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TVA-SQN-TS35 Change No. 2 JUSTIFICATION FOR CHANGE IN MAXIMUM ISOLATION TIME FOR CONTAINMENT ISOLATION VALVE During a phase A isolation, the outboard containment isolation valve (FCV-62-77) usually isolates before the inboard isolation valve closes (FCV-62-72, 73, or 74); see attachment 1 for the measured stroke times of FCV-62-72, 73, 74, and 77. This will result in overpressurizing the letdown line between the containment isolation valves and lifting the relief valve 62-662. This specification change will allow Sequoyah to adjust the stroke time of FCV-62-77 to approximately 10 - 20 seconds, thereby allowing the inboard isolation valve to close before FCV-62-77 closing. | TVA-SQN-TS35 Change No. 2 JUSTIFICATION FOR CHANGE IN MAXIMUM ISOLATION TIME FOR CONTAINMENT ISOLATION VALVE During a phase A isolation, the outboard containment isolation valve (FCV-62-77) usually isolates before the inboard isolation valve closes (FCV-62-72, 73, or 74); see attachment 1 for the measured stroke times of FCV-62-72, 73, 74, and 77. This will result in overpressurizing the letdown line between the containment isolation valves and lifting the relief valve 62-662. This specification change will allow Sequoyah to adjust the stroke time of FCV-62-77 to approximately 10 - 20 seconds, thereby allowing the inboard isolation valve to close before FCV-62-77 closing. | ||
Additional justification for this proposed change is documented in the NRC May 5,1981 letter from Darrel G. Eisenhut to All Licensees of Operating Plants regarding the " Engineering Evaluation of the H. B. Robinson Reactor Coolant System Leak on January 29, 1981." The engineering evaluation was prepared by Wayne D. Lanning of the NRC and is dated March 23, 1981. This evaluation discusses the RCS leak at H. B. Robinson and one of the problems experienced during the event was the rupture of the CVCS relief valve bellows. The discussion on pages 10 ar.d 12 of the repor t was: | Additional justification for this proposed change is documented in the NRC {{letter dated|date=May 5, 1981|text=May 5,1981 letter}} from Darrel G. Eisenhut to All Licensees of Operating Plants regarding the " Engineering Evaluation of the H. B. Robinson Reactor Coolant System Leak on January 29, 1981." The engineering evaluation was prepared by Wayne D. Lanning of the NRC and is dated March 23, 1981. This evaluation discusses the RCS leak at H. B. Robinson and one of the problems experienced during the event was the rupture of the CVCS relief valve bellows. The discussion on pages 10 ar.d 12 of the repor t was: | ||
. . .from an operational consideration, overpressurizing the CVCS could be prevented, provided the orifice isolation valves were closed before the outboard isolation valves. Correcting the valve closing sequence for isolation would also reduce the challenge to the relief valve. | . . .from an operational consideration, overpressurizing the CVCS could be prevented, provided the orifice isolation valves were closed before the outboard isolation valves. Correcting the valve closing sequence for isolation would also reduce the challenge to the relief valve. | ||
The closing sequence for the isolt. tion valves appears to cause part of i | The closing sequence for the isolt. tion valves appears to cause part of i | ||
Latest revision as of 04:39, 1 June 2023
| ML20062N848 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/16/1982 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20062N833 | List: |
| References | |
| NUDOCS 8208230375 | |
| Download: ML20062N848 (52) | |
Text
{{#Wiki_filter:-- e * , e4 . ENCLOSURE 1 SEQUOYAH NUCLEAR PLANT TVA-SQN-TS-35 CHANGE NO. 1 DELETION OF FLOW RATE MONITORS FROM TABLES 3.3-13 AND 4.3-9 i i l l- 8208230375 820816 l' PDR ADOCK 05000327 i P PDR _ , , . L .i .)
1"
)
l l Enclosure 1 (Cont.) ! PROPOSED TECHNICAL SPECIFICATIONS TVA-SQN-TS-35 CHANGE NO. 1 i I l l l I
t TABLE 3.3-13 (Continued) El - jj , RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION 1
- MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION c: .
