qualified and certified in accordance with the applicable requirements of the ASME Code. The Principal Metallurgist maintains records of the welding procedure specifications, including essential variables under which the welders are examined, and the results of the examinations. A welder is not permitted to weld SR and designated NSR items until an appropriate performance qualification record, a letter of certification, or, in an emergency, verbal clearance from the Principal Metallurgin, is on file at CCNPP. Each welder is required to be requalified as specified in the applicable code.
- 2. Non-BGE welders are not pennitted to weld SR and designated NSR items until they are qualified and certified in accordance with the applicable requirements of the ASME Code.
l I l l l Page 32 of 69 !
QUALITY ASSURANCE POLICY Revision 42R l
- 3. Nondestructive Examination personnel employed by or responsible to BGE are certified according to applicable sections of the ASME Code and/or SNT-TC-1 A and its Supplements. BGE employees ,
are trained and certified in accordance with a written procedure. Non-BGE personnel are qualified to procedures approved by BGE, and their qualifications and certifications of personnel are verified according to written procedures. Qualification records of procedures, equipment, and personnel associated with special processes conducted by BGE are filed and kept curr:nt by the Principal Metallurgist or Principal Engineer - Nondestructive ! Examination - Nuclear. The Manager-NQAD provides independent verification that special processes are performed by qualified personnel. I B.10 INSPECTION ! Activities that affect the quality of SR and designated NSR items are inspected as specified in approved instructions, procedures, and plans which set forth requirements and acceptance criteria to ensure that work is done in conformance with particular requirements. Controls exercised during inspections ensure that:
- 1. Personnel who perform quality verification inspections are independent of the personnel who performed the activity being inspected.
- 2. Inspection procedures or instructions, with necessary drawings and specifications for use, are available before inspection operations are performed.
- 3. In the case of special processes, inspectors are qualified, and their qualifications comply with applicable codes and standards.
- 4. Test and measuring equipment is calibrated within required limits.
- 5. Inspection procedures, as applicable, specify objective acceptance criteria, prerequisites for performing inspections, limiting conditions, requirements for special equipment and Quality Verification (QV) hold-points at which inspections are to be witnessed. ,
- 6. Appropriate inspection requirements are established for modification, repair, and replacement.
- 7. Personnel who perform quality verification inspections are qualified in accordance with appropriate codes, standards, and Company training programs, and their qualifications and certifications are kept current.
- 8. Procedures for maintenance and modification are reviewed by QV personnel, or others authorized by QV, to detennine the need for independent inspection and the degree and method if such an inspection is required, and to ensure the identification ofinspection personnel and the documentation ofinspection results.
Page 33 of 69
QUALITY ASSURANCE POLICY Revision-4LG l
- 9. Procedures for Nondestructive Examination, excluding visual examination procedures performed on nuclear fuel, are reviewed by qualified personnel in FED. Procedures for nuclear fuel inspection and visual examination on nuclear fuel are reviewed by qualified personnel in NED. Review is to determine the adequacy of procedural controls and of inspection criteria, the need for independent inspection, and the degree and method, if such inspection is required; and to ensure the identification of qualified inspection personnel and the documentation ofinspection results.
- 10. Inspection results are recorded, evaluated, and retained. ;
Inspection procedures, instructions, and checklists used by inspection personnel provide the following:
- 1. Identification of characteristics to be inspected.
- 2. Acceptance and rejection criteria.
- 3. Description of the method ofinspection.
- 4. Identification of required procedures, drawings and specifications.
- 5. Identification ofinspector or data recorder.
- 6. Verification of completion and certification ofinspection.
- 7. Record of results ofinspection.
- 8. Provision for identifying mandatory inspection hold-points for witness for an authorized inspector or BGE inspection personnel.
- 9. Provision for indirect control by monitoring processing methods, equipment, and personnel if direct inspection is not possible.
- 10. Specification of necessary measuring and test equipment including requirements for accuracy.
The General Supenisor-Quality Verification (GS-QV) is responsible for the preparation and implementation of procedures for inspection and surveillance activities performed by or for QV. (11) Other inspections are conducted randomly to verify that overall plant operations are being conducted according to approved procedures and to ensure that the use of jumpers is properly documented; that equipment is returned to operating status after test, modification, or repair; that instruments are properly calibrated; and that personnel who perform tests are properly trained and qualified. In-senice inspections are perfonned on pressure-containing components within the reactor coolant system boundary according to requirements of the TSs. In-senice inspections and examinations on components designated Class I or Class 11 by the ASME Code are witnessed or otherwise verified by an authorized Code Inspector who is responsible for ensuring that the work is performed by qualified personnel according to written qualified procedures. Records of in-senice inspections, results, corrective action required and taken, inspection standards required for repair, and results of inspection of repairs are maintained and compared with the results of subsequent examination. i Page 34 of 69 l
QUALITY ASSURANCE POLICY Revision 42.M l 1B.11 TEST CONTROL To demonstrate the ability of SR and designated NSR structures, systems, and components to function as designed, they are subjected to a program of surveillance and operational testing. Procedures specify the systematic development, review, approval, and conduct of tests and review of test results. Conditions such as failures, malfunctions, deficiencies, deviations, and non-conformances discovered during testing are documented and evaluated. Whenever testing is required to demonstrate that SR and designed NSR material, parts, components, or systems will perform satisfactorily in sersice, a test program is established and procedures are used that have been written and approved in accordance with basic requirements. Nuclear Engineering Department, and CCNPPD conduct tests to verify that plant behavior conforms to design criteria, ensure that failure and substandard performance are identified and controlled, and demonstrate satisfactory performance after plant modification and maintenance actinties. Written test procedures are developed, reviewed, and approved before testing is performed. They specify j instructions for testing, methods of test, test equipment, and instrumentation; and for the following as applicable:
- 1. Adequate and appropriate equipment. !
1
- 2. Preparation, condition, and completeness ofitem to be tested.
- 3. Suitable and controlled environmental conditions.
- 4. Mandatory inspection hold-points for witness by BGE inspection or . uthorized inspector personnel.
l S. Provision for data collection and storage. { l
- 6. Acceptance and rejection criteria. j
- 7. Methods of documenting or recording test data and results.
- 8. Provision for ensuring that test prerequisites have been met.
l Test results are documented and evaluated; they are accepted or rejected by a qualified, responsible individual or group. Results of completed tests on SR and designated NSR structures, systems and components (per Q-List) that identify a malfunction or were out of specification are reviewed and evaluated by the POSRC and accepted and approved by the Plant General Manager. Test records are kept in sufficient detail to nuke possible an evaluation of test results and to show how individual tests demonstrate that SR and designated NSR structures, systems, and components and the plant as a unit can operate safely and as designed. SR l and designated NSR test records are retained as plant history records. Results of testing performed as part of receipt inspection are evaluated, accepted and approved by qualified NSSD personnel. (5) Page 35 of 69
QUALITY ASSURANCE POLICY Revision 4U3 { 1B.12 CONTROL OF MEASURING AND TEST EOUIPMENT Calibration controls have been established to prescribe the technique and frequency of calibration, maintenance, and control of measuring and test instruments, tools, gauges, fixtures, reference and transfer standards, and nondestructive test equipment used in measuring, inspecting, and monitoring SR and designated NSR components, systems, and structures during the operations phase of CCNPP. Personnel of the following functional organizations control, calibrate, and adjust measuring and test equipment: System Operation and Maintenance Department Calvert Cliffs Nuclear Power - Radiation Safety Plant Department - Performance Engineering Unit Mechanical Maintenance
- Electrical and Controls
- Nuclear Operations Chemistry Calibration controls require cach group to identify measuring and test equipment and calibration test data related to it.
Written procedures are prepared and implemented to ensure that tools, gauges, instruments, and related test and measuring devices are of proper accuracy to verify conformance to established requirements. Manufacturer's Procedures are used for calibration or a procedure is prepared for each category of measuring and test equipment as necessary. These Calibration Procedures contain the following infomiation:
- 1. Identification of the item to be calibrated and its period of calibration.
- 2. Standards to be used, specific test-points, and checks, tests, and measurements to be made.
- 3. Acceptance criteria to be used and special precautions to be taken when necessary.
