ML20150A646
| ML20150A646 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/22/1978 |
| From: | Stewart W FLORIDA POWER CORP. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 3--17, 3-0-17, TAC-11484, NUDOCS 7809270212 | |
| Download: ML20150A646 (28) | |
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September 22, 1978 ~ ~ ~ ~
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f 3-0-17 Director
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Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Crystal River Unit No. 3 Docket No. 50-302 Operating License No. DPR-72
Dear Sir:
Enclosed are three (3) originals and forty-(40) copies of Technical Specifi-cation Change Request No. 33 requesting amendment to Appendices A and B of Operating License No. DPR-72.
As pa rt of this req ue s t, proposed replacement pages for Appendices A and B are enclosed.
Florida Power Corporation considers Change Request No. 33 to be a Class 11 amendment per 10 CFR 170.22 as it involves issues that are pro f orma and admin-istrative in nature and that have no safet-or environmental significance.
Enclosed is FPC's licensing fee, in the ami unt of one thousand two hundred dollars ($1,200) for Change Request No. 33.
Also included 4, one signed copy of Cert.ificate of Service for Technical Speci-fication Change Request No. 33 to the Chief Executive of Citrus County, Florida.
Please advise if further discussion is desired.
Very truly yours, A.9\\.
towiH W.P. Stewart Enclosures cc: Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission 101 Marietta Street, Suite 3100 -
Atlanta, Ga 30303 (2 copies _
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6 STATE OF FLORIDA COUNTY OF PINELLAS W.P. Stewart states that he is the Directo*r, Power Production, of Florida Power Corporation; that he is authorized on the part of said company to sign and file 'with the Nuclear Regulatory Commission the information attached hereto; and that all such j
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statements made and matters set forth therein are true and correct to the.best of his knowledge, information and belief.
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W.Pi Stewart Subscribed and sworn to before me, a Notary Public in end for the State and County above named, this 22nd day of September, 1978.
1978.
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Notary Publicj Notary Public, State of Florida at Large, My Commission Expires:
July 25, 1980
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l UNITED STATES. OF AME'.ICA l
NUCLEAR REGULATORY C0% MISSION IN Tile MATTER OF
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DOCKET.No. 50-302 FLORIDA POWER CORPORATION
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CERTIFICATE OF SERVICE W. P. Stewart deposes and says that the following has been served 'on the Chief Executive of Citrus County, Florida by deposit in the United States mail, addressed as follows:
Chairman, Board of County Commissioners of Citrus County Citrus County Courthouse Inverness, Florida 32650 An original copy of our September 22, 1978 submittal.
FLORIDA POWER 00RPORATION f
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C h t Obl W. P. Stewart Director, Powe r Production SWORN TO AND SUBSCRIBED BEFORE ME Tills 22nd DAY OF SEPTEMBER, 1978
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b Notary Public jf' Notary Public State of Florida at Large My Commission expires: July 25, 1980 (NOTARIAL SEAL) e a
RMB/1dd/Rd J
. Technical Specification Change Request No. 33 [Ap.nendices A & B]
Delete and insert, as indicated, the following pages in Appendices A and 8 of Operating License DPR-72:
Delete Insert Appendix A_
I I
XIII XIII 1-7 1-7 B2-2 B2-2 3/4 2-13 3/4 2-13 3/4 4-11 3/4 4-11 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 3/4 6-19 3/4 6-19 3/4 7-23 3/4 7-23 3/48-4 3/4 8-4 3/4 8-5 3/4 8-5 3/4 9-1 3/4 9-1 3/49-4 3/49-4 B3/4 9-2 B3/4 9-2 6-2 6-2 6-9 6-9 6-14 6-14 Appendix B 11 11 4-1 4-1 4-2 5-2 5-2 5-4 5-4 5-7 5-7 5-8 5-8 Procosed Change These changes will update the Offsite Organization, delete requirements that are no longer applicable, clarify some requirements, and make miscellaneous administrative corrections.
Reasons for Proposed Changes Appendix A a)
Page I:
The definition of SHIELD BUILLING INTEGRITY is in the B&W Standardized Technical Specification but is not applicable to Crystal 0+
River, Unit 3.
