ML20212K388
| ML20212K388 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/15/1985 |
| From: | Houston D NRC - TECH SPEC REVIEW GROUP |
| To: | NRC - TECH SPEC REVIEW GROUP |
| Shared Package | |
| ML20209E496 | List:
|
| References | |
| FOIA-85-511 NUDOCS 8608220042 | |
| Download: ML20212K388 (5) | |
Text
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July 15, 1985 ( i Docket No. 50-458 ) NOTE TO: File ! FROM: Dean Houston, TSRG:DL
SUBJECT:
RIVER BEND TECHNICAL SPECIFICATION NOTE OF i JULY 9,.1985 } In my note of July 9,1985, I made reference to errors, discrepancies and i problems in the current River Bend Technical Specifications that allegedly- , j 3 were known to GSU staff members but were not pursued for resolution due to ! i time constraints. During the noted conversation, the source was specifically asked about details concerning these areas and the question was evaded and- , ! no specific examples were furnished. The basis' for my concern is that GSU has already demonstrated a weakness in their certification (under oath) of-l the Final Draft. Following that certification, Bob 8enedict and I issued i j memorandums, dated May 17 and 21,1985, respectfully, in which we identified
- numerous errors that were not uncovered in their review (Enclosures 1 and 2).
j To the best of my knowledge, I am unaware of specific details that would ; indicate errors, deficiencies, or problem areas in tte current Technical i Specifications. However, it would not surprise me if there were. For i example, in Amendment 20 to the FSAR, that has been the basis of our review, * { dated late June 1985, eight new ventilation systems and one new building were identified for the first time. While it may be true that these systems did not affect the Technical Specifications, it is an indication that the FSAR has j not truly reflected the as-built plant. If similar discrepancies exist for safety systems, the Technical Specifications could be in error. : 1 1
- As further evidence of uncertainties in the ability of the utility to review Technical Specifications, note the differences between their markup from the ,
i certified copy dated April 26, 1985, and a markup dated June 28, 1985 (Enclosures 3 and 4). Three key instrumentation setpoints for the HPCS system were before? changed, hopefully the =last set is correct but why wasn't the matter caught' { , }' One area that has previously been discussed and withdrawn that might represent I i a problem Fuel Handling. is in Technical Specification 3.6.1.2. Primary containment Integrity - i As written, all primary containment penetrations required to , be closed during accident conditions are closed by some mechanism. GSU had ' j infomally requested that some of these penetrations be put into operable status rather than closed. Their request was conditioned that if it took t l some time to review and approve, that we should forget it. Since pytting valves into other than a closed category would have reopened the review of i-l this area, we did not grant the change or pursue it further. There is some B indication that the GSU concept for this Technical Specification is fine in j principle, but may not work in the real world. This may be an area that GSU should be pursuing now. 1 B600240042 POR FOIA 860015 ' ' ' PLETTIN85-511 PDR y , -
'l File -
2- July 15, 1985 The extent of errors and discrepancies is not known other than noted above.
& ??---
Dean Houston Technical Specification Review Group, DL
Enclosure:
As stated
- cc: D. Crutchfield ,
M. Virgilio e S 9 e O
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File July 15, 1985 The extent of errors and discrepancies is not known other than noted above. Original signed by ' Dean Houston i Technical Specification Review Group DL
Enclosure:
As stated cc: D. Crutchfield M. Virgilio 1 . ~ DISTRIBtlTION TSRG Rdg . DHouston l; i NintMel a D . 3-( ,
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,/ July 15, 1985 Docket No. 50-458 NOTE TO: File FROM: Dean Houston, TSRG:DL
SUBJECT:
RIVER BEND TECHNICAL SPECIFICATION NOTE OF ' JULY 9, 1985 In my note of July 9,1985, I made reference to errors, discrepancies and probleas in the current River Bend Technical Specifications that allegedly were known to GSU staff members but were not pursued for resolution due to time constraints. During the noted conversation, the source was specifically asked about details concerning these areas and the question was evaded and no specific examples >were furnished. The basis for my concern is-that GSU has already demonstrated a weakness in their certification (under oath)_of the Final Draft. Following that certification, Bob Benedict and I issued memorandums, dated May 17 and 21,1985, respectfully, in which we identified numerous errors that were not uncovered in their review (Enclosures 1 and 2). To the best of my knowledge, I am unaware of specific details that would ' indicate errors, deficiencies, or problem areas in the current Technical Specifications. However, it would not surprise me if there were. For , example, in Amendment 20 to the FSAR, that has been the basis of our review, dated late June 1985, eight new ventilation systems and one new building were. identified for the first time. While it may be true that these systems did not affect the Technical Specifications, it is an indication that the FSAR has not truly reflected the as-built plant. If similar discrepancies exist for safety systems, the Technical Specifications could be in error. - As further evidence of uncertainties in the ability of the utility to review Technical Specifications, note the differences between their markup from the certified copy dated April 26, 1985, and a markup dated June 28, 1985 (Enclosures 3 and 4). Three key instrumentation setpoints for the HPCS system were changed, hopefully the last set is correct but why wasn't the matter caught before? One area that has previously been discussed and withdrawn that might represent a problem is in Technical Spectfication 3.6.1.2, Primary Containment Integrity - fuel Handling. As written, all primary containment penetrations required to be closed during accident conditions are closed by some mechanism. GSU had infonnally requested that some of these penetrations be put into operable status rather than closed. Their request was conditioned that if it took some time,to review and approve, that we should forget it. Sinceputting valves into other than a closed category would have reopened the review of this area, we did not grant the change or pursue it further. There is some indication that the GSU concept for this Technical Specification is fine in principle, but may not work in the real world. This may be an area that GSU should be pursuing now.
~
File
- 2- July 15, 1985 m.
The extent of errors and discrepancies is not known other than noted above.
& a / _ h.
a Dean Houston Technical Specification Review Group, DL
Enclosure:
As stated cc: D. Crutchfield M. Virgilio t O
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.. uz [+muc - o, UNITED STATES 8 3 .c ,p, NUCLEAR REGULATORY CO* ilSSION WASHINGTON. D. C. 20555
%, ...../ May 21, 1985 MEMORANDUM FOR: Edward J. Butcher, Group Leader Technical Specification Review Group, DL FROM: R. A. Br. edict, Reactor Engineer Technical Specification Review Group DL SUB1ECT:
MORE CEFICIENCIES IN REGARD TO GSU CERTIFICATI0t' . OF RIVER BEND TECHNICAL SPECIFICATIONS (FINAL DRAFT) Further to Dean Houston's May 17, 1985 memorandum to you on this same subject, there are an additional 38 pages in which editorial errors were not identified by GSU. The missed errors are circled on the enclosed pages. . . All told, 79 pages out of 509 had editorial errors that GSU missed. R. A. Benedict, Reactor Engineer Technical Specification Review Group, DL
Enclosure:
As stated cc: D. Crutchfield D. Houston S. Stern 't e O
2.1 SAFETY LIMITS 2 BASES l
2.0 INTRODUCTION
i The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel riamage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-b'ack approach is used to establish a Safety Limit such that the MCPR is not less than 1.06. MCPR greater than 1.06 represents a conservative margin relative to the conditions reeJired to maintain fuel cladding integrity. The fuel cladding is one of tle physical barriers which separate the radioactive saterials from the environs. The integrity of this cladding barrier is . related to its relative freedom from perforations or cracking. Although some - corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurab]e. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use i C.,relateocracking,thethermallycausedcladdingperforationssignalath incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is tion defined boiling,with MCPR a margin of 1.0.to the condition which would pr'bduce onset of transi-These conditions represent a significant departure from the condition intended by design for planned operation. 2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calculations flow. at pressures below 785 psig or core flows less than 10% of rated other means. This the Therefore, fuel cladding integrity Safety Limit is established by is done by establishing a limiting condition on core THERMAL POWER with the following basis. i Since the pressure drop in the bypass region ially all elevation head, the core pressure drop at low power and flows ill A 28 x 0 be greater than 4.5 psi. Analyses show that with a bundle flow of s/hr, bund and ha a, alue of 5 ps re drop is nearly independent of bundle power . ill greater tha 28 us, the bundle flow with a 4.5 psi driving head l bs/hr. Full scale ATLAS test data taken at og
- pressures from 14.7 sia
~
power at this flow i )r psia indicate that the fuel assembly critical ately 3.35 MWt. With the design peaking facters, [tM.Y this corresponds to a THIRMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a below THERHAL 785 POWER limit of 25% of RATED THERMAL PCWER for reactor pressure psig is conservative. APR 2 61965 RIVER BEND - UNIT 1 B 2'1 O
- l
,,. REACTIVITY CONTROL SYSTEMS ' 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity equivalence of the difference between the actual R00 DENSITY and the predicted ROD DEN 5ITY shall not exceed 1% delta k/k. APPLICABILITY: OPERATIONAL CONDITIONS I and 2. ACTION: Withthereactivityequivalence[ e ceeding 1% delta k/k:
- a. Within 12 hours perfo an an is to determine and explain the cause d of the reactivity differe ; operation may continue if the difference is explained and corrected.
- b. ~ Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
(.s SURVEILLANCE REQUIREMENTS 4.1.2 The reactivity equivalence of the difference between the actual ROD DENSIIY and the predicted ROD DENSITY shall be verified to be less than or equal to 1% delta k/k:
- a. During the first startup following CORE ALTERATIONS, and b.
At least once per 31 effective full power days during POWER OPERATION. 4
~.
p t. 6 19BS F.IVER BEND - UNIT 1 3/4 1-2
/
REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1. 3. 2 The maximum scram insertion time of each control rod from the fully withdrawn position, based on de energization of the scram pilot valve solenoids as time zero, shall not exceed the following limits: Maximum Insertion Times to Notch Position (Seconds) Reactor Vessel Dome ' Pressure (psic)* 43 29 13 950 M M M 1050 0.32 0.86 1.57 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. . ACTION: - a. With the maximum scram insertion time of one or more control rods exceeding the maximum scram insertion time limits of Specification 3.1.3.2 as deter-mined by Surveillance Requirement 4.1.3.2.a or b, operation may continue provided that:
- 1. For all " slow" control rods, i.e. , those which exceed the limits of C'. Specification 3.1.3.2, the indisidual scram insertion times do not exceed the following limits:
Maximum Insertion Times Reactor Vessel Dome to Notch Position (Seconds) Pressure (psic)* 43 29 13 950 1050 0.38 M M 0.39
. 1.14 2.22
- 2. For " fast" control rods, i.e. , those which satisfy the limits of Specification 3.1.3.2, the average scram insertion times do not exceed the following limits:
Maximum Average Insertion Times to Notch Position (Seconds) Reactor Vessel Dome Pressure (psic)* 43 29 13 950 0.30 0.78 1.40 1050 0.31 0,84 1.53 3. g The sum of " fast" control rods with individual scram insertion times in excess not exceedof5. the limits of ACTION a.2 and of " slow" control* rods does "For intermediate reactor vessel dome pressure, theriteri scram time % l determined by linear interpolation at each notch position. fis Q
.- g RIVER BEND - UNIT 1 3/4 1-6 APR 2 6 E
L/ REACTIVITY CONTROL SYSTEMS R00 PATTERN CONTROL SYSTEM a uns ab Na =- - LIMITING CONDITION FOR OPERATION 3.1.4.2 The rod pattern control system (RPCS) shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS I and 2*# . ACTION: a. With the RPCS not satisfied andinoperable with: or with the requirements of ACTION b, below
- 1. fo ] ,
THERMAL POWER less than or equal 320% rod movement shall not be p i RATED THERMAL POWER control xcept by a scram. ( ES
- 2. n THERMAL POWER greater than 20% of RATED THERMAL POWER control' rod withdrawal shall not be permitted.
b. With an inoperable control rod (s), OPERABLE control rod movement m continue by-bypassing provided that: the inoperable control rod (s) in the RPCS 1. With one control rod inoperable due to being immovable, as a known to be untrippable, this inoperable control ro bypassed in the rod gang drive system (RGDS) and/or the rod action control system (RACS) provided that the SHUTDOWN MARGIN Specification 3.1.1.has been determined to be equal to or greater than 2. With up to eight control rods inoperable for causes other than addressed in ACTION b.1, above, these inoperable control rods may be bypassed in the RACS provided that:
- a)
The control rod to be bypassed is inserted and the direc-tional control valves are disarmed either:
- 1) Electrically, or *
- 2) Hydraulically by closin water isolation valves.g the drive water and exhaust b)
All inoperable control rods are separated from all other inoperable all control rods by at least two control cells in directions, c) There are not more than 3 inoperable control rods in any RPCS group. *
"See Special Test Exception 3.10.2 .
1
# Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control r is permitted for the purpose of determining the OPERABILITY e to of the RPCS withdrawal RIVER BEND - UNIT 1 of control rods for the purpose of bringing the reactor .
to critic 3/4 1-17 APR 2 61985 l
TABLE 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS AND SENSOR LOCATIONS INSTRUMENTS RANGE OPERABLE 1. Triaxial Time-History Accelerogra hs
- a. Reactor 81dg
- b. 70'0" '
0 1 1.0 g ~~ Reactor Bldg Ex Shield Wall 1 EL 232'0"
- c. Reactor Bldg Drywell EL 151'0" 0 2 1.0 g 1
- d. O t 1. 0 g Free Field - Grade Level 0 t 1.0 g I
1
- 2. Triaxial Peak Accelerographs *
- a. Reactor 81dg SLCS Stora 0 1 10.0 g
- b. Reactor Bl_dg - RHR Inj.ge Tank .1
- c. Piping 0 1 10.0 g Aux. Bldg Service Water Piping 0 1 10.0 g I
1 3. Triaxial Seismic Swite
- a. Reactor Bldg Ma Eh 70' '
O.025 to 0.25 g 1(*) [
- 4. Triaxial Response-Spec Recorders
- a. Reactor 81dg Mat EL 70'0"
- b. Reactor Bldg Floor EL 141'0" 0i2g I I")
- c. Auxiliary Bldg Mat EL 70'0" 012g I
- d. Auxiliary 81dg Floor EL 141'0" 0i2g I 012g 1
(*)With reactor control room indication a on. and annunci ti ' I o I RIVER BEND - UNIT 1 3/4 3-71 LK t 6 G85
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_ TABLE 3.3.9-1 (Continued) FINAL DRAFT l ACTION 150 - a. Withonechannelinoperagle,placetheinoperablechannel in the tripped condition within one hour or declare the associated system inoperable. b. With system more than one channel inoperable, declare the associated inoperable. l ACTION 151 - a. i With the number of OPERA 8LE channels one -lestthan required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERA 8tE status within 7 days or be in at least within HOT SHUTDOWN the following 24 hours. within 12 hours and COLD SHUTDO b. 4
! With the number of OPERA 8LE channels two less than required
- by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE status within 12 hours 72anhours or be in at least NOT SHUT 00WN within SHUTDOWN within the following 24 hours.
i ACTION 152 - Declare t soc inoperabl lated Containment Ventilation System - ! ACTION 153 - ' 1. g 4 With the number of OPERA 8LE channels one less than required by the Minimum OPERA 8LE Channels requirement,
! k. restore the inoperable channel to OPERA 8LE status within 7 days or be in at least STARTUP within the next 6 hours. ,
2. With the number of OPERA 8LE channels two less than
' required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to .
! OPERABLE status within 72 hours or be in at least .
- STARTUP within the next 6 hours.
i
#Provided this does not actuate the system.
- i
< i , 1 i {~ - i I i RIVER BEND - UNIT 1 3/4 3-110 APR 16 #
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4 6 1 1 I i 1600-A - SYSTEM MYDROTEST LIMIT WITM FUE L IN VESSE L A 8 - NONWUCLE Am HE ATING LIMIT geon C
- 8.
- I I bl C - NUCLE AR (CORE CRITICALI LIMIT -
RE $ S A5ED ON G.E. CO.8wA LICEN51NG
- l TOPICAL REPORT NEDO 2t?78A
{ AFTER g e5MtFT l8 A*. 8*. C'.- CCRE SE LTtyMITS AFTtR
. e AN A55 JMED 48 F TEMP.5MIFT C 1200 FROM mN INITIAL PLANT RT l
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NDT 5 I I l OF 8*F ~ g 3 8 VESSEL Ol5 CON-i l d TINUITY ' l l
- N y
1000 - LIMITS l I i I ~ 5 I C - 1 I I
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E 800 - s sb4L& W
- NEw iOCFRs0 iistai 7 APPENDI A G LIMIT 5 400 -
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- BOLTUP 312 oug LIMIT A 70 8, 8 gon' - C L
I i 1 I I 0 100 200 1 300 400 500 600 700 MtNLMUM REACTOR VE55EL METAL TEMPER ATURE t 't I l l i NOTE CURVES A.8 & C ARE PREDICTED l
- TO APPLY AS THE LIMITS FOR 11 YE ARS 18.8 E FPV) OF OPER ATION. .
q$, MINI '[MEA '
/ (TOR PRESSURE Fi gure i3. 4. 6.,J-1 VESSEL b METALf TEPIPERATURE [
t RIVER BEND - UNIT 1 3/4 4-22 APR 2 6 sas -
PLANT SYSTEMS
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t l l= a *Asd SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months or (1) after any structur painting, fire.or chemical release cating with the subsystem by: 2 followingin any muni-ventil 1. Verifying that the subsystem satisfie andbypass1akagetestingacceptance[criterithe 0.05% and of hess than in D e penetration tions C.5.a. C.S.c and q.5.d of Regula'ihe e latory' Posi- test proced X
)(
- 2. March 1978, and th(fystem flow rate is 1.52 4000 Revision cfm +,10%
- 2. .
X Verifying within 31 days after removal that analysis laboratory a of a representative carbon sample Regulatory Position C.6.b of Regulatory Guide 152 obtained in accordanc ' March 1978, meets the laboratory testing criteria o Revision 2, Position C.6.a of Regulatory ~ Guide 1.52, Revision 2f Regulato f for a methyl iodide penetration of less than
~
, March 0175%; 1978, and '
3. Verifying a subsystem flow rate of 4000 cfm g+ sub- 10% durin
- d. system operation when tested in accordance with ANSI A N5
. \'
A' fter every 720 hours of charcoal adsorber erifying operation
, within 31 days after removal that a laboratory analysis a repre-of sentative ton carbon sample C.6.b of Regulatory obtained Guide 1.52, Revisionin2accordance osi- with Regu Regulatory Guide 1.52, Revision 2the laboratory . .a of testing c penetration of less than 0.175%. . March 1978, for a methyl iodide e.