Measuring and test equipment that require calibration are assigned an identifying serial number. Instruments are calibrated at specified intervals according to the required accuracy, purpose, degree of usage, stability characteristics, and other conditions that alTect the measurement. When equipment is found out of calibration, an evaluation is made by the supenisor responsible for that equipment to determine any adverse effect on items previously accepted on the basis of using that equipment. Test and measuring equipment that cannot be adjusted to required tolerances during calibration is identified and placed in a designated segregated area; if the equipment can be used in limited applications, the limitations are identified. The status of each item controlled under the calibration system is recorded and maintained. Equipment is marked or records of calibrations are maintained to indicate calibration status. An interval of calibration is established for each item of measuring and test equipment and recorded on a master record of calibrations prepared as a calibration schedule. Page 36 of 69
QUALITY ASSURANCE POLICY Revision 4L41 } Measuring and test equipment is controlled to prevent the use of uncalibrated or defective equipment, the spread of radioactive contamination, the introduction ofimpurities into high-purity systems, and damage to or loss of equipment. Identification tags are placed on measuring and test equipment to indicate such special conditions as radioactive cleanliness, special limitations, or failure to meet established calibration requirements. Measuring and test equipment is calibrated and adjusted at specified intervals, or before use, against I certified standards. Reference and transfer standards are traceable to nationally recognized standards; or, where national standards do not exist, provisions are established to document the basis for calibration. i' 18.13 HANDLING. STORAGE. AND SHIPPING Appropriate and special requirements for handling, preservation, storage, cleaning, packaging, and shipping of SR and designated NSR items are specified in procurement documents. Procedures have been established to ensure that the handling, preservation, storage, cleaning, packaging, and shipping of SR and designated NSR items are performed in accordance with specified requirements to reduce the likelihood of damage, loss, or deterioration by such environmentas conditions as temperature or humidity. Special handling, preservation, storage, cleaning, packaging, and shipping activities associated with SR and designated NSR items are perfonned by suitably trained personnel in accordance with specific written procedures. Controls have been established for the safe storage of hazardous materials. Items with a limited shelf-life , are controlled to ensure that they will not be used in SR and designated NSR applications after expiration of designated shelf-life periods. 18.14 LN_SPECTION. TEST. AND OPERATING STATUS Controls have been established for the application and removal of status indicators such as tags, markings, labels, and stamps to ensure that the inspection, test, and operating status of SR and cesignated NSR structures, systems, and components is clearly indicated at all times. Procedures / instructions are prepared to identify and control inspection, testing, and operating status by the use of logs, forms, and tags that identify the inspection, test, and operating status of structures, systems, and components; control the use of indicators, including the authority for their application and removal; control bypassing operations, such as jumping or temporary removal of electrical leads; and identify non-confonning, inoperative, or malfunctioning structures, systems, or components. Senior shift personnel are responsible for aligning, isolating, and appropriately tagging installed equipment and systems so that activities affecting quality can be perfbrmed. The Manager-NQAD is responsible for the perfonnance of surveillances to verify that the inspection, testing, and operating status of structures, systems, and components are properly identified and controlled during operation, maintenance, and testing of the plant. The bypassing of required inspections, tests, and other critical operations is controlled to ensure that bypassed inspections or tests are properly documented and that the effect of bypassing the inspection or test is evaluated by the organization responsible for specifying the inspection or test. Controls have been established to ensure that the status of non-conforming, inoperative, or malfunctioning SR and designated NSR structures, systems, or components is identified to prevent inadvertent use. Page 37 of 69
1 QUALITY ASSURANCE POLICY Revision 42.43 l l 1B.15 NONCONFORMING MATERIALS. PARTS. OR COMPONENTS (6) j i Controls have been established for identifying, documenting, segregating, reviewing, dispositioning, and notifying affected organizations ofIssues afTecting materials, parts, or components (i.e., items). Issues affecting nuclear plant items are referred to as nonconformances. Nonconformances are hardware deficiencies which render the quality of an item unacceptable or indetenninate. Any individual identifying an actual or suspected nonconforming item is responsible for documenting and reporting such nonconforming item promptly to supenisory or Nuclear Quality Assurance Department personnel. Nonconfonning items are controlled by documentation, marking, logging, tagging, or physical segregation to prevent inadvertent installation or use. Nonconfonnance control documents are submitted to responsible departments for resolution. Designated personnel have the responsibility and authority for approving the resolution of nonconformances. Nonconformance control documents are not closed until corrective actions have been completed. Nonconforming items are dispositioned as rework, repair, reject, or accept-as-is. The disposition of a repair or accept-as-is nonconformance is treated as a design change and is evaluated and approved or rejected by the RDO. Reworked, repaired, and replacement items are inspected and/or tested in accordance with the original inspection and/or test requirements or acceptable alternatives to ensure that critical characteristics possibly affected by the nonconformance remain acceptable. Nonconforming items may be conditionally released for installation, test, energization, pressurization, or use if the conditional release will not adversely affect nor preclude identification and correction of the nonconformance. Nonconfonning items required for Technical Specification operability may be released for use following verification that the nonconforming item meets all operability requirements specific to its function and is approved for use by authorized Operations personnel. Conditionally released items will be resolved in accordance with this Section. Conditional release evaluations are documented, reviewed, and approved prior to implementation. IB.16 CORRECTIVE ACTION (6) Controls have been established to ensure that Issues are identified, documented, reviewed, and corrected. These controls are applied to deficiencies associated with the programmatic content, process, and implementation of the Quality Assurance Program as well as nonconfonnances (ref Section 1B.15). Corrective actions are implemented by responsible personnel and may include immediate actions, remedial actions and/or actions to prevent recurrence, based on the significance and extent of the issue. Issues identified as potentially impacting the safe production of nuclear power are evaluated for Technical Specification Operability, NRC Reportability, Nuclear Safety Significance, and if the activity should be stopped. The VP-NED, or designated alternate, is informed ofIssues which require NRC notification. Corrective action verification is perfo med for Significant issues prior to the close-out of the corrective action document. Verification is performed and documented by individuals not directly involved with implementing the corrective action (s). Unacceptable corrective action (s) are reported to supenitory or ' management personnel directly responsible for resolving the issue and to progressively higher levels of management until the issue is resolved. Page 38 of 69
I QUALITY ASSURANCE POLICY Revision 4M3. I Significant Issues require a root cause analysis and the implementation of corrective actions to prevent ) recurrence and are reported to management for review and assessment. l Issues are periodically analyzed for the identification of adverse quality trends. The existence of an adverse l quality trend is resolved in accordance with this section. A Trend Report is issued to management at j intervals specified in approved procedures. 18.17 OUALITY ASSURANCE RECORDS l Controls have been established to ensure that quality assurance records are maintained to provide ; documentary evidence of the quality of SR and designated NSR items and activities. Applicable design specifications, procurement documents, test procedures, operational procedures, QAPs, TSs, and other documents specify records that should be generated, supplied, or maintained by and for BGE. I Quality assurance records are classified as lifetime or non-permanent. l Lifetime records, maintained for particular items for the life of CCNPP, for particular items have significant value in relation to demonstrating capability for safe operation; maintaining, reworking, repairing, replacing, or modifying an item; detennining the cause of an accident or malfunction of an item; and providing required baseline data for in-senice inspection. 1 Non-permanent records, which show evidence that a SR and designated NSR activity was performed in ! accordance with applicable requirements, are retained for periods sufficient to ensure BGE's ability to reconstruct significant events and to satisfy applicable regulatory requirements. Retention periods are : based on requirements specified in QA Program documents. Retention periods shall be documented. Procurement documents specify vendor responsibilities for the generation, retention, and submission to l BGE of quality assurance documentation related to the fabrication, inspection, and test of SR and designated NSR items and senices. Inspection and test records contain the following as appropriate:
- 1. Description of the type of observation. i 1
- 2. Date and results ofinspection or test. j 1
- 3. Information related to noted discrepancies, including action taken to resolve them.
- 4. Identification ofinspector or recorder of data.
