ML20041E000
| ML20041E000 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/02/1982 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20041D997 | List: |
| References | |
| NUDOCS 8203090778 | |
| Download: ML20041E000 (29) | |
Text
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no, UNITED STATES
{.h y., f)/;~h NUCLEAR REGULATORY COMMISSION S h t, WASHINGTON,0. C. 20555
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PUBLIC SERVICE COMPANY OF COLORADO DOCKET NO. 50-267 FORT ST. VRAIN NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING. LICENSE Amendment No. 25 License No. DPR-34 1.
The Nuclear' Regulatory Commission (the Commission) has found that:
The applications for amendment by Public Service Company of A.
Colorado (the licensee) dated January 20, April 14, May. 9,1978, June 19,1979, July 24, and December 31, 1980 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations y
. set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of i
the Commission; There is reasonable assurance (i) that the activities authorized C.
by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commisrsion's regulations; The issuance of this amendment will not be inimical to the common D.
defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part E.
51 of the Commission's regulations and all applicable requirenents have been satisfied.
8203090778 820302
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PDR ADOCK 05000267 P
r
.8
,I 2.
Accordingly, Facility Operating License No. DPR-34 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending Paragraph 2.C.(4) and
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2,0.(2) to read as follows:
2.C.(4) Pursuant to the Act and to 10 CFR Part 30, " Rules of General Applicability to Licensing of Byproduct Material," to receive, possess and use in connection with operation of the facility:
1.
Any byproduct material with Atomic Numbers 1 through 83, inclusive, not to exceed 5 millicuries per radio-nuclide; 2.
Americium 241, not to exceed 2.01 curies;
_3. Americium 243, not to exceed 5 millicuries; 4
Cesium 137,.not to exceed 11 curies; 5_. Hydrogen 3, not to exceed 15 curies; 6_. Krypton 85, not to exceed 110 millicuries; 7.
Neptunium 237, not to exceed 5 millicuries;
_8. Polonium 210, not to exceed 10 microcuries; 9.
Thorium 228, 230, 232, not to exceed 1 millicurie per nuclide; 10.
Radium 228, 226, not to~ exceed 5 millicuries per nuclide; l
2.0.(2) Technical Specificatio,n,s,s The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
,are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This l'icense amendment is ef fective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b
Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: March 2, 1982
,s ATTACHMENT T0 LICENSE AMENDMENT ~
AMENDMENT NO. 25 TO FACILITY OPERATING LICENSE NO. DPR-34 DOCKET WO. 50-267 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identiff e4 by Amendment number and contain ve'rtical liner indicating the area of change.
Remove Insert 1
i (no change) ii ii v
v vi vi 4.3-7 4.3-7 4.3-8 4.3-8 4.3-11 4.3-11 4.4-8 4.4-8 4.6-1 4.6-1 5.4-4 5.4-4 E.4-5 5.4-5 5.4-8 5.4 5.5-3 5.5-3 5.5-4 5.5-4 (repositioned) 5.5-E (repositioned) 5.5-6 7.4-2 7.4-2 7.4-3 7.4-4 7.4-5 7.5-6 7.4-7 7.5-2 7.5-2 7. 5.-3 7.5-3 7.5-4 7.5-4 7.5-5 7.5-5 (repositioned) 7.5-6 7.5-6 (repositioned) 7.5-7 7.5-7 7.5-8 7.5-8 7.5-9 7.5-9 7.5-10 7.5-10 7.5-11 i
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s FORT ST. VRAIN NUCLEAR GENERATING STATION TECHNICAL SPECIFICATIONS TABLE OF CONTENTS PAGE 1-1 INTRODUCTION..................................................
1.0 D EF I N IT I ON S.....................................
2.0 3 0-1 SAFEIY LIMITS AND LD41 TING SAFETY SYSTD4 SETTINGS.............
30 3.1-1 REACTOR CORE - SAFETY LIMIT...................................
31 3.1-1 Specification SL 3.1 - Reacter Core Safety LL=it..............
3.2-1 REACTOR VESSEL PRESSURE - SAFETY LD417........................
3.2 Specificatica SL 3 2 - Reacter Vessel Pressure
~3.2-1 Safety Limit.............................................
3.3-1 LIMITING SAFETY SYSTEM SETTING S................................
33 Specification LSSS 3 3 - L1=iting Safety Syste 3.3-1 Settings.................................................
h.0-1 LE4ITING CONDITION S FOR 0?IRATION.............................
'-.C REACTOR CORE AND REACTIVITY CONTROL - LIMITING h.1-1 L.I CONDITIONS FOR 0FERATION.................................
h.1-1 Speci ficatien LCO k.1.1 - Core Irradiation....................
h.1-2 Specifica:icn LCO L.1.2 - Operable Cent.-:1 Reds...............
h.1-3 Specification LCO h.1 3 - Red sequence........................