5
-4
~'
- 5. AUXILIARY BUILDING VENTILATION SYSTEM ,
- 42
- a. Noble Gas. Activity Monitor 1
- b. Iodine Sampler 1 44
- 44
- c. Particulate Sampler 1 I
- 4I d.. Sampler Flow Rate Monitor I
a.
- 6. SERVICE BUILDING' VENTILATION SYSTEM ,
* ** 42
$$ a. Noble Gas Activity Monitor 1 Flow Rate Monitor 1 41 w b,.
h!. . e
.i
- 4 8
B.-
-8 n
W
- g 9
e
u, TABLE 4.3-9 (Continued) j! RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 x
- CHANNEL MODES IN WHICH cE CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE Ej INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED
~~
- 5. AUXILIARY BUILDING VENTILATION SYSTEM
~
- a. Noble Gas Activity Monitor D M R(3) Q(2) *
- b. Iodine Sampler W N.A. N.A. N.A.
- c. Particulate Sampler W N.A. N.A. N.A.
- d. Sampler Flow Rate Monitor D N.A. R Q i
t
- 6. SERVICE BUILDING VENTILATION SYSTEM b a. Noble Gas Actvity Monitor D M R(3) Q(2)
- eg b. Flow Rate Monitor D N.A. R Q M
l l' s
$ ?
I' i
I Atti I 3.3-13 (Continneil) '
"m RADIDACTIVE CASE 005 EfI tuENT MON!!ORiflG lilSTRt!MENIAI10N 5
- r
' MINIHtit1 CilANNELS ACI10l1 OPERABLE APPLICABIL1TY g INS 1Ruf1LNI U 5. AUXILIARY Bult0 LNG VENTILATION.
,. y SYSTEM
- 42
- a. Noble Gas Activity Monitor 1
- 44 1
- b. Iodine Sampler
- 44
- c. Particulate Sampler 1
- 41 I
- d. Sampler Flow Rate Monitor f
, 6. _ SERVICE BUILDING VENTILAll0N SYSTEM s
- 42 Noble Gas Activity Monitor' I .
- a. ^ 41
[ b'. Flow Rate Monitor 1 4 co I Q S e
m TABIE 4.3-9 (Continneil) ~ E 8 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENIATION SURVEILLANCE REQUIREMENTS Y x
' CilANNEL MODES FOR WilICil CllANNEL SOURCE CilANNEL FUNCTIONAL SURVEILLANCE IS E lEST y INSTRUMENT CilECK CilECK CAllBRATION REQUIRED
- 5. -AUXILIARY BUILDING VENTILATION SYSTEM
- a. Noble Gas Activity Monitor D M R(3) Q(2) *
- b. Iodine Sampler W N.A. N.A. N.A.
- c. Particulate Sampler W N.A. N.A. N.A.
^
- d. Sampler Flow Rate Monitor D N.A. R Q f
, 6. SERVICE BUILDING VENTILATION SYSTEM x
^
[ a, Noble Gas Actvity Monitor D M R(3) Q(2) *
- b. Flow Rate Monitor N.A.
~
R a r D Q 8 9
l TVA-SQN-TS'35 Change No. 1 Justification for Deletion of Auxiliary Building Flow Monitor The auxiliary building gas treatment system starts and the auxiliary building vent (refer to figure) isolates on:
- 1. phase A containment isolation signal from either unit,
- 2. high radiation signal from fuel handling area radiation monitors, 3
high radiation signal from auxiliary building exhaust vent monitors, or
- 4. high temperature in the auxiliary building intakes for general supply fans.