- 5. Statement as to acceptability of results. )
1 Controls have been provided to ensure that records are protected from possible destruction. Within i established time-intervals, completed lifetime records are transmitted to the Records Management Unit for incorporation into the Long Term Records Storage and Retrieval System. l Page 39 of 69 ! I II
QUALITY ASSURANCE POLICY Revision 42.41 l 1B.18 AUDITS Internal audits are performed by BGE's Quality Audits Unit to ensure that activities and procedures established to implement the requirements of 10 CFR 50, Appendix B, comply witii BGE's overall QA Program. These audits are performed under the cogni7nnce of the OSSRC and provide a comprehensive l independent verification and evaluation of quality-related activities and procedures. Audits ensure the effective and proper implementation of BGE's QA Progrant Audits of selected aspects of operational phase activities are nerformed with a frequency commensurate with their safety significance and in such a manner as to assure that an audit of all safety-related functions is completed within a period of two year 2. In addition to the audilsubiects specified in the regulations and in ANSI N18.7-1976/ANS-3.2. audits shall also encomnass: the Facility Fire Protectiou Program and imolementing orocedures and an indcoendent fire prolgetion and loss orevention nrogratn audit and inspeglion utili7ing a qualified outside fire consultant: the radiological environmental monitoring program and the results thereof: the Offsite Dose Calculation Manual and imnlementing procedures: the Process Control Program and implementing procedures for processing and packaging of radioactive wastes: the perfonnance of activities required by the OA Program for effluent and environmental monitoring the perfonnance of_ activities required by the OA Program to meet the criteria of 10 CFR 50. Appendix B; and any other area of facility operation considered appropriate by the OSSRC or the VP-NED. (191-They are wheduled on Se has of Se : pertance-to safetref+etivitiesbeing-performed: Vendor audits are performed to evaluate QA programs, procedures, and activities. Audits of major vendors are made car!y enough to ensure compliance with all aspects of BGE's procurement documents. Additional r audits are perfonned as required to ensure that all requirements of BOE's QA Program are properly implemented according to procurement documents. Auditeef<lesignatedeetivitiesenequired-by4ho-T& rare-performedender cognizaneeef4he-OSSRG: Audits are performed in accordance with preestablished written procedures or checklists by qualified NQAD personnel who have no direct responsibility for the work being audited. Technical specialists from other BGE departments and outside consultants may assist as necessary in performing audits. Audits include objective evaluation of quality-related practices, procedures, instructions, activities, and items, as well as review of documents and records. < Reports of audits are analyzed and documented. Results that indicate the QA Program to be inadequate, ineffective, or improperly implemented, including the need for re-audit of deficient areas, are reported to the Manager and Supervisor of the audited activity. Controls have been established for verifying that corrective action is taken promptly to correct noted deficiencies. To ensure that BGE's NQAD complies with the requirements of BGE's QA Program, an independent management audit of NQAD activities is performed annually by a Joint Utility Management Audit (JUMA) Tean,. Page 40 of 69 l
, - .-= . - - - . ._ .- -- . -. - - - .-. ,
i 5 s. QUALITY ASSURANCE POLICY ; Revision 42_43 l' ; I TABLE 18-l ' BALTIMORE GAS AND ELECTRIC COMPANY'S POSITION ON GUIDANCE CONTAINED IN-ANSIINDUSTRY STANDARDS AND REGULATORY , GillD.ES Revision ofIndustry Standards Anolicable to the , E_altimore Gas and Electric Ouality Assurance Procram i Reauirement l i Seme of the Industry Standards listed in Section 18.2 identify other Standards that are required, and some ! Regulatory Guides define the revisions of those Standards that are accMahle to the NRC. p Response ;
, a m ;cj '
BGE's QA Program was developed to respond to the specific revision of the ex:uments listed in Section f IB.2 and is not necessarily responsive to other documents listed in the referenced Edustry Standards. ! i t ANSilM18.7/ANS 3.2 - 1976 l item 1 j r Requirsmall ! Section 5.2.15 requires that plant procedures sha!! be reviewed by an individual knowledgeable in the area j affected by the procedure no less frnanentiv than every two years to determine if changes are necessary or .; desirable. A revision to a oroceAnre ennctitutes a nroceAnre review. (19) ? Resrqn.lsq [ BGE app!!= ii; mici =;en cf-sr:wa y= :: != ic ;!! p!=: p cecirs ;=cp t=: p c;;&r= pedr.ed f
!=: c^.= in . y twc y== = :: :=p=iS;d f=;;ce!=. '!kee-=: :=ir;d cc == i= 5^ iy:; j befer; perfumanee-Plant ornceAntes are subicet to programmatic enntrols which enntinually identifv needed orocedure revisions. (19) i Rest!@l) [
i i Eng;;= ring Test-Proc;&r= (ETPs)-and4:Serr !!!.; i=; are writt= f= c= tim; =!y pei.r.n= =d l kept4iw : fer= e far-fu::+e-stmi!= :=::;. Ificy ::: =d-again, icy = ::v;;wed ;;d =cdiS;d c ==: i condnionse::::ing : Se tim; af p; form =ce. !i Some-Sm veillance-Test-P ce;&r= (STfs) =: ped:=:d ==y 1:= to Sv: y== 'P;;y :cc =: :=;;wed ; befwe-each ped mne :c === in: Sey =: :=:patib!: . id ni:::ing :=di:!c= =d :=p;=iv: te l cc:=: =t. 1 i Ihe programmatic enntrols meet the intent of the biennini review orncecc from both a inchnieni and j pactical ctandnnint breance they constitute dvnnmic,, rather thnn statie nrnecAnre review methndninov. l Thus the biennini review process is redundant to the established programmatic controls nnd is no long;;r considered nececurv (19) f 1 I Page 4I of 69 t
QUALITY ASSURANCE POLICY Revision 42_43 l Item 2 f10) E.cquirement Section 5.2.2 specifies that temporary procedure changes that clearly do not change the intent of the , approved procedure shall as a minimum be approved by two members of the plant staff knowledgeable in , the areas affected by the procedure; and at least orie of these individuals shall be the supenisor in charge of the shift and hold a senior operators' license on the unit affected. Bespons.c BGE does not require the Shift Supenisor to be the Senior Reactor Operator (SRO) approving temporary - changes to procedures; any active SRO (either on-shift or on-staff) may provide the SRO approval for procedure changes. Reason Many proposed temporary procedure changes do not require the Shift Supenisor's immediate attention or knowledge of the change since they do not affect plant safety. Other SROs are available and qualified to perform this task since the Shift Supenisor's detailed review of the proposed change is not necessary to ensure plant safety. Requiring the Shift Supenisor to review all changes is burdensome and contrary to plant safety in light of the total number of procedures that exist and the time the Shift Supenisor must dedicate to ensuring the plant is safely operated and maintained. Additionally, our Technical Specification requires this approval be from someone holding an SRO license (not necessarily the Shift Supenisor), REGULATORY GUIDE 1.33-1978 (19) Rcquiruncnt Section C.. Regulatorv Position. item 4. states; Srction 4 5 " Audit Program." of ANSI N18.7-1476/ANS 3.2 states that audits of selected aspests of opsrational phastactivities shall be nerformed with a frequency commensurate with their safety significance and in such a manner as to ensure that an audit of all safeln rclated functions is comnleted within a period of 2 years. In amnlifration of this Icquirement. the followine piegram_cicments should be audited at the indicated ficqucacies;
- a. The results of actions taken to correct deficiencies that affert nuclear safety and occur in facility caninment, . structures systems. or method of operation - at leastanctper 6 months.
- b. The confonnance of facility operation to provisions containcd within the technical spsfificalionsmd_ applicable license conditions - at leastonce per 12 months.
- c. The prrfonnance. traimd_gualifications of ti el facility staff
- at least once ocr 12 montht Page 42 of 69
QUALITY ASSURANCE POLICY ! Revision 42S l . ECS99nSC Ihs andit freauenev for all safetv-related functions is at least once everv two vears (excent as otherwise reauired in regulationst 2 Reason The more freaugnt audit intervals do not allow BGE manneement the ficxibility to devote auditing resources consistent with the strencth of oerformance and safety significance of an activity. Exocrience has shown that some audits are performed more frequently than deemed noorooriate for the function. ANSI N18.1 - 3/8/71 Item I Egguirement Paragraph 4.2.2 states that at the time of initial core loading or appointment to the active position, the Operations Manager shall hold a Senior R: actor Operator's (SRO) License. I Paragraph 3.2.1 states that positions at the functional level of Manager are those to which are assigned broad responsibilities for direction of major aspects of a nuclear power plant. This functional level generally includes the plant manager (plant superintendent, or other title), his line assistants, if any, and the principal members of the operating organization reporting directly to the plant manager and having overall responsibility for operation of the plant or for its maintenance or technical senice activities. EcEponsg , Baltimore Gas & Electric has two positions in its organization, Superintendent-Nuclear Operations and General Supervisor-Nuclear Plant Operations. Neither of these positions needs to individually meet all of the requirements of both paragraphs 3.2.1 and 4.2.2. The Superintendent-Nuclear Operations will satisfy paragraph 3.2.1 and most of 4.2.2 except that he will not maintain an SRO license. Instead, the Superintendent-Nuclear Operations will hold or have held an SRO license. The GS-NPO will hold and . maintain an SRO license. The GS-N!-O satisfies paragraph 4.2.2, but he does not satisfy 3.2.1 because he does not report directly to the plant manager. Page 43 of 69 i
I QUALITY ASSURANCE POLICY Revision 42_43 \ Reason The Superintendent-Nuclear Operations will hold or have held an SRO license, as opposed to having a license at the time of appointment to the position. He will have an excellent understanding of plant operations. The GS-NPO will not only hold an SRO license at the time of appointment to the position, but he will maintain the license. The GS-NPO directly supenises the operating shift organization, whereas the Superintendent-Nuclear Operations is also responsible for operations procedure development, modifications acceptance, and operations / maintenance coordinations. The Superintendent-Nuclear Operation's level of supenision does not require current in-depth and plant specific knowledge which results from maintaining an SRO license. Item 2 (17) R.qquirement Pr saph 3.2.2 states that supenisors are persons principally responsible for directing the actions of q,euw 'echnicians, or repairmen. Those positions usually designated as intermediate and first line wer m,e are included in this category. i',uagrapS 4.3.2 states that supenisors not requiring Atomic Energy Commission (AEC) licenses shall have a high school diploma or equivalent and a minimum of four years of experience in the craft or discipline he supenises. Bg.sportss Baltimore Gas and Electric has three supenisory positions in its organization - Supenisors, and in some cases Assistant General Supenisors and General Supervisors - which are organizationally equivalent (when supenising technicians / repairmen) to the positions described in paragraph 3.2.2 of ANSI N18.1-3/8/71. All these individuals need not possess the four years of craft / discipline experien,;e required by paragraph 4.3.2. Instead, at least the first line supenisor shall possess four years experience in the craft / discipline he supenises while other supenisors in the organization may be selected to fill supenisory positions based on possessing a minimum of an Associate's Degree, with four years of related technical experience, and demonstrated supenisory ability. (18) Additionally, all first line and intermediate supenisors shall have at least a high school diploma or equivalent. Rea;on To provide a balanced and broad base of supenisory ability within the site organizations made up of technicians / repairmen, it is desirable to include as supenisors both individuals with extensive craft / discipline experience accrued through field work and individuals with related education and experience who have demonstrated the ability to effectively supenise. ANSI N45.2.1 - 1973 Requirement Subsection 3.2 outlines requirements for demineralized water. EffpArts.g BGE specifications for demineralized water are different than the specifications outlined in the st.7ndard. Page 44 of 69
QUALITY ASSURANCE POLICY Revision 42_43. l Beason BGE specifications for demineralized water are consistent with guidelines prosided by the Nuclear Steam Supply System supplier. BGE specifications are generally more restrictive than those specified by ANSI '; N45.2.1. ANSI N45.2.2 - 1972 Item 1 Requirement Subsection 2.4 could be interpreted to mean that on-site and off-site personnel who perform any - inspection, examination, or testing activities related to the packing, shipping, receiving, storage, and handling ofitems for nuclear power plants shall be qualified in accordance with ANSI N45.2.6.