L.1-7 h.1. h - P ar ti ally I ns er t ed R ed s.............
Specification LCO Trecification LCO h.l.5 - Reactivity Change with h.1-8 h.1-10 Te=perature.....
Specification LCO h.1.6 - Reserve Shutdevn Syste=.............
h.1-11 Specification LCO h.l.7 - Core Inlet Crific e Valves...........
h.1-13 Specification LCO h.1.S - Reactivity Status...................
h.1-lh Specification LCO h.1 9 - Ccre Region Te=perature Rise........
i
.e s
- fage, 4.2 PRIMARY COOLANT SYSTEM - LIMITING CONDITIONS 4.2-1 FOR OPERATION...........
4.2-1 Specification LCO 4.2.1 - Number of Operable Circulators....
4.2-2 Specification LCO 4.2.2 - Operable Circulator...............
4.2-3 Specification LCO 4.2.3 - Turbine Water Removal Pump........
4.2-3 Specification LCO 4.2.4 - Service Water Pumps...............
SpecificationLCO4.2.5-CirculatingWaterMakeu'$ System...
4.2-3 4.2-4 Specification LCO 4.2.6 - Firewater Pumps 4.2-5 Specification LCO 4. 2.7 - PCRV Pressurization...............
4.2-7 Specification LCO 4.2.8 - Primary Coolant Activity..........
4.2-11 Specification LCO 4. 2. 9 - PCRV Closure Seals................
Specification LCO 4.2.10 - Loop Impurity Levels, 4'.2-13 High Temperatures......................................
Specification LCO 4.2.11 - Loop Impurity Lev'els, 4.2-13 Low Temperatures.......................................
4.2-14 Specification LCO 4.2.12 - Liquid Nitrogen Storage...........
4.2-14 Specification LCO 4.2.13 - PCRV Liner Cooling System........
4.2-15 Specification LCO 4.2.14 - PCRB Liner Cooling Tubes Specification LCO 4.2.15 - PCRV Cooling Water System 4.2-16 Te=peratures...........................................
4.3 SECONDARY REACTOR COOLANT SYSTEM - LIMITING CONDITIONS 4.3-1 FOR OPERATION.............................................
4.3-1.
Specification LCO 4.3.1 - Steam Generators..................
Specification LCO 4.3.2 - Boiler Feed Pumps................., 4.3-2
~4.3-2 Specification LCO 4.3.3 - Steam / Water Dump Tank Inventory...
i-Specification LCO 4.3.4 - Emergency Condensate and Emergency 4.3-3 Faedwater Headers......................................
'4. 3 Specification LCO 4.3.5 - Storage Ponds.....................
4.3-4 Specification LCO 4.3.6 - Instrument Air System.............
4.3-4 Specification LCO 4.3.7 - Hydraulic Power System............
4.3-5 Specification LCO 4.3.8 - Secondary Coolaat Activity........
Specification LCO 4.3.9 - Deleted 4.3-7 I
Specification LCO 4.3.10.- Sheck Suppressors (Snubbers)
Table 4.3.10-1 Class I Hydraulic Snubbers 4.3-9 4.4 INSTRUMENTATION AND CONTROL SYSTEMS - LIMITING 4.4-1 CONDITIONS FOR OPERATION..................................
Specification LCO 4.4.1 - Plant Protective System 4.4-1 Ins tr um e n t a t i o n........................................
4.4-13 Specification LCO 4.4.2 - Control Rocm Temperature..........
4.4-13 Specification LCO 4.4.3 - Area Radiation Monitors 4.4-15 Specification LCO 4. 4.4 - Seismic Instrumentation...........
11 Amendment No. 25 2
E PAGE 5.5-1 5.5 CONFINEMENT SYSTEM - SURVEILLANCE REQUIREMENTS............
5.5-1 Specifica'. ion SR 5.5.1 - Reactor Building.................
Specification SR 5.5.2, Reactor Building Pressure 5.5-1 Relief Device...........................................
l.
5.5-3 Specification SR 5.5.3 - Reactor Building Exhaust Filters.
5.6-1 5.6 EMERGENCY POWER SYSTCIS - SURVEILLANCE REQUIREMENTS.......
Specification SR 5.6.1 - Standby. Diesel Generitor.........
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5.6-1 5.6-2 Specification SR 5. 6. 2 - Statien Bat tery..................
5.7 FUEL HANDLING AND STORAGE SYSTDiS - SURVEILLANCE 5.7-1 REQU IR C ENTS............................................
Specif ication SR 5.7.1 - Fuel Handling Machine............ ~ 5.7-1 5.7-2 Specif ication SR 5.7.2 - Fuel Storage Facility............
5.8 RADIOACTIVE EFFLUENT DISPOSAL SYSTCIS - SURVEILLANCE 5.8-1 REQUIREMENTS............................................
Specification SR 5.8.1 - Radioactive Gaseous Effluent 5.8-1 System..................................................