At least once per 18 months by:
- 1. ,
and charcoal adsorber banks isHEPA less than 7 in filters while operating the subsystem at a flow rate of+ 10%. 4 00 cfmaterM X e e RIVER BEND - UNIT 1-3/4 7-6 APR 2 61985 L,,.,_..,...,... ..;. - " ' * - ~ ' ~
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PLANT SYSTEMS g SURVEILLANCE REQUIREMENTS (Continued) i Testing equipment failure during functional testing may invali-j date that day's testing and allow that day's testing to resume anew at a later time, providing all snubbers tested with the j The representative sample selected for the f . i plans and shall before reviewed be randomly beginningselected th from the snubbers of each typ!I ensure as far as practical that--g geertj
- z. p The review shall various configurations, operating enviro representative of the y nments, range of size, and capacity of snubbers of each type. Snubbers placed in the same locations as snubbers test shall be ratested at ich failed the previous functional but shall not be included timesample the of theplan.
next functional test y functional testing, additional sampling is requi d due to fail- I during the y: ure of only one type of snubber, the functional testing f results / i shall be reviewed at the time to determine if additional samples should betesting. functional limited to the type of snubber which has failed the 4) 88 snubbers shall be functionally tested.For each type Three (3) snubbers i of each ance crit type are allowed not meeting the functional test accept-is grea rt 3, the number of snubbers that failed the test ( equal t 2 in additional sample of that type of snubber A-3) shall be functionally tested, where "A" is thethe of totr Iresen number tiveoff snubbers failed during the functional test sample. tlamL N the funct test of the resample, an additional sample of 22 snubbers of the same type shall be functionally tested. The all snubbers of that type have been functionally t . f. Functional Test Acceptance Criteria The snubber functional test shall verify that: 1) Activation (restraining action) is achieved within the specified range in both tension and compression; 2) For mechanical snubbers, the force required to initiate or main-tain directionsmotion of the and of travel; snubber is within the specified range in both 3) 1 For snubbers uous load specifically required not to displace under contin-2 displaceme,nt.the ability of the snubber to withstand Toad without meters other than those specified if those results to the specified parameters through established methods . RIVER BEND - UNIT 1 3/4 7-13 APR 2 6125 seem@ 8" *
'.., - , - +-
PLANT SYSTEMS HALON SYSTEMS LIMITING CONDITION FOR OPERATION
=
- 3. 7. 6. 3 system shall be OPERABLE with the storage tanThe Halon main c charge weight and 90% of full charge pressure.ks having at least 95% of full APPLICABILITY:
to be OPERABLE. Whenever equipment protected by the Halon u redsystems is ACTION:
- a. ;
With the above fire watch patrol. required Halon system inoperable
, establish an hburly b.
The provisions of Specifications 3.0.3 and 3.0 . 4 are not a pplicable. _ SURVEILLANCE REQUIREHENTS 4.7.6.3 a. The above required Halon system shall be demonstrated : O At least once per 31 ca operated or automatic, ys by verifying that each valve, manual, power
- b. in the flow path is in its correct position.
At least once and pressure, per 6 months by verifying Halon storage tank weight c. At least once per 18 months by: 1. Verifying the system actuates, manually n a l receipt of a simulated actuation signal . omatically, upon Halon bottle initiator valve actuation, a ctupl Halon release, burning may be excluded from the test), lectro-thermal link'g M !'
- 2. .
Performance no blockage. of a flow test through headersoand assure nozzles t { j t O i RIVER BEND - UNIT 1 ! 3/4 7-23 APR 2 6 %
e ELECTRICAL POWER SYSTEMS l , SURVEILLANCE REQUIREMENTS (Continued) sta),e c- enter voltage and frequency shall be maintained within ese limits d ging this test. Within 5 minutes after com-pleting thig 24-)our test, perform Surveillance Requirement g
- 4. 8.1.1. 2. y. 4. a )2 ) and b)2)*. ,
>W p
- 9. MiYn; L;. he auto-connected loads to each diesel generator do not exceed for diesel generator330 kwIC.
for diesel generator A and B and 2600 kw 10. Verifying the diesel generator's capability to: a) Synchronize with the offsite power source while the geoerator is loaded with its emer tion of offsite power, gency _, loads upon a simulated restora-b)
~
_ Transfer its loads to the offsite power source, and c) Be restored to its standby status. 11. Verifying that ywith the diesel generator 4 operating in a test mode and connected to its bus, a simulat"ed ECCS actuation signal N overrides the test mode by (1) returning the diesel generator tostandbyoperationgand(2)automaticallyenergizestheemer- X gency loads with offsite power. 12. Verifying that the automatic load sequence timers are OPERABLE with designthe interval interval between for diesel eachIAload generators and block
- 18. within i 10% of its 13.
Verifying that the following diesel generator lockout features - prevent diesel generator starting only when required: a) ForDieselGeneratorsIAand18:fcot 1) 2) Diesel control panel loss o Starting air pressure below 50 psi. n rol power. [
- 3) Stop solenoid energized.
4) Diesel in the maintenance mode (includes barring device engaged).
- 5) Overspeed trip device actuated.
6) Generator backup protection lockout relay tripped. b) For Diesel Generator IC:
- 1) Diesel generator lockout relays not reset.
2) Diesel engine mode switch not in "AUT0" Ptsition.
"If Surveillance Requirements 4.8.1.1.2.e(4).a)2 and b)2) are not setisfactorily completed, it is not necessary to repeat the preceding 24 hour test. Instead, .
the diesel generator may be operated at 3130 kw for diesel generators IA and 18 and 2600have temperatures kw stabilized. for diesel generator 1C for one hour or until operating RIVER BEND - UNIT 1 3/4 8-7 APR 2 61985
RADI0 ACTIVE EFFLUENTS LIQUID HOLOUP TANKS k .
~
LIMITING CONDITION FOR OPERATION _ i 3.11.1.4 outdoor tank shall be limited to less than or eThe quantity of radioac tritium and dissolved or entrained noble gases. qual to 10 curies, excluding _ APPLICABILITY: At all times. ACTION: a. With the quantity of radioactive material in any of the above unprotected outdoor tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank; within . 48 hours reduce the tank contents to within the limit; and edescrib K the events active Efflue,nt leading ReleasetoReport. this condition in the next Semiannual Radio- y b, The provisions of Specifications 3.0.3 and 3.0.4 are not applicable . SURVEILLANCE REQUIREMENTS b 4.11.1.4 The quanti ofr unprotected outdoor tank sha 1 be determined analyzingarepresetatthe 7 days when radioact to be within mple of the tank's contents at least once per erlais are being added to the tank. Mg @ e e. j . . . l RIVER BEND - UNIT 1 3/4 11-6 ! APR 2 61985
. . . . . . . . . . . , - , - - - - - - - - - - ~ " ~ '~
L.
. RAD 10 ACTIVE EFFLUENTS '
L l. * '7m 3/4.11.2 GASEOUS EFFLUENTS EmEWth y *' d a J: 8 OOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 . effluents from the site to areas at and beyond the SITE i Figures 5.1.1-1 and 5.1.3-1) shall be linited to the following: ,
- a. For noble gases:
body and less than or equal to 3000 arems/yr , and to the skinles 1 b. For iodine-131, in particulate for iodine-133, form with half lives greafor tritium, and for all r dionuclide or equal to 1500 meems/yr to any or,gan. ter than 8 days: ss tha,n [ APPLICABILITY: At all times. g ACTION: With the dose rate (s) exceeding the above limits release rate to within the above limit (s).
, without delay restore the
_ SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The se rate 7 } determined to be within th to noble gases in gaseous effluents shall be NM.p$b and parameters in the above limits in accordance with the methodology CM K 4.11.2.1.2 , The dose ratej due'to iodine-131, iodine-133, tritium, and all radionuclides a in particuTate form with half lives greater than 8/ days in nce with the me,thodology and parameters tive samples and performing analyses in accordance with the sa a-inXthe 0
>(
analysis specified in Table mpling and i 4.11.2.1.2-1. 4 9 I e
- RIVER BEND - UNIT 1 3/4'11-7 APR 2 61985 k
RADIOACTIVE EFFLUENTS , GASEOUS RADWASTE TREATMENT U LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM shall be in operation. APPLICABILITY: Whenever the main condenser air ejector system is in operation. ACTION: ' p
- a. With GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTE in operable for more o
than 7 days, prepare and submit to the Commissi n within days,
' pursuant to Specification 6.9.2, a Special Repor =t ludes the following information: -
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability.
- 2. Action'(s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent b.
recurrence. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. [ SURVEILLANCE REQUIREMENTS 4.1 hh ~'
$ ul M**
4 The instruments specif wh ver t h ' denser air ejector M n operationyto ensure that thed [X in th GASE0 S RADWASTE TREATMENT (0FFGAS) SYSTEM is functioning. - 1 I l i RIVER BEND - UNIT 1 3/4 11-13
9 TABLE 3.12.1 N RADIOLOGICAL ENVIRONNENTAL NONITORING PROGRAM E i 5 Number of
, Representative c Exposure Pathway Samples and sampilng and and/or Sample _ Sample locations, Type and Frequency Collection Frequency of Analysis H
- 1. DIRECT RADIATION '40 routine monitoring stations
. Quarterly' Gauuna dose quarterly.
(DRI-DR40) either with two or more dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:
. an i r ing T stations, one in
} eac met logi tal sector in the g gene al ao the SITE BOUNDARY
'? (DRI- - ,
W an outer ring of stations, one in each meteorological sector in , the.6- to 8-km range from the i site (DR17-DR32); the balance of the stations Y w'" (DR33-DR40) to be placed in special interest areas such
* ~ m" as population centers, nearby c:23 residences, schools, and-in 1 W or 2 areas to serve as control stations. g
- 2. AIR 90RNi v.)
,3
.".; J J i
Radiofodine and . Samples from 5 locations (Al-AS): Continuous sampler Particulates Radiciodine cannister: ' *'8 [ 3 samples (Al-A3) from close operation with sample collection weekly,' or I-131 analysis weekly. W to the 3 SITE BOUNDARY locations, more frequently if h y in different sectors, of the required by dust Particulate Sampler: p_ , highest calculated annual average loading. Gross beta radioactivity 4c *~"-vel D/Q. a -lysis following (C
~ ~ -~
TABLE 4.12.1-1 (Continued) FINAL DRen TABLE NOTATION It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posterori (after the fact) limit for a particular measure-
, ment.
_ Analyses shall be performed in such a manner that the stated LLDs f,willbe#unavoidablgsmallsamplesizes,thepresenceofinterfering I l nuclides, or other uncontrollable circumstances may render these LLDs,
<unachievable. In such cases, the contributing factors shall be identi- _
fied and described in the Annual Radiological Environmental Operating [ Report pursuant to Specification 6.9.1.7. o, S- LLO fer J..e - 4 e te- ---ai-e 7f e 1,,,a2-dhg wa;;7_--thu;3 2 ;;;, the u _a ; f : - -< , - + m m e ,. . . . , _ m . . . . m . a.eb;eal u.nler ro u.W e conhhiss. ' Occasinually hukgr L FluM%3 N) l93lh Mb-(
/
RIVER BEND - UNIT 1 3/4 12-12 6 7985
/
D APPLICABILITY
, FINAL DON 8Wd BASES 4.0.1 This specification provides that surveillance activities necessary to ensure the limiting Conditions for Operation are met and will be performed j
during the OPERATIONAL CONDITIONS or other conditions for which the Limiting Conditions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL CONDI-TIONS or other conditions are provided in the individual Surveillance Require-ments. Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification. 4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operationa.1 flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance; instead, it permits the more frequent performance of surveillar.ce activiti_es. The tolerance values, taken either individually or consecutively over three
- test intervals, are sufficiently restrictive to ensure that the reliability [ ,
associated with the surveillance activity is not significantly degraded beyond ' that obtained from t.he nominal specified interval. 4.0.3 The provisions of thi fication set forth the criteria for determination of compliance with ABILITY requirements of the Limiting
~
Conditions for Operation. Under th Leria, equipment,systemsorcomponents[1 are assumed to be OPERABLE if th ass ted surveillance activities have been satisfactorily performed within ified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still . meeting the Surveillance Requirements. 4.0.4 This specif' at'on ensures that surveillance activities associated ti j with a time intervalLimitingprior Condit' to n try fo Operation have been performed within the specified h nto an applicable OPERATIONAL CONDITION or other specified applicability ition. The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation. Under the terms of this specification, for example, during initial plant startup or following extended plant outage, the applicable surveillance activ- ,, i ities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status.
- i e
1 APR 2 61965 RIVER BEND - UNIT 1 B 3/4 0-2
\
REACTIVITY CONTROL SYSTEMS BASES CONTROL R005 (Continued) Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and,therefore3this check must be performed prior to achieving criticality after jN ' comkleting CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demonstration. In order to ensure that the control rod patterns can be followed and X therefore that other parameters are within their limits, the control rod # )( position $ndicationsystemmustbeOPERABLE. - The control rod housing support restricts the outward movement of a control . rod to less than 3 inches in the event of a housing failure. The amount of i rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage i to the primary coolant system. The support is not required when there is no .. pressure to act as a driving force to rapidly eject a drive housing. The required surveillance intervals are adequate to determine that the C components. rods are OPERABLE and not so frequent as to cause excessive wear on the sy i 3/4.1.4 ROD PATTERN CONTROL SYSTEM s ) tA Y-g i s The rod withdrawal limiter system input powe i nal o inates from the , first stage turbine pressure. When operating with t a bypa valves open, this signal indicates a core power level which is les the true i' core power. . Consequently, near the low power setpoint and high power setpoint of the rod pattern control system, the potential exists for non- ) conservative control rod withdrawals. Therefore, when operating at a
- 4 sufficiently high power level _.there small probability of violating fuel Safety Limits during a(licensino-basis rod withdrawal error transient. a +42.Xw . n .l?
To ensure that fuel Safety Limits are violated, this specification prohibits control rod withdrawal when a biased power signal exists and core power exceeds the specified level. , I I Control rod withdrawal and insertion sequences are established to assure that the maximum iniequence individual control rod or control rod segments which 8 are withdrawn at afy time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater ! than dropped20%atof RATED THERMAL POWER, there is no possible rod worth which, if pps jnrate
- e velocity limiter, could result in a peak enthalpy o c 7gm.
POWER is le than or equal t htfiring the RPCS to be OPERABLE when THERMAL 20% of RATED THERMAL POWER provides adequate
, pp i ,'
i control. - l RIVER BEND - UNIT 1 B 3/4 1-3 APR 2 61985 i l . . .. l l .-. . .. _ - _ .
/
REACTIVITY CONTROL SYSTEMS FINE DNT BASES ROD PATTERN CONTRm ev qEM (Continued) The RPCS (rovidef au)bmatic supervision to assure that willnotbew'thdraw[o inserted. 9 (
?
The analysis of the rod drop accident is presented in Section /15.j of the FSAR and the techniques of ganalysis are presented in a topical reportu K
":'.. - f , and two supplementtj..... - _ _ . . _ . .
The RPCS is also designed to automatically prevent fuel damage in the event
' of erroneous power operation. rod withdrawal from locations of high power density during higher A dual channel system is provided that, above the low power setpoint, -
restricts the withdrawal distances of all non peripheral control rods. This { restriction is greate,st at highest power levels.
; 3/4.1.5 ~ STANDBY LIQUID CONTROL SYSTEM The standby liquid control syst i
the reactor from full power to a col , peno o '-free desshutdown, a backup capability assuming that the for bringing
; withdrawn controi ro'ds remain fixed n he ated power pattern. To meet this ; ([ objective it is necessary to inject a tity of boron which produces a concen-tration of 660 ppa in the reactor core in approximately g90 to 120$ minutes. %
A minimum available quantity of 3542 gallons of sodium pentaborate solution
- containing a minimum of 4246 lbs. of sodium pentaborate is required to meet a i
shutdown requirement of 3% Ak/k. Thereisanadditionalallowanceof(150 ppm )( ' in the reactor core to account for imperfect mixing and the filling of oth)er piping systems connected to the reactor vessel. The time requirement was selec J ed to override the reactivity insertion rate due to cooldown following theg enon poison peak,and the required pumping rate is 41.2 gpm. The minimum , storage volume of the' solution is established to allow for the portion below / the pump suction that cannot be inserted. The temperature requirement is necessary to ensure that the sodium pentaborate remains in solution. With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to j continue for short periods of time with.the system inoperable or for longer periods of time with one of the redundant components inoperable. l 1. C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis I for Large BWR's," G. E. Topical Report NEDO-10527, March 1972 2. C.
'1972 J. Paone, R. C. Stirn and R. M. Young, Supplement I to NEDO;10527, July 3.
J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Co_res,"
- Supplement 2 to NE00-10527, January 1973 RIVER BEND - UNIT 1 APR 2 61985 B 3/4 1-4
_, ..-,,.-.-,--.,--,,_r . . . . . _ . . . , . - - . ~ . - _ , . . - . ._
POWER DISTRIBUTION LIMITS . {S-2o BASES i 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derive from the established fuel cladding integrity operational transients. Safety Limit MCPR of 1.06 and an analysis of abnormal
)<
For any abnormal o erating transient analysis r. i r % 4een with the initial condition of the reactor being at the steady state - operating limit, it is required that the resulting MCPR does not decrease below K setting given in Specification 2.2.the Safety Limit MCPR at any time during the To assure that the fuel cladding integrity Safety Limit is not exceeded [ i,
- sients have been analyzed to determine which result in the lar in CRITICAL POWER RATIO CPR . The type of transients evaluated were loss of flow, increase in pressur(e an)d power, pcsitive reactivity insertion, a temperature decrease. - The limiting transient yields the largest delta MCPR.
i When added to the Safety Limit MCPR of 1.06, the rpquired minimum operating limit MCPR of Specification 3.2.3 is obtained ane nt c h in re 3.2.3-1. The power-flow map of Figure B 3/4 2.3-1 sho J - operation. 4typicalregionfof ^ lant The evaluation of a given transient begin A. La N O
/' n
\ metersshowninFSARTable15.0-2thatareinputtoaGhre.m transient computer prog dynamicinitialbehavior para-fX described in NED0-24154 y . The code used to evaluate pressurization events is described in NED0-10802(2) and the program used in non pressurization events is MCPR form thedingle the input for further analyses of the thermalTy limiting % bu l NEDE-25149gannel.transientthermalhydraulicTASCcodedescribedin /
MCPR caused by the transient.The principal result of this evaluation is the reduction in The purpose of the MCPR, and pMCPR of Figures 3.2.3-1 and 3.2.3-2 is to define operating limits at other than rated core flow and power conditions. At of the less MCPR than and100% MCPR of rated flow and power the required MCPR is the larger value i f p at the existing core flow and power state. The MCPR s f t are that the established 99.9% MCPR to limitprotect requirementthecan core be assured, from inadvertent core flow increases [ The MCPR the correspon$s ing THERMAL were calculated POWER along th such tha/105% of-t for the maximum core flow r steam-flowg ontrol line, X the limiting bundle's relative power was adjus;ted" rated ntilkheMhRwasslightly above the Safety Limit. Using thisrelat we lated at different points along the 105%-of raundle power, the MCPAs were calcu-
-steam flow X
corresponding to different core flows. 3 / of core flow is defined as MCPRf . The calcuiated kPR bcontrol a giten point line ' i h Y RIVER BEND - UNIT 1 8 3/4 2-4 b?R 2 6198'c l .._ __ _ _ _ _ _ _ _ _ . _ . _ _ . . _ _ , _ . _ _ _ _ _ _ - _ . _ _ _ _ . _ _ _ _ , - _ , _ _ , _ _ _ . __
l- .