The definition was delet wf but its Index reference was q$
not.
d b) l Page XIII:
The correct title of Specification 3/4.9.7 is "CRANE TRAVEL-SPENT FUEL STORAGE P0OL BUILDING."
b Page 1-7:
The footnote to the APPLICABILITY of Specification 3/4.9.1 fg requires that "the reactor shall be maintained in MODE 6 when the q
reactor vessel head is unbolted or removed." This change will make 5 T
[
g the requirements consistent.
d Page B2-2:
Technical Specification Change Request No.12, dated November 8,1977, proposed to delete all references to operation with two reactor coolant pumps.
This reference was not deleted at that time and is being deleted now.
Page 3/4 2-13:
The plant reactor coolant flow instrumentation is in
/
,lbs/hr. Therefore this table is being changed to make the requirements y
clearer for plant to follow. Also, Technical Specification Change Request No.12, dated November 8,1977, proposed to delete all p
references to operation with two reactor coolant pumps.
This e Fg reference was not deleted at that time and is being deleted now.
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Page 3/4 4-11:
Crystal River - Unit 3 has two steam generators which did receive a preservice inspection.
This standardized table is being p,
changed to apply specifically to CR3.
I $ g [ Page 3/4 4-31:
Q There is no second part to Specification 3/4.4.10 so it is extraneous to have a first part.
h[Page3/44-32:
The first periodic surveillance has been completed.
This exemption is no longer required.
p g o.f i
Valve MVV-27 is closed during Reactor Building 9 }Y Page 3/4 6-19: Isolation.not High Pressure Injection.j Page 3/4 6-19 also co
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4 changes proposed in Technical S ecif catio:
Change Request No. 24
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dated March 17, 1978.
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Page 3/4 7-23:
"Ventilation" was misspelled in Surveillance Requi rement 4.7.8.1.
cb AOY Page 3/4 8-4 2nd 5:
The first surveillance demonstration of Section 4.8.1.1.2.c.4 has been performed.
This exemption is no longer required.
Page 3/4 8-5 also contains a change proposed in Technical Specification Change Request No. 6 dated July 21, 1977.
Inserting the word "interval" in Specification 4.8.1.1.2.c.6 clarifies the intent of the requirement.
9 T)f9'l Page 3/4 9-1:
The applicablility of Mode 6 was revised in item c above.
This footnote is redundant.
m)
Page 3/4 9-4; Technical Specification Change Request No. 24 dated
& March 17, 1978 clarified the term "isolation valve" in Specification A,. 1 3/4.6.3 as "containment automatic isolation valve." The clarification U# '
was not made in Specification 3/4 9.4 at that time and is being made
- (l>!
C now for consistency.
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Page B 3/4 9-2:
See item b above.
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p Page 6-2:
The offsite organization supporting the operation of Crystal River-Unit.3 has recently undergone a reorganization.
This figure has been changed to reflect the new organization.
,1;/,d p)
Page 6-9:
It has been more than one year since fuel loading, p
therefore this requirement is no longer applicable.
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Page 6-14:
Footnote 1 has been deleted because there is only a single unit at the Crystal River Site and Footnote 2 has been rer.ambered.
A t,# ( Also the initial Annual Report has been submitted so its Tyf submittal requirement is extraneous.
Appendix B
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Page 11:
See item s below. Page 11 also contains changes proposed in
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Technical Specification Change Request No. 3, dated July 21, 1977.
A-k s) Pages 4-1 and 2:
The Special Surveillance, Research, or Study W-Activities have been completed and the report submitted.
The Specifications are no longer applicable.
[ Page 5-2:
See item o above.
))CPage5-4:
See item o above.
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, S C Page 5-7: The initial reports for the Nonra'diological and the t/ ' Radiological Volumes of the Annual Environmenta' Report and the Special Surveillance, Research, or Study Activi:les Report have been submitted therefore their. reporting requirements are no longer applicable.
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Page 5-8:
The initial Semiannual Effluent Report has been submitted O) therefore its reporting requirement is no lo'.ger applicable.