Specification SR 5.8.2 - Radioactive Liquid Effluent 5.8-1 System..................................................
5.9 ENVIRONMENTAL SURVEILLANCE - SURVEILLANCE REQUIREMENTS....
5.9-1 5.9-1 Specification SR 5.9.1 - Environmental Radiation..........
5.10 FIRE SUPPRESSION SYSTEMS - SURVEILLANCE REQUIREMENTS......
5.10.
Specification SR 5.10.1 - Three Room Control Complex 5.10-1 HVAC System.............................................
Specification SR 5.10.2 - Halon Fire Suppression 5.10-1 System..................................................
5.10-2
]
Specification SR 5.10.3 - Smoke Detectors and Alarm.......
5.10-3 Specification SR 5.10.4 - Fire Barrier Penetration Seal...
5.10-3
- Specification SR 5.10.5 - Breathing Air System............
5.10-3 Specif ication SR 5.10.6 - Fixed Water Spray System........
Specification SR 5.10.7 - Carbon Dioxide Fire Suppression 5.10-4 System..................................................
5.10-5 Specification SR 5.10.8 - Fire Hose Stations Specification SR 5.10.9 - Yard Fire Hydrants and Hydrant 5.10-5 Hose Houses.............................................
6.0-1 6.0 DESIGN FEATURES...........................................
6.1-1 6.1 REACTOR CORE - DESIGN FEATURES.............................
6.1-1 Specif ication DF 6.1 - Reactor Core.......................
Ft. St. Vrain v
Amendment No. 74, 25
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PAGE 6.2 REACTOR C00i.AUT SYSTEM AND STEAM PLANT SYSTEM -
6.2-1 DESICN FEATURES.........................................
6.2-1 Specification DF 6.2.1 - PCRV.............................
6.2-3 Specification DF 6.2.2 - Steam Generator Orifices.........
6.2-3 Specification DF 6.2. 3 - Steam Saf ety Valves..............
6.3-1 6.3 SITE - DESIGN FEATURES.......................,............
6.3-1 Specification DF 6.3 - Site...............................
7.0-1 7.0 ADMINISTRATIVE CONTROLS...................................
7.1 '
ORGANIZATION, REVIEW AND AUDIT - ADMINISTRATIVE CONTROLS..
7.1-1
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7.1-1 Specification AC 7.1.1 - Organization.....................
Specification AC 7.1.2 - Plant Operations Review 7.1-5 Committee...............................................
Specification _AC 7.1.3 - Nuclear Facility Safety 7.1-11 Committee...............................................
7.2-1 7.2 S AFETY LIMITS - ADMINISTRATIVE CONTROLS...................
Specification AC 7.2 - Action to be taken if a Safety 7-2-1 Limit is Exceeded.......................................
7.3-1 7.3 ABNORMAL OCCURRENCE - ADMINISTRATIVE CONTROLS.............
Specification AC 7.3 - Action to be Taken in the Event of 7.3-1
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an Abnormal occurrence..................................
7.4-RECORDS - ADMINISTRATIVE CONTROLS..........................7.4-1 7.4-1 Specification AC 7.4 - Records............................
7.5 REPORTING REQUIREMENTS....................................
7.5-1 Specification 7.5.1 - Routine Reports.....................
7.5-1.
Specif ication 7. 5. 2 - Reportable Occurrences..............
7.5-4 Specification 7. 5.3 - Environmental Qualification.........
7.5-9 Ft. St. Vrain vi Amendment No. 44, 25 1
e 4.3-7 Specification LCO 4.3.10 - Shock Sup' pressors (Snubbers) - Limiting Condition
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[
for Operation
- 4 The reactor shall not be operated at power unless all shock suppre'ssors -
(a)
(snubbers) on Class I piping systems (listed in Table 4.3.10-1) are operable except as noted in (b) through (d) of th'is LCO.
From and af ter the time that a shock suppressor is determined to be in-(b) operable, continued reactor operation at power is permissible only during the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless the suppressor is sooner made operable or replaced.
an orderly (c)
If the requirements of (a) and (b) of this LCO carnot be met, shutdown shall be initiated and the reactor shall be in a low power con-l dition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
(d)
If a shock suppressor is determined to be inoperable while the reactor i's in the shutdown or refueling mode, the suppressor shall be made ' operable or replaced prior to reactor operation at power.
(c)
Shock suppressors may be added to Class I systems without prior License to Table 4.3.10-1, provided a revision to Table 4.3.10-1 is in-Amend ent cluded with a subsequent License Amendment request.
Easis for specificatica LCO 4.3.10 Shock suppressors (snubbers) are designed to prevent unrestrained pipe motion under dynamic loads, as might occur during an earthquake, while allowing normal ther=al motion during startup and shutdown. The consequence of an in-operable snubber is an increase in the probability of structural damage to It is piping resulting frem the dynamic loads produced by a seismic event.
therefore necessary that all snubbers required to protect-the Class I syste=s or components be operable during reactor power operation.
