ML17263A643
| ML17263A643 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/13/1994 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17263A642 | List: |
| References | |
| NUDOCS 9405170068 | |
| Download: ML17263A643 (181) | |
Text
ATTACHMENT B Marked Up Copy of R.E.
Ginna Nuclear Power Plant Technical Specification 6.0 Included pages:
11 3.1-21 3.5-2a 3.6-3 3.15-1 3.16-1 3.16-2 6.1-1
~ 6.2-1 6.2-2 6.2-3 6.2-4 6.3-1 6.4-1 6.5-1 6.5-2 6.5-3 6.5-4 6.5-4a 6.5-5 6.5-6
, 6.5-7 6.5-8 6.5-8a 6.5-9 6.5-10 6.5-11 6.5-12 6.6-1 6.7-1 6.8-1 6.8-2 6.9-1 6.9-2 6.9-3 6.9-4 6.9-5 6.9-6 6.9-7 6.9-8 6.9-9 6.10-1 6.10-2 6.10-3 6.11-1 6.12-1 6.13-1 6.13-2 6.14-1 6.15-1 6.16-1 6.17-1 6.17-2 9405i70068" 940513.
PDR 'DQCK 05000244 P
4.8 4.9 4.10 4.11 4'2 4.13 4.14 4'5 4.16 TABLE OF CONTENTS (cont d)
Auxiliary Feedwater Systems Reactivity Anomalies Environmental Radiation Survey Refueling Effluent; Surveillance Radioactive Material Source Leakage
- Test, Shock Suppressors (Snubbers)
Deleted Overpressure Protection System Pacae 4.8-1 4.9-1 4.10-1 4.11-1 4 '2-1 4.13-1 4.14-1 4.15-1 4.16-1 5.0 DESIGN FEATURES 5.1 5.2 5.3 5.4 5.5 Site Containment Design Features Reactor Design Features Fuel Storage Waste Treatment Systems
- 5. 1-1
- 5. 2-1
- 5. 3-1 5.4-1 5.5-1 6.0 ADMINISTRATIVECONTROLS 6.1 6.2 V,n'i+
6.3 Responsibility Organization 6.2.1 Onsite ~d Offsite Organization 6.2.2
'5taff Staff Qualification~
6.% Safety Limit Violation Q~
NK.- 4.o-(2 ce~g-Rege~
Amendment No. gg, 49 13.
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3.1.4 Maximum Coolant Activit S ecifications 3.1.4.1 3.1.4.2 3.1.4.3 Whenever the reactor is critical or the reactor coolant average temperature is greater than 500 F:
0 a.
The total specific activity of the reactor coolant shall not exceed 84/E pCi/gm, where E is the average beta and gamma energies per disintegration in Mev.
b.
The I-131 equivalent of the iodine activity in the reactor coolant shall not exceed 0.2 pCi/gm.
c.
The I-131 equivalent of the iodine activity on the secondary side of a steam generator shall not exceed 0.1 pCi/gm.
If the limit of 3.1.4.1.a is exceeded, then be subcritical with re ctor coolant average temperature less than 500 F
within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
a.
If the I-131 equivalent activity in the reactor coolant exceeds the limit of 3.1.4.1.b but is less than the allowable limit shown on Figure 3.1.4-1, operation may continue for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />.
If the I-131 equivalent activity in reactor ant exceeds the limit
.1.4.1.b for more than 500 ho any consecutive 6-month period, th repare and submit ort to the ission pursuant to Specification 6.9.2-Amendment No.
27 3
~ 1-21
3.5.5.2 If the setpoint for a radioactive effluent monitor alarm and/or trip is found to be higher than required, one of the following three measures shall be taken immediately:
the setpoint shall be immediately corrected without declaring the channels inoperable; or immediately suspend the release of effluents monitored by the effected channel; or.
(iii) declare the channel inoperable.
3 '.5.3 If the number of channels which are operable is found to be less than required, take the action shown in.
Table 3.5-5.
Exert best efforts to return the instruments to OPERABLE status within 31 days and, if unsuccessful, explain in the next Radio-3.5.6 3.5.6.1 active Effluent Release Report why the inoperability was not corrected in a timely manner.
Control Room HVAC Detection Systems During all modes of plant operation, detection systems for chlorine gas, ammonia gas and radioactivity in the control room HVAC intake shall be operable with setpoints to isolate air intake adjusted as follows:
3.5<<2a Amendment No.
29
Basis:
The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.
The shutdown margins are selected based on the type of activities that are being carried out.
The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances.
When the reactor head is not. to'be removed, a cold shutdown margin of 14 k/k precludes criticality in any occurrence.
Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident wexe as much as 1 psig.<'>
The containment. is designed to withstand an internal vacuum of 2.5 psig.~"
The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.
Zn order to minimize contai.nment leakage during a design basis accident involving a
significant fission product
- release, penetrations not required for accident mitigation are provided with isolation boundaries.
These isolation boundaries consist of either passive devices or active automatic valves and are isted in a proce ure un er e contro o
Closed manual valves, deactivated automatic valves secure xn t ea.r closed position (including check valves with flow through
~ the valve secured),
blind flanges and closed systems are considered passive devices.
Automatic isolation valves designed to close following an accident without operator action, are considered active devices.
Two isolation devices are provided for each mechanical penetration, such that no single credible failure or malfunction of an active component can cause a loss of isolation, or result in a leakage rate that exceeds limits assumed in the safety analyses<'>.
Xn the event that one isolation boundary is inoperable, the affected penetration must be isolated with at least one boundary that is not affected by a single active failure.
Xsolation boundaries that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed manual valve, or a blind flange.
The opening of closed containment isolation valves on an intermittent basis under administrative control includes the following considerations:
(1) stationing an individual qualified in accordance with station procedures, who is in constant communication with the control room, at the valve controls, (2) instructing this individual to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to isolate the boundary and that this action will prevent the release of radioactivity outside the contaihment.
Amendment No.
- 3. 6-3
1
3
~ 1 3 Over ressure Protection S stem A licabilit Applies whenever the temperature of one or more of the RCS cold, legs is
< 330'F, or the Residual Heat Removal System is 1n operation.
Ob ective To prevent overpressurization of the reactor coolant system and the residual heat removal system.
S ecification 3.'5.1 Except during secondary side hydrostatic tests in which RCS pressure is to be raised above the PORV setpoint, at least one of the following over-pressure protection systems shall be operable:
b.
Two pressurizer power operated relief valves (PORVs) with a lift setting of
< 424 psig, or A reactor coolant system vent of 1.1 square inches.
1
~ '5sa
~ 1 3.15.1.2 3.-5.1.3 Basis With one PORV inoperable, either restore the inoperable PORV to operable status within 7 days or depressurize and vent the RCS through a 1.1 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a
vented condition until both PORVs have been restored to operable status.
With both PORVs inoperable, depressurize and vent the RCS through a 1.1 square inch vent(s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both PORVs have been restored to operable status.
Use of the overpressure protection system to mitigate an RCS or RHRS pressure transient shall be reported in accordance with WWW s~C ~o~
c. (..9 Ar RCS vent opening of greater than 1.1 square inches ensures that) the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the'RCS cold legs are
< 330'F'~'.
This relief capacity will Amendment No. p,JJ, 48
- 3. 15-1
3-16 Radiolo ical Environmental Monitorin Aoolicabilit Applies to routine testing of the plant environs.
Ob 'ective To establish a program which will assure recognition of changes in radioactivity or exposure pathways in the environs.
Saecifi. ation 3'16, 1 Monitorin Pr ram 3 ~ 16-1. 1 The radiological environmental monitoring program shall be conducted as specified in Table 3.16-.1 at the locations given in the ODCM.
3.16.1.2 Zf the radiological environmental monitoxing program
~ is not conducted as specified in Table 3.16-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
(Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal availability, or to malfunction of automatic sampling equipment.
Xf the latter, ef orts shal3.
be 3.16.1.3 made to complete corrective ac ion prior to the end of the next sampling period.
)
Zf the level of radioactivity in an environmental sampling m~.dium at one or more o the locations speci-fied 'n tne ODCM exceeds the reporting levels 3 '6-1
when averaged over any calendar quarter a
Special Report shall be submitted to the Commission within thirty days which includes an evaluation of any release conditions, environmental factors or other aspects which caused the reporting levels to be exceeded.
s~~
When mo..e than one of the radionuclide are detected in the sampling medium, this eport shall be submitted if:
concentration 1
+
concentration (2
+
....> 1.0 liml.t level (1) limit level (2
When radionuclides othe than those are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is greater than the calendar year Ilimit of Specifications 3.9.1.2.a or 3.9.2.2.b..
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and d scribed in the Annual Radiological-Environmental Operating Report.
If milk or fresh leafy vegetable samples are unavailable for more than one sample period from one or more of the sampling locations indicated by the
- ODCM, a dis-hllb d
Re.i~~
Effluent Report which identifies th cause of the unavailab'lity of sa...yles and ident-'fies locations fo" 3 '6-2
6.0 6.1 6.1.l The Plant V.
hh'verall P&miwiS ~NAmiUE ~MT~~
Res pan e iQ~Xi~
shall be responsible for operation and shall delegate in writing the succession to this responsibility during his absence.
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6.2 6 ~ 2.1
'Q gMiKli8TP-&~iVF ~~~it.S 0 ~can,a.W ddii Onsite and Offsite Or anization 5 h
O
~PtTsite and -aa offsite organization shall be established p
for unit operation and corporate management, The onsite and offsite organization shall include the positions for activities affecting the. safety of the nuclear power plant.
a.
Lines of authority, responsibility and communication
~rsvp~
J*hl'd~d~h'h o.n5 management levels, ~agh intermediate levels, ~ and opera.'ring cia ga~noSow" 11 positions.
Thy'se relationships shall be documented and
- updated, as h
- h Dha Kl organization charts ~M be documented in the UPSAR~
b.
he Senior Vice President, Production and Engineer'hal ave corporate responsibility for ove Plant nuclear safe and shall take any me res needed to maintaining, and providin assure acceptable p
ormance of e staff in operating,
!l chnical support in the Plant so that continued lear safety assured.
c.
The Plant
- anager, Ginna Station shall have-resp ability for overall unit operation.and s
have d
control over those resources necessary for safe operatio and maintenance of the Plant.
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he. persons responsible for the tra'ing, nealt.
physics quality assurance functions y report to an appropriat anager onsi
, bu-shall have direct access to res z
corporate management at a level w
e action appropriat o the mitigation of zning, health physics and quality ass ce concerns can be accomplished.
MnA.
6.2.2
~i+ aAo.
inc Ku4M The
'rganization shall the following:
a woMQ
+
b.
e licensed Operato om when fuel is in the re zn the At least two licensed Operators shall be pres the control ro reacto t-up, scheduled reactor shutdow uring recover ro caused by transients or emergencies 1 core alterations shall be directly superv'ither a lice Senior Reactor perator or Senior Reactor Operator Lim'o andling who has no other c operation.
rent responsibilities during thxs An individual qualified 'in radiation protection. pro-cedures shall be on site when fuel is in the reactor.
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A ate shift coverage shall be maintained
'out routine h vy use of overtime. Administrat'rocedures shall be develop and implemented limit the working hours of unit staff who rfo safety-related functions including senior react ope ors, reactor operators',
health physici auxiliary op
- tors, and key maintenanc ersonnel.
Changes to the guide 's for the adm'trative procedures shall be submitted to th C
for review.
MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION RCS Above Cold Shutdown Cold Shutdown
No 1
None SS - Shift Supe visor with a Senior Reactor 0
rators License SRO - individual w th a Senior Reactor Opera rs License RO individual wit a Reactor Operators icense AO - Auxiliary Operat STA - Shift Technical Ad 'sor Except for the Shift Supervis
, the Shift Crew Composition may be one less than the minimum requi ment of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in der to commodate unexpected absence of on-duty shift crew members rovided. i diate action is taken to restore the Snift Crew Composi on to within e minimum requirements of Table 6.2-1.
his provisi does not permit ny shift crew position to be unmanned upon shift change due to an onc ing shift crewman being late or absent.
During any abs ce of the Shift Supervisor f m the Control Room while the unit is ove Cold Shutdown, an individual (other than the Shift Tecnnical visor) with a valid SRO license sha be designated to assume th Control Room command function.
During any absence of the Shift S'rvisor from the Control Room while the u 't is in Cold Shuzdo n or Refueling, an individual with a valid S or RO license shal be designated to assume the Control Room comman function.
A,QM,i&iawRA~v~
6.3 6.3.1 Vo,i+ ZM% ~<i<~~~
vnjp sW c~EO Each member of'the
'hall meet or exceed the minimum qualifications of ANSI Standard N18.1-1971, gsLultsow Lp as supplemented by Regulatory Guide 1.8, September 1975, for comparable positions
~
1 1
6,4 TRAINING 6.4.1 6.4.2 A
retraining and replacement training program for t fa 'ty staff shall be maintained under the di tion of the Divis' Training Manager and shall or exceed the r q"irements an recommendations Section 5.5 of ANSI Y18.1-1971 and Appendix 0
CFR Part:
SS.
The training progr shall mee r exceed NFPA No. 27, 1975 Se "ion 40 cept that (1) training salvage operations nee not be provided and (2) the Fire B 'e training se.s
'ons shall be held at least quarterly.
Dr ar considered to be training sessions.
6.
Three separate organizational units shall be est ishedh for the purpose of review and audit of plant oper ions and a
P fety-related matters.
One of these will be an on-site ope ations review
- group, the Plant Oper tions Review Commi tee (PORC).
A second is the Quality ssurance
({}.A.)
- group,
'sponsible for the audit of safety-related activities associated with plant op ations.
A third is the independ t audit and review,
- oup, the Nuclear Sa'ety Audit and Revie
- Board, (NSARB).
This group is responsible for the periodic review of he activities of the Plant )
Operations Review ommit e,
for directing audits and evaluating their resul and for the management, evaluation of the status and de acy of the Quality Assurance program.
6.5.1 PLANT OPERATIONS iaaf CORI PORC FUNCTION 6.5.1a 1 The Plant Ope ations Review Co ittee shall function to advise the lant
- Manager, Ginna tation on all matters related t nuclear safety and for re erral of appropriate matter to the Nuclear Safety Audit and eview Board.
Corrected Amendment No.