Response
BGE requires that only persons who are responsible for approving items for acceptance shall be qualified in accordance with Regulatory Guide 1.58 (which endorses ANSI N45.2.6) and that personnel who verify ; that storage areas meet requirements will be qualified to either Regulatory Guide 1.58 (which endorses ANSI N45.2.6) or ANSI N45.2.23. Reason Our receipt inspection procedures require persons who approve items for acceptance to be qualified in [ accordance with Regulatory Guide 1.58 (which endorses ANSI N45.2.6). QV technicians, inspectors or QA auditors verify that storage areas meet requirements. All other inspection, examination, and testing activities are subject to review by persons qualified to Regulatory Guide 1.58 (which endorses ANSI N45.2.6). Item 2 8_e_quirement The second sentence of Subsection 2.4 requires that: Off-site inspection, examination, or testing shall be audited and monitored by personnel who are qualified in accordance with ANSI N45.2.6. _R_tsyp_nz BGE uses personnel qualified in accordance with ANSI N45.2.23 to perform auditing and monitoring functions. Beason , The qualification requirements for auditors cannot always be met by persons qualified to Regulatory Guide 1.58 (which endorses ANSI N45.2.6). Page 45 of 69
QUALITY ASSURANCE POLICY Revision 42_41 l Hem 3 Reauirement Subsection 2.7 requires that activities covered by the Standard shall be divided into four levels, though . recognizing that within the scope of each level there may be a range of controls depending on the : importance of the item to safety and reliability. Essponse
- 1. The level of protective measures dermed by Subsection 2.7 are applied to Basic Component purchases.
- 2. Personnel of BGE's Nuclear Engineering Department (NED) will determine the level of protective measures to be applied to Commercial Grade purchases. :
Reamti BGE's position is as follows:
- 1. For Commercial Grade items, it is not ahvays possible to assign a level of classification in accordance with ANSI N45.2.2, as mar.y items are purchased after they have been packaged by the manufacturer and shipped to his local agent, the wholesaler.
- 2. Experience has shown that the level of protection assigned to Commercial Grade items by vendors is adequate.
Item 4 Requirement , Subsection 3.0 specifies detailed requirements for packing items for each level defined in Subsection 2.7. EtEpanz BGE has replaced Section 3.0 with the following:
- 1. Packaging for Shipment to BGE Personnel of BGE's NED or NSSD shall ensure that procurement documents for Basic Component and Cemercial Grade item purchases either indicate that the normal methods of packaging and shipment used by industry in general are acceptable for the items being procured or specify the level of protection assigned to the item and the requirement that the vendor conform to applicable requirements for items in that classification defined in Regulatory Guide 1.38, Rev. 2 - March 1977.
- 2. The nonnal methods of packaging used by the industry in general are acceptable for items being procured as Commercial Grade.
Page 46 of 69
QUALITY ASSURANCE POLICY Revision 42_43 l
- 3. Packaging for Storage by BGE 1
In general the packaging used by the vendor to ship items for all types of purchases to BGE need not , be retained after the item is received by BGE, provided that the item is stored in an area that meets the requirements for a storage area for the level of protection assigned to the item. Special or unique , items, however, may require special protective measures. For such unusual items, the Department that initiated the purchase, together with NED, or NSSD shall identify if any of the requirements of Section 6.4.2 of ANSI N45.2.2 - 1972 apply. : Reason
- 1. This substitution will ensure that the item will receive adequate protection during shipment and storage, thus eliminating unnecessary restrictions and enabling BGE to use .:ommercial sources to !
the utmost.
- 2. Experience shows that industrial practices for packaging Commercial Grade items are adequate for most applications.
Item 5 Rpguirement Section 4.0 defines shipping requirements related to the protection levels assigned to items.
Response
BGE has replaced Section 4.0 with the following:
- 1. Shipping to Baltimore Gas and Electric BGE will invoke the requirements for shipping specified in Section 4.0 of ANSI N45.2.2 - 1972 on Ba ic Component purchases only when NED or FSSD personnel have specified in procurement documents that the item shall be packaged in conformance with ANSI N45.2.2, Section 3.8.
BGE will not invoke the requirements of ANSI N45.2.2 -(1972, Section 4.0, on Commercial Grade , item purchases. ;
- 2. Shipping from Baltimore Gas and Electric items shipped from BGE need not conform to any of the requirements of ANSI N45.2.2, but the organization that packs and handles the item shall provide roughly the same level of protection that the item was given during shipment F BGE.
Reason if engineering personnel have determined that the vendor's methods of packaging are acceptable, they have already determined that the supplier's methods of shipping are adequate. As items are shipped from BGE ! only for repair, the detailed requirements specified in Section 4.0 of ANSI N45.2.2 are not necessary. j 1 l l J I I Page 47 of 69 l l
i QUALITY ASSURANCE POLICY Rwision42_43 l Item 6 Reauirement Subsection 6.4 gives detailed requirements for care of items in storage, according to the protection levels assigned to the items.
Response
BGE does not require items to be stored in the packing used for shipment if the storage level in the area provides the same protection as the level of packing assigned to the items. Caps, covers, etc., will be required only if specified by NED or NSSD personnel during the procurement process. If an item is taken , from one storage area to another, however, the persons who move it are responsible for ensuring, as applicable, that additional packing is supplie.d to give adequate protection during transportation. Reasom 'Ihe degree of protection given an item during storage should be tailored to the importance of the item to safety and the probability of deterioration during storage; to base storage requirements purely on the categories in Subsection 2.7 of ANSI N45.2.2 - 1972 is impractical. BGE requires NED or NSSD , personnel to specify requirements more closely related to the actual function of items and to storage conditions. Item 7 Reauirement Subsection 7.3.3 requires compliance with a series of ANSI documents. Rqsponse BGE controls for the use of hoisting equipment are compatible with the Standards listed in Subsection 7.3.3 of ANSI N45.2.2, although at the discretion of the Plant General Manager, they need not be compatible with documents referred to in these documents. E m 9B Lower-level documents referred to in the documents listed in Subparagraph 7.3.3 will not necessarily affect the ability of BGE personnel to properly handle SR items and could lead to confusion. " Page 48 of 69
r QUALITY ASSURANCE POLICY Revision 42_43 l ANSI N45.2.3 - 1973 Item 1 , Emui_remem Subsection 2.1 outlines housekeeping cleanliness requirements for five designated zones. Resnonse BGE has established three classes for cleanliness requirements. There is no class equivalent to the ANSI Zone 1. Requirements of ANSI Zones 4 and 5 have been consolidated into BGE's class 3. Reason
- 1. ANSI Zone I level of cleanliness applies to new construction activities.
- 2. Where required, smoking restrictions are posted for BGE's class 3 areas, item 2 Enluircniggi Subsection 2.1 requires for Zones I, II, and Ill, that a written record of the entry and exit of all personnel and material shall be established and maintained.
Responss BGE has established the following methods for personnel and material accountability:
- 1. Written accountability.
- 2. Where possible tethering of tools and materials to permanent plant structures or persons.
- 3. Post-maintenance closcout inspections.