Amendment No. 25
,e 4.3-8 Because snubber protection is required only during relatively low proba-In bility events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is. allowed for repair or replacement.
case a shutdown is required, the allowance of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach a low power condition will pemit an orderly power reduction consistent with standard Since reactor operation at power should not be conducted operating procedures.
with.def ective saf ety-related equipment, reactor power operation is prohibited' vich inoperable snubbers.
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Amendment No. 25
,s 4.3-11 TAEI.E 4. 3. "0-1 (centinued)
CLASS I HYDRACL!C SNUBBERS (continued)
Boiler Feed Snubbers Generators BFS-B1-1 BFS-B1-4 BFS-32-1 BF5-E2-4 BFS-B1-2 BFS-B1-5 B FS-B2-2 BFS-32-5 B FS-al-3 B FS-B1-6 BFS-32-3 B75-32-6 Boiler Feed Snubbers BFS-54 BFS-421 BFS-526 BFS-577 BFS-138 BFS-422 BFS-52 8 BFS-614 BFS-139 BFS-425 BFS-529 BFS-641 BFS-142 BFS-424-1 BFS-530 BFS-679 BFS-149 B FS-434-2 BFS-532 BFS-711 B FS-152 B FS-434-3 BFS-534 B FS-763 BFS-153 BFS-435 BFS-536 BFS-764 B FS-297 BF5-437 BFS-537 BFS-796 BF5-352 BFS-451 BFS-553-1 BFS-820 BFS-397 BFS-477 BFS-553-2 BFS-823 BFS-398 BFS-479 BFS-556 BFS-824 BF5-400 BFS-4 98 BFS-563 BFS-843 BFS-402 BF5-500 BFS-564 BFS-844 BFS-412 BFS-501 BFS-566 BFS-870 B FS-416 B FS-516 BFS-572 BFS-871 BFS-420 BFS-523 BFS-573 B FS -166 l
Boiler Feed Snubbers - E=arrency BFS-14E BFS-89E BFS-219E BFS-399E BFS-ISE BFS-122E BFS-228E BFS-405E BFS-15E BFS-141E BTS-229E BIS-414E BFS-26E BF5-142E BFS-243E BFS-417E BFS-29E BFS 1.43E BFS-244E BFS-419E BFS-30E 3 FS-15SE B FS-245E BFS-421E BFS-312 BFS-167E B FS-257E BFS-422E BFS-4 7E BTS-181E B FS-260E BFS-423E BFS-53E BFS-197E BFS-263E BFS-430E BFS-56E BFS-203E B FS-264E BFS-431E BFS-57E BFS-204E BFS-268E BFS-432E B FS-74E BFS-110E BFS-269E BFS-442E BFS-76E BFS-216E B FS-444E BFS-77E B FS-218E.
BFS-39 8E BFS-154E l
Amendment No. 25
4.4-8 Specificaticn LCO 4.4.1 NOTES FOR TABLES 4.4-1 THROUGH 4.4-4 (a)
See Specification LSSS3.3 for trip setting.
Two ther=ocouples from each loop, total of four, constitute one channel.
(b)
For each channel, two ther=occuples must be operable in at least one operating loop for that channel to be considered operabl,er With one pri=ary coolant high level moisture monitor tripped, trips of (c) either loop pri=ary coolant moisture monitors will cause full scra=.
Hence, nu=ber of operable channels (1) minus =ini=u= *n=ber required (0) equals one, the =inimum degree of redundancy.
to cause scra:
buses lA and 1C loss of voltage for no longer than 35 (d)
Both 480 volt seconds.
One channel consists of one undervoltage relay from each of the two (e)
These relays 480 volt buses (two undervoltage relays per channel).
f ail open which is the direction required to initiate a scra=,
The inoperable channel must be in the tripped condition, unless the (f) trip of the channel vill cause the protective action to occur.
RWP bypass per=itted if the bypass also causes associated single (g) channel scra=.
(h) Permissible Bypass Conditions:
Any circulator buffer seal malfunction.
I.
Loop het reheat header high activity.
II.
As stated in'Lco 4.9.2.
III.
Ite=s la. or Ic. or ld acco=panied by 2a., 2b., 2c., or 2d. on Table (j )
4.4-2 are required for loop 1 shutdown.
Ite=s lb. or Ic. or lf.,
acco=panied by 2a., 2b., 2c., or 2d. on Table 4.4-2 are required for loop 2 shutdown, c
One operable helic = circulator inlet thermocouple in an operable loop (k) is required for the channel to be considered operable.
4% af ter reactor power initially
(=)
Low Power RWF bistable resets at exceeds 5*.