1'5, 32 6.5-1
COMPOSITION
~ 5.1.2 The PORC shall be composed of the; Chairman:
Superintendent, Ginna Production Vice Chairman/Member:
Superintendent, Ginna Su ort Services Vice Chairman/Member:
Operations Manager mber: Maintenance Manager Me er: Maintenance Planning/Schedulin Manager Membe:
Instrument
& Control Superv or Member:
echnical Manager Member:
Re ults and Test Superv'sor Member:
Reac or Engineer Member: Health hysics and hemistry Manager Member: Nuclear A suranc Manager tdemher: Quality Can ro Engi.neer ALTERNATES 6 ~ 5 ~ 1.3 Alternate membe s shall b
designated by name, in writing by the Chairm n.
6.5.1e4 The PORC hall meet at least onc per calendar month and as conven by the PORC Chairman.
QUORUM:
6.5.1.5 A
orum of the PORC shall consist of t e Chairman or Vice hairman and four members including alte nates.
No more than two shall be alternates.
ndment No. XS, 32
- 6. 5-2
'ESPONSIBILITIES 6.
.1.6 The PORC shall be responsible for:
a
~
Review of
- 1) all procedures required by Spec'fication 6.8 and changes thereto,
- 2) any other propose procedures or changes thereto as determined by the P ant
- Manager, nna Station to affect nuclear, safety.
b.
Revx w of all proposed tests and exper'nts that affect c
nuclear safety.
Review o
all proposed change to the Technical Specification d.
Review of all pr osed changes or modifications to plant systems or equipmen that a
ect nuclear safety.
e.
Investigation of all v'olations of the Technical Specifications and sha repare and forward.
a report covering evaluation nd re ommendations to prevent
\\
recurrence to the S
ior Vice Pr sident, Production and Engineering, and o the Chairman o
the Nuclear Safety Audit and Revie Board.
f.
Review of fac'ty operations to detect po ntial safety hazards.
g.
Performan e of special reviews and investigate ns and reports thereon as requested by the Chairman of the Nucl r Safety Audit and Review Board.
Amen nt No. 3g, 3S
- 6. 5-3
h.
Review of the Plant Security Plan and shall submit recommended changes to the Chairman of the Nucle Safety Audit and Review Board.
Review of the Radiation Emergency Plan and s all submit recommended changes to the Chairman of th Nuclear Safety Audit and Review Board.
R view of implementing procedures for he Plant Sec ity Plan and the Radiation Eme gency Pian and propos d changes thereto.
k.
Review o all Reportable Event Review of t e Fire Protectio Program and Implementing Procedures an submittal o
recommended Program changes to the Chairman f the N clear Safety Audit and Review Board (NSARB).
AUTHORITY 6.5.1.7 The PORC shall:
a 0 b.
Recommend in riting to the ant Manager, Ginna b
Station app oval or disapproval of items considered under 6.
.1.'6(a) through (d) and
) above.
Render eterminations in writing wi regard to whether or n t each item considered under 6.5.
.6(a) through (d
and (l) above constitutes an unrevie d safety uestion as defined in 10 CFR Section 50.59.
Amendm nt No.
49 6.5-4
C.
Provide immediate written notification to the Senior V'ce President, Production and Engineering, and the N
lear Saf ety Audit and Review Board of disagreement etween the PORC and the Plant Manager, Ginna Station
- however, the Plant
- Manager, Ginna Station hall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The P
C shall maintain written inutes of each meeting and cop'es shall be provi d
to the Senior Vice President, Production and E gineering, the Chairman of the Nuclear afety Audit and Review
- Board, and such others as the C airman ay designate.
Amendment No.
32, 38 6.5-4a
~.C..S."-';!'L!CLFAR ShF" > Y AU
$i AND i!::VIL'A'Q'.lI')QV9ARii}
>.',, Xw~TQ" 6
~
2 ~ l The NSARB sha.'l function to provide indeper:dent re ieiv and audit do igna.d activities in the. areas oi:
a ~
uclear pov.er plant operations b.
nu 'ear eng!neering c ~
chemi ry and radiochemistry d.
meta llurg e.
instrumentati and cont. ol f.
radiological -afe 0 ~
mechanical
<<nd elec ical engi cering h.
qualit; ass..rance practz es CCRC!'COITION 6.5.2.?
The compositio of the NSAR!3 shall be es
. blished as folio'vs Chairman an Vice Chairman <<ppointed by na ie by the Chairmaa of th.
13oa d ar:d Chief Executive Gfficer of the rporation or officer f his designation.
At le st four ter.",nically'qualified persons:vho ar not embeis cf he plant staff t.o provide expertise in the function:.l are.
t escrib. d in 6.5.2.1.
- 6. S-5
C POSITION (Continued
)
/
c.
At least one qualified non-company affiliated technical consultant and others as required Duly appointed consultant members shall have e ual vote with company affiliated members of the oard.
d.
Three members from the staff of the
.E. Ginna uclear Power Plant.
e.
Me ers in (b) and (d) above to e designated by the airman of the Board and ief Executive Officer ALTERNATES 6.5.2.3 Alternate members hall b appointed in writing by the NSARB Chairman to serve on a te o ary basis;
- however, no more than two alternates shall partic'e in NSARB activities at any c ne time.
QUALIFICATIONS 6.5.2.4 The minimum alifications of t Nuclear Safety Audit and Review Board with reg'ard to the individua members shall be maintained at a lev l ~qual to or higher than t e following:
a.
R ctor Eng ineering ngineer
.".g graduate or equivalent wx h over eight years ex"ec;ence in the nuclear power eld and over four
~ ea s responsible engineering m nagement.
- 6. 5-6 Amendment No.
2
UALZFICATEONS (Continued (b
UtilityOperations Engineering graduate or equivalent with over eight years exp ience in utility operations and with over four years responsible engineering mana ment.
(c)
Reactor hysics Physics gr uate or equivalent with over five years erience in reactor physi work.
(d)
Heat and Fluid F w
Engineering or Phys s graduate or equivalen wi.th four years experience in heat and fluid flow alysis.
(e)
Environmental Analysis Engineering graduate or equi lent wit over five years experience in environmental hazard analysis.
(f)
Reactor Control and Enstrumentati Engineering graduate or equival t wi over five years experience in nuclear engineering.
(g)
Power Plant Operations Engineering graduate or uivalent with over ve years experience in power plant operations (h)
Safety Analysis Engineering grad e or equivalent with over five year experience in nuclear enginee g.
(i)
Chemistry and diochemistry Engineerin grauuate or equivalent with over five years experie e in nuclear gine erin g (j)
Radiol gical Safety Engi eer graduate or equivalent with over five years experience in health p
sics and/or radiological safety.
.5-7
EETXNG FRE UENCY 6
~.2.5 At least semi-annually and as required on call of the Chairman.
QUORVN 6-5.2.6 A quorum shall consist of a majority of th principals a
will include the Chairman or Vice C
irman. At least one f the quorum shall be a non-comp y affiliated techni al consultant and no more th n a minority of the quorum s
1 1 be members of the Pl nt staff.
REVIEW 6.5. 2.7 The NSARB sha review:
a.
The safety e aluations for 1) changes to procedures, equipment or s tems as described in the safety analysis report a
2 ) tests or experiments completed under the provi ion f 10CFR Section 50.59 to verify that s ch aetio s did not constitute an unreviewed afety questi b.
Proposed changes to procedu es, equipment or system which have been dete ined by the -PORC to invo ve an unreviewed safety qu tion as defined i
10CFR Section 50.59.
c.
Proposed tests or experiments which ve.been determined by the PORC to involve an un eviewed safety cuestion as defined in 10CFR Secti 50.59.
d.
Proposed changes in Technical Specifications or licenses.
- 6. 5-8 I
Amendment-No.
1
e.
Violations of applicable statutes,
- codes, tegu atio'ns~
- orders, Technical Specif ications, license r uirements,,
or of internal procedures or instructions aving clear safety significance.
f.
Sig ificant operating abnormalities o
deviations from ormal and expected performanc of plant equipme t that affect nuclear sa ty.
g.
All Repor able Events.
6.5-8a Amendment No.
21
Continu h.
Any indication of an unanticipated defici<<cy aspect oi design or operation of saf ety rilated atructure3,
- systems, or components, eports and meeting minutes of the E la pperatfong Re
'ew Committee.
AUDITS 6,5.2.8 The NS shall direct the est@bi hment oi an audit program a
evaluate audits per rmed to ensure safe facility ape tion.
Audits sha 1 encenpass:
r The conformance of facility operation to a11 provisionl contained wf.thin the T
hnical Speci fica tioaa and applicable license ndi iona at least once poz yeaz.
b.
The par formance, tz ing and qualificationl of the operating and tee al taff at least onc! a yeaz.
C ~
d.
The results of all actions ea to correct doficieaciea.
occurriny in fa laity equiyne t, structurII, system or method of ope tioa that affect uclear safety at leaat onca per s months.
The perfo e of all activitiea requ reel-hy the Quality-Assur o Prograa foz R. E. Oinna Nuclea Power Plant t meet ho criteria of Appoa4hc.3, 10 CFR-50.
e least.one p
24 monthl.
e Radiation Emergency Plan aa4 impleawntane p
edurea-.
at least at the frequency required by 10 CFR 50 tt) ~
Tho Station Securi~ %lan an4 iay1ementtm Icoce4ure at leaat at the frepsancg zepLtx4 hy 10 CFP.'3.10(4)
~
'ne cent No.
46
ITS (Continued) g.
The Facility Fire Protection Program and implem ting procedures at least once per two years.
h.
An independent fire protection and loss revention program inspection and, audit performed at 1 ast once per 12 months utilizing. either qualified o
site licensee ersonnel or an outside fire protectio firm.
i.
A inspection and audit of the fire rotection and loss pre ntion program performed by n-licensee personnel at le st once per 36 months.
The personnel may be represe tatives of ANl, an i urance brokerage firm, or other qua 'fied individual j.
The radiolo ical enviro ental monitoring program and the results th reof a
least once per 12 months.
k.
The Offsite Dos lculation Manual and implementing procedures at le once per 24 months.
1.
The Process Co rol rogram and implementing procedures at least onc per 24 m nths.
m.
Any othe area of cility operation considered appropr te by the NSARB the senior vice president, Produ tion and Engineering.
endment No. gZ, 38 6.5-10
UTHORITY 6.
.2.9 a.
The Chairman of the Nuclear Safety Audit and Revi Board is responsible to the President on all acti ities for which the Review Board is responsible.
The NSARB shall report to and advise t e Senior Vice President, Production and Engineering, n those areas of sponsibility specified in Sections
.5.2.7 and 6.5.2.8.
RECORDS 6.5.2.10 Records o
NSARB activities shall b
- prepared, approved, and distributed indicated below:
a.
Ninutes sha 1 be recorded o all meetings of this Board.
Copies of the inutes sh 1 be forwarded within 14 days following each m eting to the Corporate Chairman of the Board, Senior Vice
- esident, Production and Engineering and such others as the Chairman of the NSARB may designate.
b.
Reports of re iews encompa sed by Section 6.5.2.7 e,f,g and h abov
, shall be prepar d, approved and forwarded to the Senior Vice President, P
duction and Engineering
/
withirv'14 days following completi of the review.
/
. l c.
Audit reports encompassed by Section 5.2.8 above, shall he forwarded to the senior Vice presiden production and Engineering and to the management positio responsihle for the areas audited within 30 days after c pletion of the audit.
endment No./P, )g,38 6.5-11
ROCEDURES 6.
.2.11 6.5.3 6.5.3.1 Written administrative procedures for committee oper ion l'I shall be prepared and maintained describing the me od of submission and the content of presentations t the committee, provisions for use of subcommittees, review and approval by members of written committe evaluations nd recommendations, distribution of minu s,
and such o
er matters as may be appropriate.
UA TY ASSURANCE GROUP The o ganization, qualifications, r sponsibilities and trainin of quality assurance per onnel for audits of safety re ated activities are d scribed in the Quality Assurance ogram.
6.5-12
6.6 REPORTABLE EVE%I'CTION 6.
.1 The following actions shall be taken for Reportable Events:
a.
The Commission shall be notified and a report submitted pursuant to the requirements of Section 50a73 to 10 R
Part 50, and b
Each Reportable Event shall be reviewed by the P
RC the results of this review shall be submit d to and the N
and the senior Vice President, Pr uction and Engi eering.
Amen ent No.
H. 82, 38
- 6. 6-1
t'
J ibk~ow The following actions shall be taken in the event a Safety Limit is violated:
(a) a.
The provisions of 10 CFR Section 50.36(c)(1)(i) shall be complied with immed'tely.
b.
The Safety Limit violation shall be reported to the h NAg c.
A Safety Limit Violation report shall be reviewed immediate ly.
, to the Report shall be prepared.
The mc bbA, by the M~ This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective ac.ion taken to prevent recurrence.
d.
The Safety Limit Violation Report shall be submitted to hJRC.
o%s Ae ~~~
~Mo~
h h ~
d h
within two weeks of the violation.
~ C54LCsA Z'Q.
~r po~ ~a cu.&@4
t..w.i ~i.
~
~
~
~
~
Pco~~
Written procedures shall be established, implemented, anc
~oilo~iw~
maintained covering theactivx,ties:
a.
The applicable procedures recommended in.
~Regulatory Guide 1.33, Rc.us.iIo~ 2. Apy~~gqi, g
+<'4ru~ g qq g 8.2 Each procedure and administrative policy of 6.8.1 ab the changes
- thereto, shall be reviewed by t QRC and appro d
by the
- Plant, Manager, Ginna ation prior to implemen+.ati and reviewed per'cally as set forth in the applicable proce es.
6.8.3 Temporary changes proce s of 6.8.1 above may be made provided:
a.
e intent of the original procedures ot b.
The change is approved by two members of management staff, at least one of whom is Supervisor who holds a
Senior Reactor License.
altered.
plan the Shi Operator's
u P G 6:o3.Z1, s~W~
c.
The document,ed, reviewed b
- 0RC, and:,
approved bp the Plan implementation.
er, ion within 10
I ~
3'
Programs and Manuals 6.5
- 6. 0 ADMINISTRATIVE CONTROLS 6.5 Programs and Manuals 6.5.1 The following programs shall be established, implemented, and maintained.
Offsite Dose Calculation Manual (ODCM) a.
The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and b.
The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Specification 6.6.2 and Specification 6.6.3.
Licensee initiated changes to the ODCM:
a.
Shall be documented and records of reviews performed shall be retained.
This documentation shall contain:
1.
sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),
2.
a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.. 106, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; b.
Shall become effective after review and acceptance by the onsite review function and the approval of the Plant Manager; and
Offsite Dose Calculation Hanual (ODCH)
(continued)
C.
Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCH as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCH was made.
Each change shall be identified by markings in the margin of the affected
- pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e.,
month and year) the change was implemented.
Post Accident Sampling This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive
- gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions.
The program shall include the following:
'a ~
Training of personnel; b.