/1 FINAL DR2n l 3/4.3 INSTRUMENTATION !
BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION i I The reactor protection system automatically initiates a reactor scram to: I
- a. Preserve the integrity of the fuel cladding X j ,
- b. Preserve the integrity of the reactor coolant system j /C
- c. Minimize the energy which must be adsorbed following a loss-of-coolant 2
- accident, and b
- d. Prevent inadvertent critica .
S . g , Thisspecificationprovideste[atir itions [ ation.necessary ! to preserve the ability of the sys em per ora its inte ed etion even during periods when instrument chan s may out of se decause of main-tenance. When necessary, one channel may be made inoperable for brief intervals
- to conduct required surveillance.
The reactor pro'tection system is made up of four logic channels. The logic
- channels A(A1) and C(A2) comprise one trip system and the logic channels B(B1)
J , and D(B2) comprise the other trip system for determining compliance with technical specifications. Placement of either logic channel of a trip system in the tripped condition places the trip system in the tripped condition. The trip systems; y as defined above,are independent of each other. There are usually four instrument K channels (one is each logic channel) to monitor each parameter. The tripping of a logic channel in each trip system will result in a reactor scram. The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-
- plated within the time limit assumed in the safety analyses. No credit was
- l taken for those channels with response times indicated as not applicable. i
- Response time may be demonstrated by any series of sequential, overlapping or !
1 total channel test measuremen provided such tests demonstrate the total l channel response time as defi d. Sensor response tima verification may be )< l t demonstrated by either (1) inplace, onsite or offsite test measurements, or l (2) utiliting replacement sensors with certified response times. I W APR 2 61985 RIVER SEND - UNIT 1 8 3/4 3-1
, - - - . - - - - - - , - - , . . , - , , -.,,.----,,.r . - - - ,-,.en .r y .
INSTRUMENTATION BASES l 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY require-ments, trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Some of the trip settings ]g have tolerances explicitly stated where both the high ow values are crit < cal and may have a substantial / i effect on safety. The se oin s f other instrumentation, where only the high ' or low end of the setting a irect bearing on_ safety, are established at a level away from the noraal ope ting range to prevent inadvertent actuation of the systems involved. a"' h Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors i are connected. For D.C. operated valves a 3 second delay is assumed before , ] the valve starts to move. For A.C. operated valves, it is assumed that the A.C. pow ~er supply is lost and is restored by startup of the emergency diesel ! generators. In this event, a time of 10 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C. opeR - X gated valve is assumed; thus the signal delay (sensor response) is concurrent 7 )( i i with the 10 second diesel startup. The safety analysts considers an allowable inventory loss _in each' case which in turn determines the valve speed in conjunc-
. tion witn tne 10 second delay. It follows that checking the valve speeds and g ' the 10 second time for emergency power establishment will establish the response time for the isolation functions. However, to enhance overall system reliability and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as part of the ISOLATION SYSTEM RESPONSE TIME.
Operation with a trip set less conservative than its Trip Setpoint but ' within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trip Setpoint and the Allowable Value is equal to or less than the drif t allowance assumed for each trip in the safety analyses. l 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION t i The emergency core cooling system actuation instrumentation is provided
' to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the ,
! OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time. i Operation with a trip set less conservative than its Trip Setpbint'but within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. i RIVER BEND - UNIT 1 8 3/4 3-2 F "_- ___ _ --- - t ,
- - - - - - -- ~ --- - - - - - - - _ _ _ _ _ _ _ _-
INSTRUMENTATION . BASES ( - 3/4.3.4 RECIRCULATION PUMP 'R!o AC'.7 3N INSTRUMENTATION The anticipated transient without scram (ATVS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of I study events in General Electric Company Topical keport NEDO-10349, dated Maren 1971 and NEDO-24222, dated December 1979, and Section 15.8 of the FSAR. , The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of l the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the E0C-RPT is to recover the loss of thermal , margin which occurs at the end-of-cycle. The physical phenomenon involved is ; that the void reactivity feedback due to a pressurization transient can add . I positive reactivity to the reactor system at a faster rate than the control. rods add negative scram reactivity. Each EOC-RPT system trips both recircu- 1 l lation pumps, reducing coolant flow in order to reduce the void collapse in g i the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of . the turbine stop val'ves and fast closure of the turbine control valves. -- I A fast closure sensor from each of two turbine control valves provides F: input to the EOC-RPT system; a fast closure sensor from each of the other two QN turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop vaTves. The operation of alther logic will actuate the EOC-RPT system and - - trip both recirculation pumps. Each EOC-RPT system may be manually bypassed by use of a keyswit ch dministratively controlled. The manual bypssses and the automat ating ss at less than 40% of RATED THERMAL POWER are annunciated in th m. rol l The E0C-RPT system response time is the time assumed in the analysis between initiation of valve action and complete supp f the electric arc, i.e., 140 ms. Included in this itse are: the espo 'me of the sensor, the time allotted for breaker arc suppression and t e time of the system logic. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip-Setpcint and the Allowable Value is ecual to or less than the drift allowance assumec for each trip in the safety analyses. APR 2 61985 RIVER BEND - UNIT 1 6 3/4 3-3 _' 1
INSTRUMENTATION ! 8ASES
- 1 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION
', The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core j cooling equipment. j . i i Operation with a trip set less conservative than its Trip Setpoint but within ita specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the trift allowance assumed for each trip in the safety analyses. 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-
; ments of the specifications in Section 3/4.1.4, Rod Pattern Control System, Section~3/4.2, Power Distribution Limits and Section 3/4.3, Instrumentation.
The trip logic is arranged so that a trip in any one of the inputs will resul.t j in a control rod block. Operation with.a trip set less conservative than its Trip Setpoint but l ~. within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERASILITY of the r ist n oring instrumentation ensures that; (1) the radiation levels are ontin individual channels; (2) the larm or u asured in the areas served by the y radiation level trip setpoint atic action is initiated when the d; and (3) sufficient infomation is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63 and 64. 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION i The OPERA 8ILITY of the seismic ij9 ring instrumen atiop ensures that sufficient capability is avai ble t roept3) deterni he* magnitude of a X seismic event and evalua respon e of those features important to safety. i This capability is require to permit comparison of the measured re ponse to K that used in the design basis for the unit. This instrumentation ts' cons'istent with the recommendations of Regulatory Guide 1.12 " Instrumentation for Earth-quakes", April 1974. li
- l RIVER BEND - UNIT 1 3 3/4 3 4 APR.2 61985 i.- -
l l 1 l 3/4.4 REACTOR COOLANT SYSTEM FlMLDrH l l l BASES l 3/4.4.1 RECIRCULATION SYSTEM
- Operation with one reactorMIcoolant recirculation loop inoperable is prohibited until an evaluation rthe performance of the ECCS during one loop [
operation has been performed. r'Mrd andVdeterstined to be acceptable. K suk ofenman ha been An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump perfomance on a
, prescribed schedule for significant degradation. Recirculation loop flow i
mismatch limits are in compliance with ECCS LOCA analysis design criteria. The limits will ensure.an adequate core flow coastdown from either recircu-lation loop following a LOCA. In order to prevent undue stress on the vessel nozzles and bottom head - region, the recirculation loop temperatures shall be within 50*F of each other
- p- prior to startup of en idle loop. The loop temperature must also be within '
50*F of the reactor pressure vessel coolant temperature to prevent thermal shock
'( --
to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of t essel is at a lower temperature than the coolant in the upper regions of hp re, undue stress on the vessel would result if the temperature. Lh differenc' Mg eater than 200*F. 3/4.4.2 5 A rtMK 6 >
/ RELIEF VALVES y
s l The safety valve function of the safety / relief valves (SRV) [p;&is to ' prevent the reactor coolant system from being pressurized above t Safeky imit n of 1375 psig in accordance with the ASME Code. A total of 9 OPERAB -.=ty-relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient. Any combination of 4 SRVs operating in the relief mode and 5 SRys operating in the safety mode is acceptable. Demonstration of the safety-relief valve lift settings will occur only during shutdown and will be, performed in accordance with the provisions of Specification 4.0.5. .. 7 The low-low set M ' 'en'sures that safety / relief valve discharges are N minimized for a secon pening of these valves, following any overpressure transient. This is ac leved by automatically lowering the closing setpoint of 5 valves and lowering he opening setpoint of 2 valves following the initial opening. In this way the frequency and magnitude of the containneht blowdown duty cycle is substan tally . reduced. Sufficient redundancy is provided for
- the low-low set ste such that failure of any one valve to open or close at
~ its reduced setpoin does not violate the design basis. )( i APR 2 61985 RIVER BEND - UNIT 1 8 3/4 4-1
,_._,.-_.m . %- . , . , . . - _ . , - - _ _ , , _ _ - , , n ., -
, , , . . , _ . . , . _ ., # ..,._,._,_,y, - - . , - .
l 4 REACTOR COOLANT SYSTEM g 8ASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS ' The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detecti6n Systems", May 1973. 3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based i on the predicted and experimentally observed behavior of cracks
.p
~
- s. The . .
normally expected background leakage due to equipment design an ;ection' n, capability of the instrumentation for determining system leakag Lo considered. The evidence obtained from experiments suggests th akage somewhat greater than that specified for UNIDENTIFIED LEAKAGE,the probability X is small that rapidly. Ho the imperfection or crack associated with such leakage would grow i or the 1 a ver, located in all cases, if the leakage rates exceed the values specified and known to be PRESSURE BOUNDARY LEAKAGE, the reactor 7 will be(shu own o, allow further investigation and corrective action. , ( The thaal Eh lance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross i valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. 3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the j coolant is low thus the 0.2 ppm limit on chlorides is permitted during F0WER OPERATION. Dubngskutdownandrefuelingoperations,thetemperaturenecessary b '. for stress corrosion to occur is not present so a 0.5 ppm concentration of. chlorides is not considered harmful during these periods. Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure thaR. the chlorides are not exceeding the limits. * RIVER BEND - UNIT 1 B 3/4 4-2 APR 2 61985
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L 3/4. 7 PLANT SYSTEMS FDPtl *
=
. f d.s.l I
j 1 BASES t 3[4.7.1 STANDBY SERVICE WATER SYSTEM ; The OPERABILITY of the service water system and ultimate heat sink ensure that suf ficient cooling capacity is available for continued operation of safety- ! related equipment during normal and accident conditions. The redundant cooling X assumptions used in the accident gcapacityofthesesystems,assumin g ithin acceptable limitsg [ 3/4.7.2 MAIN CONTROL ROOM AIR CONDITIONING SYSTEM The OPERABILITY of the main control room air-conditioning system ensures that(1)theambientairtemperaturedoesnotexceedtheallowabletemperature M for continuous uty rating for the equipment and instrumentation cooled by this system an 2) the control room will remain habitable for operations per- / sonnel during following all design basis accident conditions. Continuous ; operation of the system with the heaters OPERABLE for 10 hours during each % 31 day period.is suffic,ient to reduce the buildup of moisture on the adsorbers % j and HEPA filt'ers. The OPERABILITY of this systasy i n conjunction with control room design provisions i # s based on limiting the radiation exposure to personnel 7)f-. occupying the control room to 5 rem or less whole body, or its equivalent. This 19 of limitation Appendix "A", is consistent with the requirements of General Design Cr erida /C 10 CFR Part 50. g 3/4. 7. 3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure t adequate core cooling in the event of reactor isolation from its primary heat sink and theofloss of feedwater flowSystem to theequ without requiring reactor vessel'ipment y actuation of any the Emergency Core Cooling . The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring the RCIC system. The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2 and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized. With the RCIC system inopera adequate core cooling is assured by the OPERABILITY of the HPCS system tifies the specified 14 day out-of-service period. / A [ 6 e. . . l . l RIVER BEND - UNIT 1 8 3/4 7-1
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3/4.8 ELECTRICAL POWER SYSTEMS , FINAL yM BASES 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES and DNSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient pow will be available to supply the safetyJeyted equipment iequired for the safe X shutdown of the facility and.5iT the mitigation and control o accident condi-tions within the facilit . he minimum specified independent and redundant % ! A.C. and 0.C. power so es a d distribution systems satisfy the requirements ofGeneralDesignCritrigl7 hf Appendix "A" to 10 CFR 50.
,p '
4 TheACTIONrequirMggt specified for the levels of degradation of the. sur tesources er provide restriction upon continued facility operation commen-w th the level of degradation. The OPERABILITY of the power sources iS ace- . nsis ent with the initial condition assumptions of the safety analyses X an afb d . C .- ed upon maintaining at least Division I or II of the onsite A.C. ~ i ower sources and associated distribution systems OPERABLE during acci en conditions coincident with an assumed loss of offsite power and 6]7L% single failure of the other onsite A.C. or 0.C. source. Division III supplies the high pressure core spray (HPCS) system only. The A.C. and D.-C. source allowable out-of service times are based on l C Regulatory Guide 1.93, " Availability of Electrical Power Sources" December 1974. ACTIONWhen diesel to requirement generator ve 1A or 18 is inoperable, there is an additional i components and device fy that all required systems, subsystems, trains, 1A or 18 as a source at depend on the remaining OPERABLE diesel generator J@ , rgency power, are also OPERABLE. This requirement is intended to provide assurance that a loss of offsite power event will not 3 periodhiesel generator IA or 18 is inoperable. result 1.a complete loss of s ! this cohtextj neans to administrative 1y check by examining logs or otherThe ,M X, term v information to determine if certain components are out-of-service for mainte-nance or other reasons. It does not mean to perform the surveillance require-ments needed to demonstrate the OPERABILITY of the component. ass The OPERABILITY of the minimum specified A.C. and D.C. power sources and ated distribution systems during shutdown and refueling ensures that the facility can be maintain d in the shutdown or refueling condition for extended time periods and ( % is available for monitori g and maintaining the unit status.ufficient instrumentation Y an The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby tower Supplies", March 10, 1971, and Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power , Plants", Revision 1, August 1977. ! l APR 2 61985 RIVER BEND - UNIT 1 8 3/4 8-1
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ELECTRICAL POWER SYSTEMS
= y; BASES 3/4.8.4 ELECTRICAL ECUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers by periedic surveillance.
The surveillance requirements applicab lower voltage circuit breakers and fuses provides assurance of breaker ar fuse . one representative sample of each manufact eliability by testing at leastg - ure$ b and of circuit breaker.aastor fuse. Each manufacturer's solded case anc met' case circuit breakers and@ fuses are grouped.into representative sample % basis to ensure that all breakers and@r fuses are tested.which are h variety If a wide tested on a rotating %
- exists within any manufacturer's brand of circuit breakers sygfor fuses, it is K necessary to divide that manufacturer's breakers gorfuses into groups and K y.
treat each group as a separate type of breaker or fusf for surveillance purposes. 'K
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The reactor protection system (RPS) electric power monitor n assemblies provide redundant protection to the RPS,and other systems M ceive power 'W' from the RPS buses,by r acting to disconnect the RPS from the power source cir-cuits in the presence of an electrical fault in the power supply. K e e APR 2 61985 RIVER BEND - UNIT 1 B 3/4 8-3 l [ . .. . .. . . . . _ . . . _ . . . _ _ . _ . . -
REFUELING OPERATIONS BASES 3/4.9.7 CRANE TRAVEL - SPENT AND NEW FUEL STORAGE, TRANSFER AND UPPER CONTAINMENT FUEL POOLS ' The restriction on movement of loadsjn excess of the nominal weight of a T fuel assembly over other fuel assemblies Tn the pools ensures that.in the event this load is droppe 9 the activity release will bedimited to ttfat contained M in 123 fuel rods 7
$ any possible distortion of fuel in the storage racks X will not result n a critical array. This assumption is consistent with the activity release assumed in the safety analyses.
3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL
,a STORAGE AND UPPER CONTAINMENT FUEL POOLS The restrictions on minimum water level ensure that sufficient water ddpth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consis-tent with the assumptions of the safety analysis.
3/4.9.16 CONTROL ROD REMOVAL These specifications ensure that maintenance or repair of control rods or control of inadvertent rod drives will be performed under conditions that limit the probability criti-cality. The require nts fpr simultaneous removal of more C than one control rod are more stringent provides for the core to remain subcritical with only one control rod fully
' ihe SHUTDOWN MARGIN specification b withdrawn.
3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE , and in operation or that an alternate method capable of decay heat removal be demonstrated ensures that(and that an alternate method of coolant mixing beY in opera and maintain the water in the reactor pressure vessel below 140'F as required during REFUELING, and(.2) sufficient coolant circulation would be available X through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby ifquid control system. The requirement to have two shutdown cooling mode loops OPERABLE when
- there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of residual heat removal capability the reactor vessel head removed and 23 feet of water above the reactor el lange, a large heat sink is available for core cooling. Thus, in the ev adequate time is provided to initi e al f lure of the operating RHR loop, r removal or emergency procedures to rnate methods capable of *ecay he core.
d heat o %g APR 2 61985 RIVER BEND - UNIT 1 B 3/4 9-2
1 D 3/4.10 SPECIAL TEST EXCEPTIONS m~ 'a,].1 h a-).')O I BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY /DRYWELL INTEGRITY The requirements for PRIMARY CONTAINMENT INTEGRITY and ORYWELL INTEGRITY ~ are not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS. 3/4.10.2 ROD PATTERN CONTROL SYSTEM In order to perform t required in the technical specifications it is necessary to bypass se ene restraints on control rod movement. The f{ g additional surveillance requi nts ensure that the specifications on heat generation rates and sh down r requirements are not exceeded during the period when these tests a. performed and that individual rod worths do not exceed the values assumed in the safety analysis. 3/4.10.3- SHUTDOWN MARGIN DEMONSTRATIONS { Performance of shutdown margin demonstrations with the vessel head removed - requires occur. These additional restrictions in order to ensure that criticality does not additional restrictions are specified in this LCO. 3/4.10.4 RECIRCULATION LOOPS
- This special test exceptigpermitsgeactor criticality under no flow conditionsandisrequiredtfgerforffertainstartupandPHYSICSTESTSwhile at low THERMAL POWER levels. [
3/4.10.5 TRAINING STARTUPS This special test exception permits training startups to be performedy with the reactor vessel depressurized at low THERMAL POWER and temperaturej while shutdown controlling cooling mode.4-RCS0-de-temperatg1 IIOEIGeontaminated water dischaewi radioactive waste disposal system rge to the [
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APR 2 s 1995 RIVER BEND - UNIT 1 B 3/4 10-1 t _ j
RADI0 ACTIVE EFFLUENTS BASES
*4 l
/'"=+ % RWS".@ w.h ":.s t f 5 3 ib liquid effluents are consistent with the methodology provided in Mr.