Page 5-8 also contains changes proposed in Technical Specification Change Request No. 3, dated July 21, 1977.
g Safety' Analysis of Proposed Changes The revisions proposed ir. this Change R'equest are strictly administrative in nature and none of the requirements applicable to the Safety Analysis are diminished'by the proposed changes.
No unreviewed safety question is involved.
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INDEX DEFINITIONS SECTIO _N PAGE 1.0 DEFINITIONS DEFINED TERMS.........................................<>
'v 1 THERMAL P0WER...........................................
1-1 RATED THERMAL P0WER.....................................
1-1 OP E R AT I O NA L M0 D E........................................
1-1 ACTI0N..................................................
1-1 OPERABLE - 0PERABILITY..................................
1-1 R E PO R TAB L E 0C C UR R E NC E...................................
1-2 CO NTA I NME NT I NTEGR I TY...................................
'l-2 CHANNEL CALIBRATION.....................................
1-2 C HA N N E L C H E C K...........................................
1-2 CHANNE L FUNC T IO NAL TE S T.................................
1-3 CORE ALTERATIUN.........................................
1-3 SHUTDOWN MARGIN.........................................
1-3 I DE NT I F I E D LE A KAGE......................................
1-3 UN I DE NT IF I E D L E AKAGE....................................
1-4 PRESSURE BOUNDARY LEAKAGE...............................
1-4 CO NTR O LLE D LE A KAGE......................................
1-4 QUADRANT POWER TILT...............,.....................
1-4 DO S E EQU I V ALE NT I -131...................................
1-4 E-AVERAGE DISINTEGRATIUN ENERGY.........................
1-4 STAGGERED TEST BASIS....................................
1-5 FREQUENCY N0TATION......................................
1-5 AX I AL POW E R I MB ALANC E...................................
1-5 REACTOR PROTECTION SYSTEM RESPONSE TIME.................
1-5 ENGINEERED SAFETY FEATURE RESPONSE TIME.................
1-6 PHYSICS TEST............................................
1-6 OPERATIO NAL MODE S (TABLE 1.1 )...........................
1-7 FREQUENCY NOTATION (TABLE 1.2)..........................
1-8 CRYSTAL RIVER - UNIT 3 I
INDEX BASES PAGE SECTION 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY...................
B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P0OL BUILDING......B 3/4 9-2 3/4.9.8 CO O LA N T C I R CU LA T I O N................................
B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM.....
B 3/4 9-2 3/4.9.10 WATER LEVEL - REACTOR VESSEL.......................
B 3/4 9-2 3/4.9.11 S T OR AGE P0 0 L.......................................
B 3/4 9-2 3/4.9.12 S TOR AGE PO O L VE NT I LAT IO N.........................'.'.
B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUT ION LIMITS........................
B 3/4 10-1 3/4.10.2 PHYSICS TESTS......................................
B 3/4 10-1 3/4.10.3 NO F L OW TE S T S......................................
B 3/4 10-1 3/4.10.4 SHUTDOWN MARGIN....................................
B 3/4 10-1 CRYSTAL RIVER - UNIT 3 XIII 1
i
J TABLE 1.1 OPERATIONAL MODES 2
REACTIVITY
% RATED AVERAGE-COOLANT MODE J"cit 50N, Keff THERMAL POWER
- TEMPERATURE
- 1. POWER OPERATION
> 9.99
> 5%
> 280*F
- 2. STARTUP
> 0.99
< 5%
> 280*F
- 3. HOT STANDBY
< 0.99 0
> 280*F
- 4. H0T SHUTDOWN
< 0.99 0
280*F > Tavg > 200 F
~
,_ 200*F
- 5. COLD SHUTDOWN
< 0.99 0
- 6. REFUELING **
< 0.95 0
< 140*F
- Excluding decay heat.
- Reactor vessel head unbolted or removed.
CRYSTAL RIVER - UNIT 3 1-7
SAFETY LXMlTS BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about 30 psi less than core outlet pressure, providing a more conservative margin to the safety limit.
The curves of Figure 2.1-2 are based on the raore restrictive of two thermal limits and account for the effects of potential fuel densification and potential fuel rod bow:
1.
