ML19249E922
| ML19249E922 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/06/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19249E920 | List: |
| References | |
| TAC-13056, NUDOCS 7910020705 | |
| Download: ML19249E922 (38) | |
Text
J
.f 4o UNITED STATES 8
g NUCLEAR REGULATORY CO'AMisSION g
wAssitaaTou. o. c. ro.es
\\, ***** /
FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSIMEEE CITY OF LEES 8URG CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW SMY'RNA SEACH CITY OF OCALA ORLANDO 'JTILITIES C0ttMISS10N AND CITY OF ORLANCO SEBRING UTILITIES COMMISSION SEMIt:0LE ELECTRIC COOPERATIVE, INC.
CITY OF TALLAHASSEE COCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICE."SE Amendment No. 24 License No. DPR-72 1.
The Nuclear Regulatory Ccmission (the Comission) has found that:
A.
The applications for amendment by Florida Power Corporation, et al (tne licensees) dated November 21cano 23,1977, July 21,1977, March 17,1978, and September 22, 1978, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; S.
The facility will operate in confamity with the applications, the provisions of the Act, and the rules and "egulations of the Cemission; C.
There is reasonable assurance (i) that the activities authorized by trf amendment can be conductud without endangering the health
..d.,afety of the public, and (ii) that such activities wfil be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the cxman defense and security cr to tne health anc safety of tne puolic; and E.
The issuance of this amendment is in acccedance sith 10 CFR ?ar:
51 of the Comission's regulations and all apclicable requirem nts have been satisfied.
4 910 0 'NO E 1973 251
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in tne attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. CPR-72 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 24, are hereby incorporated in the license.
Florida Power Corporation shall operate the facility in accordance with the technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMJiISSI0ti 74 Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 6,1979 O
G 1073 252
ATTACHMENT TO LICENSE AMENDMENT NO. 24 FACILITY OPERATING LICENSE N0. OPR-72 DOCKET NO. 50-302 Replace the following pages of the Appendices "A" and "B" Technical Specifications with the enclosed r ges. The revised pages are identified a
by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Appendix "A" Pages Appendix "B" Pages I
l-4 XIII 3-1 8 2-2 3-2 3/4 4-11 3-3 3/4 4-31 3/4 4-32 3/4 6-8 3/4 6-9 3/4 6-9a (added) 3/4 6-19 3/4 7-23 3/4 8-4 3/4 8-5 3/4 9-1 B 3/4 6-2 B 3/4 9-2 1073 253
INDEX DEFINITIONS PAGE_
SECTION l.0 DEFINITIONS 1 -1 DEFINED TERMS..............................................
1-1 THERMAL P0WER..............................................
1 -1 RATED THERMAL P0WER........................................
1-1 O P E RA T I O NAL M0 D E...........................................
1 -1 ACTION.....................................................
OPERABLE - OPERABILITY.....................................
1 -1
~~~
1 -2 REPORTABLE OCCURRENCE......................................
C O NTA I NME NT I NTE GR I TY......................................
1-2 CH ANN EL C AL I B RATI O N........................................
1 -2 1 -2 CHANNEL CHECK..............................................
C H ANNE L FU NCTI ONAL T EST....................................
1 -3 CORE ALTERATION............................................
1 -3 S H U TD OWN MAR G I N............................................
1 -3 I D ENT I F I ED L EA KAG E.........................................
1 -3 U N I D E NT I F I E D L EA KAG E.......................................
1 -4 P RESSURE B OU NDARY L EAKAGE..................................
1 -4 CO NT R O L L E D L EA KA GE.........................................
1 -4 14 Q U AD RANT POW ER T I L T........................................
1 -4 DO S E EQU I V A L ENT I -131......................................
E-AV ERAGE DI S I NT EGRATIO N E NER GY............................
1 -4 STAGGERED TEST BASIS.......................................
1 -5 1 -5 FREQUENCY N0TATION.........................................
AXIAL POWER IMBALANCE......................................
1 -5 REAC TOR PROTECTION SYSTEM RESPONSE TIME..................
-5 ENGINEERED SAFETY FEATURE RESPONSE TIME....................
1 -6 PHYSICS TESTS.............................................
1 -6 OPE RAT IONAL MOD E S ( TAB L E 1.1 )..............................
1 -7 F REQUENCY NOTATION ( TABL E 1. 2).............................
1 -8 CRYSTAL RIVER - UNIT 3 I
AmendmentNo.2Li 10/3 254
INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Reactor Core..............................................
2-1 Reactor Coolant System Pressure...........................
2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Setpoints......................
2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS Reactor Core..............................................
B 2-1 Reactor Coolant System Pressure...........................
B 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Recctor Protection Sys tem Setpoints.......................