= Regulator Guide 1.109,"CalculationofAnnualDosestoManfromRoutineReleases[
ac Ap i x 1,"Effluents Revisionfor 1,the Purpose October 1977ofand Ey luating Compliance with 10 CFR Part 50, Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases f egulatory Guide 1.113 " Estimating N for the Purpose of Implementing Appendix I," April 1977. ' 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in . liquid effluents will be kept "as low as is reas,onably achievable". This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design - Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The m "'M limits governing the use of appropriat'e portions of the liquid radwaste treatment system were [ specified as a suitable fraction of the dose design objectives set forth in ' 1 Section II. A of Appendix I,10 CFR Part 50, for liquid effluents. 3/4.11.1.4 A UID HOLDUP TANKS 7L%
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tanksThe th ttan a 1 ted in this Specification include those unprotected outdoor 'n ot surrounded by liners, dikes, or walls capable of holding l the tank
- ents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
Restricting the quantity of radioactive material contained in the specified tanks provides assurance that f in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of ,/ 10 CFR Part 20, Appendix B, Table II, Column 2,.at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE
- h This specification is provided to ensure that tf dose #at any time at and X beyond the SITE BOUNDARY 3 from gaseous effluents wilrbe within the annual dose X limits of 10 CFR Part 2U to UNRESTRICTED AREAS. The annual dose limits are the Table dosesII, Column associated 1. with the concentrations of 10 CFR Part 20, Appendix B,
' These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outs:Ide the SITE BOUNDARY, to annual average concentrations exceeding the limits spetified'in Appendix B Table II of 10 CFR Part 20.00 CN foii NO"M . For. MEMBERS % OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of RIVER BEND - UNIT 1 B 3/4 11-2 v--,-m g+- - - --- , - , - - - - - - - - - - - - - . , , , -,,- ,- ,-r -
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l_ FIN E D r y 3/4.12 RADIOLOGICAL ENWRONNENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specifi-cation provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the f highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting tation operation. This monitoring program implements Section IV.B.2 f Appe'ndi I to 10 CFR Part 50 and thereby supplements the radiological i efflyuent onitoring program by verifying that the measurable concentrations e of radio tive materials and levels of radiation are not higher than expecte N, asis of the effluent measurements and the modeling of the environmen,dtal r( exposure pathways. Guidance for this monitoring program is provided by the l Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience. The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measurements C, in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement particular system and not as an a posteriori (after the fact) limit for a measurement. Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually); C for Qualitative Detection and Quantitative Determinatio$ urrie, L. A. , " Limits chemistry," Anal. Chem. Application to Radio- [ and Hartwell, J. K., for Radioanalytical Counting Report ARH-5A-215 (June 1975). 40,586-93(1968);ktlanticRichfieldHandfordCompany Techniques," " Detection Limits f
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1 l {- .. l RIVER BEND - UNIT 1 8 3/4 12-1 0E l
t l; RADIOLOGICAL ENVIRONMENTAL MONITORING de BASES 3/4/12.2 LAND USE CENSUS g, Thisbeyondspecification is provided to ensure that changes in the use of areas Uh the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the f , { results of this census. The best information from the door-to-door survey, ! or from consulting with local agricultural authorities - l frombe shall aerial used.survey,is Th census satisfies the requirements of Section IV fB 3 of .. Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy 4 vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine ~ this minimum garden size, the following assumptions were made: 1) 20% of the garden was used for gr ingbroadleafggetation(i.e.,simila(rtolettuce K andcabbage),and(2) 'egetationyieldfbe2kg/m, 2 y NTERLABORATORY COMPARISON PROGRAM p equirement for participation in an approved Interlaboratory Comparis n Program is provided .to ensure that independent checks on the precision and accuracy of the saasurements of radioactive material in environmental sample 9 matrices art performed as part of the quality assurance program for environment monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.? cf Appendix I to 10 CFR Part 50. t e I e APR 2 e 1995 RIVER BEND - UNIT 1 B 3/4 12-2 l i
. 5.0 DESIGN FEATURES htiRL BY p u._ g 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.
LOW POPULATION ZONE I 5.1. 2 The low population zone shall be as shown in Figure 5.1.2-1. MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIV LIQUID EFFLUENTS , 5.1.3 Informatio regarding radioactive gaseous arid liquid effluen will allow identi ication of structures and release points - d js, definition X of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible f OF THE PUBLIC, shall be as shown in Figures 5.1.1-1 and 5.1.3-1. to MEMBE
- 5. 2 CONTAINMENT -
PRIMARY CONTAINMENT - 5.2.1 cylinderThe andprimary containment a torispherical dome.is aInsi steel structure composed of a vertical right is a reinforced concrete drywell composei nd at the bottom of the containment steel ad. Primary cont vertical right cylinder and a roximately 20 feet Mwater in yhN. r-)
~
h-' uppression poolfD.airm*ent dllriected contain; to the 11 through a series of horizontal A. vents. The primary containment has a minimum net free air volume of 1,190,000 cubic feet. The drywell has a minimum net free air volume of 236,000 cubic feet. DESIGN TEMPERATURE AND PRESSURE 5.2.2 for: The primary containment and drywell .tre designed and shall be maintained-
- a. Maximum internal pressure:
- 1. Drywell 25 psig.
- 2. Primary Containment 15 psig. -
- b. Maximum internal temperature:
- 1. Drywell 330*F.
- 2. Suppression pool 185*F.
- c. Maximum externalgtoxinternal differential pressure:
1. 2. Drywell 20 psid. [ Primary Containment 0.6 psid. SECONDARY CONTAINMENT , 5.2.3 The secondary containment consists of the shield building, the auxiliary building and the fuel building. Secondary containment has a minimum free volume of 2,278,000 cubic feet.
. i RIVER BEND - UNIT 1 5-1 6 1985 l
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DESIGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.
- 5. 6 FUEL STORAGE I l
I gf 5.6. . e spent fuel storage racks are designed and shall be maintained with:
- a. A k,ff mi '- t tr iess than or equal to 0.95 when j flooded with )< '
unborated water, including all calculational uncertainties and biases as described in Section 9.1 of the FSAR. {
- b. A fuel assembly minimum center-to-center storage spacing of 7 in.
within rows and 12.25 in, between rows in the Low Density Storage M j Racks. -
- c. I AJuelassemblyminimumcenterpoce~nt s acing of 6.28 i g X. 6 Cwith a neutron noison saterisi oewee storse' acje in the High E
_ Density Storage Racks. e T The storage of spent fuel in the upper contai ent fuel storage pool is prohibited during normal operation. , O 5.6.1.2 The K for new fuel for the first core loadin Y (. spentfuelstofkeracksshal/beadministrativelycontrbstoreddryinthe led to not exceed : 0.98 when optimum moderation (foam, spray, fogging, or small droplets) is assumed. 5.6.1.3 Provisions shall be taken to avoid the entry of sources of optimum moderation new fuel, stored (foam,in spray, the new fogging, fuel storageor small droplets) facility, couldtoexceed preclude that K,ff for 0.98. - DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 95'. CAPACITY ! 5.6.3 The spent fuel storage pool in the fuel building is designed and shall l! be maintained with a storage capacity limited to no more than 2680 fuel l assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1. .. l l RIVER BEND - UNIT 1 5-6 l APR 2 61985 l
FINAL DRAFT ADMINISTRATIVE CONTROLS 6.3 UNITSTAFF.QUkLIFICATIONS 6.3.1 tions of Each member ANSI /ANS of the unit staff shall meet or exceed the minimum qualifica-3.1-1978 for comparable positions, except for the Supervisor-Guide 1.8, September 1975.Radiolegical Programs who shall meet er exceed th The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all lice'nsees. 7 6.4 TRAINING 6.4.1 A retraining and replacement training pro a be maintained under the direction of the Managergram for the ug taff shall (} Administra exceed the requirements and recommendations of Section 5.5 6 ion 4sh 1 meet or [f and Appendix A of 10 CFR .Part 55 and the supplemental requirem AN /ANS 3.1-1978 ts specified in Sections A and C of Enclosure 1 of the March, 28 1980 NRC letter to all licensees, experience. and shall _ include familiarization with relevant industry operational 6.5 REVIEW AND AUDIT . 6.5.1 FACILITY REVIEW COMMITTEE (FRC) FUNCTION 6.5.1.1 related to nuclear safety.The FRC shall function to advise the Plant Manager on a COMPOSITION 6.5.1.2 The F"RC shall be composed of the: Chairman: Member: Assistant Plant Manager-Technical Services Member: Assistant Plant Manager-Operations, Radwaste and Chemistry Member: Assistant Plant Manager-Maintenance and Material General Operations Supervisor Member: Member: Supervisor-Radiological Program
. Reactor Engineering Supervisor ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the FRC Chairman i pate as voting members in FRC activities at any one time.to serve MEETING FREQUENCY ~
6.5.1.4 by the FRC Chairman or his designated alternate.The FRC shall meet at RIVER BEND - UNIT 1 6-7 APR 2 61985
dl i ADMINISTRATIVE CONTROLS FINAL DRAFT l l QUORUM 6.5.1.5 The quorum of the FRC necessary for the performance of the FRC responsibility and authority provisions of these Technical Specifications shall no moreofthan consist the twoChairman or his designate! alternate and four members including alternates. RESPONSIBILITIES 6.5.1.6 The FRC shall be responsible for: a. Review of all plant general administrative procedures and changes thereto; b. Review of all proposed tests and experiments that affect nuclear safety; c. Review of all proposed changes to Appendix A Technical Specifications; d. Review of all proposed changes or modifications to structures, compo-nents, systems or equipment that affect nuclear safety;
- e. -
p Investigation of all violations of the Technical Specifications, g (g . h includino the preparation and forwardino of reoorts covering evaluation X. andrecommendationstopreventrecurrence,tothNuclearReviewBoard; y Review of all REPORTABLE EVENTS; h_ N. g. evie' of unit operations to detect potential hazards to nuclear saf .y. j/ tem I that may be included in this review are NRC inspection reports, QA .
, pp its/ surveillance reports of operating and maintenance activities, results; results, and American Nuclear Insurer (ANI) inspection audit h.
Performance of special reviews, investigations, or analysesjand reports N thereonI as requested by the Plant Manager or the Nuc1sar Review Board; and K i. Review of initial start-up testing phase start-up procedures and revisions. 6.5.1.7 The FRC shall: a. Recommend in writing to the Plant Manager approval or disapproval of itemsimplementation. their considered under Specif.ication 6.5.1.C.a. through d. prior to b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6.a. through e. constitutes an unreviewed safety question. c. Provide written notification within 24 hours to the Vice ' President - , RBNG and the Nuclear Review Board of disagreement between.the FRC and the Plant Manager; however, the Plant Manager shall have respon-sibility for resolution of such disagreements pursuant to'Specifica-tion 6.1.1. RIVER BEND - UNIT 1 6-8 APR 2 61985 l , . . . . . . . . . .--.,,-n------ - - . . . . -
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t. ADMINISTRATIVE CONTROL FINAL DRMT ,
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t 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation. 6.13.2 Licensee-initiated changes to'the PCP: n [ 1. Shall be submitted to the Commission in the Semiannual Radioactive EffluentThis made. Release Report submittal shallforcontain: the period in which the change Q was K
/'
a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental
, information;
- b. '
A determination that the change did not reduce the overall confor-mance of the solidified waste product to existing criteria for solid wastes; and _ c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to Specification 6.5.2. 2. Shall become effective upon review and acceptance pursuant to Speci-fication 6.5.2. 6.14
. 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
2 6.14.2 Licensee initiated changes to the ODCH: 1. y Shall be submitted to the Commission in'the Semiannual Radioactive Effluent Release Report for the period in which the change % was ' made effective. This submittal shall contain: K a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be M with each page numbered and provided with an approval and a t! tox, together with appro-
' V priate analyses or evaluations ju ig the change (s); M N. l b.
A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to Specification 6.5.2. -
- 2. Shall b ficatio 6. e 2. eff'ctive upon review and acceptance pursuant to Speci- M hN' RIVER BEND - UNIT 1 6-23
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D . M c, u rt a
, s ousq g
k a UNITED STATES ENCLOSURE 2 NUCLEAR REGULATORY COMMISSION
~f wAsHWGTON D.C.20555 May 17, 1985
% * . . . * .c!
l l l l 1 MEMORANDUM FOR: Edward J. Butcher, Group Leader Technical Specification Review Group i Division of Licensing : 1 FROM: M. Dean Houston, Reactor Engineer Technical Specification Review Group
, , Division of Licensing
SUBJECT:
DEFICIENCIES IN REGARD TO GSU CERTIFICATION OF RIVER BENO l TECHNICAL SPECIFICATIONS (FINAL DRAFT) j 1
~~
By letter dated May 2,1985. Gulf States Utilities (GSU) was requested to review the final draft of River Bend Technical Specifications and submit by May 13. 1985, a certification under oath and affirmation that the final l draft accurately reflects the FSAR, SER and as-built configuration of the
' plant. By letter . dated May 6,1985. GSU submitted their certification, under oath and affirmation, of the final draft of River Bend Tech Specs.
Included in their submittal were: (1) identified editorial changes (173 items), (2) proposed changes to the Tech Specs (38 items), (3) proposed revisions to the SER (7 items to be revised with some propos)ed FSAR revisions (55 areas).and (4) identifi t In my review of their submittal, numerous deficiencies, acts of omission and
! comission, have been identified. Examples of editorial errors that were not identified by GSU are presented in Enclosure 1. The attached pages are ,
from the GSU submittal, some with their markup, and the errors that were i not detected are circled. These errors take many foms - typos, missed headings, non-existent trip signals, nomenclature, etc. While misspelled' words can be properly interpreted, many of the other unidentified errors would have contributed to operator confusion. The enclosure is not intended l to be a complete compilation of overlooked errors but does establish that the GSU review process was less than thorough. Approximately 20% of the errors in the final draft were not detected during their review. This is an unacceptable level by any standard and their review process for certi-fication at licensing must be improved. Examples of deficiencies by acts of comission are presented in Enclosure 2. Two of these are editorial errors and one is based on a possible false statement regarding their as-built plant. (1) GSU proposed changes to page 3/4 8-5 of the River Bend - Tech Specs as shown. As proposed, the change was to be inserted between 10 and seconds. On page 3/4 8-6 of the 4;sGG5M3.LG D 52e h
f Edward J. Butcher May 17, 1985 same submittal, GSU properly proposed an identical change with the clause to be inserted after 10 seconds. As proposed on 3/4 8-5, the revision makes no sense and would confuse the operator. j (2) In Attachment B of the GSU submittal of May 6,1985 Item 29 l requests a deletion of a surveillance requirement because -
"There are no valves in the flow path of any PGCC subsystem."
In past discussions, GSU has resisted this requirement on the basis that the valves did not have a position indicator ' although, in fact, the valves do have a trip indicator. A copy of FSAR Figure 9.5-13 is enclosed which shows numerous l solenoid operated valves as well as a couple of check valves j m - in the flow path. Therefore, the GSU statement of "no valves l in the flow path" appears to be a false representation. - 1 (3) Also, in Attachment B. Item 30 refers to adding a footnote to TS 3/4.7.6.4, Table 3.7.6.4-1. This is in error as the proposed l footnote was identified with TS 3/4.7.6.5 Table 3.7.6.5-1. 1 In addition to the deficiencies noted above, I would like to comment briefly on other uncertainties associated with the River Bend Tech Spec review. GSU has submitted a listing of 55 areas in the FSAR that need revision to support Tech Spec sections. Amendment 19 to the FSAR was delivered on May 14, 1985 and only 12 of these areas were addressed. Therefore, in the other 43 areas, the NRR Technical Reviewer has not seen the necessary documentation to support the current Tech Spec section or a proposed revision to a section. There is also the potential for additional FSAR revisions resulting from the reviewer's evaluation. This lack of timely infomation will impact the accelerated schedule for issuance of the Tech Specs with the River Bend license in June, 1985. There seem to be some values in the FSAR and Tech Specs that are constantly being changed. For example, the DBA activity release to the environment i following a LOCA (used for containment Tech Spec review) were revised in ' Amendment 18 to the FSAR dated April 1985 and revis'ed again (increased) in Amendment 19 on May 13, 1985. In the Tech Specs, GSU has pro the water level for the Ultimate Heat Sink be 112'4" (2nd Draft) posed , 108"6"that (Final Draft) and 111'10" (current revision). Changes of this frequency would indicate that the utilities review process has not settled down. All of the above matters should be given due considerations when discussing comitments and completion schedules for the River Bend Tech Specs. N. Y M. Dean Houston, Reactor Engineer. 1 Technical Specification Review Group Division of Licensing - cc: D. Crutchfield R. Benedict S. Stern
ENCLOSui.i 1 FINAL DRAFf I I TECHNICAL SPECIFICATIONS RIVER BEND - UNIT 1 Markup Pages From _, , GSU Submittal of 5/6/85 ,
~
Errors not identified by GSU are circled and noted in margin with L Not Intended To Be Complete April 26, 1985 ) l i
l
- S - us FINAL DRAFT l BASES 2.1.3 REACTOR COOLANT SYSTEM PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is desi Section III of the ASME Boiler and Pressure Vessel Code 1971 Edition,gned to including Addenda through Summer 1973, which permits a maximum pressure transient of 110%,
1375 psig, of design pressure,1250 psig. The Safety Limit of 1325 psig, as measured by the reactor vessel steam done pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the ASME Boiler and Prassure vessel Code, Section III, Class I. 2.1.4 REACTOR VESSEL WATER I.EVEL _ With fuel in the reactor vessel during periods when the reactor is shut sideration must be given to water level requirements due to the effect ecay eat. If the water level should drop below the top of the active irradi-I during this period, the ability to remove decay heat is reduced.
- This reduction in cooling capability could lead to elevated cladding tempera-
[% tures and clad perforation in the event that the water level became less than
.two-thirds of the core height. The Safety Limit has 'been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.
O 6 M APR 2 6 585 RIVER BEND - UNIT 1 B 2-5 e
r;
, LIMITING SAFETY SYSTEM SETTINGS nun E wn a
~
. BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOIN the water level has reached a point high enough to indicate that it is indeed filling the up, butofthe movement the volume rods whenisthey stillare great tripped. enough to accommodate the water from The trip setpoint for each 17 gallons of water. scram discharge volume is equivalent to a contained volu
- 10. Turbine Stop Valve-Closure
~. flux, and heat flux increases that would result from clo valves.