The 1.30 DNBR limit produced by a nucle:r power peaking factor of F = 2.57 or the combination of the radial peak, axial peak and position of the axial peak that yields no less than 1.30 DNBR.
2.
The combination of radial and axial peak that causes central fuel melting at the hot spot.
The limit is 19.7 kw/ft.
Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.
The specified flow rates for curves 1 ard 2 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps and three pumps, respectively.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.11.
The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation.
These curves include the potential effects of fuel rod bow and fuel densification.
The DNBR as calculated by the BAW-2 DNB correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher.
Extrapolation of the correlation beyond its published quality range of 22%
is justified on the basis of experimental data.
CRYSTAL RIVER - UNIT 3 B 2-2
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bh TABLE 3.2-1 M
DNB MARGIN po LIMITS C
35 Four Reactor Three Reactor
-4 Coolant Pumps Coolant Pumps
')
Parameter Operating Operating Reactor Coolant Hot Leg Temperature, Tn*F
< 605.2
< 605.2(1)
Reactor Coolant Pressure, psig.(2) > 2062.7
> 2058.9(1)
Reactor Coolant Flow Rate, Ib/hr
> 137.89x106
> 103.00x106 o,
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C (1)Appplicable to the loop with 2 Reactor Coolant Pumps Operating.
(2) Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER per minute or a THERP.AL POWER step increase greater than 10% of RATED THERMAL POWER.
TABLE 4.4-1 O
MINIMUM NUMBER OF STEAM GENERATORS TO BE "l
INSPECTED DURING INSERVICE INSPECTION 3'
ai
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Preservice Insgection YES w
No. of Steam Generators per Unit Two First Inservice Inspection One Second & Subsequent Inservice Inspections Onel w1 a
cL Table Notation:
- 1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are perfonning in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY: All MODES ACTION:
a.
With the structural integrity of any ASME Code Class 1 compo-nent(s) not confoming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the mini-mum temperature required by NDT considerations, b.
With the structural integrity of any ASME Code Class 2 compo-nent(s) not confoming to the above requirements restore the structural integrity of the affected component (sf to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200'F.
c.
With the structural integrity of any ASME Code Class-3 compo-nent(s) not confoming to the above requirements, restore the structural integrity of the component (s) to within its limit or isolate the affected component (s) from service.
d.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5:
a.
The reactor coolant pump flywheels shall be inspected per the recommendations of Regulatory Position C.4.b. of Regulatory Guide 1.14, Revision 1, August 1975.
CRYSTAL RIVER - UNIT 3 3/4 4-31 l
e REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.
Each internals vent valve shall be demonstrated-0PERABLE at least once per 18 months during shutdown, by:
-l 1.
Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation 2.
Verifying the valve is not stuck in an open position, and 3.
Verifying through manual actuation that the valve is fully open with a force of < 425 lbs (applied vertically upward).
1 CRYSTAL RIVER - UNIT 3 3/4 4-32 Amendment No. 14
i 1
l TABLE 3.6-1 (Continueu)
CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds) 9.
-(Continued) l MUV-163 check #
Open during HPI and NA MUV-25 #
iso dur. nor operat Mn 60 MUV-164 check #
NA NJV-26 #
60 MUV-160 check #
open during HPI and NA MUV-23 #
iso. dur. nor, operation 60 MVV-161 check #
open during HPI and NA MUV-24 #
iso. dur. nor. operation 60 MVV-27 #
open dur. nor. operation 60 and closed during RB Isolation l
10.