Rcasari , BGE's three methods of accountability ofter the same level of control as that required by the standard. ANSI N45.2.4 - 1972 Rgquirement The last paragraph of Subsection 6.2.1 (Equipment Tests) states: Items requiring calibration shall be tagged or labeled on completion indicating date of calibration and identity of person that performed the calibration. Page 49 of 69 <
QUALITY ASSURANCE POLICY Revision 483 {
Response
The new calibration program at Calvert Cliffs does not use calibration stickers that contain date of calibration and identity of person that performed the calibration. The new calibration stickers indicate that : the instrument is periodically calibrated according to the calibration program. The sticker, a green "C," means the instrument is in the program. Reaso_ n In the past, the date of calibration noted on the instrument was important because the calibration history of preventive maintenance was not kept on computer. Computer tracking systems and trending programs did not exist. In the new system, the date of calibration being on the sticker is not necessary because the date of calibration and the identity of the person that performed the calibration is retrievable in the PM history in Nucleis according to equipment ID. Calibrations ofinstruments are scheduled and tracked by computer. We are going into a real predictive and preventive maintenance calibration program. Calibration frequencies will be shifted based on calibration history, PRAs, vendor's recommendations, and instrument use. A database eHsts which controls what instruments are added to or deleted from the program. By ! maintaining the database, we ensure that no instruments are identified as calibrated that are not. In the new program instruments identified as calibrated are kept up to date and specific information is kept on computer with no need for that information to be on the sticker. ANSI N45.2.6 - 1978 Item i BeguiremerJ1 f i Subsection 1.2 states in part, The requirements of this standard apply to personnel who perform inspection, examination, and tests during fabrication prior to and during receipt ofitems at the construction site, during construction, during preoperational and startup testing, and during operational phases of nuclear power plants. Besponse-A Personnel of BGE's Quality Assurance organizations who perform independent verification through inspections, examinations, or tests at the plant site during operational phases of the nuclear power plant are required to be qualified in accordance with Regulatory Guide 1.58 (which endorses ANSI N45.2.6) or to ANSI N18.1,1971. All other BGE personnel who perform inspection, examination, and testing ftmetions associated with normal operations of the plant are qualified either to Regulatory Guide 1.58 (which endorses ANSI N45.2.6) or to ANSI N18.1 - 1971. Ecason-6
- 1. The individuals who perform inspection, examination, and testing functions associated with normal operation of the plant, such as maintenance and certain technical reviews, are normally qualified to ANSI N18.1 - 1971.
Page 50 of 69
QUALITY ASSURANCE POLICY Revision 4231 l
- 2. Some testing activities conducted during normal operation of the plant, such as surveillance testing, do not require that test personnel meet the requirements specified in Paragraph 4.5.2 of ANSI N 18.1 for technicians. Personnel qualified to Regulatory Guide 1.58 (which endorses ANSI N45.2.6) are adequately qualified to conduct such testing.
flesponse-B BGE does not always require vendor personnel performing inspection or test activities to comply with the requirements of Regulatory Guide 1.58 (which endorses ANSI N45.2.6) but evaluates the need for invoking Regulatory Guide 1.58 (which endorses ANSI N45.2.6) on the vendor during the review of procurement documents. The requirements are not applied to procurement classified as Commercial Grade. Reason-B BGE's position is as follows:
- 1. For replacement items purchased as Commercial Grade items, the purchaser may not impose nuclear unique requirements on the vendor. Additionally, items may be manufactured before placement of the purchase order and the vendor may not be required to maintain records of the performance ofinspections or tests.
- 2. For Basic Component Purchases, the qualification requirements for inspection, examination, and test personnel are determined by;
- a. Item status (new or replacement).
- b. Complexity and importance ofitem.
- c. Manufacturer's QA program approval level (Appendix B, ANSI N45.2, etc.).
Refponse-C BGE does not require pei:nnnel who perform specific limited and repetitious inspection functions, such as inspection for removal or replacement of snubbers, to be trained as required by Regulatory Guide 1.58 (which endorses ANSI N45.2.6). Esason-C Inspections, examinations, or tests that are repetitious or of limited scope need not be performed by ; individuals qualified to the requirements of Regulatory Guide 1.58 (which endorses ANSI N45.2.6) provided that they receive instruction in the following:
- 1. Activities to be verified.
- 2. Acceptance criteria.
- 3. Method of documenting results.
- 4. Method of reporting deficiencies.
I The person responsible for the inspection activity ensures that such instruction is given to inspectors before l they perform specific inspection functions, and that both this training and the acceptability of the results of the inspection are documented. Page 5Iof69 l l
QUALITY ASSURANCE POLICY Revision 42.43 l ; EggggnatD When it is necessary to monitor the activities of a vendor, BGE uses personnel qualified as auditors in ; accordance with ANSI N45.2.23 or inspectors in accordance with Regulatory Guide 1.58 (which endorses ANSI N45.2.6). t Egitson-D Both Regulatory Guide 1.58 (which endorses ANSI N45.2.6) and ANSI N45.2.23 establish training requirements suitable for moniu r'.ng vendor activities. Item 2 Requirgmsnt Table I specifies that Level 111 personnel shall be capable of qualifying Level III personnel,
Response
When there is only one Level til position or when a new Level Ill position is created, BGE personnel with the title General Supenisor, or higher, qualify Level III personnel. Reason BGE personnel in these grades are capable of certifying Level III personnel without being trained as Level 111 inspectors. NOTE: Regulatory Guide 1.58 (which endorses ANSI N45.2.6-1978) statesin part, und:r item 6 of Regulatory Position, that..."In addition to the recommendations listed under Section 3.5 (of ANSI N45.2.6-1978) for Level I, II, and III personnel, the candidate should be a high school graduate or have earned the General Education Development equivalent of a high school diploma.. " Based on the NRC letter dated January 17, 1985 from Thomas T. Martin to A. E. Lundvall, Jr., the above educational requirements will be implemented for inspection, examination, and testing personnel hired or assigned after November 27,1984, in addition to the present commitment to ANSI N45.2.6-1978 for the i qualification of such personnel. ANSI N45.2.9 - 1976 Item I l I Requireingni Section 4.0 titled " Receipt" gives instructions for receipt controls. Besonsg BGE applies these requirements only to the receipt of records by the Plant llistory File. j l Page 52 of 69 l
QUALITY ASSURANCE POLICY Revision 4L41 l
& awn Most records received by such organizations as Receiving Inspection, Engineering, etc., are not shipped in a manner that makes these requirements applicable. These requirements are applicable, however, when the records are finally turned over to the Plant History File.
Item 2 L Be_quirement Subsection 5.6.1 reads as follows, " Design and construction of a single record storage facility shall meet - the following criteria:" Items a) and b) of the subsection state that: ,
"a) Reinforced concrete, concrete block, masonry, or equal construction."
"b) A floor and roof with drainage control. If a floor drain is provided, a check valve (or equal) shall be included."
8.esponse/ Reason Item a The intent of this requirement is both structural integrity and fire resistance. This vault is entirely enveloped by a structurally sound, fire resistive building. Second, the vault rests on a reinforced slab on ; grade and its walls extend fully to the underside of the structural deck. Third, the walls of the vault are constructed of gypsum wallboard on metal studs per Underwriters Laboratory Test Number U412, assuring the equivalent of 2 hour fire resistive construction. This is equal construction to concrete block in terms of fire protection. The walls carry no structural load; hence, they provide equivalent structural integrity to that needed of concrete block.I (See footnote following page). Response / Reason Item b Agam, the vault is contained within an environmentally protected building. As such, it has no roof, or need for floor drain.1 (See footnote). Item 3 kquiremenni Subsection 5.6 allows only the dual facility defined in Subsection 5.6.2 as an alternative to the single facility defined in Subsection 5.6.1. iThese responses have been fonvarded to the NRC by the BGE letter dated 02/11/83 from Robert G. Nichols, Sr. Facilities Project Administrator, Real Estate and Office Senices Department, to Terry L. liarpster, Chief QA Branch, Division of QA, Safeguards and Inspection Programs, IE, USNRC. These responses have also been accepted by the NRC in their letter dated 04/22/83 from Walter P. Haass, Deputy Chief, QA Branch, Division of Quality Assurance, Safeguards, and Inspections Programs, Office ofInspection and Enforcement. Page 53 of 69
QUALITY ASSURANCE POLICY Revision 42_41 \ ' Enans i BGE allows the following alternative storage requirements for orga bns other than the Records Management Unit: Organizations that originate records and do not transfer them to the Records Management Unit within 30 days of completion shall establish one of the following three controls as alternatives to the requirements specified for the R.ecords Management Unit:
- 1. Duplicate Storage Either A or B.
A. Within 30 days of completion of a record, a duplicate record file shall be established. This activity shall be controlled by procedures which provide for the following:
- 1. Assignment of responsibility for records.
- 2. Description of storage area.
- 3. Description of filing system.
- 4. An index of the filing system.
- 5. Rules governing access to and control of files.
- 6. Methods for maintaining control of and accountability for records removed from the file.
- 7. Method for filing supplemental information and 6tsposing of superseded or obsolete records.
- 8. Method for preserving records to prevent deterioration.
- 9. Method for maintaining specially processed records that are sensitive to light, pressure, or temperature.
- 10. Transfer of duplicates to the Records Management Unit within two years of completion of records.
B. Make arrangements with at least one other department that receives a copy of each document to subject this other copy to the controls specified above.
- 2. Fire-resistant Building Storage Records shall be stored in steel cabinets located in a fire-resistant building or a non-combustible building with a fire suppression system.