Procedures for sampling and analysis; and C.
Provisions for maintenance of sampling and analysis equipment.
i
('. o p R Wi&sha~)uC 6.$V
~g~g ~~~@,
In
'tion to the applicable reporting require Title 10,.Code o
eral Regulatio e following ident'ied reports sha s 'd to the Regional Administra of the USHRC, Region 1, unless ise ed.
6.9.3..1 Star uo Report.
A summary report of plant startup an <
power escalation testing shall be submitted follow' (1) rece'pt of an operating license, (2) amen nt to
)
e license involving a planned increase in ower
- leve, (3) installation of fuel that has a different design has been manufactured by a 'f erent fuel
- supplier, (4) modifications t may have significant y altered the nuc ar, thermal, r )hydraulic performance o
the plant.
The eport all addzess each'of the tests performed and sha in general include a description of the measured. val s of operating conditions or characte istics ta'ned durin th test program and a
comparison of ese values with de 'gn predictions md specca
'ons.
Any corrective action that were reauir to obtain satisfactory operation all also be des ihed.
Any additional specific details re ized in icense conditions based on othe-commitments sha be included 'n this report.
~AOmam~ ~O lO ( FP Sn,g
artup reports shall be submitted within (.1) $0 d follows completion of the startup test pr am, o
(2 )
90 days fo wing resumption of,ercial power operation, whichever x earli
'. If the Startup Report does not cover b ev s (i.e., completion of startup test pro
, and resumption commercial power ope ion), supplementary reports sha e submitte at ast every three months until both events have n
completed.
c.. 4.'I Mo lv Ooeratina eoozt.
Routine reports of operating
)in~4.<~~
statistics and shutdown experience shall be subm'tted,.
ca.O~gg, +o
~~ P~~u~ ~ p~~
Op~~
ow~ /~tow<~
I
~
~
o ~~+4!~
E-4~ ~ >SA of each month following the calendar month covered by the report.
The mon ly report. shall include a narrativ s
of operating experience describing operation of the faci including major safe related maintenance for the monthly period, ep at safety related maintenance performed xng the z ling outage may be reported in e monthly repozt or the mon ollowing the en the outage rather than each month during the tage.
a ual Radioloaical =nvironmental Ocerati Resort
'r}nmc9-A'~diological /vironmental generating gCport cove i..g the oper'tion of the unit during the previous calendar bg 5
y h llb.'~
y f
h
~U ~~c3-pQ+ih,eh S4~ ~ co*g,<sic.n+
~ i~Svm
~~inn.h sn ~ 0%iiX, Ca.h~k ~'~~
~~~~~ QQC.PAL a.eh l< ao C PR. '50, A~44iz
~~g~g hV Q,+ l hv, Q ~ 3, ~ tv.c.
The report sPxall include summaries, interpretations,.
and analyst of trends of the results of the radiological environmental activities for the report period zncludi oarison with background (control) samp~
and previous enviro 1 surveillan eports and an assessment.
of the obse
~pa the plant operation on the envir nt.
The reports shall also inc the re s of land use censuses as reauired.
The pnnual radiological gnvizonmental gpenating gepott
+~
04 cLm~4M shall include results ~hhe-of all radiological environmental
~ng~
samples In the event (a4.u iuo3-that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missin results.
The
- a. g.vyplC~N.
b missing data shall be submitt as oon as possible.~
In addition, the annual re s
1 include a discussion which identifies "
circum-stances w '.prevent any required det ion limits for environmental sam analyses f~
being met, and a
discussion of all deviate from the sample schedu'e of Table 3. 16-1.
report sha iso include the following:
summary description of the iological envir en"al monitoring program includ:.rg a map all ampling locations keyed to a table giving distances
E S~mt' G WMM4Ku OJ4 4~&-Z,,<e. M +M 4H,.>
a4~~~M'L~
L.aa Wa~,.~mi.o~ <, No@.~eat
~>'U.
ar rom the reactor, and n an interlaboratory corn Radioactive Effluent Release Ressort o
the
~C Rbmhwmz%dioactive~fluent release
@%ports covering the operation of the. unit lo C Pg. SOTS(
shall be submitted shall include a summary Thy'- report of the quantities of radioactive liquid and gaseous ef luents
~~ ~u~+. ~
and solid waste released of radar tion doses from the radioactive liquid d
du 'ng each of lined in gaseous e
uents released from the mnt P
the previous ur calendar quarters as o
Regulatory Guide
.21, Revision 1.
addition, the site boundary maxim noble gas ma ai" and beta air doses shall be evaluate T
assessment of radiation doses shall be performed 'cordance with the ODCM.
This same report shal include n annual summary of hourly meteorolog al data collect d over the previo"s calendar year Alternatively, the li nsee has the option of etaining this summary on site 'n a fil that shall e provided to the NRC upon request.
A o, the sem'nnual report. shall include any ne location(s) identified by the land use census which he radioactive effluent release report submitted wi
'n 60 days of January 1 shall include an assess nt
pm~
R~l M~
4
.4 Q, C-V=..R S'0. 3+a, ~Q, hQ C HR.,~ gp~~
yield a calculated dose or dose commi.ment creater than those rming the basis of Specifications
- 4. 12.2.2 or 3. 16. 1.
Th repo i shall also contain a discussion which identifies he causes the unavailability of milk or leafy veget-le samples 0
and identif s locations for obtaining replacem t samples in accordance with pecification 3.16.1.4.
The radioactive eff1 nt release repor shall include a
discussion which identi s the ci umstances which prevent any required detection limi f r effluent sample analyses from being met.
The radioactive efflu t release repo ts shall include any
'changes made duri the reporting period o the ODCH as specified in S ction 6.15, and to the Proces Control Program as specifi d in Section
- 6. 16.
The radioactive e fluent
.releas reports shall also include a discussion of a major ch ges to radioactive waste treatment systems in accor nce ith Specification 6.17.2.1.
ssurizer Relief and Safet Valve Challen es Challenges to the p
or safety valves n annual basis.
r power e reported no less relief valves than
nnually:
Results of required leak te source es s
e resence of 0.005 r
rie or mo e
able co 'o Dc ~@BOA to~om &~~ ~aM number of statio.i, utility and other personnel (including mrem/yr and their a'ssociated man-rem exposure according to work and job functions~ e.g.,
reactor operations and surveillance, i+service inspection, routine maintenance, special maintenance (describe maintenance),
waste processing, and refueling.
The dose assignments to various dut functi~og a
be mtimatep'ased on
+a,mo<v,~s~
s,~~o pocket d osimeter,(TLD~, or film badge measurements.
'/
~gC+ "a4aML~
~lobP a~4. '0% bC ~lb~
A.q~ 3+ cap 2a3 total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.
-(HGTB-This tabulation supplements zo,zzca<>>~~
~.the requirements of Section Annually:
The results of specific activity analysis the primary coolant exceeded the limi of Specificatio 1.4.1.a and b.
The lowing information shall be included:
(1 or power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to e first sample
'. which the limit was excee
- (2) Results of the last isotope alysis radioiodine performed prior to exceeding the limit,
results of analyses while the limit was exceeded and re its /
o one analysis after the radioiodine activity was red ced to less than the limit.
Each result should include th date and time o
sampling and the radioiodine concent tions; (3)
Clean-up ystem flow history starting 48 hou prior to the first sampl in which the limit was exceeded (4) Graph of the I-131 concent ation and one other dioiodine isotope concentration as function of time f r the duration of the specific activity a ove the steady tate level; and (5)
The time duration when t e specif'ctivity of the primary coolant exceeded the rad iod' limit.
Reactor Overpressure Prote son System Operation In the event either the P RUs r the RCS vent(s) are used to mitigate a
RCS pressur transien a Special Report shall be prepared and submit d to the Comm sion within thirty days.
The report shall describe the circu tances initiating the transient, the ffect of the PORUs or ven (s) on the transient and any o
er corrective action nece ary to prevent recurrenc Specia reports shall be submitted tvi.t l0 C FF'Q. 4 tn a ceo d.anc.e g
~
within the time period specified for each report.
h
TABLE 6.9-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM
SUMMARY
Name of Facility R. E. Ginna Nuclear Power Plant Docket No. 50-244 Location of Facility Wa ne Count New York Reporting Period Medium or Pathwa Sampled (Unit of Measurement)
Type and Lower Limit Total Number of All Indicator $ocations Locations with Highest Annua
~an Control Loc~tion f Analyses Detection Mean LI1)
Name M
n(()
Mean($ )
P ormed (LLD)
Range Distance and Dxrectxo ange Range r
Nomin ower Limit of Detection (LLD) as defined in Table Notation a. of Table 4.12-1.
b ean and range based upon detectable measurements only.
Fraction of detectable measurements at specified locat.un i
indicated in parentheses 1
TABLE 6.9-2 REPORTING LEVELS FOR RADIOACTIVITYCONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Analysis H-3 Water (pCi/1) 2 x 10 Airborne Particulate Fish r Gases (pCi/m
)
(pCi/Kg, wet)
Mii/1)
Broad Leaf Vegetables (pCi/Kg, wet)
Mn-54 Fe-59 Co<<58 Co-60 Zn-65 Zr-Nb-95 I-131 Cs-134 Cs-137 1000 400 1000 300 300 400(a) 30 0.9 10 20 3 x 10 10 1 x 10 2 x 10 1 x 10 2 x 10 60 70 1 x 10 1 x 10 l
2 x Ba-La-140 200 a)
Total for parent and daughter 300 I
RECORD RETENTION In accordance with Rochester Gas and Electric Co ration policy, operating charts for the first year's o eration will be permanently stored.
The following records shall be retained fo at least ive years:
a.
Records and logs of facility opera ion, including wer levels and periods of oper tion at each i
I I
Recor and logs of princip maintenance activities,,'nspection, rep r
su,bstitution or items of equipment pertaining pow r level.
b.
includin replacemen of principa to nuclear sa ety.
c.
Reportable Even R
orts.
d.
Records of surve'nce activities, inspections, and calibratio s re ired by these Technical Specificatio s.
e.
Records o
reactor tests r experiments.
f.
Records f changes made in e Operating Procedures.
g.
Recor s of sealed source leak ests and results.
h.
Rec rds of annual physical inven ry of all sealed urce material of record.
Th following records shall be retained f the duration f the Facility Operating License:
a.
Records and drawing changes reflecting fac ity design modifications made to systems and equi ment described in the Final Safety Analysis Report;
~ J'
changes shall also be periodically incorporate Co into the as-built file.
Records of new and irradiated fuel invent y< fuel transfers, and assembly burnup historic cords of plant radiation and contam'nation su eys.
d.
Recor s of off-site environmental monitoring surveys e.
Records o
radiation exposure of all plant personnel<
ncluding all c tractors and visitors to the plant ho enter r iation control areas.
f.
Records of radi ctivi in liquid and gaseous material released to he environmental and radioactive waste
'pments.
g.
Records of trans'ent operational cycles for those facility componen designed for limited I
number of tr nsients or c les.
h.
Records of training and qua fication for current station echnical and operati s staff members.
i.
Recor s of in-service inspectio peformed pursuant to ese Technical Specifications.
j.
R cords of reviews performed for cha ges made to procedures or equipment or reviews of ests and k.
experiments pursuant to 10 CFR Section 5
59 Records of meetings of the PORC and the NS B
Records of Quality Assurance activities as r uired by the QA Nanual not listed in 6.10.1.
Ame
Records of the service lives of,all hyd xc and mechanic snubbers listed in the rvice 1nspection Program including e
da which the service life commences and a
ciated in lation and maintenance record
- 6. 11 RADIATION PROTECTION PROGRAM Radiation c
rol procedures shall be prepared and made avail e to all station personnel or ot ersons who may be subject adiation exposure at the station.
These procedures s
'show pe ble radiation exposure, and shall be consistent with the requ ts 0 CFR protection program shall be sniped -nd maintained to Part 20.
The radiation t the requirements of 10 CPR Part 20 th exceptions set forth in Section 6.13 of t e
Technic pecifications.
The'rogram shaLl be adhered to for all operat olvtng personnel radiation exposure.
'k l 0
.Q.. 0 44bh.ahJ<c it-ATluu ~K TP DGQ
'7 6.~
HIGH KQBIATION ARE'A ieu of the "contxol device" or "alaxm signal" required by 20.203 (c) (2)
CFR Part 20:
a.
Each High Radiate radiation is 1000 XI1 wh3.
hr intensity of ess shall be.-barri-caded and spicuously posted as a hx diation a
and entrance thereto shall be controlled by re-lP quiring issuance of a Health Physics Work Permit Any individual or group of individuals permitted to enter such areas shall be provided with one or'ore of the following:
A radiation monitoring device which con-tinuously indicates the radiation dose rate in the area.
-49 A radiation monitoring devicewhieV continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel~
,t of them.
4 in~.vill qualified in radiation protection procedures with Health P y 'nel shall be exempt from the HPWP iss requirement during the pe of radiation protection duties, r
are lant radiation prote ures for entry into high radiation
2 o.~<Oi(.ab 2n. 'l(oOl (c)
Pursuant to 10 CFR 20, paragraph
, in lieu of the requirements of 10 CFR
, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is
> 100 mrem/hr but
< 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).
Indi 'duals q~lifie Qn.radiation protection procedures (e.g.,
e'chns cs an~or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates
< 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
a radiation dose rate monitoring device who is 3
responsible for providing positive control over sWah$
h d
perform periodic radiation surveillance at the d
frequency specified in the/%~
Q.us'P Each High Radiation Area in which the inten of radia is greater than 1000 mrem/hr s
lacked doors shall be locked oors shall be maintaine nder the be subject I eo hhe provia of 6.13.1 a.
- ove, and in addition/
ided to prevent unauthorized entry into the areas and the ys to unlock these inistrative control of the Shift Supervisor on
In addition to the requirements of Specification 5. 11.1, areas with radiation levels Z 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the ke s shall be maintained ugde th a~j istrative control of the Shift Meemies-on duty or supervision.
Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas.
In lieu of the stay time specification of the RWP, direct or r emote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.
For individual high radiation areas with radiation levels of h 1000 mrem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously
- posted, and a flashing light shall be activated as a
warning device.
e
1 ~
6.15
- 6. 5.1 6.15.1.a Offsite Dose Calculation Manual (ODCM)
Any changes to the ODCM shall be made by the fol owing method:
Licensee initiated changes shall be submitte to the C
ission with the Semi-annual Radioactiv
- Effluent, Rele e Report for the period in which e change(s) was mad and shall contain:
6.15.1.b (i) suffi 'ently detailed informati n to support the rational for the change.