SWV-39 #
iso. NSCCC from AHF-1C 60 SWV-45 #
60 SWV-35 #
iso. NSCCC from AHF-1A 60 SWV-41 #
60 SWV-37 #
iso. NSCCC from AHF-1B 60 SWV-43 #
60 SWV-48 #
to isolate NSCCC from 60 SWV-47 #
MUHE-1A & IB and WDT-5 60 SWV-49 #
60 SWV-SC #
60 SWV-80 #
iso. NSCCC from RCP-1A 60 SWV-84 #
60 SWV-82 #
iso. NSCCC from RCP-1C 60 SWV-86 #
60 SWV-81 #
iso. NSCCC from RCP-1D 60 SWV-85 #
60 SWV-79 #
iso. NSCCC from RCP-1B 60 SWV-83 #
60 SWV-109#
NSCCC to DRRD-1 60 SWV-110#
60 CRYSTAL RIVER - UNIT 3 3/4 6-19 Amendment No. 11
E-5 PLANT SYSTEMS 3/4.7.8 AUXILIARY BUILDING VENTILATION EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8.1 The auxiliary building ventilation exhaust system shall be OPERABLE and shall consist of a minimum of two independent pairs of exhaust fans and four filter systems.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
With one pair of exhaust fans or one filter system inoperable, restore the inoperable pair of fans or system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in COLD SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEll. LANCE REQUIREMENTS 4.7.8.1 Each auxiliary building ventilation exhaust system shall be demonstrated OPERABLE:
a.
At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes.
b.
At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:
1.
Verifying that with the system operating at a flow rate of 156,680 cfm + 10% and exhausting through the HEPA filters and charcoal adsiirbers, the total bypass flow of the system to the facility vent, including leakage through the system diverting valves, is < 1% when the system is tested by admitting cold 00P at the system intake.
- The air flow distribution test Section 8 of ANSI N510-1975 may be performed downstream of the HEPA filters.
i CRYSTAL RIVER - UNIT 3 3/4 7-23
ELECTRICAL POWER SYSTEMS SVRVEILLANCE REQUIREMENTS (CONTINVED) 4.
At least once per 18 months, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual emergency loads for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when the battery is subjected to a battery service test.
5.
At least once per 60 months, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.
This performance discharge test shall be perfonned subsequent to the satisfactory completion of the required battery service test.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
a.
At least once per 31 days on a STAGGERED TEST BASIS by:
1.
Verifying the fuel level in the day fuel tank, 2.
Verifying the fuel level in the fuel storage tank, 3.
Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day tank, l
4.
Verifying the diesel starts from ambient condition and accelerates to at least 900 rpm in f 10 seconds, 5.
Verifying the generator is synchronized, loaded to 11500 kw, and operates for ?_60 minutes, and 6.
Verifying the diesel generator is aligned to provide standt y power to the associated emergency busses.
b.
At least once each 92 days by verifying that a sample of diesel fuel from the fuel tank storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water and sediment.
c.
At least once per 18 months during shutdown by:
1.
Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufac'urer's recommendations for this class of standby service, CRYSTAL RIVER - UNIT 3 3/48-4 Amendment No. 6
~~
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (CONTINUED) 2.
Verifying the generator capability to reject a load of 2,515 kw without tripping, 3.
Simulating a loss of offsite power in conjunction with reactor building high pressure and reactor building high-high pressure test signals, and; a)
Verifying de-energization of the emergency busses and load shedding from the emergency busses, b)
Verifying that the 4160 v. emergency buss tie breakers open.
c)
Verifying the df isel starts from.tmtient condition on the auto-start signal, energized tne emergency busses with permanently connected loads, energizes the auto-connected emergency loads through the load sequencer and operates for > S minutes while its generator is loaded with the emergency loads.
4.
Verifying the diesel generator operates for > 60 minutes while loaded to 2,3000 kw, 5.
Verifying that the auto-connected loads to each diesel generator do not exceed the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of 3000 kw, and 6.
Verifying that the automatic load sequence timers are OPERABLE with each load sequence time interval within + 10%.
CRYSTAL RIVER - UNIT 3 3/4 8-5 Amsadment No. 6
3/4.9 REFUELING 0FERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concen-tration of all filled portions of the Reactor Coolant system and the ' refuel-ing canal shall be maintained unfform and sufficient to ensure that the more restrictive of the following reactivity condition is met:
a.
Either a Keff of 0.95 or less, which includes a 1% A k/k conserva-tive allowance for uncertainties, or b.
A boron concentration of 2,1925 ppm, which includes a 50 ppm con-servative allowance for uncertainties.
APPLICABILITY:
MODE 6.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity i
changes, and initiate and continue boration at > 2700 gpm of 2270 ppm boron solution or its equivalent until Keff is reduced to < 0.95 or the boron concentration is restored t) 2,1925 ppm, whichever is the more restrictive.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
a.