The procedural controls defined for duplicate storage shall be applied.
- 3. Non-fire-resistant Building Storage Within non-fire-resistant facilities, records shall be stored in UL one-hour-minimum fire-rated storage cabinas and be subject to the procedural controls defined for duplicate storage Page 54 of 69
QUALITY ASSURANCE POLICY Revision 42_43 l BGE dermes a Fire-resistant Building as follows: A facility constructed to resist the initiation or spreading of fire; non-combustible and/or fire-suppressive materials used; building certified as fire-resistant by the Risk , Management Unit of BGE's Corporate Finance Group. Reason Although these alternatives are compatible with standard methods of handling records, they do not materially decrease the level of protection afforded to the records. ANSI N45.2.23 - 1978 Item 1 Ec51uirement 2.3 Qualification of Lead Auditors Section 2.3.1 requires prospective Lead Auditors to obtain a minimum of ten credits under the scoring system defined in paragraphs 2.3.1.1-2.3.1.4.
Response
BGE has revised the scoring system as follows: Education and Experience The prospective Lead Auditor shall have accumulated a minimum of ten credits under the following scoring system: 1.0 Education (4 credits maximum) 1.1 For the Associate degree for an accredited institution, score one credit, if the degree is in engineering, physical sciences, mathematics, or quality assurance, score two credits. Or, for the Bachelor degree from an accredited institution, score two credits; if the degree is in engineering, physical sciences, mathematics, or quality assurance, score three credits. 1.2 For the Master degree in engineering, physical sciences, business management, or quality assurance from an accredited institution, score one credit. 1.3 For the successful completion of part of the required curriculum for an Associate, Bachelor, or Master degree, score a corresponding percentage of the credits specified above for the degree. 1.4 For the successful completion of Navy Nuclear Training, its equivalent in another armed service, or the training required for becoming a licensed operator in a conunercial nuclear power plant, score two credits. 1 Page 55 of 69
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QUALITY ASSURANCE POLICY Revision 4231 l 2.0 Experience (9 credits maximum) 2.1 Technical Experience (5 credits maximum) For experience in engineering, manufacturing, construction, operation, or maintenance, score one credit for each full year. 2.2 Nuclear Experience If two years of technical experience have been in the nuclear field, score one additional credit. 2.3 Quality Assurance Experience If two or more years of the technical experience have been in quality assurance or quality control, score two additional credits. Persons whose work activities are controlled by the QA Program but who are not full-time members of the QA organization may be awarded half the credits that would be given to a person with specific quality assurance experience. . 2.4 Audit Experience if two or more years of the technical experience have been in auditing, score one additional credit. 2.5 Supplemental Experience Persons who have a proportion of the experisace specified in 2.1-2.4 may be awarded a corresponding percentage of the credits specified. 2.6 Time exclusively spent in training does not apply as credit toward experience requirements for lead auditors. ,
. 3.0 Training (2 credits maximum) i Persons who have successfully completed the training requirements of ANSI N45.2.23 may be given two credits. ,
4.0 Rights of Management (2 credits maximum) The Manager-NQAD, may grant additional credits for other performance factors applicable to auditing as follows: 4.1 For certification of competence in engineering or science related to nuclear power i plants, or in quality assurance specialties, issued and approved by a State Agency or National Professional or Technical Society, score two credits. 4.2 For nuclear experience in excess of 2 years, score one credit for each two years experience.
)
4.3 For practical experience that can be related to power plants, in excess of 5 years, score one credit for cach two years of experience. l i Page 56 of 69 i l
QUALITY ASSURANCE POLICY I
, Revision 42_43 l Emm BGE is in agreement with the basic purpose of ANSI N45.2.23-that is, to establish minimum educational or experience requirements for Lead Auditors. We think, however, that the system of credits outlined in ANSI N45.2.23 tends to reduce the size of the pool of potential replacement auditors without making redeeming improvement in the capabilities of persons selected.
We calculated the credit score of 11 of our present Lead Auditors at the time they were appointed Lead Auditors. Six had completed Navy Nuclear Training and spent several years in the Navy Nuclear Program. Four of these scored only 8 credits total, including 2 credits allowed by paragraph 2.3.L4 of ANSI N45.2.23 for rights of management based on their having completed the BGE QA training programs for Lead Auditors. One of our auditors, with neither nuclear nor power plant experience, had a credit score of 12 because he held a Bachelor's degree in engineering and was a professional engineer with over 5 years design experience. Because all of these individuals have acted as Lead Auditors satisfactorily for several years, it appears that the credit system should be revised slightly to allow for the differences in education and experience of prospective Lead Auditor candidates. We consider the flaw in the current system to be the emphas4 on educational requirements that will allow a person with a Master's degree and no nuclear or power plant experience to become a Lead Auditor, but will exclude a person who has no degree, even though he may have 20 years' experience in operating or maintaining nuclear or power plant systems. The practical balance between education and experience will vary with individuals and particular work assignments. Any attempt to establish rigid requirements is likely to allow some unsuitable candidates to meet the qualification requirements while excluding some acceptable candidates. For these reasons, we think that the supenision of prospective Audit Team Leaders should be given more flexibility in determining whether, for a particular individual, educational or professional qualifications are more significant and valuable than past experience. The present credit system, while recognizing the Associate degree, gives no credit for completion of the nuclear training programs. We think that someone who has taken Navy Nuclear Training or its equivalent in another anned senice, or someone who has completed the training required to become a licensed operator in a conunercial nuclear power plant, should receive the same credit as a person who has an Associate degree from an accredited institution in engineering, physical sciences, mathematics, or quality assurance. The points now awarded for education are related to the effect that formal courses might have on the ability ofindividuals to comprehend the regulations or the technical aspects of activities being audited. The point system makes no allowance for the fact that such knowledge comes gradually and not upon receipt of a degree. Persons who have completed part of a degree course should receive a percentage of the credits allowed for that course. The requirements for training specified in ANSI N45.2.23, paragraph 2.3.2, would seem to ensure that prospective Lead Auditors will meet the requirements of paragraph 2.3.1.4 dealing with the rights of management. We think, therefore, that all prospective Lead Auditors should qualify for these two credits. Page 57 of 69
r QUALITY ASSURANCE POLICY RaisionM l Similarly, the present system recognizes the effect that working in a QA Program will have on the ability of a person to comprehend regulations and technical requirements. Persons who are not assigned as full-time members of the QA Organization, however, receive similar exposure if they perform activities controlled by a QA Program. We therefore allow such persons half the credits specified for quality assurance experience. Item 2 Reauiremgn.1 3.3 Requalification Lead Auditors who fail to maintain their proficiency for a period of two years or more shall be required to requalify. Requalification shall include retrdning in accordance with the requirements of paragraph 2.3.3, reexamination in accordance w% paragraph 2.3.5, and participation as an Auditor in at least one nuclear quality assurance audit.
Response
BGE requalifies Lead Auditors on the basis of the satisfactory performance of one audit, as obsened by a qualified Lead Auditor. Reason The purpose of the training specified in paragraph 2.3.3 of the Standard is to ensure that candidates understand the fundamentals of auditing and the requirements for activities to be audited. The fact that persons have not maintained their proficiency does not mean that they need complete re-training; it means only that they have not been able to review and study the applicable Codes, Standards, Procedures, instructions, and other documents related to QA Programs and program auditing. BGE considers that the satisfactory perfonnance of an audit under the observation and guidance of a qualified Lead Auditor should ensure that persons with lapsed certification will review and understand the pertinent documents. ANSI N101.4 - 1972 Requirement Section 1.2 specifies applicability requirements for the Standard.
Response
BGE requires that only activities performed inside containment structures and related to protective coatings applied to ferritic steels, aluminum, stainless steel, zinc-coated (galvanized) steel, concrete, or masonry surfaces shall confonn to applicable Sections of ANSI N101.4. Reason Deterioration of protective coatings applied to surfaces outside containment structures would have no detrimental effects on the safe operation of the plant. Page 58 of 69
_ QUALITY ASSURANCE POLICY Revision 42._41 l l i ANSI N45.2.13 - 1973 ! t Reauirement ,
' ANSI N45.2.13 could be interpreted to mean that all requirements of this standard are applicable to all I safety-related items or senices. i Empsts f . BGE has two approaches for safety-related and designated non-safety related procurement as' described in }
Sections,1B.4 and 1B.7. Controls established for Basic Component Purchases correspond to the , requirements of ANSI N45.2.13. .The extent to which the individual requirements of ANSI N45.2.13 are applied to Commercial Grade Purchases depends on the nature and scope of the work to be perfonned and : the importance to nuclear safety and the items or senices purchased. This approach is consistent with the i introductory discussion in Section 1.3 of ANSI N45.2.13 - 1973. i r t i 6 i I i i i [
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j ~ QUALITY ASSURANCE POLICY jy Revision 42_4.2 l ATTACHMENT A-1
+
BASES FOR QA POLICY REVISIONS (1) Entry PRF-Q ! No. No. Bases for R.evision(s) : i
- 1. 771 Procedure Uoarade Action Plan (PUAP). cer L. B. Russell letter 1-20-89. - -j i
2, 783 10 CFR Part 21 reauirements. !