Power range RWP bistables automatically reset at 10% after reactor (n) power is decreased from greater than 30%.
The RWP may be manually reset between 10% and 30% power.
(p) Ite: 7a. =ust be acccupanied by ite: 7c for Loop 1 shutdown.
7 7b. =ust be acco=panied by ite= 7c. for loop 2 shutdown.
Ite:
l A=end=ent No. 25 L
e
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h.6-1 h.6 AUXILIARY E:.ECTRIC PO'.EB SYS24 - LIMITING CONDITIONS FOR OPE?ATION Attlicability Applies to the mini =u= operable equipment supplying electric power to the plant auxiliaries.
Objective To ensure that the capability of supplying electric power to the plant auxiliaries is =aintained by defining the =ini=um operable equi;=ent.
Conditions Steeification LCO h.6.1 - Auxiliary Electric Syster, Li=iting fer Operation The reactor shall not be operated at power unless the following conditions are satisfied:
a) Both the Unit Auxiliary and Reserve Auxiliary Transformers are operable.
The Unit Auxiliary Transformer ot; the Reserve Auxiliary Trans-forcer can be made inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided both diesel generator sets (two engines and associated generator per set) are started'immediately prior to taking the affected trans-l former out of service to verify their operability, are shut devn and their controls lef t in the auto =atic mode and all three 480 volt a-c essential buses are operable.
b) h160 V a-c Bus 13 =ust be operable, h160 V a-c Bus 13 =ay be =ade incperable for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> providing the h80 V a-c Essential buses and both diesel generater sets are operable, cperability to be proved as in a) abcve.
Amendment No. 25
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5.4-4 Table 5.4-1 MINIMUM PREOUE!JCIES FOh CilECKS, CALIBRATIONS, AND TESTING OF SCRAM SYSTEM,(Continued)
Method Channel Description Function Frequency (1)
Inject moisture laden gas into sample lines 6.
Continued c.
Calibrate R
c.
d.
Check D
d.
Verification of eight separate monitor's sample, flow, per Item (t) of Notes for Tables 4.4-1,
through 4.4-4.
Verify that each of the eight monitors will c.
Test M
e.
alarm on low and high sample flow.
Trip one high Icvel, one low level channel, a.
Test M
a.
7.
Primary Coolant Moisture pulse another low level channel.
(ligh Level Channels)
Comparison of the averaged therm, couple 8.
Reheat Steam Temperature a.
Check D
a.
channel input indications 4
b.
Test M
b.
Trip channel, verify alarms and indications.
Internal test signal to verify trips and alarms.
to an NBS Compare each thermocouple output c.
Calibrate R
c.
traceable standard.
Internal test signal to adjust trips and indicators.'
Comparison of six separate channel indicators.
a.
Check D
a.
9.
P'rimary Coolant Pressure b.
Test M
b.
Trip channel, internal, test signal to verify trips and alarms.
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S Known pressure applied to sensor. Internal El c.
Calibrate R
c.
test signal to adjust trips and indicators, S
a Comparison of eight separate indicators.
=
10.
Circulator Inlet a.
Check D
a.
b.
Test M
b.
Trip channel, internal test signal to verify
- Temperature na trips and alarms.
c.
Calibrate R
c.
Compare each thermocouple output to an NBS traceable standard.
Internal test signal to adjust trips and indicators.
5.h-5 Table 5.h-1 MINIMUM FREQUENbIES POR CHECKS, CALIBRATIONS, AND TESTING OF SCRAM SYSTEM (continued)
' Channel Description Function Frequency (1)
Method Reduce pressure at sensor to t_ rip channel, 11.
Hot Reheat Header a.
Test M
a.
verify alarms and indications.
Pressure b.
Calibrate R
b.
Known pressure applied at sensor to adjust trips.
Reduce pressure at sensor to trip channel, 12.
Main Steam Pressure a.
Test M
a.
verify alarms and indications.
4 Known pressure applied at sensor to adjust trips.
b.
Calibrate R
b.
Special test module used to trip channel by n.
Ter. t M
a.
energizing each of four appropriate pairs of 13.
Two Loop Trouble two-loop trouble relays.
b.
Test R
b.
Trip logic to cause two loop trouble scram.
Trip each channel by applying simulated a.
Test M
a.
14.
Plant 480 V Power Loss loss of voltage signal; Mer.ify alarms and indicationis g
Comparison of three separate channel indicators.
k 15 High Reactor Building a.
Check D
a.
K Temperature (Pipe Cavity)
Trip chanrrel, verify' alarms and indications.
b.
Test M
b.
g Internal test signal to verify trips and alarms.
O to a NBS traceable Compare each thermocouple output c.
Calibrate R
c.
standard to adjust temperature trip point.