With a trip setting of 5% of valve closure from full open, the - resultant increase maintained during theinworstheatcase flux transient.- is such that adequate thermal margins are 11. Turbine Control Valve Fast Closure, Trip 011 Pressure-Low The turbine control valve fast closure trip anticipates the pressure
- k r : 'ailure tbineof the control.
turbine bypass valvesvalves. due to load rejection w The Reactor Protection System (' imustes a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 20 milliseconds after the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. logic input to the Reactor Protection System.is sensed by pres This trip setting, a slower clairure time, and a different valve characteristic from that of the turbine stop the stopvalve, valve.combine to produce transients which are very similar to that for . of the Final Safety Analysis Report. Relevant transient analyses are discussed 12. Reactor Mode Switch Shutdown Position automatic protective instrumentation channels and prov reactor trip capability.
- 13. Manual Scram .
instrumentation channels and provides manual reacto . M RIVER BEND - UNIT 1 B 2-9 yit 2 61S85
-" a .
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) -
- 4. No " slow" control rod, " fast" control rod with individual scram inser-tion time in excess of the limits of ACTION a.2, or othemise inoperable i control rod occupy adjacent locations in any direction, including the diagonal, to another such control rod.
Otherwise, be in at least HOT SHUTDOWN within 12 hours. i
- b. With a " slow" control rod (s) not satisfying ACTION a.1, above: ,
- 1. Declare the " slow" control rod (s) inoperable, and
., 2. Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or '
more " slow" control rods declared inoperable. *
~
Othemise, be in at:1 east HOT SHUTDOWN within 12 hours. c. With the maximGm scram insertion time of one or more control rods exceed-
~
ing the maximum scram insertion time limits of Specification 3.1.3.2 as determined by Specification 4.1.3.2.c, operation may continue provided that: .
~
- 1. " Slow" control rods, i.e., those which exceed the limits of C. ~.-~ Specifica' tion 3.1.3.2, do not make up more than 20% of the 10%
sample of control rods tested. 2. Each of these " slow" control rods satisfies the limits of ACTION a.1. 3.
.- The eight adjacent control rods surrounding each " slow" control rod are:
a) Demonstrated through measurement within 12 hours to satisfy the ' maximum scram insertion time limits of Specification 3.1.3.2, and i b) . OPERABLE. 1 4. The total number of " slow" control rods, as determined by Specifica- { tion 4.1.3.2.c, when added e sum of ACTION a.3, as determined by Specification 4.1.3.2. an/b does not exceed 5. g l Otherwise, be in at least HOT SHUTDOWN within 12 hours.
- d. The provisions of Specification 3.0.4 are not applicable.
O RIVER BEND - UNIT 1 s a sas 3/4 1-7 l
. I 1
i _ _ __ _ __
REACTIV2TY CONTROL SYSTEMS F
] ]=
CONTROL ROD SCRAM ACCUMULATORS $" LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.
- 2. With mor X9 scram cu$ato ne withdrawn control rod with the associated operating, immediately place the reactor mode switch in theino 3
j Shutdown position. c. The provisions of Specification 3.0.4 are not. applicable. 4.1.3.3 .Each-control rod scram accumulator shall be determined OPERAB , a. At least once per 7 days by verifying that.the indicated pressure is
- greater than or equal to 1520 psig unless the control rod is inserted and disarmed or scrammed. .
. b. At least once per 18 months by:
, 1. Performance of a:
a) CHANNEL FUNCTIONAL TEST of the leak detectors, and b) CHANNEL CALIBRATION of the pressure detectors, and verifying
~~
an alarm setpoint of 1520 psig on decreasing pressure.
- 2. '
Verifying th'at each individual accumulator check valve maintains the associated accumulator pressure above the alarm set point for greater than or equal to 10 minutes, starting at normal system operating pressure, with no control rod drive pump operating. l l l RIVER BEND - UNIT I 3/4 1-10
. - , . _ , -.- -. _ _ , . - - , _ _ . . - 4 - _
N METEOROLOGICAL MONITORING INSTRUMENTATION FINLL DMFT LIMITING CONDITION FOR OPERATION 3.3.7.3 Themefteorologicalmonitoringinstrumentationchannelsshownin -- Table 3.3.7.3-1 shall be OPERABLE.
^ " ' " " ' ' T TY: At all times.
k ith one or more seteorological sonitoring instrumentation chann'els ' inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status. r- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab1t. SURVEiLLANCEREQUIREMENTS
- h. - 4.3.7.3 Each of the above required meteorological monitoring instrumentation
' channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHEC and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.3-1.
'AIVER BEND - UNIT 1 3/4 3-73 D 2 61985
( g FINAL DRAFT W t *
- y,
,ISE i Ely"a '
g 5
=
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==
w j g' "" g' l .. I e E E C
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==
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- ((g t -
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- : 22 g
g [. -a" rcf I s
- 3 $d
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-, 2 a:
. e : -
0,
= '
q' e - x x ~ l
- x :x a
o . [, E Euu ee . -- ll E, 4 ci 4 4 4 .4 22 4 4 4 g ,G G RIVER BEND - UNIT 1 3/4 3-77 i APR2619e5
i
' FIN E D W T TABLE 3.3.7.4-2
- REMOTE SHUTOOWN SYSTEM CONTROLS I
. 1 e 1 MINIMUwcumuurte^ay[E) Q U Z G DIV. L/ ;
- 1. RCIC Suction from CST MOV (IE51*MOVF010) 1 NA R \
- 2. RCIC Injection Shutoff MOV i
(IE51*MOVF013) 1 NA 1
- 3. RCIC Nin. Flow to Suppression i
1 NA
,, 4 Pool MOV (1E51*MOVF019)
RCIC Test typass to CST MOV (1E51*MOVF022) 1 NA
- 5. RCIC Gland Seal Air Compressor -
(IE51*PC002C) 1 NA
- 6. RCIC Pump Suction from Suppression
- 7. Pool MOV (1E51*MOVF031) .. e RCIC Steam to Turbine M0V WP 46A-i 1 j (1E51*MOVF045) NA 8.
if RCIC Turbine Lube Oil Cooling MOV (1E51*MOVF046) 1 NA "
- 9. '
RCIC Test typass to CST MOV F Ndr' i (1E51*MOVF059) 1 NA
.'\ 10.
RCIC Steam Supply Inboard Isolation MOV(1E51*MOVF063) 1 NA 11. i RCIC Steam Supply Outboard Isolation 1 NA i- 12. MOV(1E51*MovF064) i ~~ RCIC Turbine Exhaust to Yo,ol,, MOV(IE51*MOVF068) 1 NA i 13. * , RCIC Steam Line Warsup Line Isolation 1 NA ' i
- 14. MOV(1E51*MOVF076)
RCIC Vacuum Breaker Outboard Isolation 1 NA i 15. MOV(1E51*MOVF077)
- j RCICMOV(1 Vac um Breaker Inboard Isolation 1 NA s
- 16. VF078)
I RCIC Turbine Flow Controller . (1C61*FICR001) 1 NA 17. j w RCIC Turbine Trip & Throttling MOV (1E51*M0VFC002) 1 NA 1 81 W. RHR Pump (1E12*PC002A,28,2C) 1 2g,)
; ~10 N. '
! RHR Nx Shell Side Outlet MOV
; u (IE12*MOVF003A,B) 1 1 3e RHR Pump Suction MOV i (1E12*M0VF004A, 8; 1E12*M0VF105) 1 2I ,) !
1 u M. 1 RHR Shutdown Cooling)MOV
% (IE12*M0VF006A,65 2(*) NA
- 64. Reic Tv eme ~
g.4
, se gua 3,*,4,5, g
.(a) One per control equipment gf4 I
i RIVER SEND - UNIT 1 j 3/4 3 78 N C__._ , _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . , . _ . _ _ . _ _ _
~
TABLE 3.3.7.4-2 (Continued) FW$iin D W*T
\ REMOTE $ HUT 00WN SYSTEM CONTROLS 7e gygyg;wcuawurcnq M
{ )lv. Y.I f DIV. l AS.pt: RHR Oatboard shutdown Isolation MOV 1 NA (IE12*MovF008) w 33. RHR Inboard Shutdown Isolation MOV
- 1 NA NN/
gec.dr/w ., (IE12*MOVF009) g mis .. Ar 44 fm,, ,f RHR Hx Flow to h [ool MOV (1E12*MovF011A,8) 1 1 My ss M. RHR Reactor Head Spray MOV (1E12*MOVF023) 1 NA - 7/M u M. RHR Test Line MOV I"I 1 (IE12'MOVF024A,8) 1 gy tr 67. RHR Nx Flow to RCIC MOV i O' ' (1E12*MOVF026A) 1 NA J-E 11 M. RHRInjectionShutoffMOV
- 1 1
* (1E12*MOVF027A,8) 3 49. RHR Upper Pool Cooling Shutoff MOV ~
1 , 1 (1E12*MOVF37A 8) 3: De. RHR Injection-MOV 2g,)
.1 ' (IE12*MovF042A,8,C) 1s M. RHR Hx Shell $fde Inlet MOV 1 1 (1E12*M0VF047A,8) 37 M.
( RHR Mx Shell Side 8ypass MOV (IE12*MOVF044A,8) 1 . 1
.\ sy M. RHR Discharge to Radwaste MOV i 1 NA (IE12*MOVF040) sr 44. RHR Steam Isolation MOV 1 1 (1E12*M0VF052A,8) st M.- RHRInjectionMOV 1 1 i
(1E12*MOVF053A,8) 17 Mr. RHR Pump Mininum Flow *MOV 2g,) l 1 ' i (IE12*MOVF064A, 8, C) SF W. RHR Nx Water Discharge MOV 1 1 (1E12*MOVF068A,8)
. si D&. Safety Relief Valves 3g,) 3g,)
(1821*RVF051,C,G,D)
- 4. H. $$W Pump (ISWP"P2A,2C.T 28, 20) 14-N 2(a) w 46. Normal Service Water Isolation MOV 1 1 (15WP"MOV96A, 8) j 'v t.41. $$W Cooling Tower Inlet MOV 1 1 (ISWP*MOV55A,8) 1 (a)[0nepercontrolequipment
- M ssw e-,
3/4 3 79 r-.a is . L.u.I w,,l g ! 61985 RIVER SEND - UNIT 1esw. .eac is ,. e ...... zr e. pr.wid a.
s 83 f f . t
/
~
' TAett 4.3.7.s-1
)=
m n EW ACCIDENT' MONiiORING INSTRUMENTATION $URVElttANC * E INSTRUMENT
- CHAML. CHANNEL APPLICA8tE
- 1. ' _ CHECK OPERATIONAL E Reactor Vessel Pressure _
CALIBRATION
- 2. M CONDITIONS M Reactor Vessel Water Level
- a. Wide Range R
w 1, 2
- b. Fuel Zone M '
3 R Suppression Pool Water tevel M, 1, 2 4 R Suppressipn Pool Water Temperature M 7.:-., C...^.: - ..^ L !.- ,-..^ .. M R
. 1, 2 1, 2,y 6 " .. R Primary Containment Pressure 1-- 1, 2 /
7 Drywell Pressure . M :, 2-
- 8. Drywell Air Temperature M R
1, 2
- 9. R M 1, 2 Drywell Analyzer andand Primary Monitor Containment j ";;;;;. Concentration M R
1, 2 R
- 10. qa 1, 2 Safety Area / Relief Valve Position Indicators Radiatic,..* ;
y + ,11. .,7,." M
, .. _. . ,..> _., C.,..^.!: ~ .^. O "-_:4 R 2
~ _. ..__w . 4 1, 2 I?. 4-
";_-s._
' --c'=:' 'd'n;'F:: ? ": . ? " ; "- ;:
~ ~ ~ ' - ~ "::':: .
t, 2 , I'. n :- r_,' - '
" "; " : '- - ; ;" !,;" ; !:" __;", " ;. " m # -
- -: ": i n': " '?d' ; " x^_:: tr'.:_." , , ,,
N w , ,__...,u,__
._._..__....,___.,,_.,__,.__-,__.____,m
=
. =
!, 2, 3 Om Fusing sample gas containing:
g,/ M a. One volume percent hydrogen, balance nitrogen. N enum
- pid- Four volume percent hydfogen, balance nitrogen. ,
.M.
**detector,he {HANNEL for range decades cat!8 above RATION 10 R/hr shall consist of an channel, electronic calibration not including the of b the "JD #with an installed or portable gamma source.and a one point calibration check of the detector below
- y 10 R/hr
,. High range c ".!: '
! gem go;,ame1 monitors.
- g v'
_ f a, peinary ten 4as,,,u,,+ Aren dc. m 1,D R
- k. prv elf Ar:A ' ?- 3 M i
3 l m.: A l,23 W
, espo soem * * * . =
TABLE 3.3.7.8-1 (Continued) yEdgin 6J6A m y[ ,l FIRE DETECTION INSTRUMENTATION . TOTAL INSTRUMENTS OPERABLE" INSTRUMENT LOCATION FLAME SMOKE HEAT Tx7y) (x/y) (x/y) I. CONTROL BUILDING g (Continuec) ,
. 0/9 17/0 50-143 PGCC PANEL MODULE, EL 136'0" 17/0 PGCC PANEL MODULE, EL 136'0" 0/9 50-144 0/8 8/0 50-145 PGCC PANEL MODULE, EL 136'0" 8/0 >-
PGCC PANEL MODULE, EL 136'0" 0/8 50-146 0/12 14/0 } 50-147 PGCC PANEL MODULE, EL 136'0" 18/0 PGCC PANEL MODULE, EL 136'0" 0/12 -l 50-148 0/10 14/0 50-149 PGCC PANEL MODULE, EL 136'0" 15/0 PGCC PANEL MODULE, EL 136'0" 0/9 50-150 0/10 10/0 SD-151 PGCC PANEL MODULE, EL 136'0" 8/0* PGCC PANEL MODULE, EL 136'0" 0/8 50-158 10/0 NDH PANEL MODULE AREA NORTH, EL 135'0" SD-152 10/0 . 50-153 NON PANEL AREA SOUTH, EL 135'0'" 84/0 SD-154 50-162 GENERAL AREA, EL 136' REMOTE SNUTDOWN PANE V 1/0
/r t -
EL 98'0" . E 50-163 REMOTE SHUTDOWN PA IV 1/0
~- EL 98'0" FD-26 CNARC0AL FILTER 1HvC*FLT38, ; ' EL 115'0" 1/0 '
FD-27 CHARC0AL FILTER 1HvC*FLT3A, t EL 115'0" 1/0 II. REACTOR BUILDING *
~~ ZONE ,
Ws- 1Vo 50-57 iCONTAINMENT AREA, EL 114'0" GN& 2Vo l 5D-102 ANNULUS AREA, EL 186'3" 17/0 ! SD-104 # CONTAINMENT AREA, EL 186'3" WO ~ 7/o 50-117 # CONTAINMENT AREA, EL 162'3" 13/0 , 50-119 # CONTAINMENT AREA, EL 141'0" 2/0 50-156 # CONTAINMENT AREA, EL 95'9" FD-13 #RECIRC PUMPS
- DRYWELL, EL 70'0"
& 98'0" 2/0 1
* (x/y): x is number of Function A (early warning fire detection and notifi-cation only) instruments.
y is number of Function B (actuation of fire suppression syst' ems and early warning fire detection). :
#The fire detection instruments located within the Containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate Tests.
i RIVER BEND - UNIT 1 3/4 3-89 l
I , 1 TA8LE 3.3.7.10-1 (Continued) gl l 5 BEL 3m 33 .1 m TABLE NOTATION , 1 ACTION 100 - With the number of channels OPERABLE less than required by the Minimum Channels CPERA8LE requirement, effluent releases may continue for up to 14 days provided that prior to initiating a release: a
- a. At least two independent samples are analyzed in accordence l
4
'with Specification 4.11.1.1.1,and
(
- b. At least two technically qualified members of the facility t staff independently verify the release rate calculations I and discharge Ifne valving; l Otherwise, suspend release of radioactive affluents via this pathway. .
i ACTION 101 - With the number of channels OPERA 8LE less than required by the i Minfeum Channels OPERA 8LE requirements, effluent releases via l 1 this pathway may continue for up to 30 days provided that, at least once per 12 hours, grab samples arie collected and analyzed i for gross radioactivity (beta or gamma) at a limit of detection of at least 10 7 aferocuries/m1. ACTION 102 - With the number of channels OPERABLE less than required by the ';- Minimum Channels OPF88" quirement, effluent releases via : this pathway mayfontinue/3or up to 30 days provided the flow / _ rate is estimatec .6 ....T. once per 4 hours during actual releases. Pump curves generated in situ may be used to estimate flow. h j 4 6 L RIVER SEND - UNIT I 3/4 3 96 EIO
, . , - - , - , , - - , _ , , - ,---,.,,._,,..,,--,,,.n... .,_--n,. _ _ _ ,, . - , - . . . , . . -. . . . ,,,,.,.,..--n_--,-,.,,,.,--,-.n,,,,... ,.._, p ..e--aw,., ,__ ,_~ - - ,,,,,._,
w-- --
TABLE 4.3.7.10-1 (Centinusd) - TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
" - ^ FUNCTIONAL TEST shall also demonstrate that control room alarm (2) ~
annf.. ?"becurs if any of the following conditions exists: iatie ,
- 1. Instrument indicates measured levels above the alare setpoint, j i
- 2. Circuit failure. j 3.
Instrtment indicates a downscale failure.
- 4. Instrument controls not set in operate' mode.
~
t (3) The initial CHANNE'L CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or , using standards that have been obtained from suppliers that participate f > in seasurement assurance activities with NBS. These standards shall permit
; calibrating the system over its intended range of energy and sensurement , , range. For subsequent CHANNEL CALIBRATION, sources that have been related [
to the initial calibration shall be used. ; (4) CHANNEL CHECK shall consist of verifying indication of flow during periods ' of release. CHANNEL CHECK shall be made at least once per 24 hours on
.. days which continuous, periodic, or batch releases are made. !
. ~
m RIVERBEND-UNITk 3/4 3-98 W26E e
TA8LE 4.3.7.11-1 (Continued) a TA8tE NOTATIONS
- 4 At all times.
During main condenser offgas treatment system operation.
.. lr., ;;;.ia67. ;f tt =' :::in::r :'r ;'::t:r.
(1) [annpiatiopeurs FUNCTIONAL if sq of the TEST following shall alsoconditions demonstrate that control exists:" room 7'_"""
- 1.