(ii) a determina
'on that the c ge will not reduce the accuracy o reliabil' of dose calculations or setpoint dete
'nat's; and (iii) documentation of the act that the change has been reviewed and found ccep able by the PORC.
Licensee initiated changes shall ecome effective after review and acceptan e by the PORC o
a date specified by the licensee.
6.15-1
16 6.1 Process Control Program (PCP)
Any changes to the PCP shall be made by the fol wing method:
6.16.1.a Lz ensee initiated changes shall be sub tted to.the Commi ion with the Semiannual Radi ctive Effluent Release R
ort for the period 'hich the change(s) was made and all contain:
(i) sufficiently eta'd information to support the rationale for change; (ii) a determi tion tha the change will not reduce the o rail conformance f the solidified waste oduct to existing criteri for solid wastes; and
('i
documentation of the fact that e change has been reviewed and found acceptable by th PORC.
6.1.l.b Licensee initiated changes shall become effe ive after review and acceptance by the PORC on a date spec ied I
by the licensee.
6.
FUNCTT,ON 6.17.1 6.17.2
- 6. 17. 2. 3.
Ha'or Chances to Radioact've Waste Treatment S st (Ziquid, Gas ous and Solid)
T radioact'v waste treatment systems (li id, gaseous and olid) are those systems defined in T hnical Speci
- z. ation 5.5.
Major ch es to the radioactive w'ast systems (liquid and gaseous) hal3. be reported by following method.
For the pu=ose f this specifica on, "major changes" is def'd
.". Spec'~ication 6.17.3 be'w.
The Commission sh ll e info d oz all major changes by the 'lus'n of a su'tab e discussion or by re e ence to a suitable discussion o
each change in the Semiannual Radioact've
"-f luent Rel ase port for the period in which the changes we e made.
Th discussion of each change shall contai a) a su-~a=~
oz e evaluation that ed to the determination that "~e c."
ge could be made (in a c dance with 10 C:-R 50 59);
b) suff'i nt.detailed information to suppor the zeaso for Me change; c) a d tailed description of the equ pment, compon rts a d proc sses involved and the 'nterfaces with othe p'n" systems;
d) an evaluation of the cha".ge wn'ch shows the re eases o= radioactive materials in licuid quand' gn I
gaseous e fluerts from ~".ose previously p e
ct d; e) a evalua=ion oz the cha".g wh'ch shows
.e ex
" d max' exposures to individual in the unrestricted a"
to th gene al populat'on oz viousl estimated-
-:om Nose f) doc aentati of ~~e fact that change was
'ev'ewed
~.d und. acceptabl by
~~e POP.C.
"Na-.'cr C.".a.c s" to r
'oact've aste svste
.s (1'c-"'c, gaseous 2nd sol'd.) shal'nc "de the following:
a)
Major cha ces in pro s ecu'=ment, components, and structures fro thos in use (e.g.,
deletion o
evapora" ors d installa 'n of demineralizers );
b)
Major change in Ze design o~ radwaste t e me svs" ms (1'cgxid, gaseous and sol'd).that could s'.
"tly alte the charact 'cs and/or can>>aa a
~ mes c)
Ch= ges
'.". system design wn'" may 'nva 'date
~~e cc'ent analysis (e.g.,
changes. in tank c ac'ty that would alter th cur's re'eased).
ATTACHMENT C Proposed Revised R.E.
Ginna Nuclear Power Plant
.Technical Specification 6.0 Revise the pages as follows:
Remove Insert 11 3.1-21 3.5-2a 3.6-3 3.15-1 3.16-1 3.16-2 Entire Section 6.0 11 3.1-21 3.5-2a 3.6-3 3.15-1 3.16-1 3.16-2 6.0-1 6.0-2 6.0-3 6.0-4 6.0-5 6.0-6 6.0-7 6.0-8 6.0-9 6.0-10 6.0-11 6.0-12
4.8 4.9 4.10 4.11 4.12 4.13 4.14 4.15 4.16 TABLE OF CONTENTS (cont'd)
Auxiliary Feedwater Systems Reactivity Anomalies Environmental Radiation Survey Refueling Effluent Surveillance Radioactive Material Source Leakage Test Shock Suppressors (Snubbers)
Deleted Overpressure Protection System Pacae 4.8-1 4.9-1 4.10-1 4.11-1 4.12-1 4.13-1 4.14-1 4.15-1 4.16-1 5.0 DESIGN FEATURES 5.1 5.2 5.3 5.4 5.5 Site Containment Design Features Reactor Design Features Fuel Storage Waste Treatment Systems 5.1-1 5.2-1 5.3-1 5.4-1 5.5-1 6.0 ADMINISTRATIVECONTROLS 6.1 6.2 6.3 6.4 6.5 6.6 6.7 6.8 Responsibility Organization 6.2.1 Onsite and Offsite Organization 6.2.2 Unit Staff Unit Staff Qualifications Procedures Programs and Manuals 6.5.1 Offsite Dose Calculation Manual (ODCM) 6.5.2 Post Accident Sampling Program Reporting Requirements 6.6.1 Occupational Radiation Exposure Report 6.6.2 Annual Radiological Environmental Operating Report 6.6.3 Radioactive Effluent Release Report 6.6.4 Monthly Operating Reports High Radiation Area Safety Limit Violation 6.0-1 6.0-2 6.0-2 6.0-2 6.0-4 6.0-5 6.0-6 6.0-6 6.0-7 6.0-8 6.0-8 6.0-8
- 6. 0-9
- 6. 0-9 6.0-10 6.0-12 Amendment
~
~ll Proposed
3.1.4 Maximum Coolant Activit S ecifications
- 3. 1.4. 1 a ~
b.
c ~
- 3. 1.4.2 Whenever the reactor is critical or the reactor coolant average temperature is greater than 500'F:
The total specific activity of the reactor coolant shall not exceed 84/E pCi/gm, where E is the average beta and gamma energies per disintegration in Mev.
The I-131 equivalent of the iodine activity in the reactor coolant shall not exceed 0.2 pCi/gm.
The I-131 equivalent of the iodine activity on the secondary side of a steam generator shall not exceed 0.1 pCi/gm.
If the limit of 3.1.4.1.a is
- exceeded, then be subcritical with reactor coolant average temperature less than 500'F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- 3. 1.4.3 a 0 If the I-131 equivalent activity in the reactor coolant exceeds the limit of 3.1.4.1.b but is less than the allowable limit shown on Figure 3.1.4-1, operation may continue for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />.
Amendment No.
g7 301-21 Proposed
3.5.5.2 If the setpoint for a radioactive effluent monitor alarm and/or trip is found to be higher than required, one of the following three measures shall be taken immediately:
(i) the setpoint shall be immediately corrected (ii) without declaring the channels inoperable; or immediately suspend the release of effluents 3.5.5.3 3.5.6 3.5.6.1 monitored by the effected channel; or (iii) declare the channel inoperable.
If the number of channels which are operable is found to be less than required, take the action shown in Table 3.5-5.
Exert best efforts to return the instruments to OPERABLE status within 31 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
Control Room HVAC Detection Systems During all modes of plant operation, detection systems for chlorine gas, ammonia gas and radioactivity. in the control room HVAC intake shall be operable with setpoints to isolate air intake adjusted as follows:
Amendment No.
pp 3.5-2a Proposed
Basis:
The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.
The shutdown margins are selected based on the type of activities that are being carried out.
The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances.
When the reactor head is not to be removed, a cold shutdown margin of 14~k/k precludes criticality in any occurrence.
Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident were as much as 1 psig.<'~
The containment is designed to withstand an internal vacuum of 2.5 psig.+
The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.
In order to minimize containment leakage during a
design basis accident involving a
significant fission product
- release, penetrations not required for accident mitigation are provided with isolation boundaries.
These isolation boundaries consist of either passive devices or active automatic valves and are listed in a
procedure under the control of the Quality Assurance Program.
Closed manual valves, deactivated automatic valves secured in their closed position (including check valves with flow through the valve secured),
blind flanges and closed systems are considered passive devices.
Automatic isolation valves designed to close following an accident without operator action, are considered active devices.
Two isolation devices are provided for each mechanical penetration, such that no single credible failure or malfunction of an active component can cause a loss of isolation, or result in a
leakage rate that exceeds limits assumed in the safety analyses+.
In the event that one isolation boundary is inoperable, the affected penetration must be isolated with at least one boundary that is not affected by a single active failure.
Isolation boundaries that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed manual valve, or a blind flange.
The opening of closed containment isolation valves on an intermittent basis under administrative control includes the following considerations:
(1) stationing an individual qualified in accordance with station procedures, who is in constant communication with the control room, at the valve controls, (2) instructing this individual to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to isolate the boundary and that this action will prevent the release of radioactivity outside the containment.
Amendment No.
3.6-3 Proposed
Over ressure Protection S stem A licabilit Applies whenever the temperature of one or more of the RCS cold legs is
< 330 F, or the Residual Heat Removal System is in operation.
Ob'ective To prevent overpressurization of the reactor coolant system and the residual heat removal system.
S ecification Except during secondary side hydrostatic tests in which RCS pressure is to be raised above the PORV setpoint, at least one of the following overpressure protection systems shall be operable:
a 0 Two pressurizer power operated relief valves (PORVs) with a liftsetting of < 424 psig, or b.
A reactor coolant system vent of 1.1 square inches.
3.15.1.1 With one PORV inoperable, either restore the inoperable PORV to operable status within 7 days or depressurize and vent the RCS through a 1.1 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both PORVs have been restored to operable status.
3.15.1.2 With both PORVs inoperable, depressurize and vent the RCS through a
1.1 square inch vent(s) within 8
hours; maintain the RCS in a vented condition until both PORVs have been restored to operable status.
3.15.1.3 Use of the overpressure protection system to mitigate an RCS or RHRS pressure transient shall be reported in accordance with Specification 6.6.4.
Basis An RCS vent opening'f greater than 1.1 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are
< 330'F<".
This relief capacity will Amendment No-g,
$ $ I NP Proposed
5 ~
Radiolo ical Environmental Monitorin A licabilit Applies to routine testing of the plant environs.
To establish a program which will assure recognition of changes in radioactivity or exposure pathways in the environs.
S ecification Monitorin Pro ram 3.16.1.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.16-1 at the locations given in the ODCM.
3.16.1.2 If the radiological environmental monitoring program is not conducted as specified in Table 3.16-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating
- Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
(Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal availability, or to malfunction of automatic sampling equipment. If the latter, efforts shall be made to complete corrective action prior to the end of the next. sampling period.)
3.16.1.3 If the level of radioactivity in an environmental sampling medium at one or more of the locations specified in the ODCM exceeds the reporting levels provided in the Proposed
ODCM when averaged over any calendar quarter, a Special Report shall be submitted to the Commission within thirty days which includes an evaluation of any release conditions, environmental factors or other aspects which caused the reporting levels to be exceeded.
When more than one of the radionuclides specified in the ODCM are detected in the sampling medium, this report shall be submitted concentration 1
limit level (1)
+
concentration 2
limit level (2)
+
~
~
~
~ Q 1 ~ 0 When radionuclides other than those specified in the ODCM are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is greater than the calendar year limit of Specifications 3.9.1.2.a or 3.9.2.2.b. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
3.16.1.4 If milk or fresh leafy vegetable samples are unavailable for more than one sample period from one or more of the sampling locations indicated by the
- ODCM, a discussion shall be included in the Radioactive Effluent Release Report which identifies the cause of the unavailability of samples and identifies locations for 3.16-2 Proposed
Responsibil ity 6.1
- 6. 0 ADMINISTRATIVE CONTROLS
- 6. 1 Responsibility 6.1.1 6.1.2 The Plant Nanager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.
The Plant Hanager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to structures, systems or components that affect nuclear safety.
The Shift Supervisor (SS) shall be responsible for the control room command function.
During any absence of the SS from the control room while the unit is above Cold Shutdown, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function.
During any absence of the SS from the control room while the unit is in Cold Shutdown or Refueling, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.
6.0-1 Proposed
'I
Organization 6.2
- 6. 0 ADMINISTRATIVE CONTROLS
- 6. 2 Organi zati on 6.2.1 Onsite and Offsite Or anizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.
The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
a
~
Lines of authority, responsibility, and communication shall be defined and established throughout highest management
- levels, intermediate levels, and all operating organization positions.
These relationships shall be documented and updated; as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.
These requirements shall be documented in the UFSAR; 6.2.2 Unit Staff The unit staff organization shall include the following:
a
~
A non-licensed operator shall be assigned to the shift crew with fuel in the reactor.
An additional non-licensed operator shall be assigned to the shift crew above Cold Shutdown.
b.
c ~
Shift crew composition may be less than the minimum requirement of 10 CFR 50.54 (m) (2) (i) and Specifications 6.2.2.a and 6.2.2.e for a period of time not to exceed 2
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 requirement.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected
- absence, provided immediate action is taken to fill the required position.
(continued) 6.0-2 Proposed
Organization 6.2 6.2 Organization 6.2.2 Unit Staff (continued) d.
The amount of overtime worked by unit staff members performing safety related functions shall be limited and controlled in accordance with a NRC approved program.
e.
The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.
The STA shall be assigned to the shift crew above Cold Shutdown and shall meet the qualifications specified within a NRC approved STA training program.
6.0-3 Proposed
I
Unit Staff qualifications 6.3
- 6. 0 ADMINISTRATIVE CONTROLS 6.3 Unit Staff qualifications 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI Standard N18. 1-1971, as supplemented by Regulatory Guide 1.8, Revision 1, September 1975, for comparable positions.
6.0-4 Proposed
k
Procedures 6.4
- 6. 0 ADH IN ISTRATIVE CONTROLS 6.4 Procedures 6.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:
'a ~
b.
C.
The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; 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 All programs specified in Specification 6.5.
6.0-5 Proposed
I I
Programs and Manuals 6.5
- 6. 0 ADMINISTRATIVE CONTROLS 6.5 Programs and Manuals The following programs shall be established, implemented, and maintained.
6.5.1 Offsite Dose Calculation Manual (ODCM)
'a ~
The ODCH shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and b.
The ODCH shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Specification 6.6.2 and Specification 6.6.3.
Licensee initiated changes to the ODCH:
a.
Shall be documented and records of reviews performed shall be retained.
This documentation shall contain:
1.
sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),
2.
a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20. 106, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; b.
Shall become effective after review and acceptance by the onsite review function and the approval of the Plant Manager; and (continued) 6.0-6 Proposed
H 4
Programs and Hanuals 6.5 6.5 Programs and Hanuals 6.5.1 Offsite Dose Calculation Hanual (ODCH)
(continued)
C.
Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCH as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCH was made.