D - Daily when in use NOTE 1:. M - Monthly R - Once per refueling cycle P - Prior to each start-up if not done previous week t
. 5. 2 -8 4
Table 5.14-2 MINIMl!M FREQ11ENCIES FOR CllECKS. CA1.IBRATIONS AND TPSTING OF IDOP S!!IrfpWN SYSTEM Method' Channel Description Fu nc t. l on_
Frequency (1)
Ccanparison of 3 separate temperature 7.
Superheat IIcader a.
Check D
a.
indicator's per loop.
i h.
Check D
b.
Comparison of 3 separate temperature differential indicators.
Pulse test one channel with another channel c.
Test M
c.
tripped and verify proper indications.
d.
Calibrate R
d.
Compare each thermocouple output to an NBS traceable standard. Internal test signal to adjust trips and indicators.
Trip each channel, verify proper indications.
8.
Primary Coolant, a.
Test M
a.
Moisture (iov Trip each channel, pulse test other loop Level Channels) b.
Test M
b.
to check loop identification.
. Pulse test one channel with another channel a.
Test M
n.
tripped and verify proper indications, l
9
' Primary Coolant Pressure both channels.
i 1
i NOTE 1: D - Daily when in use M - Monthly H - Once per refueling cycle P - Prior to each start-up if not, done previous week Amendment No. 25
5.5-3 Basis for Soecifica:1on SR 5.5.2 The reac:or building pressure relief device is designed to pro:ec: the building in :he event that pressure in the Reactor Building exceeds the
- urbine building pressure by 3 inches of water. The device consists of louvers ins:alled in a nu ber of individual codules opera:ed by =echanical linkages to pneu=atic actua: ors (see FSAR See:1on 6.1.3.4).
The specified frequency shall ensure the operability of the reactor building relief
- es sys:es.
Specifica ion 3R 5.5.3 - Resctor Building Exhaus: yil:ers, Surveillance fil:ers in :he reactor building ventilation systes shall be ces:ed The exhaus:
as f ollevs:
a) A laboratory ' analysis of a representative carben sa:ple obtained in
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accordance with Regulatory Position C.6.b of Regula:Ory Guide 1.52, Revisien 2, P. arch 1973, shall be perf orned af:er each 4400 h:urs of cpera: ice of the uni:, or f ollowing painting, fire, or che=ical*
f
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1 release in any ven:112:1on :ene tw__.:nica:ing vich :he uni:. The I
l
.resul:s of labora:ory carbon sa=ple analysis frc= :he unit shall shew 1 90: radicactive ::e:hyl icdide re=cved uhen tes:ed in ac-cordance with A'!SI !!510-1973 (130*C, 95: R.E.).
- Defined as any =aterial which could reascnably be e=pec:ed to in:erfere vi:h i
the charecal :o adsorb nethyl iodide.
Amendment No. 25
5.5-4 b) A halogenated hydrocarbon test shan be perfor=ed once per calendar year or af ter each replacement of a charcoal adsorber bank or af ter s truct. tral maintenance on :! e fil:er' housing. Halogenated hydro-carbon removal by the sharcoal filters shall be 199% when conducted at nor=al flow conditions in accordance with the applicable portions of ANSI N510-1975.
c)
The HIPA filters shall be leak tested in place once per calendar year, af ter each eceplete or partial replace =ent of a HI?A filter bank. or af ter any structural =aintenance on the filter housing, using cold DOP. Cold DCP re=cval by the HE?A filters shall be 1 99 when tested in accordance vi:h the applicable porrions of ANSI N510-1975.
d)
F1cv dis:ribu:Lon across the HEPA and charcoal filters will be ecs:ed with initial operation of the system and folleving any s:ruc-tural modification to the f11ter hcusings. Air distribution, shan be denenstra:ed vi:hin 120~ across the HI?A and charcoal filters when tes:ed in accordance with ANSI N510-1975.
e)
Total pressure drop across the cenbined HI?A filter and char: cal adsorber banks shall be less chan 6" H O a:. fil:er design flev 2
- 10".
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Amendt.ent No. 25
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5.5-5 3 asis f or seccification SR 5.5.3
..W The reactor building exhaust fil:er sy, tem is designed to filter the reacter s
building at=esphere prior to release to the f acili:7 ven: s tack during both nor=al and accident conditions of operation. The system consists of three
'50i capacity units, two of which are in een:inuous operation, wi:h the third en s:andby.
High ef ficiency particulate air (EEPA) filters are ins: ailed bef ore :he char-coal adsorbers to re=ove particula:e =acter f rom the air strea= to prevent clogging of the iodi=e adsorbers. The charcoal adsorbers are ins:alled to reduce the potential release of radiciodine :o the a =osphere. 3ypass leakage fer the charecal adsorbers and particulate re=cval efficiency for HE?A fil-ters are deter =ined by halogena:ed hydrocarben and DOF respec:1vely. The i
labora: cry carben sa=ple test resul:s indica:e a radicactive =e:hyl iodide re=cval ef ficiency f or expected accident ecedi:1ces. The surveilla:ce :es:
frecuencies specidied es:ablishes syste= perfer=ance capabili:tes. If sys:es conditions are as specified, the calculated doses will be less than the guide-lines sta:ed in 10 CFR 100 f er the acciden:s analy:ed, as indica:ed in See-
- 1 ens 14.5 and 14.12 of the FS AR.