B 2-4 CRYSTAL RIVER - UNIT 3 II 107,3 2St o
INDEX BASES SECTION PAGE 3/4.9.6 FUEL H ANDL ING BRIDGE OP ERAB ILITY.................. B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING... B 3/4 9-2 3/4.9.8 COOLANT CIRCULATION................................
B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM..... B 3/4 9-2 3/4.9.10 W AT ER L EVEL - REACTOP. V ESS EL....................... B 3/4 9-2 3/4.9.11 STORAGE P00L.......................................
B 3/4 9-2 3/4.9.12 STORAGE POOL VENTILATION...........................
B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POW ER D ISTRIBUTION L IMITS........................ B 3/4 10-1 3/4.10.2 PHYSICS TESTS......................................
B 3/410-1 3/4.10.3 NO FLOW TESTS......................................
B 3/4 10-1 3/4.10.4 S HUTD CW N MA RG I N.................................... B 3/4 10-1 CRYSTAL RIVER - UNIT 3 XIII AmendmentNo.2ff 1073 2566
INDEX DESIGN FEATURES PAGE SECTION 5.1 SITE Exclusion Area............................................
5-1 5-1 Low Population Zone.......................................
5.2 CONTAINMENT Configt iation.............................................
5-1 5-4 Der.ign Pressure and Temperature...........................
5.3 REACTOR CORE Fuel Assemblies...........................................
5-4 Control Rods..............................................
5-4 5.4 REACTOR COOLANT SYSTEM 5-5 Design Pressure and Temperature...........................
Volume....................................................
5-5 5.5 METEORLOGICAL TOWER LOCATION..............................
5-5 5.6 FUEL STORAGE 5-5 C r i t i c a l i ty...............................................
5-5 Drainage..................................................
5-6 Capacity................................
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........
5-6 1073 257 CRYSTAL RIVER - UNIT 3 XIV
2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding temperatures because of t.5c onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWF.R and Reactor Coolant Temper-ature and Pressure have been related to DN8 through the BAW-2 DNB correla-tion. The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially unifonn and non-unifom heat flux distributions. The local,0NB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.
This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 139.86 x 106 lbs/hr, which is 106.5% of the design flow rate for four operating reactor coolant pumps.
This curve is based on the following nuclear power peaking factors wit's potential fuel densification effects:
F = 2.57; F
= 1.71;
= 1.50 g
The design limit power peaking factors are the most restrictive calcu-lated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis.
i073 258 CRYSTAL RIVER - UNIT 3 B 2-1 Amendment No. 16, 19
~
e
SAFETY LI? LITS BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the rn.ctor 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 more restrictive of two thermal limits and account for the effects of potential fuel densifica-tion and potential fuel rod bow:
1.
The 1.30 DNBR limit produced by a nuclear power peaking factor of FN = 2.57 or the canbination of the radial peak, g
axial peak and position of t:'e axial peak that yields no less than a 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/f t.
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 and 2 of Figure 2.1-2 cor-respond 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 coola t pump-maximum themal power combinations shown in BASES Figure 2.1 1he curves of BASES Figure 2.1 represent the conditions at which a minir;m 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 minimtzn D?BR, 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 8 2-2 Amendment No 76, 17 2 @
1073 259
TABLE 4.4-1 c3 MINIMUM NUMBER OF STEAM GENERATORS TO BE p
INSPECTED DURING INSERVICE INSPECTION
,o 2
9 C
5 H
Preservice Inspection Yes No. of Steam Generators per Unit Two First Inservice inspection One R.
e S'econd & Subsequent Inservice Inspections One "s
Table Notation:
(
l.
The inservice inspection may be limited to one steam generator on a g[
rotating schedule encompassing 6% of the tubes if the results of the g
first or previous inspections indicate that both steam generators are g
performing in a like manner. Note that under some circumstances, the operating conditions in one steam generator may be found to be more severe than those in the other steam generator. Under such circumstances o
c-.;
tre sample sequence shall be modified to inspect the most severe se n,
conditions,
(,a 45 ru C7s CD
TABLE 4.4-2 m
i MMp STEAM GENERATOR TUBE INSPECTION m2 IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION rn" Sample Size Result Actu)n Required flesult Action Required Result Action flequired A minimum of C-1 None N/A N/A N/A N/A g
S Tubes per
~4 S. G.
w C-2 Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None 2S tuto M this S. G.
C-2 and inspect additional C-2 Plug defective tubes l
4S tubes in this S. G.
Perform action for C-3 C-3 result of first sample Perform action for f
w C-3 C-3 result of first N/A N/A sample j
w A
C-3 Inspect all tubes in All other i
this S. G., plug de-S. G.s are None N/A N/A N
fective tubes and C-1 inspect 2S tubes in me S. G.s Perform action for N/A 14/A each other S. G.