Each change shall be identified by markings in the margin of the affected
- pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e.,
month and year) the change was implemented.
6.5.2 Post Accident Sampling Program This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive
- gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions.
The program shall include the following:
a.
Training of personnel; b.
Procedures for sampling and analysis; and c.
Provisions for maintenance of sampling and analysis equipment.
(continued) 6.0-7 Proposed
Reporting Requirements 6.6
- 6. 0 ADMINISTRATIVE CONTROLS 6.6 Reporting Requirements 6.6.1 6.6.2 The following reports shall be submitted in accordance with 10 CFR 50.4.
Occupational Radiation Exposure Report A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures
) 100 mrem/yr and their associated man rem exposure according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),
waste processing, and refueling).
This tabulation supplements the requirements of 10 CFR 20.2206(7)(b).
The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements.
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 received from external sources should be assigned to specific major work functions.
The report shall be submitted on or before April 30 of each year.
Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.
The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring activities for the reporting period.
The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
(continued) 6.0-8 Proposed
il il e
i ~
e
Reporting Requirements 6.6 6.6 Reporting Requirements 6.6.2 6.6.3 6.6.4 Annual Radiological Environmental Operating Report (continued)
The Annual Radiological Environmental Operating 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 figures in the
- ODCN, as well as summarized and tabulated results of these analyses and measurements in the format 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 in a supplementary report as soon as possible.
Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a.
The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
The material provided shall be consistent with the objectives outlined in the ODCN and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
Nonthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
6.0-9 Proposed
Nw/%
I b
4t
- 6. 0 ADMINISTRATIVE CONTROLS 6.7 High Radiation Area 6.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c),
in lieu of the requirements of 10 CFR 20.1601(a),
each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is
> 100 mrem/hr but
< 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).
Individuals qualified in radiation protection procedures (e.g.,
Radiation Protection Technicians) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates
< 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
a.
A radiation monitoring device that continuously indicates the radiation dose rate in the area.
b.
A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.
c.
An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the RWP.
6.7.2 In addition to the requirements of Specification
- 5. 11. I, areas with radiation levels Z 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or radiation protection supervision.
Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate (continued) 6.0-10 Proposed
High Radiation Area 6.7 6.7 High Radiation Area 6.7 (continued) levels in the immediate work areas and the maximum allowable stay times for individuals in those areas.
In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.
- 6. 7.3 For individual high radiation areas with radiation levels of z 1000 mrem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously
- posted, and a flashing light shall be activated as a
warning device.
6.0-11 Proposed
Safety Limit Violation 6.8
- 6. 0 ADMINISTRATIVE CONTROLS 6.8 Safety Limit Violation 6.8.1 The following actions shall be taken in the event a Safety Limit if violated:
a ~
b.
c ~
d.
The provisions of 10 CFR 50.36(c)(l)(i)(A) shall be complied with immediately; The Safety Limit violation shall be reported to the corporate executive responsible for overall plant nuclear
- safety, to the offsite review function and the NRC immediately A Safety Limit Violation Report shall be prepared.
The report shall be reviewed by the onsite review function.
This report shall describe:
(1) applicable circumstances preceding the violation, (2) effects of the violation upon the facility components, systems or structures, and (3) corrective action take to prevent recurrence; and The Safety Limit Violation Report shall be submitted to the NRC, the offsite review function, and the corporate executive responsible for overall nuclear plant safety within two weeks of the violation.
6.0-12 Proposed
ATTACHMENT D Marked Up Copy of Improved Technical Specifications (NUREG-1431)
Technical Specification 5.0 Included pages:
5.0-1 5.0-2 5.0-3 5.0-4 5.0-5 5.0-6 5.0-7 5.0-8 5.0-9 5.0-10 5.0-11 5.0-12 5.0-13 5.0-14 5.0-15 5.0-16 5.0-17 5.0-18 5.0-19 5.0-20 5.0-21 5.0-22 5.0-23 5.0-24 5.0-25 5.0-26 5.0-27 5.0-28 5.0-29 5.0-30 5.0-31 5.0-32 5.0-33 5.0-34 5.0-35 5.0-36
- 5. 0-'37 5.0-38 5.0-39 5.0-40 5.0-41 5.0-42 5.0-43 5.0-44
Responsi bility Cu WO ADMINISTRATIVE CONTROLS g g.U M1 Responsibil ity Wl. 1 c 4'.t.
Msassisa.~
The gPl ant shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.
thavv.a~
Th F1 t,~
I l* 1 lh shall approve, prior to 111 tt.l, ht 1
ht h~t 1 t~
modificationar to anviat"systems or that affect nuclear safety.
&mayo.&fata
CoWl.2 The /Shift Supervisor (SS)g shall be responsible for the control room command function.
C, fo,'s
.M.ua.<
During any absence of the SSJ from the control room while the unit is ~
an individual with'AC@P.Senior Reactor Operator (SRO) license shall be designated to assume the control room command function.
During any absence of the+S~rom the control room while the unit is i an individual with
~~ ~~+'" ~&Hl SRO license or Reactor Op rator license shall be designated to assume the control oom command function.
Cc,hh SM r ASS a Oc P
uaXavtg 8.0-1
Organi zati on A.O AONINISTRATIVE CONTROLS
~A".2 Organization Onsite and Offsite Or anizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.
The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
a.
Lines of authority, responsibility, and communication shall be defined and established throughout highest management
- levels, intermediate levels, and all operating organization positions.
These relationships shall be documented and
- updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.
These requirements shall be documented in the~g>>
v CESAR.
b.
The [Plant Superintendent]
shall be responsible for over e operation of the plant and shall have control o
thos nsite activities necessary for safe opera and maintena of the plant; C. Z.I,L.
C.
The [a specific porate executive sition] shall have corporate responsibi 'or ove plant nuclear safety and shall take any measu ed to ensure acceptable performance of the staff 'ting, maintaining, and providing technical ort to t lant to ensure nuclear safety; and Ui 2 ~ VS d.
The indi 'ls who train the operating staf arry out heal
- hysics, or perform quality assurance func 's may r
ort to the appropriate onsite manager; however, t individuals shall have sufficient organizational freedom ensure their independence from operating pressures.
'I M.2.2 'nit Staff amehu4 ~
4'5manIOI'f 1
- ff 1
11 *111~
'a ~
C -2.i.
A me-L'L~Q OymnhOC'446L 4 OQ>yak ~ ~
%<& im ~ ~C~r.
AW O-~4<<W~ mW 4~~
OPa nh'CSC
~SI~~k +e ~ g~~4+ ~~ ~a~
~%4 X4u)~mIN, con inue
Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) b.
At e
~c.i.-
nsed Reactor Operator (RO present in the control room w
e reactor.
In addition, wh's in NODE at least nsed Senior Reactor Operator (SRO shall be pre i
the control room.
A'n aw4uv4sq9. go~wQ~ iw ~h~b4 f~+ ~4~
c ~
shall be on site when fuel is in the reactor.
The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected
- absence, provided immediate action is taken to fill the required position.
CB d.
s
'RO or licensed SRO ue handling who has no c sibilities during this o er e present during fuel an irectly su ervise all CORE ALTERATIO Administrative procedures shall be developed and implemente to limit the working hours of unit staff who perform saf y
ated functions (e.g.,
licensed
- SROs, licensed
- ROs, alth phy 'cists, auxiliary operators, and key maintenanc person 1).
Adequate sh t coverage shall be maintained ithout routine heavy use of o rtime.
The objective sh be to have operating person 1 work an [8 or 12]
ur day, nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week while e unit is oper ing.
However, in the event that unforeseen oblems r uire substantial amounts of overtime to be used, du extended periods of shutdown for refueling, ma'aintenance, or major plant modification, on a tempo ry b is the following guidelines shall be followed:
l.
An individ should not be perm>
ed to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> raight, excluding shift mover time; 2.
An dividual should not be permitted to rk more than hours in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more th 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 'ny 7 day period, all excluding shift turnover time; 3.
A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should be allowed between work periods, including shift turnover time; (continued)
0, 09/28/92
0
b.
Shift crew composition may be less than the minimum requirement of 10 CFR 50.54 (m) (2) (i) and Specifications 6.2.2.a and 6.2.2.e for a period of time not to exceed 2
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 requirement.
Organization
~2 Co A.2 Organization Unit Staff (continued) 4.
Except during extended shutdown periods, the use of overtime should be considered on an individual ba
'ot for the entire staff on a shift.
Any deviation om the above guidelines sh be authorized in advance by the ant Superintenden or his designee, in accordance with approv administr ve procedures, or by higher levels of managemen
, 'ccordance with established procedures and with docum at>
of the basis for granting the deviation.
Controls sha e included in the procedures h that individu overtime shall be reviewed monthly by IPlant Supe 'ndent] or his designee to ensure that excess rs have not been assigned.
Routine deviation from the above guidelines is not authorized.
OR C. g, you
~
~
~
- d. The amount of overtime worked by unit staff members performing safety related functions shall be limited and controlled in accordance with
< NR< ae~u ~w C. t.. iC~
The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.
O A
C l.U.
~
~
C ~ 2" <
~ s~5 s,+oSLsL ~ ~ig~ W ~ ~<4 Z~W~Mw~S-z.K~ ~% ~ g<~a~~oAz S~V'%N ~a~i~ ~ WRC, ~pPb~
S, fA pwg~~
Table 5.2. 2-1 (page 1 of 1)
Minimum Shift Crew Composition(a)
[Single, Unit Facility]
Organizati n
.2 OSITION(b)
MINIMUM CREW NUMBER UNIT IN MODE 1, 2, 3, OR 4 UNIT MODE 5
OR 6 SS SRO RO AO STA(c) 1 None 1
1 None (a)
The shift crew composi
'on may be one le than the minimum requirements of Table 5.2.2-1 for not, more than 2 ho s to accommodate unexpected absences of on-duty shift rew members provided immediate action is taken to restore the shift rew comp ition to within the minimum requirements of Table 5.2.2-Thi provision does not permit any shift crew position to be unmanned on ift change due to an oncoming shift crewman being late or absent.
(b)
Table Notation:
- [Shift Supervisor]
wi h
a Seni Reactor Operator license; SRO - Individual with a S
ior Reacto Operator license; RO Individual with a eactor Operato license; AO
- Auxiliary Operato STA - Shift Technical dvisor.
(c)
The STA position ma be filled by an on-shift S or SRO provided the individual meets t e Commission Policy Statemen on Engineering Expertise on Shif G STS 5.0-5 Rev.
0, 09/28/9
Organization Table 5.2.2-1 (page 1 of 1)
Hiniaua Shift Crew Composition
[Two Units Nith a Cotmen Control Room)
(Totals for Both Units)
OSITION(b)
HININNI CREM XUHBER ONE UNIT IN HOOE 1g 2t 3,
OR 4, AND ONE UNIT EACH UNIT IN HOOE 1, 2, IN HOOE 5, HOOE 6, OR EAC UNIT IN HOOE 5 OR 3,OR4 DEFUELED 6 OR DEFUELED SS SRO RO AO STA(')
1 None 2
3 None (a)
The shift crew composition ma one less than the mini requirements of Table 5.2.2-1 for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to accotmedat unexpected absence of duty shift crew members, provided imnediate action is taken to restore the sh ft crew composition t within the miniaxin requirements of Table 5.2.2-1.
This provision doe not permit any sh t crew position to be uwenned upon shift change due to an oncoming shift cre n being late o absent.
(b)
Table Notation:
[Shift Supervisor]
with a Senior ea or Operator license for each unit whose reactor contains fuel.
Individual with a Senior Reactor tor license for each unit whose reactor contains fuel.
Otherwise, provide an individua for ch unit who holds a Senior Reactor Operator license for the unit assigned.
During ORE ALT TIONS on either mit at Least one licensed SRO or licensed SRO limited to fueL
- andling, wh has no other concurrent responsibilities, axet be present.
Individual with a Reactor perator license o
a Senior Reactor Operator license for unit assigned.
At least one shaLL be assigned each mit whose reactor contains fuel and one RO shall be assig as relief operator for mit(s) in HODE 1, 2, or 3.
Individuals acting as relief oper tors shall hold a license r both mits.
Otherwise, for each mit, provide a relief o rator who holds a license for e mit assigned.
At least one auxi ary operator shall be assigned t each mit whose reactor contains fueL.
Shift Technical dvisor.
(c)
The STA position may filled by an on-shift SS or SRO provided e individual meets the Comnission Policy Statement on gineering Expertise on Shift.
0, 09/28/92
Table 5.2.2-1 (page 1 of 1)
Minimum Shift Crew Composition(
)
[Two Units With Two Control Rooms]
(Numbers for Each Unit)
Organizatio 5
POS ION(b)
HINIHUM CREW NUMBER UNIT IN HODE 1, 2, 3, OR 4 UNIT IN HODE 5
OR 6 UNIT IN MODE UNIT IN MODE 1, 2, 3, OR 4; 5
OR 6; OTHER OTHER UNIT UNIT IN MODE IN MODE 5 5
OR DEFU ED DEFUELED SS SRO RO AO STA(c) 1(d) 1 1(d)
None 1
1 None 1(d) 1 2
2 1
1(d)
None 12(e)
None (a)
The shift crew composition ma be one ess than the minimum requirements of Table 5.2.2-1 for not more an hours in order to accommodate unexpected absence of on-duty s 'f crew members provided immediate action is taken to restore the sh'ft crew composition to within the minimum requirements of Table 5.
l.
This provision does not permit any shift crew position to be mann d upon shift change due to an oncoming shift crewman being ate or sent.
(b)
Table Notation:
[Shift Supervisor with a Senior Reac or Operator license; Individual with Senior Reactor Opera r license; Individual wit a Reactor Operator lice e;
Auxiliary Ope ator; Shift Techni al Advisor.
(c)
The STA position may be filled by an on-shift SS or 0 provided the individual mee the Commission Policy Statement on En ineering Expertise on ift.
(d)
Individual ay fill the same position on the other unit i licensed for both.
(e)
One of he two required individuals may fill the same position n the other nit.
W STS 5.0-7 Rev.
0, 09/28/92
4A.0 ADMINISTRATIVE CONTROLS W3 Unit Staff gualifications Unit Staff gual ificati ons W3 Revie Ninimum qualifications for members of the unit be specified by use all qualification s erencing an ANSI Standard acceptable to the NRC s ecifying individual position qualifications.
G e first method is
- however, the sec is adaptable to those unit staffs requiring spec qualification statements because of unique organizational structures.
5.3.1 Each member of the unit staff shall meet or exceed the minimum ualifications of 4%0-8
Traini 5.