Removing or unbolting the reactor vessel head, and b.
Withdrawal of any safety or regulating rod in excess of 3 feet from its fully inserted position.
4.9.1.2 The boron c'ncentration of the reactor coolant system and the re-fueling canal shall be determined by chemical analysis at least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
CRYSTAL RIVER - UNIT 3 3/4 9-1
REFUELING OPERATIONS CONTAINMENT PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The contaimnent penetrations shall be in the following status:
a.
The equipment door closed and held in place by a minimum of four
- bolts, b.
A minimum of one door in each airlock closed, and c.
Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
1.
Closed by a containment automatic isolation valve, blind flange, or manual valve, or 2.
Be capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve.
APPLICABILITY:
During CORE ALTERATIONS or movement of irradiated fuel within the containment.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment penetrations snall be detennined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic containment purge and exhaust valve within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment by:
a.
Verifying the penetrations are in their isolated condition, or b.
Testing the containment purge and exhaust valves per the applicable portions of Specification 4.6.3.1.2.
CRYSTAL RIVER - UNIT 3 3/49-4
REFUELING OPERATXONS BASES 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY.
The OPERABILITY requirements of the hoist bridges used for movement of fuel assemblies ensures that:
- 1) fuel. handling bridges will be used for movement of control rods and fuel assemblies, 2) each hoist has sufficient load capacity to lift a fuel element, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.
This assumption is consistent with the activity release assumed in the accident analyses.
3/4.9.8 COOLANT CIRCULATION The requirement that at least one decay heat removal loop be in operstion ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.
3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system enstres that the containmant purge and exhaust penetrations will be automatically isolated upon detection of high rr aion levels within the containment.
The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.
3/4.9.10 WATER LEVEL - REACTOR VESSEL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.
The minimum water depth is consistent with the assumptions of the safety analysis.
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ADMINISTRATIVE CONTROLS CONSULTANTS 6.5.2.5 Consultants shall be utilized as determined by the NGRC Chairman to prcvide expert advice to the NGRC.
MEETING FREQUENCY 6.5.2.0 iht NGRC shall meet at least or.ce per six months.
QUORUM 6.5.2.7 A quorum of NGRC shall consist of the Chairman or his designated alternate and five additional NGRC members,-including alternates.
No more than a minority of the quorum shall have line responsibility for operation of the facility.
REVIEW 6.5.2.8 The NGRC shall review:
a.
The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments canpleted under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unre91ewed safety question.
b.
Proposed chrages to procedures, equipment or systems which involve an unrevie. sed safety question as defined in Section 50.59, 10 CFR.
c.
Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
d.
Proposed changes in Technical Specifications or this Operating
- License, e.
Violations of codes, regulations, orders, Technical Specifica-tions, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
CRYSTAL RIVER - UNIT 3 6-9 d
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ADMINISTRATIVE CONTROLS i
ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.
6.9.1.5 Reports required on an annual basis shall include:
a.
A tabulation of the number of station, utility, and other person- (
nel (including contractors) receiving exposures greater than 100 ssociated man-rem exposure according to work mrem /yr and their g/ e.g., reactor operations and surveillance, l
and job functions._
inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.
The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.
Small expo-sures totalling less than 20 percent of the individual total dose need not be accounted for.
In the aggregate, at least 80 percent of the total whole body dose received from 2xternal sources should be assigned to specific major work functions.
b.
A list of the reactor vessel material surveillance capsules in-stalled in the reactor at the end of the report period and a summary of any withdrawals or insertions of capsules during the report period.
In supplying this information, the ownership of each capsule shall be indicated and the irradiation location in the vessel of each capsule which was inserted during the report period shall be identified.
I 1/This tabulation supplements the requirements of 920.40T of 10 CFR Part 20.
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CRYSTAL RIVER - UNIT 3 6-14 l
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11 TABLE OF CONTENTS (Cont'd)
Page No.