Pressure d::p a cross the ce=bined EE?A fil-c
- er and char: cal adscrber cf less :han 6 i=ches cf va:er a: :he fil:er de-sign fire ra:e will indicate tha: :he filters and adscrbers are ::: ciegged by excessive a:cu= s of f oreign =a::er.
Amendment flo. 25
5.5-6 The activated carbon adsorber in the aff ected uni: should be replaced if a representative sa=ple f ails to pass the iodine re= oval efficiency es:. Any EI?A fil:ers f ound def ective should be replaced.
If painting, fire, or che=ical release occurs such tha'$ the HI?A fil:er or
.charcoal adsorber could becene centa=inated f rc= the fu=es, che=icals, or f oreign a:erials, the sa=e tests and sa=ple analysis should be perf or=ed, as required, f or operational serve 111ance.
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l Amendment No. 25
7.4-2 Te=porary changes to procetures of a. above =ay be =ade provided:
c.
1.
The inten of the original procedure is ne: altered.
2.
The change is approved by :vo =m_bers of the plant mage-
=ent staff, at least one of who= holds a Senior Reac:or Oper-ators License.
The change is docu=ented, reviewed by the PORC and approved oy 3.
the app cpriate Superintenden: within 17. days' of i=plementation.
l Procedures for personnel radia: ion protection shall be prepared cen-d.
sis:en: vith the require =ents of 10 CTR Par: 20, and shall be ap-proved, _ain:ained, and adhered to for a*1 opera:icas involving personnel radiation exposure.
Respira:ory protective equip =en: sha'11 be provided in accordance vi:h 10 CFR 20'.103.
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Amendment No. 25
f..
7.5-2 Startup reports shaIl be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or co==ence=ent of co=nercial power operation, or (3) 9 nonths following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of co=mercial power operation),
supplementary reports shall be submitted at least every three months until all three events have been completed.
b.
Annual Occupational Exposure Report A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mre=/ year and their associated man-rem exposure according to work and job functiors, e.g., reactor operations and surveillance, in-service inspection, routine maintenance, special maintenance (describe maintenance),
waste processing, and refueling. The dose assignment to various duty fune-tions =ay be esti=ates based on pocket dosimeter, TLD, or fil= badge Small exposures totaling less than 20% of the individual i
i
=e as ure:ents.
total dose need not be accounted for. In the aggregate, at least 80% of the whole body dose received from external sources shall be assigned to specific =ajor work functions.
Amendment No. 25
7.5-3 l
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c.
Monthly Operating Report A routine operating report covering the operation of the unit during the previous month shall be submitted prior to the fifteenth calendar day of the following mo'ntn. Submittal shall be to the Director, Office of Inspection and Enforcement, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, ATTN:
Document Control Desk with a copy to the appropriate NRC Regional Administrator.
Each monthly operating report shall include:
A narrative su= mary of operating experience during the report period 1.
relating to safe operation of the facility, including major safety-related =aintenance.
Report of.any single release of radioactivity or radiation exposure 2.
which accotnes for more than 10% of the alloyable annual values.
3.
Indications of failed fuel resulting from irradiated fuel exa=ina-tions, co=pleted during the report period.
4.
The monthly statistical infor=ation contained in Regulatory Guide c
1.16.
Amendment No. 25
7.5-4
.8 7.5.2 Reportable Occurrences Reportable occurrences, including corr.ective actions and measures to prevent reoccurrence, shall be reported to.the NRC.
Supplemental reports may be required to fully describe final resolution of occurrence.
In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
Prompt Notification with Written Followup a.
The types of events listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile trans-l mission to the appropriate NRC Regional Administrator or his designee no later than the first working day following-the event, with a written followup-report within two weeks.
A copy of the confirmation and the written foi,lowup report shall also be sent to the Document Control Desk, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555. The written followup report shall include, as a minimum, a completed copy of a Amendment No. 25
7.5-5 licensee event report form, and shzll be supp.lemented, as needed, by additional narrative, material to provide complete explanation of the circumstances surrounding the event.
1.
Failure of the reactor protection system or other systems subject to limiting safety-system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety-system setting in the Technical Specifications or failure to complete the required protective function.
Note:
Instrument drif t discovered as a result of testing need not be reported under this item but may be reportable under items a.5., a. 6., or b.1., below.
2.
Operation of the unit or affected systems when any parameter or operation subject to a li=1 ting condition is less conservative than the least conservative aspect of the limiting condition for operation established in the Technical Specifications.