ADMINISTRATIVE CONTROLS 5.4 aining 5.4.1 A retraining and replacement training program for t unit staff hall be maintained under the direction of the [po tion title]
a d shall meet or exceed the requirements and rec mendations of Se tion [
] of [an ANSI Standard acceptable to he NRC staff] and 10 R 55, and, for appropriate designated pos'ons, shall inclu e familiarization with relevant industr operational experi ce.
0, 09/28/92
Reviews and Audi
.5
.0 ADMINISTRATIVE CONTROLS 5.
Reviews and Audits Reviewe 's Note:
The licensee shall describe the method(s) estab ished to conduct dependent reviews and audits.
The methods may take a
ange of forms acceptable to the NRC.
These methods may include creating an ganizational unit or a s nding or ad hoc committee, or assigning individu s capable of conducting th se reviews and audits.
When an individual per orms' review
- function, a cr s disciplinary review determination is nec sary.
If deemed necessary, such eviews shall be performed by the review ersonnel of the appropriate disci line.
Individual reviewers shall not eview their own work.
Regardless of the thod used, the licensee shall speci y the functions, organizational arran ement, responsibilities, appropr te ANSI/ANS 3.1-1981 qualifications, and r orting requirements of each f nctional element or unit that contributes to the e processes.
Reviews and audits of acti ities affecting plant safety have two distinct elements.
The first elemen is the reviews per ormed by plant staff personnel to ensure that day to day ac 'vities are cond ted in a safe manner.
These reviews are described in Secti n 5.5. 1.
The econd element, described in Section 5.5.2, is the [offsite]
eviews and audits of unit activities and programs affecting nuclear safet that ar performed independent of the plant staff.
The [offsite] reviews and dits hould provide integration of the reviews and audits into a cohesive p og am that provides senior level utility management with an assessment of faci ty operation and recommends actions to improve nuclear safety and plant rel'lity. It should include an assessment of the effectiveness of reviews con cte according to Section 5.5. 1.
5.5.1 Plant Reviews 5.5.1.1 Reviewer's Note:
The licensee shal describe provisions for plant reviews (organi tion, reporting, re rds) and the appropriate ANSI/ANS Stand rd for personnel quali 'cation.
Functions The [pla review method specified in Speci
'cation 5.5. 1] shall, as a mi mum, incorporate functions that:
a.
dvise the [Plant Superintendent]
on all ma ters related to nuclear safety; (con nued)
W STS 5.0-10 Rev.
0, 09/28/9
Reviews and Aud' 5.5 5
Reviews and Audits 5.5.1.
5.5.1.2 Functions (continued) b.
Recommend to the [Plant Superintendent]
approv or disapproval of items considered under Specifi ations 5.5. 1.2.a through 5.5. 1.2.e prior to their plementation, except as provided in Specification 5.7.1.
c.
Determine whether each item considered der pecifications 5.5. 1.2.a through 5.5. 1
.d constitutes an u
eviewed safety question as define in 10 CFR 50.59; and d.
Noti the [Vice President Nucle Operations] of any safety signi 'cant disagreement between he [review organization or indivi al specified in Specifi tion 5.5. 1] and the [Plant Superint dent] within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> However, the [Plant Superinten ent] shall have r ponsibility for resolution of such'isagr ments pursuant o Specification
- 5. 1. 1.
Responsibilities The [plant review meth d sp cified in Specification 5.5. 1] shall be used to conduct, as nimum, reviews of the following:
'a ~
b.
C.
d.
e.
All proposed proce ur required by Specification 5.7.1. 1 and changes ther o;
All proposed p ograms req 'red by Specification 5.7.2 and changes ther o;
All propos d changes and modi 'cations to unit systems or equipmen that affect nuclear s
ety; All pr posed tests and experiments that affect nuclear safe; and Al proposed changes to these Technica Specifications (TS),
eir Bases, and the Operating License.
)
Rev.
0, 09/28/92
Reviews and Audi s
.5
.5 Reviews and Audits (continued) 5.5.
Offsite Review and Audit
'eviewer's Note:
The licensee shall describe the pr isions for reviews and audits independent of the plant's staff (organization, reporting, and records) and the appropriate ANSI/
S Standards for ersonnel qualifications.
These individuals may e located onsite o
offsite provided organizational independence from plant staff is aintained.
The [technical] review respon
- bilities, Spec'fication 5.5.2.4, shall include several ndividuals located onsit 5.5.2.1 5.5.2.2 Function The [offsit review and audit provisi s specified in Specification 5.5.2] shall, as a min'm, incorporate the following func ions that:
a.
Advise the [ ice President Nuclear Operations]
on all matters relate to nuclear safety; b.
Advise the manage ent o
the audited organization, and [its Corporate Manageme t d Vice President Nuclear Operations],
of the audit result s they relate to nuclear safety; c.
Recommend to the nag ent of the audited organization, and its management, y corr ctive action to improve nuclear safety and plan operatio and d.
Notify the [
ce President uclear Operations] of any safety
" significant disagreement betwe n the [review organization or individua specified in Specifi tion 5.5.2]
and the
[organiz ion or function being r viewed] within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
[Offsite]
view Responsibilities The [rev ew method specified in Specificatio 5.5.2] shall be respon ble for the review of:
a 0 he safety evaluations for changes to proce ures, equipment, or systems, and tests or experiments complet under the provisions of 10 CFR 50.59, to verify that suc actions do not constitute an unreviewed safety question as efined in 10 CFR 50.59; (conti ed)
0, 09/28/92
Reviews and Audi
.5 5
Reviews and Audits (continued) 5.5.
2 5.5.2.3
[Offsite] Review Responsibilities (continued) b.
Proposed changes to procedures, equipment, or s tems that involve an unreviewed safety question as defin in 10 CFR 50.59; c.
Proposed tests or experiments that involve n unreviewed safety question as defined in 10 CFR 50.5 d.
oposed changes to TS and the Operati License; e.
Vio tions of codes, regulations, or ers, license requs
- ements, and internal procedur s or instructions having nuclea safety significance; f.
All Licen ee Event Reports req red by 10 CFR 50.73; g.
Plant staff erformance; h.
Indications of nanticipa d deficiencies in any aspect of design or operat'on of s
- uctures, systems, or components that could affect ucle r safety; i.
Significant accident
, unplanned, or uncontrolled radioactive release
, 'ncluding corrective action to prevent recurrence; j.
Significant ope ating abno alities or deviations from normal and ex cted perform ce of equipment that affect nuclear safe and k.
The perfor ance of the cor rectiv action system.
Reports or r cords of these reviews sha be forwarded to the
[Vice Presi ent Nuclear Operations] wit 'n 30 days following completio of the review.
Audit sponsibilities The udit responsibilities shall encompass:
a The conformance of unit operation to provision contained within the TS and applicable license conditions, b.
The training and qualifications of the unit staff; (con nued)
MOG STS 5.0-13 Rev.
0, 09/28/9
Reviews and Audi s
.5 5.5 Reviews and Audits (continued) 5.
2.3 Audit Responsibilities (continued) c.
The implementation of all programs required by Specification 5.7.2; d.
Actions taken to correct deficiencies occur ng in equipment, structures,
- systems, components or method of operation that affect nuclear safety; an e.
Other activities and documents as requ sted by the [Vice esident Nuclear Operations].
Reports r records of these audits shal be forwarded to the [Vice President Nuclear Operations] within 0 days following completion of the revi w.
5.5.2.4
[Technical]
Re iew Responsibilitie The [technical]
view responsi lities shall encompass:
a.
Plant operatin charact
- istics, NRC issuances, industry advisories, Lice see E
nt Reports, and other sources that may indicate area fo improving plant safety; b.
Plant operations, ifications, maintenance, and surveillance to v if independently that these activities are performed s
ely an correctly and that human errors are reduced as muc as pract al; C.
d.
Internal an external opera
'onal experience information that may i icate areas for i roving plant safety; and Making tailed r ecommendations hrough the [Vice Presid t Nuclear Operations] f revising procedures, equip ent modifications or other m
ns of improving nuclear saf y and plant reliability.
5.5,3 Recor s
Mr ten records of reviews and audits shall be m intained.
As a m'nimum these records shall include:
a.
Results of the activities conducted under the p
visions of Section 5.5; (co tinued)
G STS 5.0-14 Rev.
0, 09/28 2
Reviews and Audi
.5
.5 Reviews and Audits (continued) 5.5.
. Records (continued) b.
Recommendations to the management of the organi tion being audited; An assessment of the safety significance of he review or audit findings; d.
e.
Recommended approval or disapproval of i ems considered der Specifications 5.5. 1.2.a through
.5. 1.2.e; and Det rmination whether each item cons'dered under Spec ications 5.5. 1.2.a through 5.
. 1.2.d constitutes an unrevi wed safety question as def'd in 10 CFR 50.59.
0, 09/28/92
TS Bases Contr
.6 5.
ADMINISTRATIVE CONTROLS II 5.6 Technical Specifications (TS)
Bases Control 5.6.1 Changes to the Bases of the TS shall be made under ap ropriate administrative controls and reviewed according to Specification 5.5.1.
5.6.2 Lic nsees may make changes to Bases without pr' NRC approval prov ed the changes do not involve either of he following:
a.
A hange in the TS incorporated in th license; or b.
A cha ge to the updated FSAR or Bas s that involves an unrevi wed safety question as defi ed in 10 CFR 50.59.
5.6.3 The Bases Contr l Program shall co ain provisions to ensure that the Bases are mat tained consiste with the FSAR.
5.6.4 Proposed changes tha meet th criteria of (a) or (b) above shall be reviewed and approv d by e
NRC prior to implementation.
Changes to the Bases im e
nted without prior NRC approval shall be provided to the NRC on frequency consistent with 10 CFR 50.71.
W STS 5.0-16 Rev.
0, 09/2 92
Procedure
&0 ADNINISTRATIVE CONTROLS Procedures C.S.l Written procedures shall be established, implemented, and maintained covering the following activities:
O C,'.~.<V.0 c,z.v C ~ S. sv'
~
c.t.t.v a 4 The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; The emergency operating procedures required to implement the requirements oj NUREG-0737 and 4PAUREG-0737, Supplement 1,
as stated in +Generic Letter 82-33~~~A
~C~,v" 1.2 5.7.1.3 All programs specified in Specificationp-.+-.Es.
4.s Review and Approval ach procedure of Specification 5.7. 1. 1, and changes thereto sha be reviewed in accordance with Specification 5.5. 1 proved by the nt Superintendent]
or his designee in ac ance with approved adm
'strative procedures prior to im entation and
,reviewed periodic as set forth in adm' trative procedures.
Temporary Changes Temporary changes to edures of provided:
ification 5.7.1 may be made
'a ~
T ntent of the existing procedure is not ered; The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator license on the unit affected; and (continued)
(a WO-17
A >~~&a~Ie.dpi ~~ c.g~~~
Ca ~ >
Programs~and Manual s Programs~ and Manuals 4.5 5 ~
~
~
Temporary Changes (continued)
C ~
The change nted ewed in accordance with Specificati an b
the [Plant p 'ndent] or his designee in a with approved administrative procedures within 14 days of imp e 5.7.
1 5.7.2.2 The following programs shall be established, implemented, and maintained.
Radiation Protection Program Procedures for personnel radiation protection shall be prepar sistent with the requirements of 10 CFR 20 and shall be app ed, maintained, and adhered to for all operations 'olving person 1 radiation exposure.
Process Cont 1
Program (PCP)
The PCP shall cont 'he current formula
, sampling,
- analyses, tests, and determinat s to be made t nsure that processing and packaging of solid radio tive wast will be accomplished to ensure compliance with 10 20 0
CFR 61, and 10 CFR 71; state regulations; burial ground re rements; and other requirements governing the disposal of s id r ioactive waste.
Licensee initiated ch es to the PCP:
a ~
Shall be do mented and records of revi performed shall be retai d.
This documentation shall con n:
1.
sufficient information to support the chang
) and appropriate analyses or evaluations justifying e
change(s),
and 2.
a determination that the change(s) maintain the overal conformance of the solidified waste product to the existing requirements of Federal,
- State, or other applicable regulations.
(continued)
Ca
&0-18
Programs~and Manuals Qis Programs~and Manuals 54
~
~
Ca. S. L
~ <.vs.
Control Program (PCP)
(con Sha ec ive after re e tance by the review method of Specification 5.5. 1] an e
the [Plant Superintendent].
Offsite Dose Calculation Manual (ODCM) a.
The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the~adiological
~vironmental Pd'nitoring genogram; and b.
The ODCM shall also contain the~dioactivegffluent control s and gadi ol ogi cal gfivironmental dani toring gams
~~a~.+ ~
and descriptions of the
'nformation that should be included in the Annual Radiological Environmental Operating and Radioactive dfff df df t df ddfdd fff df fk++Sl d
d fff f.f 4.4. 2 Ca-4.3 Licensee initiated changes to the ODCH:
a ~
Shall be documented and records of reviews performed shall be retained.
-This documentation shall contain:
b.
1.
sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),
2.
a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20. 106, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; Shall become effective after review and acceptance by thean~'~ ~"'~
and the approval of the lant and (continued)
Programs~and Manuals 4.S Programs~and Manuals Offsite Dose Calculation Manual (ODCM)
(continued)
C ~
Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with 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
- 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.
Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from ose ortions of systems outside containment that could co ain highly dioactive fluids during a serious transient or ac 'dent to le s
as low as practicable.
The systems includ Recirculation
- Spray, Safety Injection, Chemical and Volume Co rol, gas strippe and Hydrogen Recombiner].
The pro m shall include the following:
a.
Preventive aintenance and period'isual inspection requirements, and b.
Integrated leak t t requi ments for each system at refueling cycle int val or less.
In Plant Radiation Monit ing This program provid controls to sure the capability to accurately determ' the airborne io concentration in vital areas under ac 'nt conditions.
This ogram shall include the following:
a.
Tr 'ng of personnel; b.
Procedures for monitoring; and c.
Provisions for maintenance of sampling and analysis equipment.
(continued)
&0-20
Programs~and Manuals
&+
c..S Programs~and Manuals Post Accident Sampling, P~~~
This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions.
The program shall include the following:
a.
Training of personnel; b.
Procedures for sampling and analysis; and c.
Provisions for maintenance of sampling and analysis equipment.
Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to member f
the public from radioactive effluents as low as reasonably a
ievable.
The program shall be contained in the ODCH all be imp ented by procedures, and shall include remedial ctions to be tak whenever the program limits are exceeded.
he program shall inc de the following elements:
a.
Limitatio on the functional capabi of radioactive liquid and eous monitoring ins mentation including surveillance t t's and setpoin etermination in accordance with the methodo in the CM; b.