C_2 but no C-2 result of second additional sample Prompt notification S. G. are to NRC pursuant C-3 i
to specification Additional Inspect all tubes in 6.9.1 S. G. is C-3 each S. G. arvi plug defective tubes.
Prompt notification N/A N/A to NRC pursuant to specification 6 9.1 N
U Where N is the number of steam generators in the unit, and n is the number of steam generators inspected S-3-%
pg n
during an impection
.y
REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION The structural integrity of ASME Code Class 1, 2 an" 3 comporents 3.4.10 shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY _: All MODES.
ACTION:
With the structural integr'ty of any ASME Code Class I component t;
/
a.
not confonning to the above requirements, restore the structurai 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 minimum temperature required by NDT considerations.
With the structural integrity of any ASME Code Class 2 component (s) b.
not conforming 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 above 200*F.
With the structural integrity of any ASME Code Class 3 component (s) c.
not conforming 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 REOUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5:
The reactor coolant pump flywheels shall be inspected per the a.
recommendations of Regulatory Position C.4.b. of Regulatory Guide 1.14, Revision 1, August 1975.
CRYSTAL RIVER - UNIT 3 3/4 4-31 Amendment No. 24 10/3 262
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.
Each internals vent valve shall be demonstrated OPERABLE at least once per 18 months during shutdown, by:
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).
1073 263 CRYSTAL RIVER - UNIT 3 3/4 4-34 Amendment No JJ,24'
^
CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 130 F.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the containment average air temperature > 130 F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 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 /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS The primary containment average air temperature shall be the 4.6.1.5 arithmetical average of the temperatures at the following locations and shall be determined at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:
Location Column RB-320, elevation 100' a.
b.
Colume RB-320, elevation 125' Outside secondary shield wall, elevation 180' c.
d.
Crane access platform, elevation 235' CRYSTAL RIVER - UNIT 3 3/4 6-7 10/3-2.
4 o
CONTAINMENT SYSTEMS CONTMMMENT STRUCTURAL INTEGRITY LIMITING CONDITIONS FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consister>t with the acceptance criteria in Specification 4.6.1.6.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits' within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REOUIREMENTS 4.6.1.6.1 Containment Tendons The containment tendons' strt:ctural integrity shall be demonstrated at the end of one, three and five years following the initial containment structural integrity test and at five year interv :s thereafter. The tendons' structural integrity shall be demonstratel by:
Determining tha' ' representative sample
- of at least 21 tendons a.
(6 dew, 5 vert u
, and 10 hoop) each have a lift off force of between 1,249,000 (minimum) and 1,721,000 (maximum) pounds. This test shall include an unloading cycle in which each of these tendons is deter.sioned to determine if any wires or strands are broken or damaged.
If the lift off force of any one tendon in the total sample population is out of the predicted bounds (less than minimum or greater than maximum), an adjacent tendon on each side of the defective tendon shall also be checked for lift off force.
If both of these tendons are found acceptable, the surveillance program may proceed considering the single deficiency as unique and acceptable. More than one defective tendon out of the original sample population is evidence of abnormal degradation of the containment structure. Unless there is evidence of abnomal degradation of the containment structure during the first three tests of the tendons, the number of tendons checked for lift off force during subscquent tests may be reduced to a representative sample of at least 9 tendons (3 dome, 3 vertical and 3 hoop).
- For each inspection, the tendons shall be selected on a randon but representative basis so that the sample group will change somewhat for each inspection; however, to develop a history of tendon performance and to correlate the observed data, one tendon from each group (dome, vertical, and hoop) may be kept unchanged after the initial selection.
CRYSTAL RIVER - UNIT 3 3/4 5-8 Amendment No. 2fi 1073
,n,5 2
CONTAINMENT SYSTEMS SURVEILLANCE RFQUIREMENTS (Continued) b.
- Removing one wire or strand from each of a dome, vertical and l
hoop tendon checked for lift off force and determining that over the entire length of the removed wire or strand that:
l 1.
The tendon wires or strands are free of corrosion, cracks j
and damage, 2.
There are no changes in the presence or physical appearance I
of the sheathing filler grease, and 3.
A minimum tensile strength value of 240,000 psi (guaranteed l ultimate strength of the tendon material) for at Mast three wire or strand samples (one from each end and one at mid-length) cut from each removed wire or strand. Failure of any l one of the wire or strand samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment structure.