)r -
Instroent. indicates esasured levels above the alam setpoint. 7 . 2. Circuit failure. I
- 3. Instroent indicates a downscale failure, ~
' l l~ 4. i Instement controls not set in operate mode. - l (2) l The initial CHANNEL CALIBRATION shall be perfomed using one or more of !
the reference standards certified by the National Bureau of Standards or ; using standards that have been obtained free suppliers that participate ' A, in seasurement assurance activities with ISS. N calibrating range. the systaa over its intended range of energy and esasurementTh i to the initial calibration shall be used.For subsequent CHANNEL CAL i (3) j The CHANNEL containing a nominal: CALIBRATION shall include the use of standard gas samples
**1.
One volume percent hydrogen, balance nitrogen, and . l 2. Four volume percent hydrogen, balance nitrogen. i i I I 1 u vER aEND - UNIT 1 3/4 3-105 AM t 4 25
. _ _ _ _ , _ . . _ - - _ _ _ - _ . . - - . _ . . ~ _ . _ . _- . . _ . , _ . , _ _ , . . _ - . _ _ . . - ___.. _ _ .
-. . . . . - - . - . _ - - - - . - - - . . . . - . .. ~
, INSTRUMENTATION l - 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION 8 *m w as e a I
LIMITING CONDITION FOR OPERATION
- i i '
' 3.3.9 The plant systems actuation instrumentation channels shown in Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the ' values shown in the Trip Setpoint column of Table 3.3.9-2. APPLICABILITY: As shown in Table 3.3.9-1. ' ~ ACTION: t s. With a plant system actuation instrumentation channel trip setpoint j i 1ess conservative than the value shown in the Allowable Values . ! l column of Table 3.3.9-2, declare the channel inoperable and take # the ACTION required Table 3.3.9-1. ) . -
- b. ;
With one or mor stems actuation instrument channels inoperable, take the ACTION d by Table 3.3.9-1. i t . SURVEILLANCE REQUIREMENTS i L .l - I ( 4.3.9.1 Each plant system actuation instrumentation channel sh.. ' he
- demonstrated OPERABLE by the performance of the CHAMEL CHECK, Ci .1NEL FUNCTIONAL
{ TEST and CNANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.9.1-1. 4.TS.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least. once per 18 months. ' i . 1 i l I i !( RIVER BEND - UNIT 1 3/4 3-108 APR 2 6125 l
FliWil DRAF1 REACTOR C0OLANT SYSTEM
=
3/4.4.2 SAFETY VALVES SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 The safety valve function of at least 5 of the followin 1 es the relief valve function of at least 4 additional valves of .tb other than those satisfying the safety valve function requiremen e OPERABLE with the specified lift settings: . b Number of Valves Function Setooint" (esic) Safety 1165 2 IX r 7 . 5 Safety 1180 2 1% 1 4 Safety 1190 1 15 -
- j. Relief 1103 1 15 psig
' 1 1113 2 15 psig 8 Relief -
Relief 1123 1 15 psig , i
. - 7 j i
The scoustic monitor for each OPERA 8LE valve shall be OPERABLE. l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ' l ACTION: r
'i and/or relief valve function of one or more of the above
- a. With the safety safety/ relief valves inoperable, be in at least NOT SHUTDOWN l _
' required g.
within 12 hours and in COLD SHUTDOWN within the next 24 hours. I b. With one or more safety / relief valves stuck open, provided that suppres-sion pool average water temperature is less than 105'F, close the stuck open safety /rel' ef valve (s); if suppression pool average water temperature.
*~
is 105'F or greater, place the reactor mode switch in the Shutdown position. ' l
- c. With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable sonitor(s) to OPEP.A8LE status within 7 days or be in at j i
! . least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOW - the following 24 hours. - i il l I The lif t setting pressure shall correspond to ambient Conditions-of the l valves at nominal operating temperatures and pressures. I -s pt6@ ! AIVER SEND - UNIT 1 3/4 4-5 1
REACTOR COOLANT SYSTEM
?
l\ 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION A 3.4.5 The specific activity of the primary coolant shall be Ifmited to: I a. Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, and *~~
- b. Less than or equal to 2004 microcuries, per gram.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4. t' ' ACTION: i -
. OPE TI L
- TIONS 1, 2 or 3-with the specific activity of i 1.
i Greatirr than 0.2 microcuries per gram 005E EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram, operation may continue for up to 48 hours provided that the cumulative opera- i !.- ting time under these circumstances does not exceed 800 hours i in any consecutive 12-month period. With.the total cumulative D, operating time at a primary coolant specific activity greater
. than 0.2 microcuries per gram 005E EQUIVALENT I-131 exceeding I
500 hours in any consecutive six-month period, prepare and submit i a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the number of hours of operation above
' this limit. The provisions of Specification 3.0.4 are not appli-
.. cable.
3 2 2. Greater thin 0.2 microcuries per gram DOSE EQUIVALENT I-131 for - j more than 48 hours during one continuous time interval or for 3 more than 800 hours cumulative operating time in a consecutive 12 month period, or greater than 4.0 microcuries per gram, be in at least HDT SHUTDOWN with the main steam line isolation valves closed within 12 hours. ' 4' t 3. ! gram, be in at least NOT ! Greaterthan2004micr:..[?~- SHUTOOWN with the mai 12 hours. es ne solation valves closed within l b. ! In OPERATIONAL CON 0!TIONS 1, 2, 3 or 4, with the specific activity i of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 or greater than 100/E eTerocuries per gram, perform the sampling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to l within its limit. A REPORTABLE EVENT shall be prepared and sub-i mitted to the Commission pursuant to Specification 6.6.1. This report'shall contain the results of the specific activity analyses
- RIVER BEND - UNIT 1 3/4 4-16 ,tps t t 2 5 l
REACTOR COOLANT SYSTEM
\ 3/4.4.7 MIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main with CPERABLE steamclosing line isolation timesvalvgreater than o(. (MSIVs)
;-+ steam line shall be to 5 seconds. .,__. 3pand s,s,._than or equal k
decad.s APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION: --s -
- a. With one or more MSIVs inoperable:
- 1. .
, , Maintain line that is at openleast andone MSIV within 8 hours,OPERABLE either: in each affected mai a), Restore the inoperable valve (s) to OPERABLE status, or b) e Isolate the affected main steam line by use of a deacti-wated MSIV in the closed position.
( 2. an'd in COLD SHUTOOWN within the following 24 ho . b. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.4.7 verifying 4.0.5. Specification full closure between 3 and 5 seconds when teste for entry into CPERATIONAL CONDITIONS 2 or 3 provided the and prior to entry into OPERATIONAL CONDITION 1. performed W RIVER 8END - UNIT 1 3/4 4-25 N e
, . PLAN 1 $YSTEMS IARD FIR [ NYORANTS AND NYORANT HOSE HOUS t1MIT!E COND}T]QN FOR OPIRATION s.
3.7.6.5 , w Table 3.7.6.5-1 shall be OPIRABLE.The yard fire hydrants i APPLICABit11Y: ose houses sho n in hycrants is required to be OPERABLEWhenever equipment in th g; A y the ya'd fire a. With one or more of the yard fire h hose houses shown in Table23.7.6.5 ydrants or associated hydr s, , in an adjacent OPERABLsufficient additional lengths o inoperable er hose locptec the unprotected area (s)E hydrant hose p.therwise provide the additional hose w
\- b.
on; urs. The provisions
. s of Specifications ..
3.0.3 and 3 0 4 are not applicable. SURvlitbNiEREQUIREMENTS ,,
'F- .
\, 4.7.6.5 Each of the yard fire hydrants and
=
.- shown a. in Table 3.7.6.6-1 shall be demonstrated LE: DPERAB house to assure all required on of the equipment iAt l
- hydran 6.
s at the hose house.t hose A1 least once per 6 months, by visually in that the hdrant is not damaged arrel is dry and !
- c. .
I
,At least once per 12 months by:
1. Conducting a hose hydros 1 or at least 50 psig abov st at a pressure of 250 psig pressure, whichever is.gr...u . he maxim e fire main operating
- 2. . .
- 3. Replacement of a11 degraded gaskets ngs, in coupli Perforr.ing a flow check of each hydrant .
i i RIVER stND . Unti 1 3,4 y.gy #** l l
- . _ _ . . _ . _ _ ____ .__ _ _._ . _ _ _._ . ._.__ ._. _ _ . . _ _ _ _ _ = _ . _ _ ___ _ ._ __ _ ____i___ _ _ _ .
(LECTRICAL power SYSTEMS N $ h g un - SURVEILLANCE REQUIREMENTS (Continued) 6. Simulating a loss of offsite power in conjunction with an ECCS actuation test signal, and: a) For divisions I and II: E
- 1) Verifying deenergization of the emete'ncy busses and -
Ioad shedding from the emergency busses. '
- 2) Verifytre the diesel generator starts on the auto-start I g
signal, energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto- , connected ehe4down loads through the 4eed sequencing logic and operates for greater than or equal to . 5 minutes while its generator is loaded with the . emergency loads. After energization, the steady . state
- voltage and frequency of the emergency busses shaA1 he "
- maintained at 4160 1 420 volts and 60 2 3 Hz during this test. ;~
b) For division I!!: *"h
, ,b ;,3;
1) Verifying de energization of the emergency bus.
** t A pe r .. ,vly . .a. +eJ 2) Verifying the diesel generator starts on the auto- i start signal, energizes the emergency bus with its '
' l..Je w.fl...,to s u -Jg
; S :n ;;4 the auto-connected c r ; = ,, loads d t'-
eu e-3.e s M :: rt and operates for greater than or equal to - 5 minutes while its generator is loaded with the
- emer9ency loads. After energization, the steady state voltage and frequency of the emergency bus shall be ,
maintained at 4160 2 420 volts and 60 2 3 Hz during this test.
- 7. Verifying that all automatic diesel generator trips are automatic-ally bypassed upon E tuation signal except:
.i a) For divisto ngine oserspeed and generator differential en ,
b) For divist 'ngine overspeed and generator differential current. ( '
- 8. ~ verifying the diesel generator operates for at least 24 hours.
The diesel generators shall be loaded to 3130 kw for diesel ' generator 1A and 18 and 2600 kw for diesel generator IC. The - generator voltage and frequency shall be 4160 2 420 volts and 60 2 3 Hz within 10 seconds after the start signal;.the steady ( RIVER BEND + UNIT 1 3/4 4-6 APR 2 6 M
_ TABLE 4.8.1.1.2-1
*~ '
g , ,'
- DIESEL GENERATOR TEST SCHEDULE Numbyf Failures in Las vaQI valid Tests
- Test Frecuency J/
<1 . . - .
At least once per 31 days 2 At least once per 14 days
.s .
3
- At least once per 7 days 14 .
At least once per 3 days s - 5 -T-
. " Criteria for determining number of failures and number of valid A Regulatory Guide 1.108 Revision 1 tests shalt be in accordance wi 100 tests are determined on a per n August 1977, where the last 1
lb. uclear unit basis. For the purposes of this test schedule, only valid tests conducted after "last 100 valid tests."the OL issuance date shall be included in the computatio
,, made at the 31 day test frequency. Entry into this test schedule shall be l
- 1 i
l i APR 2 61005 RIVER SEND - UNIT 1 3f4 g.9 l
t 9. TABLE NOTATION N P ( s b continued j is the standard deviation of the background counting rate or of th e counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 108 ' { is the number of disintegrations per minute per microcurie , Y is the fractional radiochemical yield, when applicable. : l A is the radioactive decay constant for the particular radionuclide,,and i At for plant collection andeffluents is the elapsed time between the midpoint of sample time of counting. - Typica'1 values of E. V, Y, and't.t should be used in the calculation. 6 It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
,r '
l' b - A batch release is the discharge of liquid wastes of a discrete volume .
. Prior mixed to assure representative sampling.to sampling for analy c - The sivelyprincipal gamma radionuclides:
are the following esitters for which the LLD specification applies exclu-
- Mo-99, ts-134, Cs-137 Ce-141, and Ce-144.Mn-54, Fe-59, Co-58, Co-60, 2n-65, only these nuclides are to be considered. This list does not maan that Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be ,
pursuant to Specification 6.9.1.te. analyzed and reported in the Sen 6 - A composite 8 tional to th uan in which the quantity of liquid sampled is propor-of sampling e ,.:.,).. liquids released. results in a specimen that is representative of th i s j I APR 2 61985 I RIVER SEND - UNIT 1 3/4 11-3
,_ -_ - - = - -
i RADI0 ACTIVE EFFLUENTS ( J.10VID RADwASTE TREATMENT SYh1EM, LIMITING CONDITION FOR OPERATl?N ff 3.11.1.3 = radioactive materials in liquid astesoprior projected doses due to the liquid effluent reduce theto their d ' scharge when the organ in a 32 day period. Figure 5.1.3-1) a would exceed 0.06 nrem body or 0.2 arem .to any APPLICABILITY: At all times. ACTION: a. With radioactive in excess of liquid waste beinge discharg d . c.- , within 3D d bove limits, prepare and s without treatment and includes th svant to $pecification 6. to the Commission ing infornation: pecial Report that
,, 1. ).f I
h treatment, identification of any inoperab rged without u pment or
,[ 2. subsystems, and the reason for the inoperability ,
Actionand status, (s) taken to restore the inoperable equipm ent to OPERABLE
- 3. *
- b. Summary description of action (s) taken to prevent a recurrence.
The provisions of Specifications 3.0.3 and 3 0 4 .. are not applicable. IIJRVEILLANCE REQUIREMENTS 4.11.1.3. Doses due to liquid releases to UNRESTRICT at least once per 31 days in accordance with the meth d LED A the ODCM. o o ogy and parameters in RIVER BEND - UNIT 1 3/4 11-5 APR 2 61985
l l j.- . INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION . LIMITING CONDITION FOR OPERATION ' 3.3.2 shall The isolation actuation instrumentation channels shown in Table 3.3.2-1 be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE , TIME as shown in Table 3.3.2-3. APPLICABILITY: As shown in Table 3.3.2-1. . ACTION: i
- a. With an isolation actuation instrumentation channel trip setpoint t I
less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel - is restored to CPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. -
- b. With the number' LE channels less than required by the Minimum M -
CPERABLE Channel ip System requirement for one trip system, I place the inopera channel (s) and/or that trip system in the tripped condition
- within one hour. The provisions of Specification 3.0.4 f are not applicable.
l '
- c. With the number of OPERABLE channels less than required by the Minimum _
CPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
,- l
(, F a
~
cAn inoperable channel need not be placed in the tripped condition where thisIn th would cause the Trip Function to occur. shall be restored to CPERABLE status witMn 2 hours or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken. ,
**The trip system need not be placed When in system a trip the tripped condition can be placed inif the thistripped would cause the Trip Function to occur.
condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped , , , condition. WID m u ss - RIVER BEND - UNIT 1 3/4 3-10 I m
,,_ TABLE 3.3.2-1 (Ccntinued) ' Q Md M dan /.gh F M-[ Tj yj . ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 -
Be inthe within at next least24NOThours.SHUTDOWN within 12 hours and it! COL ACTION 21 -
' Close the affected system isolation valve (s) within one hour or:
a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within 24 the hou next 12 hours and in entn GU'TnnWN ! within the following ' l s/ b. In Operational Condition h, suspend CORE ALTERAnons, s handling of irradiated fuel in the primary containment and operations ~ with a potential for draining the reactor vessel, ; ACTION 22 - i Restore the manual initiation function to CPERABLE status within l 48 and hours in COLDorSHUT be in00W at least HOT within the SHUTDOW following 24 hours. within the next 12 hou ACTION 25 - Be in at least STARTUP with the associated isolation valves clositd within 6 hours or be in at least HUT SMUTDOWN within 12 hours and in COLD SHUTDOW within the next 24 hours. ACTION 24 - Be in at least STARTUP within 6 hours. ! f ACTION 25 -
\
- Establish SECONDARY CONTAINMENT INTEGRITY - OPERATIN ACTION 26 - standby gas treatment 'systenfoperating Aand Fuel Sveldy VentileWas* Sym within orte hou Restore the manual initiation function to OPERABLE status within 8 hours or be in at least NOT SHUT 00W within the next 12 hours and in COLD SHUTDOW within the following 24 hours.
ACTION 27 - Close the affected system isolation valves within one hour and declare the affected system inoperable. ' T0 - -
! ? "" C^".U .O
21 !"?!Ca !'" ' th t h: :t:r g ;:: tr::t :nt :;;r:t E; d th' ::: ':r. Initiate and maintain annulus mixing system with the reactor building annulus exhaust tohour. 1 at least one operating standby gas treatment train within. ) i ACTION 06 - 40 Lock the affected system isolation valves closed within one l hour and declare the affected system inoperable. l RIVER BEND - UNIT 1 3/4 3-17 i l
TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME
=
RESPONSE TIME (Seconds)w TRIP FUNCTION
- 1. PRIMARY CONTAINMENT ISOLATION I
- a. Reactor Vessel Water Level - Low Low, Level 2 << 10 10("))
a
- b. 7 10(a) - Higb(D)
Drywell Pressure - HighContainment Purge Isolation Radiation E
- c. NA
- d. Manual Initiation
- 2. MAIN STEAM LINE ISOLATION
- a. Reactor Vessel Water Level - Low Low Low, < 1.0 */< 10C")aa Level 1 I
- b. Main Steam Line Radiation - High(b) I 1.0 "A
- 810 10((*)aa a),, 7
. c. Main Steam Line Pressure - Low 7 0.5 8 8 10(*)aa
- d. Main Steam Line Flow - Hig RA
~
e. f. Condenser Vacuus - Low n.. Main Steam Line Tunnel y. y y_M,]v= NA
)k NA
- g. Main Steam Line Tunnel A Temperature - High NA
- h. _ Manual Initiation
- 3. SECONDARY CONTAINMENT ISOLATION -
Reactor Vessel' Water Level - Low Low, Level 2 < 10(a) a. T b. 7 10(**)) 7 10(Radiation - High(b) i I c. Drywell Pressure - HighFuel Building Ventilation Exhaust ~
- d. Reactor Building Annulus < 10(*)
Ventilation Emmaap Exhaust Radiation - High(b) 't MA
, t. Manual Initiation ;~
- 4. ,_ REACTOR WATER CLEANUP SYSTEM ISOLATION
< 10(a)## ,
l
- a. A Flow - High RA
- b. A Flow Timer NA
- c. Equipment Area Temperature - High NA
- d. Equipment Area A Temperature - High < lof ,) '
- e. Reactor Vessel Water Level - Low Low, Level 2 I l
- f. Main Steam Line Tunnel Ambient NA Temperature - High NA
- g. Main D - ' '-- Tunnel A Temperature - Hig$ NA
- h. SLCSCnit1[tiod NA f
- 1. Manuai Aniwea uon S. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
< 10I*)### l
- a. RCIC Steam Line Flow - High EA
- b. RCIC Steam Line Flow-High Timer < 10I ")'
- c. RCIC Steam Supply Pressure - Low NA
- d. RCIC Turbine Exhaust Diaphragm Pressure - High NA
- e. RCIC Equipment Room Ambient Temperature - High NA
- f. RCIC Equipment Room A Temperature - High NA
- g. Main Steam Line Tunnel Ambient Temperature - High NA
- h. Main Steam Line Tunnel A Temperature - High RIVER BEND - UNIT 1 3/4 3-24 APS I 6 slib
r ( . TABLE 4.3.2.1-1 (Continued) . 150tAff0N ACTUATION INSTRtMENTATION SURVElttANCE REQUIREMENTS
*:=
- CHANNEL OPERATIONAL i
FUNC110NAL CHANNEL CONDITIONS IN WHICH CHANNEL E TEST Call 8 Rail 0N SURVEILLANCE REQUIREO CHECK TRIP FUNCTION
- I g 3. SECONDARY CONTAINMENT ISOLATION Reactor Vessel Water IC 1,2,3 Q a. 5 M R *I w
Level - Low tow, tevel 2 M R II 1,2,3
- b. Drywell Pressure g~igh, _ , ,
5,, , 1? i
" M 8 R
- c. Fuel Building Aree ventilationr >
l Reactor Building Annulus
- d. -
i Ventilation Exha ^ R 1,2,3 ' 5 Mg ,) j g e. Radiation - Mi Manual Initiation NA M MA 1,2,3 ,
~~....