3.2 Radiological Environmental Monitoring 3-5 3.2.1 Milk and Green Leafy Vegetables Census 3-26 3.2.2 Media Other Than External Radiation 3-27 3.2.3 External Radiation 3-27 3.2.4 Reporting Requirements 3-27 4.0 SPECIAL SURVEILLANCE, RESEARCH, OR STUDY ACTIVITIES 4-1 5.0 ADMINISTRATIVE CONTROLS 5.1 Organization 5-1 5.2 Res ponsibility 5-1 5.3 Review and Audit 5-3 5.4 Action to B'e Taken if Limiting Condition for Operation is Exceeded 5-3 5.5 Procedures 5-5 5.6 Plant Reporting Requirements 5-6 5.6.1 Routine Reports 5-6 5.6.2 Non-routine Reports 5-8 5.6.3 Changes 5-11 5.7 Records Retention 5-11 5.8 Special Requirements 5-11 1
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o 4-1 4.0 SPECIAL SURVEILLANCE, RESEARCH, OR STUDY ACTIVITIES Deleted.
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5-7 submitted to the Director of Inspection and Enforcement (with one copy to Director of Nuclear Reactor Regulation) as a separate Volume (#1) of the Annual Environmental.0perating Report within 90 days after January 1 of each year. The report shall include summaries, interpretations, and statistical evaluation of the results of the nonradiological environmental surveillance activities (Section 3.0) and the environmental monitoring programs required by limiting conditions for operation (Section 2.0) for the report period.
A comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports, t
and an assessment of the observed impacts of the plant operation on the environment shall be provided.
If harmful ef fects or evidence of irreversible damage are detected by the monitoring, the licensee shall provide an analysis of the problem and a proposed course of action to alleviate the problem.
- 2. Radiological Volume A report on the radiological environmental survaillance programs for the previous 12 months of operation,shall be submitted to the Director of Inspection and Enforcement (with copy to Director, Office of Nuclear Reactor Regulation) as a separate volume (#2) of the Annual Environmental Operating Report within 90 days af ter January 1 of each year.
The report shall include l
summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period including a comparison with preoperational studies; operational controls (as appropriate) and previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.
If harmful effects or evidence or irreversible damage are detected by the monitoring, the licensee shall provide an analysis of the problem and a proposed course of action to alleviate the problem.
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,.Gs 5-8 Results of all radiological environmental samples taken shall be summarized on an annual basis following the format of Table 5.6-1.
In the event that some results are not available within the 90 day period, the report shall be submitted, noting and explaining the reasons for the missing results.
The missing data shall be' submitted as soon as possible in a supplementary report.
B. Semiannual Operating Report - Radioactive Effluents A report on the radioactive discharges released from the site during the previous 6 months of operation shall be submitted to the Director of the Office of Inspection and Enforcement (with a copy to Director, Office of Nuclear Reactor Regulation) as part of the Semiannual Operating Report within 60 days af ter January 1 and July 1 of each year.
The report shall include a l
summary of the quantities of radioactive liquid and gaseous effluents and _ solid waste released from the plant as outlined in USNRC Regulatory Guide 1.21 (Rev. 1, 6/74), with data summarized on a quarterly basis following the format of Appendix B thereof.
The report shall include a summary of the meteorological conditions concurrent with the release of gaseous effluents during each quarter as outlined in USNRC Regulatory Guide 1.21 (Rev. 1, 6/741, with data summarized on a quarterly basis following the format of Appendix B thereof.
Calculated of f site dose to humans resulting from the release of effluents and their subsequent dispersion in the atmosphere (Regulatory Guide 1.109) shall be reported in accordance with Regulatory Guide 1.21 (Rev. 1, 6/74).
5.6.2 Non-Routine Reports A. Limiting Conditions for Operation Exceeded In the event that a limiting condition for operation is exceeded including any unplanned release of radioactive material from the site, or an event involving a significant adverse environmental impact occurs, a report will be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph to the Director of the Office of Inspection and Enforcement followed by a written report with a copy to the Director, Office Nuclear Reactor Regulation within 15 days.
The telegraph report will quantify the occurrence, its causes and, if aspects of the Crystal River Unit 3 operations are among the causes, planned remedial action to the extent possible.
The written report will fully describe the occurrence and will describe its causes and corrective action as fully as possible.
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