- 3. 797 NRC Inspection #89-16/89-17 (Letter from R. E. Denton to R. P. Heibel dated j July 13.1989.)
- 4. 824 NRC letter from M. V, Hodges to G. C. Creel dated March 13. 1990. This letter !
approved a one-time exenig*ign to the periodic review reauirements for oroce_dp_res e ! scheduled to be uperaded by the Procedures Unarade Proiget. This exemption was f l discontinued and removed by PRF-O 954. ;
- 5. 844 Procurement Program Proicci upgrade. Performance improvement Plan (PIP) l Action Plan #5.3.1 and OAU Audit Findine 87-13-01
- 6. 844 18.15 and 18.16 revised to clearly establish orocram applicability and controls.
f consistent terminolony. organizational responsibilities and focused approach towards developine and imokmenting an intearated Mananement System. [ 7, 891 PIP Action Plan 5.3.1 Follow-On Activity. i
- 8. 894 18.15 and 1B.16 revised to clarify reauirements which will oermit implementation ,
of the issues Management System - PIP item 4.10.0. l
- 9. 854/907 G. C. Cree' letter to the NRC dated 7/26/91 which discussed modifications to. and ;
- acceptance of. chances to thq OA Poliev invohing OA compliance reviews of OAPs and Directives. !
- 10. 815 G. C. Creel letter to the NRC dated 10/3/90 discussing temocrary changes noI !
I affectine " Approved Procedure Intent" and the relievina of the Administrative l Burden on Shift Supervisors. i l l Page 60 of 69
QUALITY ASSURANCE POLICY Revision 42E l , ATTACHMENT A-2 t
- BASES FOR QA POLICY REVISIONS (1)
' Entry PRF-Q No. No. Bases for Revision (s) +
- 11. 887 Audit Findine No. 9026-01 (Implementation of Surveillance Reauirements).
~
I
- 12. 954 G. C. Creel letter to the NRC dated 7/3/91 discontinuing the one-time temporary chance to the periodic review interval approved in Basis (4) abcve. f
- 13. 957 PIP Action Plan follow-on activity (5.3.1).
- 14. 953 PIP Action Plan 4.1 and NUREG-0737 (TMI Action Plan Reauirements)
Item I C.5. " Procedure for Feedback of Operating Experience to Plant Staff." i
- 15. 990 OAU Surveillance 5-92-28 " Interface Between Facilities Management Department and Nuclear Encrev Division on Proiects at Calvert Cliffs." Reccagnendation 4.2.
- 16. 998 OA Audit Recommendation 92-04-R03 (ISFSI cocrational chase).
- 17. 93-06 NOAD Audit Findine 92-10-01. Facility Staff Trainina. ;
I8. 93-06 Letter from M. C. Modes of the NRC to R. E. Denton dated June 21.1994. '{
- 19. 95-02 License Amendment Ream si NRC 94-075 Dated Feb 95 (Drafil
- 1 L
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QUALITY ASSURANCE POLICY i l Revision-42_43 l ADDENDUM 1B-1 (19) ; REVIEW FUNCTIONS OF Tile POSR
C. PROCEDURE
REVIEW COMMITTEE. OUAI IFIED REVIEWERS. AND OSSRC .l 0 PLANT OPERATIONS AND SAFETY REVIEW COMMITTEE (POSRC) - 1.1 FUNCTION , The POSRC chn11 function to advise the Plant General Manager on all matters related to nucigar safety. : 12 COMPOSITION The POSRC shall be composed of at least seven. but no more than ten. members including the . Chairman Members shall collectivelv have exnerience in the following areas; Nuclear Operations Electrical and Controls Maintenance , Chemistry Mcchnnical Maintenance Nuclear Engineering Badiation Safety l Plant Engineering I Design Engineering - hiembers shall be annointed in writing by the Plant General Manager. Members shall have a Dununum of eight years oower olant experience of which a minimum of three years shall be nuclear , power exocrience. At least one member shall have an SRO license on Calvert Cliffs Units 1 and 2. 1.3 CHAIRMAN The Chairman and alternate Chaimien of the POSRC sh:dl be annointed in writing by the Plant General Manager. Chainnen shall have a minimum of 10 years nower olant pspgtience of which a minimum of three years shall be nuclear power exoerienec, 1.4 ALTERNATES All alternate members shall be apnointed in writing by_the Plant General Manager. Alternate members shall have a minimum of eight years nower olant exncrience of which a minimum of three ygars shall be nucicar power experienec. 1.5 MEETING FREOUENCY The POSRC shall meet at lerts.t once per calendaunonth and as convened by the POSRC Chaianan cr one of the designated alternates. 1.6 OUORUM A ouorum of the POSRC shall include the Chairman or one of the designntgd alternate cll ainucn and shnli consist of a maiority of the members. Including alternates No more than half of the quorum shall be alternates _ including an alternate chairman Page 63 of 69
[ QUALITY ASSURANCE POLICY 1.7 RESPONSIBILITIES The POSRC shnll be resnonsible fer the followine excent for those items desienated for review by i the Procedure Review Committee or Oualified Reviewer in accordance with ddendum sections 2 ) and 3. resnectivelv: i
- n. Review of 1) all procedures reouired by Technical Spsgification 6.4 and changes therela, and 2) any other oronosed orocedures or chnnoes thereto as determined by the Plant
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General Manager to affect nuclear safety. Cross-disciolinary reviews of these promlures are conducted in accordance with administrative orocedures in addition to the reviews conducted by POSRC. the Procedure Review Committee. or Oualified Reviewer.
- b. Review of all oroposed tests and exocriments that affect nuclear safety.
- c. Review of all pronosed changes to Anoendix .A Technical Snecification d Review of all nronosed changes or modifications to olant systems or cauioment that aggst nuclear safety.
- e. Review of the Pinnt Security Plan and implementing orocedures and shall submit recommended changes to the Off-Site Safety Review Committee.
- f. Review of the Emergency Plan and implementing orocedures and shall submit recommended changes to the Off-Site Safety Review Committee.
- g. Review of changes to the Process Control Program and the Offsite Dose Calculation Manual.
- h. Review of all 10 CFR 50.59 Safety Evaluntions that supoort nrocedures in 1.7.a and changes or modifications in 1.7.d above.
- i. Investigation of all violations of the Teshnical Spsgifications including the oreparation nnd fonvarding of reports coverine evaluation and recommendations to prevent recurrence to the Plant General Manager. the Vice President - Nuclear Energv nnd to the Chairman of the Off-Site Safety Review Committee.
- i. Review of all Reportable Events The results of this review shall be submitted to the OSSRC and the Vice President - Nuclear Encrev.
- k. Review of facility opciations to dcicctpolcatial safety hn7nrds.
- 1. Review of any aggietal. unnlanned or uncontrolled radioactive release that exceeds 25%
of the limits of Techclical Specification 3.11.1.2. 3.112 2 or 3.11.2 3. including the pIcparation of renorts covering evaluation. recommendations and disoosition of the corrective action to prevent recurrence and for forwarding of these reoorts to the Plant , General Manager and the Off-Site Safety Review Committee.
- ImRC is oniv rcouired ta review Fire Protection nrocedures and chnnm thereto which afTect nuclear safety.
Page 64 of 69
QUALITY ASSURANCE POLICY Revision 431 l m Performance of soecial reviews _ investigntions or analyses and renorts thereon as reauested by the Chairman of the Off-Site Safety Review Committee.
- n. Review of Safety I imit Violation Reoorts.
i 1.8 AUTHORITY The Plant Ooerations and Safety Review Comm.ittee shall:
- a. Recommend to the anoroval authority anoroval or diennoroval of orocedures considered under 1.7.a above.
- b. Recommend to the Plant General Mannger anoroval or disapproval of items considered under 1.7 b through h abom
- c. Render deterrmnations in writing with regard to whether or not each item considered under 1.7.a throuch h above constitutes an unreviewed safety auestion.
- d. Evaluate root causes and recommended actions to prevent reCHIrence for items considered under 1.7.i throuch I above.
- c. Provide written notification within 24 hours to the Vice President - Nuclear Enerev and the Chairman of the OtT-Site Safety Review Committee of disagreement between the POSRC and the responsible approval authority in the case of item 1.7.a above or between the POSRC and the Plant General Manacer; however. the Plant General Manager shall have resoonsibility far resolution of such disagreements oursuant to Technical Soccification 6.1.1.