~
w 4. REACTOR WATER CLEANUP SYSTEM ISOLATION R 1, 2 3 5 M
- a. A Flow - High 1, 2, 3 w NA M Q 4 b. A Flow Timer' w
- c. Equipment Area Temperature - M R 1, 2, 3 5
High ,
- d. Equipment Area M R 1, 2, 3 a Temperature - Migh $
i
- e. Reactor vessel Water M- R(c) 1,2,3 Level - tow tow, tevel 2 5 1,2,3 smyg
- f. Main Steam Line Tunnel Ashlent 5 M R M.T: L B Temperature - High '
1, 2, 3 *
' g. Main Steam Line Tunnel M R 5
a Temperature - High 1, 2, 3 NA M(b) NA F
' h. ,SLCS Initiation
- l. Manual Initiation NA M(a) NA 1, 2, 3 W
.m g
m
.==5
' '~
~U ',t
~
' 7 ' r . ' l E. g
'l ,'
'f f -
e
. o TABLE 3.3.6-2 -
?*
CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS en TRIP FUNCTION . TRIP SETPolNT d .o ' . ALLOWA8tE VALUE
- 1. R00 PATTERN CONTROL SYSTEM
, e a. Low Power setpoint 27.5 TED THERMAL POWER 27.5 1 . M E RATED. THERMAL q b. High Power Setpoint 62.5 ATED THERMAL POWER POWE 62.5 1 .
ATED TNERMAL* POWER K w 2. APRM ,
- a. Flow Biased Neutron Flux - .
- Upscale '
< 0.66 W + 42%* < 0.66 W + 45%*
- b. Inoperative NA .
- c. RA Downscale Q%ofRATEDTHERMALPOWE4,
- d. Neutron Flux - Upscale - 1 3% of RATED THERMAL POWER .
Startup 3 12% of RATED THERMAL POWER $ 14% of RATED THERMAL POWER
- 3. SOURCE RANGE MONiiORS w a. Detector not full in NA 1 b. Upscale < 1 x 10 cps NA w c. Inoperative < 1.6 x 10 5 c,,
RA g d. Downscale- 1 0.7 cps. HA 1 0.5 cps **
- 4. INTERME0! ATE RANGE fRINITORS
- a. Detector not full in NA '
NA l b. Upscale
$ 108/125 division of full 1 110/125 division of full scale scale
- c. Inoperative NA
- d. NA Downscale > 5/125 division of full scale > 3/125 division of full scale "T1 ,
- 5. SCRAM OISCHARGE VotDNE *'
- a. Water level-High < 18 inches <
' _ 22 inches
- 6. REACTOR COOLANT SYSTEM RECIRCULAT10N FLOW Upscale g
- a. '
1 10G% of rated flow $ 111% of rated flow . i O es
= *The The tripAverage Power setting of this function Range Monitorinrod must be maintained blockwith accordance function Specification is varied 3.2.2. as a function
**Provided signal to noise, ratio is > 2, otherwise setpoint of 3 cps and allowable 1.8 cps. "Tj h
- l
- TABLE 4.3.7.2-1 N _ SEISMIC MONITORING INSTRUMENTATION SUR
CHANNEL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS __ CHECK FUNCTIONAL-CNANNEL _ TEST
- 1. CALIBRATION Triaxial Time-History Accelerographs
- a. Reactor Bldg. I
- b. 70'0" M Reactor 81dg hield Wall M SA R EL 232'0" SA R
- c. Reactor 81dg. Drywell EL 151'0" M ;
d. Free Field-Grade level M SA R SA
- 2. R Triaxial Peak Accelerographs s -
- a. Reactor 81dg. SLCS Stora d
- b. Reactor 81dg. - RHR Inj. ge TankNANA NA
- c. Piping R Aux. Bldg. Service Water Piping NA NA R
~
- 3. NA R Triaxial Seisaf~c~5 witches a
Reactor 81dg. Mat EL 70'0" M(a) SA
- 4. R Triaxial Respon,se-Spectrum Recorders a.
Reactor 81dg. Mat EL 70'0' k *- - t b.
- c. Reactor 81dg. Floor EL 141'0" H SA R Auxiliary Bldg. Mat EL 7 NA SA R
- d. NA Auxiliary Bldg. Floor EL NA NA NA R v
- R %
xcept seismic trigger. O e . k b RIVER 8END - UNIT 1 l 3/4 3-72 i
RADIOLOGICAL ENVIRONMENTAL MONITORING FINRL [mg 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATIO ( --' h . 3.81. 2. t.12.1 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological
- sectors of garden" of greater the nearest than milk 50 m2animal, the nearest residence and the nearest (500 ft 8) producing broad leaf vegetation.
APPLICABILITY: At all times. ' ACTION: a.
~. .
With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, identify the new location (s) in the next Seminannual ) to Specification 6.9.1.8.Radioactive Effluent Release Report, pursuant ' I b. Withlandusecensusidentifyingalocation(s)thatyieldsacalcufated dose or dose commitment (via the same exposure pathway 20 percent- - greater than at a location from which samples are curre)ntly being f
.- obtained in accordance with Specification 3.12.1, add the new location (s) to the radiological envirbnmental monitoring program within 30 days. '
The sampling location (s), excluding the control station location, having same - the lowest calculated dose or dose commitment (s), via the woure pa6. ;
, may be deleted from this monitoring program aftehetober3 condu % . T the year in which this land use census was !
..not to Specification 6.9.L8, identify the new
.. location (s) in the next Seminannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).
, j c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLA$CEREQUIREMENTS 4.12.2 The land use census shall be. conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Specification Annual Radiolcgical Environmental Operating Report pursuant to 6.9.1.7.
- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the site boundary in each of two different direction sectors with the highest predicted 0/Qs in lieu cf the garden census. Specifications
( for broad analysis including leaf vegetation sampling of control samples. in Table 3.12.1-1, 4c shall be followed. RIVER BEND - UNIT 1 APR 2 61985 3/4 12-13
i INSTRUMENTATION BASES
\
3/4.3.4 RECIRCULATION PUMP ~RI A ACL*~ "3N INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The study events in General Electric Company Topical Report NE0 March 1971 and NEDO-24222, dated December 1979, and Section 15.8 of the FSAR . The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of the Reactor reactor trip. Protection System and is an essential safety supplement to the margin which occurs at the end of-cycle.The purpose of the E0C-RPT is to r that the void reactivity feedback due to a pressurization transient can addT positive reactivity to the reactor system at a faster rate than the control . rods add negative scram reactivity. Each EOC-RPT system trips both recircu-the core during two of the most limiting pressurization The two events) the turbine stop valves and fast closure of the turbine control . -
~ . A fast closure sensor from each of two turbine control valves provides input to the EOC-RPf system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system.
j EOC-RPT system; a position switch from each of the othe provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to fom a 2-out-of-2 logic for the fast 1 closure valves. of turbine control valves and a 2-out-of-2 logic for the turbine stop , i The oper4 tion trip both recirculation of either logic will actuate the E0C-RPT system and pumps. - - Each EOC-RPT deinistratively system may be manually bypassed by use of a keyswit controlled. ch The manual bypasses and the automat ating
- m. ss at less than 40% of RATED THERMAL POWER are annunciated rol in th between arc, i.e. ,140 ms. initiation of valve motion and complete the suppThe electric EOC Included in this tin are: the espo 'me of the sensor, the system timelogic.
allotted for breaker are suppression and t e time of the within its specified Allowable Value is acceptable on the ba difference between each Trip Setpcint and the Allowable Value is eoval to or less than the drif t allowance assumec for each trip in the safety analyses. i RIVER BEND - UNIT 1 APR 2 61985 6 3/4 3-3 l l.
*a r- FINHL llEE I i NOTE: SCALF IN INCHES l
, . ABO' /ESSELZERO WATE EVEL N*MENCLATURE MEIGHT ABOVE NO. WESSEL ZERO (IN.) READING 3 300 - -
V\ (s) C 572.523 52 (7) sse.42 .,9 s.s
, (4) 551.42 430.s 750 - -
(3) 829.52 ,e.g i 722.75 VESSEL (2 : 475.12 45.5
- FLANGE =
375.12 700 - - 145.5 450 - - MAIN
- 535.5- STEAM .
LINE ftNSTRUMENT , j S00 - - f g M"--jer MN "52.82(s)) 52(g) [,33,3
~
;52 TRIP RPS'(8) 550 - =w.uu # NPCS, RCIC COTTOM OF STEAM r=51.42(4) 5 HiALARM ygips (4)".30.s LO CRYER SW N -529.52(3) ALARM 505.52 ,- 520.52 0- - 3,9(3) 0 FEED 483.5500 -
- 0- REACTOR SCRAh WATER - 475.12(2) CORE
~
" *45.5(2) CONFIRMATORY
-465 ' ADS TRIP
'~ gpgAy 7 INITIATE RCIC, HPCS
*. 450 - - g TRIP RECIRC. PUMPS 408.56 400 - - #
.r,- s 4 <
t :$s$:N"I -1s0 :.145.5n) 350 -- 354.56 INITIATE RHR AND LPCS, 6 START DIESEL, INITIATE
.[g //c ADS AND CLOSE MSIV'S j
g /Id ACTIVE FUEL 250 - - 20s.56 gog - = 206.56 U RECIRC d RECIRC -171.5- INLET
- OUTLET 166.5 150 - - NOZZLE
' 7 NOZZLE 100 - -
)' ;
50 - - i
^~ ~
Bases Figure 8 3/4 3-1 REACTOR VESSEL WATER LEVEL - -
)
RIVER BEND - UNIT 1 8 3/4 3-8 APR 2 61985 l
RADIOLOGICAL ENVIRONMENTAL MONITORING FINA!. gya; BASES s 3/4/12.2 LAND USE CENSUS f i JA c' a is specification is provided to ensure that changes in the use of areas ra Deyond the SITE BOUNDARY are identified and that modifications to the results of this census.o ogical environmental monitoring program are made if require
)
frombeaerial shall used. survey or from consulting with local agricultural authori i Appendix I to 10 CFR Part 50.This census satisfies the requirements of Section IV.Bl3 o! than 50 m2 Restricting the census to gardens of greater provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables 2 - assumed in Regulatory Guide 1.109 for consumption by a child. this minimum garden size, the following assumptions were made: To determine garden was used for growing broad leaf vegetation (i.e., similar to lettuce 1) 20% of the and cabbage), and 2) a vegetation yield of 2 kg/m2 , 3/4.12/3 INTERLABORATORY COMPARISON PROGRAM g Program is provided to ensure that independent checks o accuracy of the measurements of radioactive material in environmental sample C s monitoring in order to demonstrate that the results are of Section IV.B.2 of Appendix I to 10 CFR Part 50. O e RIVER BEND - UNIT 1 B 3/4 12-2 I
. ~
~
Enclosure 2 Errors of Comission
r1. 1. ) y *) _ ELECTRICAL POWER SYSTEMS
~
SURVEILLANCE REQUIREMENTS (Continued) .
~
- 3. i Verifying the diesel generator capability to reject a load of 3130 kwCfor generator diesel without generators IA and IB and 2600 kw for tripping.
The generator voltage shall not exceed 4784 volts for diesel generators lA and 1B er 5824 vol! ; 4 for diesel generator 1C during and following the loac n. reject Simulating a loss of offsite power by itself, and:
! a) For divisions I and II:
- 1) .
Verifying deenergization of the emergency busses and load shedding from the emergency busses. 2) Verifying the diesel generator starts on the auto-start connected loads within 10 seconds, en and operates for greater than or equal to 5 while its generator is loaded with the :L td: ~ loads After energization, the steady state voltage and . frequency at 4160 2 420of the emergency busses shall be maintained volts and 60 2 3 Hz during this test. ( b) For division III: 1) Verifying de energization of the emergency bus . 2)
-wem,energizes signal, Verifying the the diesel m
generator starts on connected loads withi with the permanently y greater than or equal (w :0+seconos y c operates 9for
' is loaded with the & + t m loads.- J ~ while its generator After energization bus shall be maintained atthe steady state voltage a 4160 2 420 volts and 1
60 1 M Hz during this test. ! 3
- 5. l Verifying that on ari ECCS actuation test signal , without loss i of offsite power, the diesel generator starts on the auto start 5signal minutes.and operates on standby for greater than or equal to 2 420 volts and 6013 Hz within 10 seconds e auto-start after thT
- be maintained within these limits during this tes .
{ y e L 0pr , n e,, ,- a a. . . ..> . s. i.a lpgre'j . a. ,9 m w. . i.e,
\
RIVER BENO - UNIT 1 3/4 8-5 APR 2 61985
794 . M M y. I 9 ELECTRICAL POWER SYSTEMS ( SURVEILLANCE REQUIREMENTS (Continued) 6. Simulating actuation a loss test signal, of offsite and: power in conjunction with an ECCS-a) For divisions I and II: 1) Verifying deenergization of the emeYVency busses and load shedding from the emergency busses. 2) Verifyirig the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently ." - connected loads within 10 seconds, energizes the auto-connected :t h c loads through the need sequencing logic and operates for greater than or equal to 5 minutes while ~its generator is loaded with the . ! - emergency loads. After energization the steady. state voltage and frequency of the emergenc,y busses shAl be maintained this test. at 4160 1 420 volts and 60 2 3 Hz during b) For division III: *" 4 1) ggf i Verifying de energization of the emergency bus. s A perm. s4Iy 6- +ed 2) Verifying the diesel generator starts on the auto- , le..f r w, f A.- 1^ start signal, energizes the emergency bus with its
, 10 s o.a.,4r, ----.)
caeq e e s 2 ;;4 the auto-connected - : ;;r.;3 loads th 5 M :: cert and operates for greater than or equal to A j 5 minutes while its generator is-loaded with the emergency loads. After energization the steady state
- hr/NT voltage and frequency of the emergenc,y bus shall be maintained at 4160 2 420 volts and 60 2 3 Hz during '
%737 DAW this test.
i 7. ally bypassed uponECCaVerifying that all automatic diesel generato ctuation signal except: l a) .K For division an differential engine overspeed and generator rren . i b) For divisi l current. engine overspeed and generator differential 8. l i The diesel generators shall be loaded to 3130 kw fo . generator IA and 18 and 2600 kw for diesel generator IC. The generator voltage and frequency shall be 4160 1 42& volts and
. 60 2 3 Hz within 10 seconds af ter the start signal; the steady .
(. RIVER 8END - UNIT 1 3/4 8-6 APR 2 61965 i
-= -_ . . - - - - - - - '-~~ ' ~ ~
FROM ATTACHMENT B TO GSU LETTER OF MAY 6, 1985. TECHNICAL CHANGE REQUESTS 1 DESCRIPTION OF CHANGE / JUSTIFICATION:
- 28)TS 3.7.6.2 - Deleted Railroad Bay.
No sprinkler systems are identified for the railroad bay as there is no safety related equipment located in this area.
% 29) TS 4.7.6.3.a - Delete. g p,f g, 4
There are no valves in the flow path of any PGCC subsystem.
---+ 3 0 ) TS 3/4.7.6
( f ' Table 3.7.6 1 - Added footnote *. Reflects R,iver Bend design.'
- 31) TS Table 3.7.'8 Add items and revise temperatures.
Additional item have been identified for inclusien and corrections to temperatures from review of EnvironmentaT Design Criteria.
- 32) TS 3/4.7.10 -
Added Table 3.7.10-2, revised the Technical Specification accordingly and also revised Table 3.7.10-1. These changes make the Technical Specification consistent with FSAR Section 2.5.
- 33) TS 3/4.7.11 - Add new Specification.
This Specification is provided to address SER requirement in , 9.1.3 page 9-6. 34).TS 3/4.8.1, 3.8.1.1 Action c, 4.8.1.1.2.f.4.b.2, 4.8.1.1.2.f.6.b.2, 3.8.1.2 Action b, 3.8.2.1 Action b, 3.8.2.2 Action b, and 3.8.3.1 Action b.2, - Addition of C SSW pump. Revisions reflect the ~ powering of standby service water pump l ISWP*P2C and it's auxiliaries from the HPCS diesel (Div III).
- 35) TS 3.8.3.1.b.1 and 3.8.3.2.b.2 - Added panel IENB*FNLO4A.
Added in conjunction of outstanding SER open item 13, Safe / Alternate Shutdown Design Modification. 1 Page 6 of 7 l 4
g . N b
) 2.5 ~
$E; l di s =s t l f E! 1
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? 8 M:
0 ml> n:: nj y_ . si @r ? i .',. '\ . s t
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'f ~0 ifiIP 8,
i .
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4
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e = a 9 D .