4.6.1.6.2 End Anchorages and Adjacent Cancrete Surfaces The structural integrity of the end anchorages of all tendons inspected pursuant to Speci-fication 4.6.1.6.1 and the adjacent concrete surfaces shall be demonstrated b/ detennining through inspection that no apparent changes have occurred in the visual appearance of the end anchorage or the concrete crack patterns adjacent to the end anchorages.
Inspections of the concrete shall be performed during the Type, A containment leakage rate tests (Specifica-tion 4.6.1.2) while the containment is at its maximum test pressure.
4.6.1.6.3 Containment Surfaces The structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate, shall be detennined during the shutdcwn for each Type A containment leakage rate test (Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be perfomed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnonnal degradation.
4.5.1.6.4 Containment Dome The containment dome's structural integrity shall be demonstrated at tne end of 1 year,18 months, 2 years, 3 years, 40
+10 months (coincident with the first periodic integrated containment leak rate test), and 5 years following the initial containment structural integrity test. The dome's structural integrity shall be demonstrated by:
1073 266 CRYSL. RIVER - UNIT 3 3/4 6-9 Amendment No. 24
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
Measuring the elevation difference of 7 dome survey points (1 a.
at the apex; 3 at a radius of =29 feet at azimuths.
, 215 and 334', and 3 at a radius of =49 feet at azimuths 90, 215" and 334*) and 3 benchmarks (on Ring Girder at azimuths 90, 215 and 334 ) along with respective azimuths. These elevation differences shall be compared to the elevatico differences established by the Baseline Survey.
If the containment is in a normal operation / shutdown mode, the acceptable change in change in elevation differences will be based on consideration of expected movement and survey accuracy coupled with an acceptable strain level for the radial reinforcement. Changes of a greater magnitude shall require an engineering evaluation.
If the containment is in a pressurized mode for a periodic containment integrated leak rate test, the acceptable changes in elevation differences will be similar to that for the initial containment structural integrity test applied to the elevation differences during the periodic containment integrated leak rate test.
b.
Measuring crack widths and plotting crack patterns in the area of the dome 3 feet en either side of azimuths 195 from the apex to the Ring Girder. Cracks wider than 0.010 inches will be plotted and cracks wider than 0.040 inches shall require an engineering evaluation.
In addition, a general visual inspection of the entire dome surface area shall be performed.
4.6.1.6.5 Reports Any abnormal degradation of the containment structure detected during the above required tests and inspections shall be reported to the Comissicn pu rsuant to specification 6.9.1.
This report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective actions taken.
1073 267 CRYSTAL RIVER - UNIT 3 3/4 6-9a Amendment No.24
CONTAINME.lT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the BWST on a containment spray actuation signal and manually transferring suction to the containment sump.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 />; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:
At least once per 31 days by verifying that each valve (manual, a.
power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.
b.
At least once per 18 months, during shutdown, by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray actuation test signal.
Verifying that each spray pump starts automatically on 2.
a containment spray actuation test signal.
CRYSTAL RIVER - UNIT 3 3/4 6-10 id[)
2bb
TABLE 3.6-1 (Conti nued)_
CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds) 9.
(Continued)
MUV-163 check #
open during HPI and NA MUV-25 #
iso. der, nor. operation 60 MUV-164 check #
NA MUV S f 60 MUV-150 check i open during HPI and NA MUVJ3 #
1so. dur. nor. operation 60 MUV-161 check #
open during HPI and NA MUV-24 #
iso. dur. nor. operation 60 MUV-27 h open dur. nor. operation 60 and closed during RB isolation
- 10. SWV-39 i iso. NSCCC from AHF-lc 60 SWV 45 i 60 SWV-35 #
iso. NSCCC from AHF-1 A 60 SWV-41 #
60 SWV-37 #
iso. NSCCC from AHF-1B 60 SWV-43 i 60 SWV-48 J to isolate NSCCC frm 60 SWV-47 #
MUHE-1 A & 1B and WDT-5 60 SWV-49 #
60 SWV-50 1 60 SWV-80 #
iso. NSCCC from RCP-1 A 60 SWV-84 #
60 SWY-82 #
iso. NSCCC frm RCP-lC 60 SWV-86 3 60 SWV-81 3 iso. NSCCC from RCP-lO 60 SWV-85 #
60 SWV-79 #
iso. NSCCC from RCP-1B 60 SWV-83 #
60 SWV-109#
NSCCC to CRRD-1 60 SWV-110#
60 CRYSTAL RIVER - UNIT 3 3/4 6-19 Amendment No. 77,2fi 10/3 269
TABLE 3.6-1 (Continued)
CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds)
- 11. WDV-4 iso. WDT-4 from RB sump 60 WDV-3 60 WDV-60 & 61 iso. WDT-4 from WDT-5 60 WDV-94 & 62 iso. WDT-4 from WDP-8 60 WDV-406 iso. gas waste disposal 60 WDV 405 from vents in RC system 60
- 12. WSV-3 iso. containment monitoring 60 WSV-4 system from RB 60 WSV-5 60 WSV-6 60 b.