Limitations on the co trations of radioactive material released in liquid flue s to unrestricted
- areas, conforming to 10 R 20, App dix 8, Table II, Column 2; c.
Honitorin
- sampling, and analyst of radioactive liquid and gaseous fluents in accordance wit 0 CFR 20. 106 and with the hodology and parameters in the CH; d.
imitations on the annual and quarterly dos or dose commitment to a member of the public from rad>
ctive materials in liquid effluents released from each it to unrestricted
- areas, conforming to 10 CFR 50, Appen I;
continued
Procedures,
- Programs, and Manuals 5.
- Programs, and Manuals Radioactive Effluent Controls Program (continued) e.
g, h.
J
~
Determination of cumulative and projected dose co ributions from radioactive effluents for the current calen ar quarter and current calendar year in accordance with t methodology and parameters in the ODCM at least every 31 ys; Limitations on the functional capability a
use of the liquid and gaseous effluent treatment sys ms to ensure that appropriate portions of these systems ar used to reduce leases of radioactivity when the pro 'ted doses in a pe ~iod of 31 days would exceed 2% of e guidelines for the annu 1 dose or dose commitment, conf rming to 10 CFR 50, Appen ix I; Limitati'qns on the dose rate res lting from radioactive material released in gaseous e
luents to areas beyond the site bound y conforming to t e dose associated with 10 CfR 20, A
endix B, Tabl II, Column I; Limitations on he annual and quarterly air doses resulting from noble gases eleas in gaseous effluents from each unit to areas bey d t e site boundary, conforming to 10 CFR 50, Appendix I Limitations on the ann al and quarterly doses to a member of the public from i dine-1, iodine-133, tritium, and all radionuclides i particu te form with half lives
> 8 days in gaseous eff uents relea ed from each unit to areas beyond the site bou ary, conformi to 10 CFR 50, Appendix I; and Limitation on the annual dose r dose commitment to any member of the public due to rele ses of radioactivity and to radiati from uranium fuel cycle sources, conforming to 40 CF 190.
5.7.2.8 Radiolog al Environmental Monitoring Prog m
This ogram is for monitoring the radiation nd radionuclides in the virons of the plant.
The program shall ovide rep esentative measurements of radioactivity in he highest p
ential exposure pathways and verification of t accuracy of e effluent monitoring program and modeling of en 'ronmental (c
tinued)
G STS 5.0-22 Rev.
0, 09/28 2
Procedures,
- Programs, and Manu s
5.7 5.7 Procedures,
- Programs, and Manuals 5.
2.8 Radiological Environmental Monitoring Program (contin ed) exposure pathways.
The program shall be contained 'he
- ODCM, shall conform to the guidance of 10 CFR 50, Append',
and shall include the following:
Monitoring, sampling,
- analysis, and repor ng of radiation and radionuclides in the environment in ccordance with the methodology and parameters in the ODCM; b.
C.
Land Use Census to ensure that cha ges in the use of areas a
and beyond the site boundary are identified and that mo 'fications to the monitoring pr gram are made if required by t results of this census; a
Partici tion in an Interlabor tory Comparison Program to ensure th t independent chec on the precision and accuracy of the mea urements of radi active materials in environment sample matri es are performed as part of the quality assur ce progra for environmental monitoring.
5.7.2.9 Component Cyclic or ansie t Limit This program provides co rois to track the FSAR, Section
[ ],
cyclic and transient oc ences to ensure that components are
'maintained within the esi limits.
5.7.2.10 Pre-Stressed Concre Contain nt Tendon Surveillance Program This program pro des controls f monitoring any tendon degradation in estressed concret containments, including effectiveness f its corrosion prot tion medium, to ensure containment ructural integrity.
Th program shall include baseline me surements prior to initial perations.
The Tendon Surveillan e Program, inspection frequen ies, and acceptance criteria hall be in accordance with [Reg latory Guide 1.35, Revisio 3, 1989].
The p ovisions of SR 3.0.2 and SR 3.0.3 are a
licable to the Ten n Surveillance Program inspection frequenc'es.
(co inued)
G STS 5.0-23 Rev.
0, 09/28/
Procedures,
- Programs, and Hanua
.7 5.7 rocedures,
- Programs, and Manuals 5.7.2 5.7.2.11 Pro rams and Manuals (continued)
Inservice Inspection Program T 's program provides controls for inservice inspe ion of ASHE Co Class 1, 2, and 3 components, including appl'cable supports.
The rogram shall include the following:
a.
P visions that inservice inspection of HE Code Class 1,
2, nd 3 components shall be performed
'n accordance with Sect n XI of the ASHE Boiler and Pre ure Vessel Code and appli ble Addenda as required by 10 FR 50.55a; b.
c, The prov ions of SR 3.0.2 are app icable to the frequencies for perfo ing inservice inspect' activities; Inspection o
each reactor coo ant pump flywheel per the recommendatio of Regulatio Position c.4.b of Regulatory Guide 1.14, Rev'sion 1, Aug t 1975; and 5.7.2.12 d.
Nothing in the AS Boile and Pressure Vessel Code shall be construed to supers de e requirements of any TS.
Inservice Testing Program This program provides c
trol for inservice testing of ASHE Code Class 1, 2, and 3 comp ents i luding applicable supports.
The program shall include the folio 'ng:
'a ~
b.
Provisions th inservice tes ing of ASHE Code Class 1, 2, and 3 pumps,
- valves, and snubb s shall be performed in accordance ith Section XI of th ASME Boiler and Pressure Vessel Co and applicable Addend as required by 10 CFR 5.55a; Testi frequencies specified in Sect on XI of the ASHE Boil and Pressure Vessel Code and ap icable Addenda as fol ows:
(co tinued)
0, 09/28 92
Procedures,
- Programs, and Han 5.7 5.
Procedures,
- Programs, and Hanuals 5.7.2.
2 Inservice Testing Program (continued)
ASHE Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testing act'vities Required Fr quencies for perfor ing inservice test'ivities ekly Ho thly qua erly or every 3
m nths Semian ally or every months Every 9 m pths Yearly or 0 nually Biennially o every 2 years At lea t once per 7 days At le st once per 31 days At east once per 92 days t least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days 5.7.2.13 c.
The provisions o
SR 3.
.2 are applicable to the above required Frequency' f r performing inset vice testing activities; d.
The provisions of R
0.3 are applicable to inservice testing activiti
- and e.
Nothing in the ASHE Boiler and Pressure Vessel Code shall be construed to upersede the quirements of any TS.
Tube Surveilla e Program Each SG shal be demonstrated OPERABL by performance of an inservice i spection program.
The prog m shall include the following:
'a ~
SG ube sample size selection, sample
'ze expansion, and i spection result classification criteri Sample selection nd testing shall be in accordance with [
gulatory Guide 1.83, Revision [ ], date].
(con inued)
G STS 5.0-25 Rev.
0, 09/28/
Procedures,
- Programs, and Ma als 5.7 5.
Procedures,
- Programs, and Manuals 5.7.2.
Steam Generator (SG) Tube Surveillance Program (cont'nued) b.
The establishment of SG tube inspection frequ cy dependent upon inspection result classification.
Insp ction frequency shall be in accordance with [Regulatory Gui e 1.83, Revision [ ], date].
c.
SG tube plugging/repair limits.
These mits shall be [40]%
f the nominal tube wall thickness con istent with
[
gulatory Guide 1.83, Revision
[ ], date].
d.
Spec ic definitions and limits fo SG tube inservice inspe ion acceptance criteria co sistent with [Regulatory Guide 1 83, Revision
[ ], date].
The content an frequency of writt reports shall be in accordance with ecification 5.9 The provisions of 3.0.2 are pplicable to SG Tube Surveillance Program inspection f quencie except those established by Category C-3 inspectio resu s.
[Key elements to be disc ed and provided.]
5.7.2. 14 Secondary Water Chemist y This program provide controls for monitoring secondary water chemistry to inhibi SG tube de adation and low pressure turbine disc stress corros'on cracking.
he program shall include:
a.
Identifica on of a sampling s
edule for the critical variables and control points fo these variables; b.
Identif cation of the procedures u
d to measure the values of th critical variables; c.
Ide tification of process sampling poi s, which shall i
lude monitoring the discharge of the ondensate pumps for idence of condenser in leakage; d.
Procedures for the recording and management f data; Procedures defining corrective actions for all ff control point chemistry conditions; and (co inued)
W STS 5.0-26 Rev.
0, 09/28
Procedures,
- Programs, and Manuals 5
5.
Procedures,
- Programs, and Manuals 5.7.2.
Secondary Water Chemistry (continued) f.
A procedure identifying the authority responsibl for the interpretation of the data and the sequence and iming of administrative events, which is required to in'ate corrective action.
5.7.2.15 Vent lation Filter Testing Program (VFTP)
A prog m shall be established to implement e following required testing f Engineered Safety Feature (ESF) ilter ventilation systems a
the frequencies specified in [
gulatory Guide
],
and in acco dance with [Regulatory Guide
.52, Revision 2; ASME N510-19; and AG-1].
a.
Demonstrat fo} each of the ESF ystems that an inplace test of the high fficiency particu te air (HEPA) filters shows a penetratio and system byp s < [0.05]% when tested in accordance wit
[Regulatory uide 1.52, Revision 2, and ASME N510-1989] at t system f owrate specified below [i 10%].
ESF Ventila ion stem Flowrate b.
Demonstrate for ach of th ESF systems that an inplace test of the charcoa adsorber sho s
a penetration and system bypass
< [0.5
% when tested i
accordance with [Regulatory Guide 1.52, evision 2, and AS N510-1989] at the system flowrate s ecified below [i 10%
SF Ventilation System Flowrate (co inued)
0, 09/28/
Procedures,
- Programs, and Hanu s
5.7
.7 Procedures,
- Programs, and Hanuals 5.7.
. 15 Ventilation Filter Testing Program (VFTP)
(continued)
C.
Demonstrate for each of the ESF systems that a
boratory test of a sample of the charcoal
- adsorber, whe obtained as described in [Regulatory Guide 1.52, Revision 2], shows the methyl iodide penetration less than the val specified below when tested in accordance with [ASTH 3803-1989] at a
temperature of ~ [30'C] and greater than equal to the relative humidity specified below.
ESF Ventilation System Pen tration RH Reviewer,'s Not Allowable penetra on
= [100% - methyl iodide efficiency for c arcoal credited i staff safety evaluation/
(safety factor).
Safety factor
= [5] %or system with heaters.
= [7] fb syste s without heaters.
. d.
Demonstrate for eac o
the ESF systems that the pressure drop across the comb ed HEPA filters, the prefilters, and the charcoal adsorb s is less than the value specified below when tested 'c rdance with [Regulatory Guide 1.52, Revision 2, and A
E N510 1989] at the system flowrate specified below i 10%].
ESF Ve tilation System Delta P
Flowrate e.
Demons rate that the heaters for each f the ESF systems diss'te the value specified below [+
0%] when tested in cordance with [ASHE N510-1989].
ESF Ventilation System ttage (co tinued)
WOG S
S 5.0-28 Rev.
0, 09/28 2
Procedures,
- Programs, and Manual 5
5.7 Procedures, Programs,- and Manuals 5.7.2.1 Ventilation Filter Testing Program (VFTP)
(continued)
The provisions of SR 3.0.2 and SR 3.0.3 are applicabl to the VFTP test frequencies.
5.7.2. 16 E
losive Gas and Storage Tank Radioactivity Monit ring Program This rogram provides controls for potentially plosive gas mixtur s contained in the
[Waste Gas Holdup Sy tern], [the quantity of radi ctivity contained in unprotected ou oor liquid storage tanks].
he gaseous radioactivity quantiti shall be determined following he methodology in [Branch Techn'cal Position (BTP)
ETSB 11;5, "Post ated Radioactive Release du to Waste Gas System Leak or Failure"].
The liquid radwaste quan ties shall be determined in accordance ith [Standard Review Pl n, Section 15.7.3, "Postulated Rad active Release due t Tank Failures"].
The program shall elude:
a.
The limits for c ncentrati s of hydrogen and oxygen in the
[Waste Gas Holdup ystem]
nd a surveillance program to ensure the limits e ma'ntained.
Such limits shall be appropri ate to the s
s m'
design criteria (i.e.,
whether or not the system is signed to withstand a hydrogen explosion);
b.
A surveillance pr ram to nsure that the quantity of radioactivity co ained in each gas storage tank and fed into the offgas treatment sy tern] is less than the amount that would re lt in a whole dy exposure of Z 0.5 rem to any individu in an unrestrict d area, in the event of [an uncontrolle release of the tank 'ontents];
and C.
A survei ance program to ensure th t the quantity of radioac ivity contained in all outdo liquid radwaste tanks that e not surrounded by liners, di s, or walls, capable of h ding the-tanks'ontents and that do not have tank ove flows and surrounding area drains co ected to the
[
quid Radwaste Treatment System] is les than the amount at would result in concentrations less th n the limits of 0
CFR 20, Appendix B, Table II, Column 2, a
the nearest potable water supply and the nearest surface ter supply in an unrestricted
- area, in the event of an uncont oiled release of the tanks'ontents.
(co inued)
0, 09/28/
2
Procedures,
- Programs, and Manual 5
5.7
- ocedures, Programs, and Hanuals 5.7.2.16 5.7.2.17 Explosive Gas and Storage Tank Radioactivity Monitoring rogram (continued)
T provisions of SR 3.0.2 and SR 3.0.3 are applica e to the Exp osive Gas and Storage Tank Radioactivity Monit ring Program surv illance frequencies.
Diesel uel Oil Testing Program A diesel el oil testing program to implem t required testing of both new fu 1 oil and stored fuel oil shal be established.
The program shal include sampling and testi requirements, and acceptance crs eria, all in accordance ith applicable ASTM Standards.
The urpose of the progra is to establish the following:
a.
Acceptability f new fuel oi for use prior to addition to storage tanks b
determinin that the fuel oil has:
l.
an API gravit or a
absolute specific gravity within
- limits, 2.
a flash point a d inematic viscosity within limits for ASTH 2D fuel 1,
a d
3.
a clear an bright ap arance with proper color; b.
Other propert'es for ASTH 2D el oil are within limits within 30 d s following sample g and addition to storage tanks; and c.
Total p
ticulate concentration of he fuel oil is g 10 mg/l when t sted every 31 days in accorda ce with ASTH D-2276, Meth A-2 or A-3.
5.7.2.18 Fire Pr tection Program This rogram provides controls to ensure that ap ropriate fire pr ection measures are maintained to protect the lant from fire a
to ensure the capability to achieve and mainta safe shutdown
'n the event of a fire is maintained.