\ -i - .
s - 1 W O i i L =
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g, s-~ 4g , hb gamm um.g e s l , f' ac - ---
- f
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- I' I m
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Edward J. Butcher May 17, 1985 same submittal, GSU properly proposed an idertical change with the clause to be inserted after 10 seconds. As proposed on 3/4 8-5, the revision makes no sense and would confyse the operator. (2) In Attachment B of the GSU submittal of May 6,1985, Item 29 requests a deletion of a surveillance requirement because -
"There are no valves in the flow path of any PGCC subsystem." -
In past discussions, GSU has resisted this requirement on the, basis that the valves did not have a position indicator although, in fact, the valves do have a trip indicator. A copy of FSAR Figure 9.5-13 is enclosed which shows numerous solenoid operated valves as well as a couple of check valves in the flow path. Therefore, the GSU statement of "no valves in the flow path" appears to be a false representation. * (3) Also, in Attachment B. Item 30 refers to adding a footnote to TS 3/4.7.6.4, Table 3.7.6.4-1. This is in error as the proposed footnote was identified with TS 3/4.7.6.5, Table 3.7.6.5-1. In addition to the deficiencies noted above, I would like to comment briefly on other uncertainties associated with the River Bend Tech Spec review. GSU has submitted a listing of 55 areas in the FSAR that need revision to support Tech Spec' sections. Amendment 19 to the FSAR was delivered on May 14, 1985 and only 12 ef these areas were addressed. Therefore, in the other 43 areas, the NRR Technical Reviewer has not seen the necessary documentation to support the current Tech Spec section or a proposed revision to a section. There is also the potential for additional FSAR revisions resulting from the reviewer's evaluation. This lack of timely information will impact the accelerated schedule for issuance of the Tech Specs with the River Bend license in June, 1985. l There seem to be some values in the FSAR and Tech Specs that are constantly ' being changed. For example, the DBA activity release to the environment 4 following a LOCA (used for containment Tech Spec review) were revised in i Amendment 18 to the FSAR dated April 1985 and revised again (increased) in Amendment 19 on May 13, 1985. In the Tech Specs, GSU has proposed that the water level for the Ultimate Heat Sink be 112'4" (2nd Draft),108"6" l (Final Draft) and 111'10" (current revision). Changes of this frequency would indicate that the utilities review process has not settled down. All of the above matters should be given due considerations when discussing commitments and completion schedules for the River Bend Tech Specs. 1 Original signed by ! M. Dean Houston, Reactor Engineer l Technical Specification Review Group Division of Licensing -- l cc: D. Crutchfield TSRG:DL Distribution R. Benedict DHouston:jc Docket File TSRG File S. Stern 5/r)/85 7, _ _ ,, - - - . , - - . _ . _ . . . - . .
.,.,._,,.,,,...g.. , . . _ . . , , . - - _ - _ . - - .--,.._,,_ye, , - - . -
. k 31 TABLE 3.3.3-2 (Continued)
E EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENT { TRIP FUNCTIONg ' ALLOWABLE ' C. TRIP SETPOINT e DIVISION O TRIP SYSTEM _ VALUE e 1. HPCS SYSTEM
=
o.4
-e a.
e-. Reactor Vessel Water Level - (Low Low, Level 2) ,
>-45.S
~
1nches* >-47.7 inches
- b. Drywell Pressure - High '
~
- c. Reactor Vessel Water level - High, Level 8 i 1.68 psig 1 1.88 psig
- d. Condensate Storage Tank Level - Low < 52 inches *
< 54.2 inches e.
Suppression Pool Water Leve1 - High I 2' *-d::** o enches I!?':S: ,
** --4.5 inc.he d '
- f. Pump Olscharge Pressure - High , 5 15 'n:M:" */.0,,,dvs 3 -1Grg-ic " 8.0 inca e d C'
- g. HPCS System Flow Rate - Low 2 45psy _'; !?^ psig * :- ::h:p
- h. Manual Initiation iMA$^^ - r 2 W agpen 7 M5 p;49 -inc- : 8 .; 211.opsy WS-gpm 2 scogpm NA R D. LOSS OF POWER ,
b 1. Olvision nd h a. 4.16 kw Emergency 8us Undervoltage a. (Sustained Undervoltage)## 4.16 %if Basis - 3 g o,3 1 3 2970 9 sec.volts time 2970 1 148 volts b. delay j4:".055:::.timedelay v -{ 4.16 kw Emergency Bus Undervoltage a. 3 3 ,33 ,,,,, - (Degraded Voltage) 4.16 kw Basis - E3740 volts 3740/,387 volts fe0I 4 % 4-49time sec.delay d.ot(w/o M sne. S^4 .055 :: . time delay / LOCA) {'60.055 :::. time delay 31 o.3 - -+9 sec. time 3t o.33 %
' delay (w/e LOCA)
"11 v D -
sish **mu
- &s to Cps Dm
- 0 3 "
3:0
- =
"??
w
a- .. -a-w... ,w ,s - -
--s-.
- ENCLOSURE 4 e.s,
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, \ j T '
~
i mj e'Isi j.bN' k! ut 3 I J* I , s!" e .2 : 14 3! 1 3, J. n V- ! r"N',J {g g@ L4 " .
;- 10 gig i a.
E 3,![ a 1
#s d .'k,9 r J 1
\ ~ E gl 4 .E.. .- ' 2 e; d..
'. g . "" y & M .
- U R
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- j Il 5 I
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.= 1 ai $ gg l~t l8 a a li '
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. _ . _ _ IIYII !!* - - .-
EMERGENCY CORE COOLING SYST_ EMS "
\
SURVEILLANCE REQUIREMENTS 4.5.1 ECCS division 1, 2 and 3 shall be demonstrated OPERABLE by: a. At least once per 31 days for the LPC5, LPCI and HPCS systems: 1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the sy~sfeTfsolation valve is filled with water.
- 2. hri nhhat each valve, manual, power operated or automatic in position, is in its correct position.... .., . sow path that is not locked, b.
Verifying that, when tested pursuant to Specification 4.0.5, each:
* . 1.
LPCS pump develops a flow of at least 5010 gpm with a pump differential pressure greater than or equal to 281 psid. 2. LPCI pump develops a flow of at least 5050 gpm with a pump differential pressure greater than or equal to 100 psid. f 3. {- HPC5 pump develops a flow of at least 5010 gpm with a pump differential pressure greater than or equal to 399 psid.
\. c.
For the LPCS, LPCI and HPCS systems, at least once per 18 months, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in
.- the flow path actuates to its correct position.
tion of coolant into the reactor vessel may be excluded fromActual injet-this test. - d. For the HPCS system, at least once per 18 months, verifying that the suction is automatically transferred from the condensate storage tank to the suppression pool ore a condensate storage tank low water level signal and on a suppression pool hign water level signal, and verifying that the HPCS system will automatically restart on Reactor Vessel Water Level - Low Low, Level 2. i i i
~.^ )
i APR 2 61985 RIVER BEND - UNIT 1 3/4 5-4
- - - . - , , - --. ., - ,-w-- ,- - ,
- r. , . _ . . , = , - - - - - - . . -
4 CONTAI2fENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued) d. The combined leakage rate for all penetrations shown in Table 3.6.1. as annulus bypass leakage paths exceeding 13,500 cc/hr, or-e. The combined leakage rate, for all valves shown in Jah14 3.6.4-1 to be equipped with FVLCS, exceeding 170,000 cc/hr, or . f. The measured combined leakage rate for all containment isolation valve in hydrostatically tested lines per Table 3.6.4-1 which penetrate the primary valves. containment exceeding 1 gpa times the total number of such restore: ,
- a. .
, . The overalland applicable, integrated leakagear~ te(s) to less than 0.75 La as b.
c. to Type B and C tests to less than or equal to 0.60 The measured leakage rate to less than 340 scfh for each of the valve (
' d.
groupings identified in 3.6.1.3.c.1, 3.6.1.3.c.2, and 3.6.1.3 and . The combined leakage rate for all penetrations shown...in Table 3 61
- e. as annulus bypass leakage paths to less than or equal to 13,5 The combined leakage rate, for all valves shown in Table 3.6.4-1 to b equipped with PVLCS, to less than or equal to
- f. ' 170.000 cc/hr, and ,
. The combined tion valves leakage rate in hydrostatically for al1C tes1.eu . C 2.5. ; dntainment isola-
...o yu
. use 3.6.4-1 which f
the total number of such valves, penetrate the primary containm prior to increasing reactor coolant systen temperature above 200*F. SURVEILLANCE REQUIREMENTS 4 4.6.1.3 The primary containment leakage rates shall be demonstrated at the follo i ing test schedule and shall be determined in confomance with the criteria in Appendix J of 10 CFR 50 using the methods and provisions eof ANS . a. Three Type A Overall Integrated Containment Leakage Rate tests shall psig, during each 10 year service period.be conducted at 40 The third test of each set shall be inservice conducted during the shutdown for the 10 year plant inspection. s : t RIVER BEND - UNIT 1 3/4 6-4 , _C9,- - - - - ' ' ~
CONTAINMENT SYSTEMS s MSIV LEAKAGE CONTROL SYSTEM f.h Mhi w g eg[ia LIMITING CONDITION FOR OPERATION 3.6.1.5 divisions shall be OPERABLE.Two independent sain steam positive leakage cont APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. -"~~ ACTION: With one MS-PLCS division inoperable, restore the inoperable division to
- a. .
12 hours and in COLD SHUTDOWN within the following 24 ho .
- _ SURVEILLANCE REQUIREMENTS 4.6.1.5 Each MS-PLCS division shall be demonstrated OPERABLE:
a. By performing Surveillance Requirement 4.6.1.10.a.
- b. At leas't once aa- " '-~
by verifying compressor OPERABILITY by operatingth{ospres Roaded for at least 15 minutes. c. During each COLD SHUTDOWN, if not performed within the previous 92 days, by cycling each remote, manual and autamatic motor operated valve through at least one complete cycle of full travel. d. At least once per 18 months by performance of a functional test which includes simulated actuation of the division throughout its operating' sequence, and verifying that each automatic valve actuates to its correct position in each steam and that 8.513 psid sealing pressure is established line. An mm RIVER BEND - UNIT 1
- 3/4 6-10 W
CONTAINMENT SYSTEMS
.R S 9 m.,
x ..p DRYWELL BYPASS LEAKAGE " LIMITING CONDITION FOR OPERATION ' 3.6.2.2 Drywell bypass leakage shall be less than or equal to 10% of the minimum acceptable A/8 design value of 1.0 ft.2 APPLICABILITY: When DRYWELL INTEGRITY is required per Specifica ion 3.6.2.1. ACTION:
- With the drywell bypass leakage greater than 10% of the minimum acceptable A/4 design value of 1.0 ft.a. restore the drywell bypass leakage to within the limit prior to increasing reactor coolant system temperature above 200*F .
SURVEILLANCE REQUIREMINTS 4.6.2.2 The drywell bypass leakage rate test shall be per 18 months at an initial differential pressure of 3. d at least once
- i k-'* shall be calculat'ed from the measured leakage. One d s
dtheA/4 leak tested during at least every other leakage rate test. remain ! a. If any drywell bypass leakage test fails to meet the specified limit, the schedule for subsequent tests shall be reviewed and approved by the Commission.
.- If two consecutive tests fail to meet the limit, a I test shall be performed at least every 9 months until two consecutive
- tests be meet the limit, at which time the 18 month test schedule may resumed.
! b. The provisions of Specification 4.0.2 are not applicable. e e RIVER BEND - UNIT 1 3/4 6-19
CONTAINMENT SYSTEMS a LIMITING CONDITION FOR OPERATION (Continued) . c ACTION: (Continued) 4 2. I With the suppression pool average water temperature greater than: 3 a) 95'F for more than 24 hours and THERMAL POWEL egreat'r than 1% of RATED THERMAL POWER, be in at least HOT SHUTOOWN within 12 hours and in COLD SHUTOOWN within the next 24 hours,
' b)
"- 110*F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the l suppression pool cooling mode. i
~ . 3. ~
With the suppression pool average water temperature greater than - 120*F, 200 psigdepressurize within 12 hour the reactor pressure vessel to less than
- c. With only one suppressio )
water level indicator OPERABLE : and/or indicators with fewer than eigit suppression pool water temperature ' I the inopera,ble indione in each of W eight locations, OPERABLE, restore i f, ,. o verify suppressic to OPERABLE status whithin 7 days or within the limits .6p w ater t level and/or ter.perature to be ce per 12 hours. I. d. 1 With no suppressiop #;l water level indicators OPERABLE and/or with fewer than sevw wppression l coveringatleastsevenlocations,poolwatertemperatureindicators,Jf OPERABLE, restore at least one i water level indicator and at least six water temperature indicators to OPERABLE status within 48 hours or be in at least HOT SHUT within 24 hours. the next 12 hours and in COLD SHUTDOWN within the following _ SURVEILLANCE REQUIREMENTS , 4.E.3.1
- The suppression pool shall be demonstrated OPERABLE:
a. 1 By verifying the suppression pool water volume to be within the limits at least once per 24 hours,
- b. At least once per 24 hours in OPERATIONAL CONDITION 1 or 2, by verifying the suppression p,ool average water temperature to be less than or equal to 95'F, except:
1. At least once per 5 minutes, during testing which a ds heat to the suppression pool, by verifying the suppression pool average water temperature less than or equal to 105'F. l
)
RIVER BEND - UNIT 1 3/4 6-28 APF 2 61985
CONTAINMENT SYSTEMS SECONDARY CONTAINHENT AUTOMATIC ISOLATION DAMPERS LIMITING CONDITION FOR OPERATION 3.6.5.3 The secondary containment ventilation system automatic isolation dampers shown in Table 3.6.5.3-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.5.3-1. - ----- APPLICABILITY: As shown in Table 3.6.5.3-1. ACTION: With one or more of the secondary containment ventilation system automatic isolation dampers shown in Table 3.6.5.3-1 inoperable, maintain at least one isolation damper CPERABLE in each affected penetration that is open, and within 8 hours either:
- a. F Restore the inoperable damper (s) to OPERABLE status, or .
b. Isolate each affected penetration by use of at least one deactivated automatic damper secured in the isolation position, or _f. k c. Isolateoreach valve affected penetration by use of at least one closed manual blind Tb tWs. e d spu,. flange.hew T c.v ar<
- c. api.c.. 6tc. , -
Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTOOWN within the next 12 hours and in COL OWN within the followin Otherwise, in Operational Conditio (GuaWH/ZeGf.) g 24 hours. a^ar ^ uspena
' " ^
nandling or irradiated M Ss.bg ; fuelt:7.g;; in the,,, -----^-" ---+ = 4 a---* "e -ad aaa--+'--- '+' - - tion 3.0.3 are ,,,;,.i..s the .; ;i:r ::::d. The provisions of Specifica-not applicable. , i SURVEILLANCE REQUIREMENTS 4.6.5.3 l Each secondary containment ventilation system automatic isolation damper shown in Table 3.6.5.3-1 shall be demonstrated OPERABLE: , a. Prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actuator, control or power circuit, by cycling the damper through at least one complete cycle of full travel and verifying the specified isolation time. nen irraciatea fuel
#E c a m m . - ;is;: being handled in the [ [Nd!I : bi::r':rdde*;
- :. = u e e drg -< g .se -- - . > .
l RIVER BEND - UNIT 1 3/4 6-52 SPS 2 S Hi
CONTAINMENT SYSTEMS
- i FUEL BUILDING VENTILATION
- LIMITING CONDITION FOR OPERATION 3.6.5.6 i Two independent Fuel Building Ventilation Charcoal Filtration sub-systems mode.
emergency shall be OPERABLE, and in. QPERATIONAL,CpHDITION,*, one operating
~
t
. w,,,. 4 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *.
ACTION:
~ a. ~
With one Fuel Building Ventilation Charcoal Filtration subsystem inoperable, within restore 7 days, or: the inoperable subsystem to OPERABLE status 1. In-0PERATIONAL CONDITION 1, 2 or 3, be in at least. HOT SHUTDOWN within 24 the next 12 hours and in COLD SHUTDOWN within the foll) win hours.
- 2. In Operational Condition
- in the :::rf: ; -aaM'- : t, C^"E, suspend handling of irradiated fdel S
Fuel A/,Sqs"'"ions - --^of rti:' ':- : :i 'r.; tt. . ".LT Specificatio x:t:r"".TIO"5
;;;;;;. :-dThe:;;r;;i:n;provi-n 3.0.3 are not applicable.
b. With both Fuel Building Ventilation Charcoal Filtration subsystems
.. inoperable or with one not operating in the emer2ency mode in Opera-Fu. i kidq
~~ H=7 Condition tional "*'=;.J., *, suspend handling of irradiated fuel in the see-C= ALT =TI= - :;r:t f r: eith ; y- .;i t r dr;i 'n; th: 7. nter ;;; nl.
tion 3.0.3. are'not applicable. The provisions of Specifica-SURVEILLANCE REQUIREMENTS 4.6.5.6 demonstrated OPERABLE:Each Fuel Building Ventilation Charcoal Filtration subsyste
.. e...
/ %
a. At least once per 12 hours in OhERATIONAE CDHDITION'* one Fuel Building Ventilation Charcoal Filtrat hu verifying ystem eration. b. At least once per 31 days by initiating, from the cont.rorroom, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours with the heaters OPERABLE.
"When irraciated fuel is being handled in the Re cr^^f: 8w/ dong.
CCC: ALT:""! M rf :;r:t'-" ut' : ;:t r t h ! '- ydr;inin; - r tr.i ..;nt :-d R ~ s , th: rG;ter ;5ae n l l ( RIVER BEND - UNIT 1 3/4 6-61 , D#IT
CONTAINMENT SYSTEMS l - SURVEILLANCE REQUIREMENTS (Continued)
- a. Manual initiation from th_e control room, and
- b. Simulated automati signal.
4. Verifying that the filter cooling bypass dampers-een be manually opened and the fan can be manually started. 5. Verifying that the heaters dissipate >49 kw when tested in accordance with ANSI'N510-MM(1980
f.
-- After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and bypass leakage testing acceptance criterion of less than 0.05% in -
~
accordance with ANSI N510-4975 while operating the system at a flow rate of_10,000 cfm 2 10%. /T#8 ' - p
- g.
After each complete or partial replacement of a charcoal adsorber ? bank, by verifying that the charcoal adsorber bank satisfies the e inplace penetration and bypass leakage testing acceptance criterion
. .) of last than 0.05% in accordance with ANSI M510-t995 for a halo-genated at 'a flowhydrocarbon rate of 10,000 refrigerant cfm 210%. test gas while[ operating the system N ro f
I e* en e e a RIVER BEND - UNIT 1 3/4 6-63 h?9 2 : sec
~
( ULTIMATE HEAT SINK h
, b% ,
SURVEILLANCE REQUIREMENTS '
- 4. 7.1. 2 OPERABLE: The standby cooling tower and water storage basin rmined shall be a.
At least once per 24 hours bys
--- > 4e+ water verifying 1 to be within their limits. the basin 2:f * ;:= t=
' e g. At less cell fr r 31 days by starting the cooling tower fans in each
~ control room and operating the fan for at least,15 minu .
f 6, Dvein3 1ke. mniks af Ivne Hsitusk 5*fic**'k, be Mr A
.hwes of n500 ud Itcc verify the. basin waks lemprafix al
' skvaEim 9E ' t 3. ' Inst. .
(appmunniety gende eteel;Q ' 1o{ he belew H.r hMd : i (' f I. as Afbelow ka.s+
, ,5 *ence-r p 79 ykur -fk. ym.via,s m,hg 2
af leasi ence pm .1y hwes wk tk punk:s
.- n.eskn3 was grut4u-t%,7s*r .
e i ( RIVER BEND - UNIT 1 3/4 7-4 APR 2 61985 i
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