CONTAINMENT PURGE AND EXHAUST 1.
AHV-lC & 1D iso pur. si p. system 60 AHV-1B & 1A iso pur, exhaust system 60 C.
MANUAL 1.
IAV-28 iso. IA from RB NA IAV-29 NA 2.
LRV-50 iso. leak rate test system NA LRV-36 from RB NA LRV-51 iso. atmos. vent and RB NA LRV-35 & 47 purge exhaust system NA from RB LRV-49 iso. atmos. vent from RB NA LRV-38 & 52 NA LRV 45 iso. LR test panel from RB NA LRV-44 NA 3.
MSV-146#
iso misc. waste storage NA tank from RCSG-1B 4.
NGV-62 iso. NG system from NA NGV-81 #
steam generators NA NGV-82 iso. NG system from pzr.
NA 107 3
'27 f)
CRYSTAL RIVER - UNIT 3 3/4 6-20
PLANT SYSTEMS 3/4.7.8 AUXILI ARY BUILDING VENTILATION EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8.1 The auxiliary building ventilation exhaust systen 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 /> and in COLD 3 HUT-DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.8.1 Each auxiliary building ventilation exhaust system shall be demonstrated OPERABLE:
l a.
At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control roan, ficw thrc)gh the HEPA filters and charcoal adsorbers and v.erifying that the system operates for at least 15 minutes.
b.
At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chenical release in any venti-lation zone communicating with the system by:
1.
Verifying that with the systen operating at a flow rate of 156,680 cfm + 10% and exhausting through the HEPA filters and charcoal adsorbers, the total bypass flow of the system to the facility vent, iricluding leakage through tha system diverting valves, is < 1% when the system is tested by admitting cold DOP at the system intake.
- The air flow distribution test Section 8 of ANSI N510-1975 may be performed downstream of the HEPA filters.
CRYSTAL RIVER - UNIT 3 3/4 7-23 Amendment No. 24' 10/3 271
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.
Verifying that the ventilation system satisfies the in-place testing acceptance criteria and uses the te t procedures of Regulatory Positiens C.S.a. C.S.c* and C.5.d*
of Regulatory Guide 1.52, Revision 1, July 1976, and the system flow rate is 156,680 cfm 110%.
3.
Verifying, within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regula-tory Guide 1.52, Revision 1. July 1976.**
4.
Verifying a system flow rate of 156,680 cfm +10% during system operation wnen te:ted in accordance wTth ANSI N510-1975.
After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by c.
verifying within 31 days arter removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of degulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.**
d.
At least once per 18 months by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the system at a flow rate of 156,680 cfm 1 10%.
After each complete or partial replacement of a HEPA filter e.
bank by verifying that the HEPA filter banks remove i 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975* while operating the system at a flow rate of 39,170 cfm i 10%.
f.
After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 99% of a halogenated hydrocarbon refrigerant test,; s when 1they are tested in-place in accordance with ANSI N510-1975*
while operating the system at a flow rate of 39,170 cfm,+ 10%.
- The faboratory test of Table 3 for a represantative sample of used activated carben shall be per Test 5b in Table 2 at a relative humidity of 70% for a methyl iodide-removal efficiency of 1 95%.
i CRYSTAL RIVER - UNIT 3 3/4 7-24 Amendment No. 4 1073 272
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS Demonstrated OPERABLE by detennining that each battery supplying c.
DC control power to the 230kv switchyard breakers is OPERABLE; 1.
At least once per 7 days by verifying that:
a)
The elect"olyte level of each pilot cell is between the minimum and maximum level indication marks, b)
The pilot cell specific gravity, corrected to 77'F, and full electrolyte level is > 1.20.
c)
The pilot cell voltage is > 2.15 volts, and d)
The overall battery voltage is > '20 volts.
2.
At least once per 92 days by verifying that:
a)
The voltage of each connected cell is > 2.15 volts under float charge and has not decreased more than 0.10 volts from the value observed during the base-line tests, and b)
The specific gravity, corrected to 77'F, and full electrolyte level of each connected cell is > l.20 and has not decreased more than 0.01 from the value observed during the previous tests, and c)
The electrolyte level of each connected cell is between the minimum and maximum level indication marks.
3.
At least once per 18 months by verifying that:
a)
The cells, cell plates and battery racks show no visual indication of physical damage or adnont.al deterioration.
b)
The cell-to-cell and tenninal connections are clean, tight and coated with anti-corrosion materials, c)
The battery charger will supply at least 95 amperes at 125 volts for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
CRYSTAL RIVER - UNIT 3 3/4 8-3 10/3 2/3
^
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.