0, 09/28/92
.8 5.0 ADMINISTRATIVE CONTROLS 1'.8 S
ety Function Determination Program (SFDP) 5.8.1 This program ensures loss of safety function is dete ed and ppropriate actions taken.
Upon failure to meet tw or more LCOs a
the same time, an evaluation shall be made to d ermine if loss of afety function exists.
Additionally, other a propriate acti ns may be taken as a result of. the support ystem inope ability and corresponding exception to e ering supported system ondition and Required Actions.
This rogram implements the req
'rements of LCO 3.0.6.
5.8.2 The SFDP sha 1 contain the following:
a.
Provision for cross train check to ensure a loss of the capability o perform the safet function assumed in the accident ana sis does not go ndetected; b.
Provisions for nsuring the plant is maintained in a safe condition if a 1
ss of fu tion condition exists; c.
Provisions to ensu th t an inoperable supported system's Completion Time is n
inappropriately extended as a result of multiple support tern inoperabilities; and d.
Other appropriate imita ions and remedial or compensatory actions.
5.8.3 A loss single cannot safety and:
of safety f nction exists w en, assuming no concurrent
- failure, safety function a
umed in the accident analysis be perfo med.
For the purpose of this program, a loss of functi may exist when a suppo t system is inoperable, a.
A re ired system redundant to the sy tern(s) supported by the noperable"support system is also operable (Case A);
or b.
required system redundant to the system(
in turn supported by the inoperable supported syste is also inoperable (Case B); or (c
tinued)
0, 09/28 92
SFDP 5.8 5.
SFDP 5.8.3 continued) c.
A required system redundant to the support system(
for the supported systems (a) and (b) above is also inope able (Case C).
Ge ric Example:
Train A Sy tern i Syste ii System i'i System iv
~(Support System Inoperable)
Train Sy em i Mase C
ystem ii System iii
~Case A
System iv
~Case 8
5.8.4 The SFDP identifies w ere a los of safety function exists.
If a loss of safety functio is det rmined to exist by this program, the appropriate Conditio s a
Required Actions of the LCO in which the loss of safety ction exists are required to be entered.
G STS 5.0-32 Rev.
0, 09/28 2
O.O ADMINISTRATIVE CONTROLS 4.M Reporting Requirements Reporting Requirements C.4p The following reports shall be submitted in accordance with 10 CFR 50.4.
St t
artup Repor A summary report of plant startup and power escalation testing shall be submitted following:
Receipt of an Operating License; b.
mendment to the license involving a planned incr se in p
er level; c.
Insta ation of fuel that has a different d ign or has been manufac red by a different fuel supplier and d.
Hodificatio that may have significa y altered the
- nuclear, ther 1, or hydraulic perf mance of the unit.
The initial Startup Rep t shall addr ss each of the startup tests identified in FSAR, Chapt
[14],
d shall include a description of the measured values of t op ating conditions or characteristics obtained duri he test program and a comparison of these values with design e 'ctions and specifications.
Any corrective actions that wer requ ed to obtain satisfactory operation shall also be scribed.
ny additional specific details required in li nse conditions ased on other commitments shall be included in is report.
Subse ent Startup Reports shall address star p tests that are neces ry to demonstrate the acceptability of anges and modifications.
Startup Repor s shall be submitted within 90 day following completion f the Startup Test Program; 90 days fo owing resumpti or commencement of commercial power opera
'on; or 9 mont following initial criticality, whichever is e liest. If the artup Report does not cover all three events (i.e., initial cr'cality, completion of Startup Test Program, and resump ion or mmencement of commercial operation),
supplementary reports all e submitted at least every 3 months until all three events hav been completed.
(continued)
~
~
Reporting Requirements Reporting Requirements 4.w 5.9.1.2 Routine Re orts (continued) al Reports
-NOTE A single submitta 'e made for a multi it station.
The submittal should combi ections co o all, units at the station.
Annual Report ering the activities of t it as described below f e previous calendar year shall be sub 'd by H
1 of each year.
[The initial report shall be s
'ed by arch 31 of the year following initial criticality.]
Reports required on an annual basis include:
4 "'~ ~
Occupational Radiation Exposure Report A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures
> 100 mrem/yr and their associated man rem exposure according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance describe maintenance],
waste processing, and refueling).
his tabulation supplements the requirements of 10 CFR The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements.
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 received from external sources should be assigned to specific major work functions~ed WW ~oeh akahSL. MS~4~>~ ~ ~ ++~
cA ~~~~
~
(continued)
- 4. 0-34
t.
4.4 Reporting Requirements Reporting Requirements C a4 Annual Radiological Environmental Operating Report NOTE-A single submitta ade for a mu submittal should combine se to station.
station.
The all units at the The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.
The report shall include summaries, interpretations, and analyses of trengs of the results of thegtadiological ghvironmental P6nitoring 'or the reporting period.
The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
The Annual Radiological Environmental Operating 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 figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements ~n the format 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 in a supplementary report as soon as possible.
(continued) k.0-35
4o Le Reporting Requirements Reporting Requirements~9 Cg ~ C~
C
~ 4.3 Radioactive Effluent Release Report NOTE A single submit a
be made for a multiple unit n.
The submittal should comb n
'ons commo units at the station; however, for unitsw'e radwaste
- systems, the submittal shall spe releases of ra 've material from each unit Q.4e (
The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a.
The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.l.
Monthly Operating Reports Routine reports of operating statistics and shutdown experience+
including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves~shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
.6 CORE OPERATING LIMITS REPORT (COLR) a 0 Core operating limits shall be established prior to re cycle, or prior to any remaining portion a reload
- cycle, a
all be documented in the COLR the following:
[
The individual specificatio at address core operating limits must be referenc ere.
b.
The analytic ethods used to determine the c oper ating limits s be those previously reviewed and appro th
, specifically those described in the following ocuments:
(continued)
0, 09/28/92
Reporting Requirements 5.
.9 Reporting Requirements 5.9.
.6 CORE OPERATING LIHITS REPORT (COLR)
(continued)
Identify the Topical Report(s) by number, title, te, and NRC staff approval
- document, or identify the sta f Safety Evaluation Report for a plant specific methodol gy by NRC letter and date.
C.
The core operating limits shall be determi d such that all applicable limits (e.g.,
fuel thermal mec nical limits, ore thermal hydraulic limits, Emergency Core Cooling S
tems (ECCS) limits, nuclear limits ch as SDH, transient an ysis limits, and accident analysi limits) of the safety anal sis are met.
d.
The COL
, including any midcycle visions or supplements, shall be rovided upon issuance or each reload cycle to the NRC.
5.9.1.7, Reactor Coolant S
tern (RCS)
PRES RE AND TEHPERATURE LIHITS REPORT (PTLR)
RCS pressure and tempe ature imits, including heatup and cooldown rates, criticality, and ydr static and leak test limits, shall be established and documente n the PTLR.
[The individual Specifications that addre the reactor vessel pressure and temperature limits and e
atup and cooldown rates may be referenced.]
The anal ical ethods used to determine the pressure and temperat re limit including the heatup and cooldown rates shall be thos previously eviewed and approved by the NRC in [Topical Report s),
number, ti e, date, and NRC staff approval
- document, or staf safety evaluatio report for a plant specific methodology by C letter and date].
The reactor vessel pressure and temperatur limits,'ncluding tho for heatup and cooldown
- rates, shall e determined so that all plicable limits (e.g.,
heatup limi s, cooldown limits, and inse ice leak and hydrostatic testing li its) of the analysis are met.
he PTLR, including revision or supplements
- thereto, shall be ovided upon issuance for eac reactor vessel fluency period.
(conti ed)
Rev.
0, 09/28/
H h
Reporting Requirements 5.
5.
Reporting Requirements (continued) 5.9.2 S ecial Re orts Special Reports may be required covering inspection, t st, and maintenance activities.
These special reports are d
ermined on n individual basis for each unit and their prepara on and s bmittal are designated in the Technical Specific tions.
Spec'al Reports shall be submitted in accordanc with 10 CFR 50.4 withi the time period specified for each rep t.
The foll ing Special Reports shall be sub 'tted:
a ~
b.
C.
In the event an ECCS is actuated an injects water into the RCS in DE I, 2, or 3, a Special eport shall be prepared and submi ted within 90 days des ibing the circumstances of the actuat n and the total acc mulated actuation cycles to date.
The c rrent value of t usage factor for each affected safe injection no zle shall be provided in this Special Report henever it value exceeds 0.70.
If an individual erge y diesel generator (EDG) experiences four or mo valid failures in the last 25 demands, these fa ures and any nonvalid failures experienced by that in that time period shall be reported within 30 days.
Reports on EDG failures shall include the infor ation commended in Regulatory Guide 1.9, Revision 3, Reg atory Pos tion C.5, or existing Regulatory Guide 1. 108 re orting requi ment.
When a Spec'al Report is requi d by Condition B or G of LCO 3.3.[
], "Post Accident Mon'toring (PAM)
Instrume ation,"
a report shall submitted within the followi g 14 days from the time the action is required.
The repor shall outline the preplanned lternate method of moni oring, the cause of the inoperab ity, and the plans an schedule for restoring the instrume tation channels of t e Function to OPERABLE status.
(conti ued)
0, 09/28/92
Reporting Requiremen
.9 5
Reporting Requirements 5.9.2 S ecial Re orts (continued) d.
Any abnormal degradation of the containment str ture detected during the tests required by the Pre-ressed Concrete Containment Tendon Surveillance Pro am shall be reported to the NRC within 30 days.
The re rt shall include a description of the tendon condit', the condition of the concrete (especially at tendon anc orages),
the
'nspection procedures, the tolerances o
- cracking, and the c rrective action taken.
e.
Fol wing each inservice inspection f steam generator (SG)
- tubes, in accordance with the SG T e Surveillance
- Program, the nu er of tubes plugged and t es sleeved in each SG shall be eported to the NRC wit in 15 days.
The complet results of the S
tube inservice inspection shall be sub 'tted to the NR within 12 months following the completion of he inspectio The report shall include:
1.
Number and e tent o
tubes inspected, 2.
Location and p
nt of wall-thickness penetration for each indication f an imperfection, and 3.
Identificati of t bes plugged and tubes sleeved.
Results of SG t e inspect>
ns that fall into Category C-3 shall be repor ed to the NRC rior to resumption of plant operation.
is report shall rovide a description of investigati ns conducted to det rmine cause of the tube degradatio and corrective measu s taken to prevent recurren WOG STS 5.0-39 Rev.
0, 09/28/9
Record Retention 5.10 0
ADMINISTRATIVE CONTROLS 5.1 Record Retention 5.10.1 The following records shall be retained for at least 3 y ars:
a.
All License Event Reports required by 10 CFR 50.
b.
Records of changes made to the procedures req red by Specification 5.7.1.1; and Co cords of radioactive shipments.
5.10.2 The folio 'ng records shall be retained for at least 5 years:
a.
Records nd logs of unit operation overing time intervals at each p wer level; b.
Records and ogs of principal m intenance activities inspections,
- epair, and repl ement of principal items of equipment rela d to nuclear afety; c.
Records of survei ance a
ivities, inspections, and calibrations requi d by he Technical Specifications (TS)
[and the Fire Protec io Program];
d.
Records of sealed so c
and fission detector leak tests and results; and e.
Records of annua physical ventory of all sealed source material of rec rd.
5.10.3 The following re ords shall be retaine for the duration of the unit Operating icense:
a ~
Records and drawing changes reflectin unit design modif'tions made to systems and equi ent described in the FSAR b.
C.
R ords of new and irradiated fuel invento
, fuel
- ansfers, and assembly burnup histories; Records of radiation exposure for all individua s entering radiation control areas; (co inued)
0, 09/28/
Record Retention 5.10 5.
0 Record Retention 5.10.
(continued) d.
g, h.
k.
[m.
n.
Records of gaseous and liquid radioactive material r eased to the environs; Records of transient or operational cycles for t ose unit components identified in [FSAR, Section X];
ecords of reactor tests and experiments; Re rds of training and qualification for embers of the unit staff; Records of inservice inspections perf rmed pursuant to the TS; Records of uality assurance acti ities required by the Operational ality Assurance
(
) Hanual
[not listed in Specification
. 10. 1 and which are classified as permanent records by appl able regulat ons,
- codes, and standards];
Records of reviews erform d for changes made to procedures, equipment, or revie of ests and experiments pursuant to 10 CFR 50.59; Records of the revie a d audits required by Specification 5.5. 1 nd S
cification 5.5.2; Records of the s
vice lives of all hydraulic and mechanical snubbers requir d by [documen where snubber requirements relocated to]
including the da e at which the service life commences, a
associated instal tion and maintenance records; Records f secondary water sampling d water quality;]
Recor s of analyses required by the Rad logical Envi onmental Honitoring Program that wo d permit ev uation of the accuracy of the analysis t a later date
(
ese records should include procedures ef ctive at pecified times and gA records showing that t ese procedures were followed);
Records of reviews performed for changes made to e Offsite Dose Calculation Hanual and the Process Control Pr ram; (conti ued)
W G STS 5.0-41 Rev.
0, 09/28/92
Record Retention 5.
10 Record Retention 5.10.
(continued)
[p.
Records of pre-stressed concrete containment tendo surveillances;]
and Records of steam generator tube surveillances]
G STS 5.0-42 Rev.
0, 09/
92
4
+High Radiation Area@
-p~~
N 0 ADMINISTRATIVE CONTROLS Q,7.
45~ High Radiation AreaP-5.11.1 1 teQL(aa 14O\\ (%)
Pursuant to 10 CFR 20, pa agraph 20.8&(e~ in lieu of the requirements of 10 CFR 20.
, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is
> 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Wor k Permit (RWP).
Indi i u ls qpali ied in radiation protection procedures (e.g., Vechniciansi'r personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with, exposure rates g 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
'a ~
A radiation monitoring device that continuously indicates the radiation dose rate in the area.
+1.X b.
A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.
C.
An individual qualified in radiation protection procedures with a radiation'ose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified 4y-4he in the RWP.
5.11.2 5 o~"~
In addition to the requirements of Specification
- 5. 11. 1, areas with radiation levels z 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the ke s shall be maintained u de t e Qministrative control of e
i
~a~~on duty or supervision.
Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in (continued)
MO-43
~igh Radiation Are~
(, t High Radiation Are~
5.11.2 (continued) the immediate work areas and the maximum allowable stay times for individuals in those areas.
In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.
5.11.3 For individual high radiation areas with radiation levels of
~ 1000 mrem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously
- posted, and a flashing light shall be activated as a
warning device.
L) 0