1.9 RECORDS The POSRC shall maintnin written minutes of each meeting and copies shall be orovided to the Vice President - Nuclear Encrev. Chairman of the Off-Site Safety Review Committee. and the Plant General Manager. 2.0 PROCEDURE REVIEW COMMITTEE 2.1 FUNCTION Ihe Procedure Review Committee may function to review items listed in 1.7.a above in lieu of review by POSRC or Ouahfied Reviewer as directed by the Plant General Manngen 2.2 COMPOSITION The Procedure Review Committee shall be comnosed of a Chairman and eicht individuals who shall collectively have exocrtise in thcarcas contained in 1.2 above.. Members shall be annointed in writing by the Plant General Manager. Members shall have a mimmum of eight years power olant experience of which a minimum of three years shall be nudcat power experience. At least oncaember shall be a POSRC member or alternate. The charter for the Procedure Review Committee shall include a descriotion of membershin annlifications_ functions and reports and shall be__dscribed in olant administrative orocedures The Procedure Egyiew Committee may be dissolved at the discretion of theflant General Mannger. Page 65 of 69
7 QUALITY ASSURANCE POLICY Revision 42E l 23 CHAIRMAN The Chairman and alternate Chairmen of the Procedure Review Committee shall be annointed in writing by the Plant General Manager. Chairmen shnli have a minimum of eight years nower plant exoerience of which a minimum of three years shall be nuclear nower exocrience. 2.4 ALTERNATES All alternate members shall be apoointed in writing by the Plant General Manager. Alternate members shnll have a minimum of eight years nower plant exoerience of which a minimum of three ygars shall be nuclear oower exoerience. 2.5 MEETING FREOUENCY - The Procedure Review Committee shall meet at least once ner calendar month and as convened by the Chairman or his designated alternates, 2.6 OUORUM A auorum for the Procedure Review Committee shall consist of the Chairman or one of the designated alternate Chairmen and three primarv or alternate members nrovided at least four disciolines are renresented. 2.7 AUTilORITY The Procedure Revinv Committee shall:
- a. Recommend to the Anoroval Authority anoroval or disanoroval of orocedures considered ,
under 1.7 a above.
- b. Render determinntions in writing with regard to whether or not each nrocedure under 1.7.a above constitutes an unreviewed safety auestion.
- c. Provide written notiftgation within 24 hours to the Vice President - Nuclear Encrev and the Dairman of the Off-Site Safety Review Committee of disagreements between the Procedure Review Committge and the resoonsible approval authority. The Plant Ggng1;d Manager shall have responsibility for resolution of such disagreements pursuant to Technical Soccification 6.1.1.
23 RECORDS The Procedure Review Committee shall maintain written minutes of each meeting and cooies shall be orovided to the Plant General Manager 3.0 OUALIFIED REVIEWERS 3I FUNCTION The Plant Gcneral Manager may designate specific procedures or classes of procedures described in 1.7.a above to be reviewed by Oualified Reviewers in lieu ofigriew by POSRC or the ProceduIc Review Committes. Page 66 of 69
l . i
. QUALITY ASSURANCE POLICY Revision 42_4. 3 l .
3.2 ' AUTHORITY
. OnnlineA Reviewers challr ,
I 1
. a. Reenmmend to the anproval nuthority annroval or diennnroval of decionated orocedures
- and chanocc concidered under 1.7.a above and
- b. Render determinatinn in writino with reoard to whether or not each nroceAnre imder '.7.a'
~
above ennetitutec an unrevieweil safety nuestion. l
- c. Provide written nntinention wi?hin 24 haurs to the Vice Precident - Nuclear Fnergy and the -
Chairman of the Off-Site Safety Review Cnmmittee of dienoreemente between the - .; Onnlified Reviewer and the annroval nuthoritv. The Plant General Manneer chall have .
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-l resnonsibility for resolution of such dienorcemente nurcnnnt to Technical Sneci6catinn l O .d 3.3 CERTIFICATION ,
Onnlified Reviewers chall be nominnteA trnineA nnd certified in accordance with ndminicerative i procedures. Certifiention chall be by a denartment manneer. ! 3.4 CERTIFICATION REOUIREMENTS } e Certifientinn reoniremente of nersonnel deciennteA as Onnlifini Reviewers chnll be in accordance with administrative proceAnrd Onnlified Reviewers chnll have: :
- a. A Rachelnr's deoree in enoineerino reinteA ccience or technient discinline. nnd two years l of nucient nower olnnt exnerience:~ I i
- b. ' Six years nucienr nower plant experience
@ i
- c. Eouivalent combination of education and exnerience as annroved by a Denartment I Manneer.
3.5 RECORDS ! Review of nrocedures by Onnlified Revicivers chnll be documented in accordance with t adminictrative orocedures. , 4.0 OFF-SITE SAFETY REVIEW COMMITTEE (OSSRQ , 4.1 FUNCTION The Ofr-Site Safety Revinv Committee chnli function to nrovide indenendent review and audit of decienntrd activities in the areas of: , ;
- a. nuclear nour plant operations i
+
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QUALITY ASSURANCE POLICY Revision 4L41 l
- b. nuclear enoineering
- c. chemistry and radiochemistrv
- d. metallurev and non-destructive examinntion
- c. instrumentation and control
- f. radiological safetv i
- g. mechanical and electrical engineering
- h. onnlity assumace oractices 4.2 COMPOSfTION
'I The Off-Site Safety Review Committee shall be comnosed of at least seven members including the Chairman. Members of the Off-Site Safety Review Committee may be from the Nuclear Encrev '
Dwision or other BGE organi7ntion or from organi7ntions external to BGE and shall collectivdy have exnertise in all of the areas of 4.1 above. , 4.3 OUALIFICATIONS The Chairman members and alternate members of the Off-Site Safety Review Committee shall be appointed in writing by the Vice President - Nuclear Energy and shaft have an academie degree in i engineerine or a physical science. or the equivalent and in a idition shall have a minimum of five years technical exnerience in one or more areas given in 4.1 above. No more than two alternatgg shall particinate as voting members in Ofr-Site Safety Review Committee activities at any one tinlc. 4.4 CONSULTANTS Consultants shall be utilized as determined by the Off-Site Safety Review Committee Chairman to { nrovide exnert adsice to the Ofr-Site Safety Review Committeg, 4.5 hiEETING FREOUENCY l The Off-Site Safety Review Committee shall meet at least once ner six months. 4.6 OUORUM The onorum of the Off-Site Safety Review Committee necessarv for the performance of the Off-Site Safety Review Conunittee review and audit functions shall consist of more than half the Off-Site Safety Review Committee membership or at least four members whichever is greater. This quorum shall include the Chairman or his annointed alternate and the Off-Site Safety Review Committee members including appointed. alternates meeting the reouirements of 4.3 above. No more than a minority of the ouorum shalljlave line resnonsibility for oneration of the nlant-4.7 SUBCOMMITTEES The Chairmnn may establish subcommittees to perform reviews of selected items enumerated in 4.8 below. Each subcommitice shall be chartered in writing- have at least threg members / alternates. and provide renorts to the full committee on the results ofits reviews with any aporonriate recommendations. Page 68 of 69
QUALITY ASSURANCE POLICY Revision-42_41 l 4.8 REVIEW
- The Ofr-Si'e Safety Review Committee shall review:
- a. The safety evaluntions for 1) changes to ornendures. enuinment or svetems and 2) testn or exnerimmte comnleted under the orovisions of 10 CFR 50.59. to verify that such actions
. did not constitute an unreviewed safety auestiQ n,
- b. Pronosed change: to procedures. couinment or systems which involve an unreviewed safety aucction as defined in 10 CFR 50.59.
- c. Proposed tests or exoerimects which involve an unreviewed safety auestion as defined in 10 CFR 50.59.
- d. Pronosed changes in Technical Specifientions or the Ooerating License.
- c. Violation of codes. regulations orders. Technical Soecifientiont license reouiremente or ofinternal precedures or instmetions having nuclear safety significa14ce.
f Signifiennt onerating abnormalities or deviations from normal and exr,ected performnnee of plant eauinment that affect nuclear safety.
- g. All Reoortable Events.
- h. All recogni7ed indientions of an unanticinated deficiency in some asnect of design or oneration of safety related structures systems _ or comoonente
- i. Reports nnd meetings minutes of the POSRC.
4.9 AUDITS Audits of facility activities shnll be nerformed under the cogni7nnee of the Off-Site Safety Review . Committee These internal audits are discussed in Section 1 B 18 of the OA Poliev. 4.10 AUT110RITY The Off-Site Safety Revir Committee shall reoort to and advise the Vice President - Nuclear Engrev on those areas of re son-ibility soecified in 4.8 and 4.9 above. 4.I 1 RECORDS Records of Off-Site Safety Review Committee activities shall be orenared. anoroved and distributed as indiented belog
- a. Minutes cf each Off-Site Safety Review Committee meetinn shnll be nrenared anoroved and fonvarded to the Vice President - Nuclean Enerev within 14 days' followinir each a'retmg
- b. Reports of reviews encompassed by 4.8 above. chnll be prenared. anproved and fonvarded to the Vice President - Nucjgar Enerev within 14 days following comnterion of the review.
- c. Audit reports encompassed by 4.0 above. shall be forwarded to the Vice President -
Nuclear Encrev and to the tinnngement positions resoonsible for the areas audited within 30 days after comnletion of the ardit. Page 69 of 69 __}}