At least once per 18 months, by verifying that the Dattery 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 bat-tery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.
This perfonnance discharge test shall be performed 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, 4
Verifying the diesel starts from ambient condition and accelerates to at least 900 rpm in < 10 seconds,
5.
Verifying the generator is synchronized, loaded to > 1500 kw, and operates for > 60 minutes, and 6.
Verifying the diesel generator is aligned to provide standby power to the associated emergency busses,
b.
At least once each 92 days by verifying that a sample of diesel fuel from the fuel 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 shutdcwn by -
1.
Subjecting the diesel to an inspec tion in accordance with procedures prepared in conjunction with its manufacturer's reconTnendations for this class of standby service, CRYSTAL RIVER - 1] NIT 3 3/4 8-4 Amendment No. g,2h 1073 274
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.
. Verifying the generator capability to reject a load of
> 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)
Verifyinc de-energization of the emergency buses and load shedding from the emergency busses, b)
Verifying that the 4160 v. emergency bus tie breakers open.
c)
Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency busses with pemanently connected loads, energizes the auto-connected emergency loads through the load sequencer and operates for > 5 minutes while its generatar is loaded with the emergency loads.
4.
Verifying the diesel generator operates for > 60 minutes while loaded to > 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"..
1073 275 CRYSTAL RIVER - UNIT 3 3/4 8-5 Amendment No. $, gli
(
ELECTRICAL POWER SYSTEMS SHUT 00WN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
a.
One circuit between the offsite transmission network and the onsite Class lE distribution system, and b.
One diesel generator with:
1.
Day fuel tank containing a minimum volume of 400 gallons of fuel, 2.
A fuel storage system containing a minimum volume of 20,300 gallons of fuel, and 3.
A fuel transfer pump.
APPLICABILITY: MODES 5 and 6.
ACTION:
'Jith less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the minimum required A.C. electrical power sources are restored to OPERABLE status.
SURVEILLANCE REQUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by performance of each of the Surveillance Re-quirements of 4.8.1.1.1 and 4.8.1.1.2, except requirement 4.8.1.1.2.a.5.
CRYSTAL RIVER - UNIT 3 3/4 8-6 10/3 2/6
3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boren con-centration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:
servative $ffowance.95 or less, which includes a 1% ak/k con-Either a K of 0 a.
for uncertainties, or b.
A boron concentration of > 1925 ppm, which includes a 50 ppm conservative allowance for uncertainties.
APPLICABILITY: MODE 6*.
ACTION:
With the requirements of the above specification not satisfied, imediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes, and initiate and continue boration at > 2700 gpm of 2270 ppm boren is reduceH to < 0.95 or the boron solution or its equivalent until K concentration is restored to > 192$$pm, 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 concentration of the reactor coolant system and the refueling 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 />.
- ine reactor shall be maintained in MODE 6 whenever the reactor vessel head is unbolted or removed and fuel is in the reactor vessel.
1073 277 CRYSTAL RIVER - UNIT 3 3/4 9-1 Amendment No. 2@
REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, the following instrumentation shall be OPERABLE:
a.
Two source range neutron flux monitors, each with visual indication in the control room and one with audible indication in the control room, and b.
One auxiliary source range neutron flux monitor with audible indication in the containnent or one source range neutron flux monitor with audible indication in the containment.
APPLICABILITY: MODE 6.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of :
a.
A CHANNEL FUNCTIONAL TEST at least once per 7 days, and b.
A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initial start of CORE ALTERATIONS, and c.
A CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.
CRYSTAL RIVER - UNIT 3 3/49-2 Amendment Mo. 8 1073 278.
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTA..NENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
3/4.6.1.2 CONTAINMENT LEAXAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure of 49.6 psig, P. As an added conservatism, the measured overall integrated leakage r$te is further limited to < 0.75 L during perfomance of the periodic tests to account for possible degrad$ tion of the containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50.
3/4.6.1.3 CONTAINMENT AIR LOCKS The limitatiors on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate given.
Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
CRYSTAL RIVER - UNIT 3 B 3/4 6-1 1073 2/9
CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 3.0 psig and 2) the containment peak pressure does not exceed the design pressure of 54.6 psig during LOCA conditions.
The maximum peak pressure obtained from a LOCA event is 49.6 psig.
The limit of 3 psig for initial positive containment pressure will limit the total pressure to 52.6 psig which is less than the design pressure and is consistent with the safety analyses.
3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a LOCA.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the con-tainment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will sithstand the maximum pressure of 52.6 psig in the event of a LOCA.