Note:
If specified action is taken when a system is found to be operating between the most conservative and the least l
conservative aspects of a limiting condition for operation i
listed in the Technical Specifications, the limiting condi-tion for operation is not considered to have been violated and need not be reported under this ites, but it may be reportable under item b.2. below.
3.
Abnormal degradation discovered in fuel cladding or the reactor coolant pressvre boundary.
Amendment No. 25
7.56 Note: Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifica-tions need not be reported under this item.
1 4
Reactivity anomalies, involving disagree =ent with the predicted value of reactivity balance under steady-state conditions during power operation, ' greater.tfan or equal to 1 ik/k; a calculated reactivity balance indicating a shutdown margin less cor.servative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if sub-critical, an unplanned reactivity insertion of more than 0.5% ik/k; or occurrence of eny unplanned criticality.
5.
Failure or malfunction of one or more cc=ponents which prevents or could prevent, by itself, the fulfillment of the functional require =ents of syste=(s) used to cope with accidents analyzed in the FSAR.
6.
Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfill:ent of the func-tional require =ents of syste=s required co cepe with accidents analyzed in the FSAR.
For ite=s a.5. and a.6. reduced red.andancy that does not Note:
result in a loss of system funecion need not be reported under this section but =ay be reportable under items b.2.
and b.3 helow.
A=end ent No. 25
7.5-7 Conditions arising from natural or man-made events that, 7.
as a direct result of the event, require plant shutdown, operation of safety syste=s, or other protective measures l
required by the Technical SpecificatYons.
8.
Errors discovered in the transient or wecident analyses, or in the methods used for such analyses as described in I
the FSAR or in the bases for the Technical Specifications that. ave or could have per=itted reactor operation in a h
manner less conservative than assumed in the analyses.
Perfor=ance of structures, syste=s, or components that 9.
requires re=edial action or corrective measures to prevent operation in a manner less conservative than that assumed in the accident analyses in the FSAR or Technical Specifi-cations bases; or discovery during plant life of conditions not specifically considered in the FSAR or Technical i
l Specifications that require remedial action or corrective l
measures to prevent the existence or develop =ent of an l
unsafe condition.
This item is intended to provide for reporting of potentially Note:
generic problems.
I b.
Thirty Day %'ritten Reports The reportable occurrer.ces discussed below shall be the subject of written reports to the appropriate NRC Regional Administrator within thirty days of occurrence of the Amend =ent No. 25
7.58 A copy of tMe written report shall also be sent to event.
the Document Control Desk, U. S. Nuclear Regulatory Commission. Washington, D. C. 20555.
copy of a licensee event repert for=.
ikfor=ation provided on the licensee event report for= shall be supple =ented, as needed, by additional narrative =aterial to provide co=plete explanation of the circe = stances surrounding the event.
1.
Reactor protection syste= or engineered safety feature instru=ent settings which are found to be less conserva-tive than those established by the Technical Specifications, but which do not prevent the fulfill =ent of 'th.e functional require =ents of af fected systc=s.
2.
Conditions leading to operation in a degraded = ode per=itted by a IL=iting condition for operation or plant shutdown required by a li=iting condition for opera tion.
Routine surveillance testing, instrument calibration, or Ne e:
preventative =aintenance which require syste: configura-
[
tions, as described in ite=s b.1. and b.2.,need not be reported except where test results the=s, elves reveal a degraded =ede as described above.
3.
OEserved i= adequacies i= the i=ple=entation of administrative i
or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection syste=s or engineered safety feature systems.
A=endment No. 25
7.5-9 4.
Abnormal degradation of systems other than those specified in~1 tem a.3. above designed to contain radiorctive material resulting from the fiss. ten
- process, Sealed sources or calibration sources are rot Note:
included under this item. Leakage of valve packing or gaskets within the limit for identified leakage set forth in the Technical Specifications need not be reported under this item.
7.5.3 Environmental Qualification A.
By no later than June 30, 1982, all safety-related electrical-equipment in the facility shall be qualified in accordance with the, provisions of:
Division of Operating Reactors " Guide-lines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines; or, NUREG-0558 " Interim Staff Position on Environmental Qualtfication of Safety-Related Electrical Equipment". Dect-ber 1979, to the extent applicable to a gas cooled reactor. Copies of these documents are attached to Order for Modificatien of License No. OPR-34 dated October 27, 1980.
By no later than December 1,1980, complete and auditible B.
records must be available and maintained at a central location which describe the environmental qualification method used for Amend:ent No. 25
1 7.5-10 all safety-related electrical equipment in sufficient detail to docunent the degree of compliance with the DOR Guidelines or NUREG-0588, to the extent app,licable to a gas cooled Thereafter, such records should be updated and reactor.
ceintained current as equipment is replacid, further tested, or otherwise further qualified.
C Amendment No. 25