The measurement of containment tendon lift off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, the measurement of dome elevation differences and cracks, the general visual inspection of the dome and the Type A leakage test are sufficient to demonstrate this capability.
1 1073 280
=
CRYSTAL RIVER - UNIT 3 B 3/4 6-2 Amendment No. 2M
+ _ -.
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on minimcm boron concentration ensure that:
- 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity cc >ol in the water he limitation of a volumes having direct access to the reactor vessel.
of no greater than 0.95, which includes a conservative allowance Kf8 uncertainties, is sufficient to prevent reactor criticality f
during refueling operations.
3/4.9.2 INSTRUMENTATION The OPERABILITY of source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the safety analyses.
3/4.9.4 CONTAINMENT PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPEPABILITY and closure requirements are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressuriza-tion potential while in the REFUELING MODE.
3/4.9.5 COMMUNICATIONS The requirement for comunications capability ensures that refueling station personnel can be promptly infonned of significant changes in the
- acility status or core reactivity condition during CORE ALTERATIONS.
CRYSTAL RIVER - UNIT 3 B 3/4 9-1 10/3 281
REFUELING OPERATIONS 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 assenblies, 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 POOL 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 operation 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 strati fication.
3/4.9.9. CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment purge and exhaust penetrations will be automatically isolated upon detection of high radiation 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 mininum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activtty released frm the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety u,
analysis.
CRYSTAL RIVER - UNIT 3 B 3/4 9-2 Amendment No. 24 10/3 282
l-3 Known Radioactive Source - A calibration source which is traceable 1.8 to the National Bureau of Standards radiation measurement system and is espable of reproducible geometry.
Intake Area - The intake canal and all of the water area south of the 1.9 north intake dike and within two miles of the west tip of the south intake dike.
Di. charge Area - The discharge canal and all of the water area north 1.10
=
of the south discharge dike ard within two miles of the north dis-charge dike.
Inner Bay - An area as shown in. Figure 1.1-2 which is five feet or 1.11 less in depth composed of a mixture of grassy bottoms, oyster associa-tions, algal bottoms and areas of sand and mud.
Outer Bay - The outer basin as shown in Figare 1.1-2 in which the 1.12 planktonic ecosystem becomes as important as the bottom ecosystems.
1.13 Channel Calibration - The cdjustment, as necessary, of the channel such that it responds with necessary range and accuracy to output The channel known values of the parameter which the channel monitors.
calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the channel func-Channel calibration mar be performed by any series of tional test.
sequential, overlapping or total chamael steps such that the entire channel is calibrated.
~
Channel Check - The qualitative assessment of chennel behavior during 1.14 operation by, observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
1.15 Channel Functional Test - The injection of a simulated signal into the channel as close to the pri=ary sensor as practicable to verify operability including alarm and/or trip functions.
1.16 Dese Ecuivalent I-131 - That concentration of I-131 (pC1/ gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in TID-14844.
10/3 283
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. ; = CD -4 l V r0 C3 Maure 1 1-2 Inner and Outer Baya 3-1 3.0 ENVIRONMENTAL SURVEILIANCE 3.1 NONRADIOLOGICAL SURVEILIANCE Study Plan The estuary has been exposed to the influence of the operation of Units 1 and 2 for approximately seven (7) years. During this time, the systems in the area have adapted to this influence. A preopera-tional surveillance program was deorgned to determine the exact nature of the new stabilized conditions relative to control areas adjacent to the plant site. This surveillance consisted of system modeling with measurements of biomass, productivity, respiration and diversity in all major compartments. The infor=ation derived will serve as a baseline for comparison with the data taken af ter Unit 3 becomes operational. The o N rational surveillance program is designed to determine any significant environmental effects of the op.eracio.n of the power plant, particularly unpredicted and catastrophic changes. The pre m consists of 3 short-term intensive surveillance program elements and l 2'long-term program elements. A period of adjustment of the ecosystem is expected concurrent with Crystal River Unit 3's initial operation. This will be a localized perturbation limited to a portion of the inner bay associated with the higher water velocity as well as the temperature increase resulting from the condenser discharge. Any ecosystem which experiences a change in its environment will undergo a period of adaptation unless catastrophic conditions occur. With the small changes anticipated with the addition of Unit 3, no catastrophic effects are expected. However, any changes in the environmental conditions of a system will normally cause it to oscillate. An example of the oscillation of a hypothetical system's productivity is shown below.
- p. Time to Perturbation
~~ l p-Time to Minimum I I g 7-Approximate Time to Stabilization 1 I i i Stabilized Level I l l Productivi ty _ l _ _ _ _ Initial Level Minimum level Time 1073 285 Amendment No. 24 3-2 In this particular system the final stabilized level is higher than the initial level and is only obtained after a period of stabiliza-tion and etter going through a suppressed level fo' lowing the initial A perturbation. The recognition of this type of potential response is obviously important in considering any surveillance program. The models of the systems involved at Crystal River along with the data available indicate that the approximate time to stabilization should not excaed one year, Therefore, the time frame for the intensive surveillance program elements allows one year of monitoring to determine the transient response that the systems are experiencing. An additional ygar of monicoring is required to indicate the new stabilized level. If the second year's data indicate that the syste=s havc noc approached stabilization, the monitoring wil. be extended for an additional year. It is anticipated that the intensive sur-veillance program elements should not be necessary beyond three years. The areas in which intensive monitoring will be performed as indicated or until stabilization occurs include the following program elements: (1) benthos in discharge area, (2) marsh grasses, and (3) impingement l on intake screens. In addition to the short-term intensive surveillance program elements designed to determine how the systems have responded to the per-turbations, an on-going program element designed to obtain a diagnos-tic view of the condition of the environment will be continued during the cgerational life of the plant. This indicator progra= ele =ent cons 2 sts of a number of simple measure =ents which will detect any major changes in the system. A second long-term program element involves chemical-industrial waste water monitoring. 3.1.1 Benthos in Discharge Area Obj ective To determine the ecological condition of the benthic system in the area directly affected by the thermal plume. Specification Operational monitoring of productivity, respiration, diversity and biomass of the benthic system in the area adjacent to and north of the discharge canal shall be measured on a quarterly basis until the system has approached stabilization. Samples shall be taken by methods employed in the preoperational studies including harvesting quadrats, by sediment cores, and by venturi pumps. The number, frequency and location of }Q/3 2Ob Amendr.ent No. 24 3-3 samples to be taken shall be determined from a critical review of the results of the preoperational research con-ducted in this area. Samples shall be stratified by sacrophyte dominance. Productivity and respiration of the system shall be determined by the methods currently employed in the modeling work. Reporting Requirement Results of the data gathered in this program element shall be reported in accordance with Section 5.6.1. In the event that any parameter measured changes beyond two standard deviations of the value measured in the preoperational monitoring program, a report shall be submitted as specified in Section 5.6.2. Bases In the discharge area adjacent to the canal, the productivity, respiration and biomass should increase due to an increased temperature of the cooling water. If any of these parameters changes beyond 2c (two standard deviation) of that measured during preoperational monitoring, the systen should be investigated for catastrophic results. 3.1.2 Marsh Grass Objective To determine the ecological condition of the salt marsh adjacent to the discharge area. Specification The biomass, productivity, and respiration of the salt marsh shall se measured on a quarterly basis after plant operation begins until the system has approached stabilization. Quadrats shall be harvested to determine biomass and productivity. Reporting Requirement Results of the data gathered in this program element sha!1 be reported in accordance with Section 5.6.1. In the event that any parameter measured changes beyond 2c (two standard deviation) of the value measured in the preoperational monitoring program, a report shall be submitted as specified in Section 5.6.2. 1073 287 Acendment No. 24 L 3-4 Sases The metabolism of the marsh' grass is expected to increase with Any de:rease indicates a breakdown of increasing temperature. If any of these parameters changes beyond 2a structure. (two standard deviation) of that measured during preopera-tional monitoring, the system should be investigated for catastrophic results. 3.1.3 Imoingement on Intake Screens Objective To determine the quantity of impinged fish and sheilfish on the intake screens to compa m with preoperational data. Specification The fish and shellfish collected in the trash racks adjacent to the intake screens of Units 1 and 2 and Unit 3 will be sampled for 24 consecutive hcurs once weekly'. This program shall be conducted for one year after operation of Unit 3 This program. may be terminated after one year period begins. with staff!s approval. Samples shall be sorted according to species, length, and wet weight. This requirement shall not be applicable during the period of the intake canai modification addressed in State of Florida, Department of Environmental Regulation pennit No. 09-20-4006 issued September 30, 1977. The screen-wash racks shall be monitored visually daily to deter-mine any abnonnal catches. Recoctinc Recuirement Results of the data gathered in this program element shall be reported in accordance with Section 5.5.1. Any weekly sample { with fish and shellfish biomass greater than 50 kg shall be reported as specified in Section 5.6.2. Bases Freoperational data indicate that the average normal expected catch is approximately 20 kg per day for Unit No. 3. 3.1.4 General Ecological Survey _ Objective To detect changes which might occur and would be used to indicate areas requiring more detailed investigation. Specifications A series of :neasurements shall be carried out during the operational life of the plant to indicate the general con-dition of the environment. The areas to be monitored are: ~n 10/3 200 Amendment No. 12,
- haem-