The auxiliary building gas treatment system discharges to the shield building vent. For off-normal conditions, the flow rate of the auxiliary building vent is not required since it isolates. These systems are described in FSAR sections 6.2 3 and 9.4.2. At present, the auxiliary building flow rate is calculated by summation of the operating fan design flow rate (sunning the fans discharging into the auxiliary building vent). The fan flow rate was measured during the preoperational testing phase. In addition, we plan to measure the fan flow approximately once each refueling cycle and whenever modification is made to change the fan performance characteristics. Therefore, we find the summation of the operating fan measured / design (whichever is greatest) to be acceptable for use in normal offsite dose release calculations performed in requirement of technical specifications. o
-- .,- c y -- -
ENCLOSURE 2 SEQUOYAH NUCLEAR PLANT TVA-SQN-TS-35 ' CHANGE NO. 2 ' MAXI WM ISOLATION TIE FOR CONTAINMENT i ' ISOLATION VALVES 8 f i l 1 l I t l l l
Enclosure 2 (Cont.) PROPOSED TECHNICAL SPECIFICATIONS TVA-SQN-TS-35 CHANGE NO. 2 1 i , i 1 l
1 . . TABLE 3.6-2 (Coatinued) ,
~
CONTAINMENT ISOLATION VALVES , e S VALVE NUMBER FUNCTION 'MiXIMUM ISOLATION TIME (Seconds)
?
7 A.
- PHASE "A" ISOLATION (Cont.)
$ 32. -FCV-62-63 RCP Seals 10
-i
- 33. FCV-62-72 Letdown Line 10
- 34. 'FCV-62-73 Letdown Line 10
- 35. FCV-62-74 Letdown Line 10
- 36. FCV-62-77 Letdown Line 20
- 37. FCV-63-23 Accum to Hold Up Tank 10
- 38. FCV-63-64 WDS N to Accum 10 2
- 39. FCV-63-71 Accum to Hold Up Tank 10
- 40. FCV-63-84 -
Accum to Hold Up Tank 10
- 41. FCV-68-305 WDS N to PRT 10
- 42. FCV-68-307 PRTt$GasAnalyzer 10 R 43. FCV-68-308 PRT to Gas Analyzer 10
- 44. FCV-70-85.. Cd5 from Excess Lt Dn Hx *. 10
? 45. FCV-70-143 CCS to Excess Lt Dn Hx 60 g 46. FCV-77-p RCDT Pump Disch 10
- 47. FCV-77-10 RCDT Pump Disch 10
- 48. FCV-77-16 RCDT to Gas Analyzer 10 .
. 49. , FCV-77-17 RCDT to Gas Analyzer 10
- 50. FCV-77-18 RCDT and PRT to V H 10
- 51. FCV-77-19 RCDT and PRT to V H 10
- 52. FCV-77-127 Floor Sump Pump Di,sch . 10
- 53. FCV-77-128 Floor Sump Pump Disch 10
, 54. FCV-81-12 Primary Water Makeup 10 m 55. FCV-87-7 UHI Test Line 10 m 56. FCV-87-8 UHI Test Line 10
- 57. FCV-87-9 UHI Test Line 10 '
] 58. FCV-87-10 UHI Test Line 10
. 59. FCV-87-ll UHI Test Line 10 0 60. FCV-26-240 Fire Protection Isol. 20 S 61. FCV-26-243 Fire Protection Isol. 20.
TABLE 3.6-2 (Continued) ~ $ CONTAINMENI IS0lATION VALVE.S E! E FUNCTION MAXIMUM ISOLATION TlHE (Seconds)
, VALVE NUMBER c:
35 A. PHASE "A" ISOLATION (Cont.) 10 ~4 32. FCV-62-63 RCP Seals 10
- 33. FCV-62-72 Letdown Line 10
- 34. FCV-62-73 Letdown Line 10
- 35. FCV-62-74 Letdown Line Letdown Line 20
- 36. FCV-62-77 10
- 37. FCV-63-23 Accum to lloid Up Tank 10
- 38. FCV-63-64 WDS N2.to Accum 10
- 39. FCV-63-71 Accum to lloid Up Tank l Accum to Hold Up Tank 10
- 40. FCV-63-84 -
10
- 41. FCV-68-305 WDS N to PRT PRTt$GasAnalyzer 10
., 42. FCV-68-307 10 10 , 43. FCV-68-308 PRT to Gas Analyzer 10
- 44. FCV-7i)-85 DCS from Excess Lt On l{x ,
as CCS to Excess Lt On lix 60 n', 45. FCV-70-143 10
- 46. RCOT Pump Disch FCV-7J-9 10
- 47. FCV-77-10 RCDT Pump Disch RCDT to Gas Analyzer 10
- 48. FCV-77-16 10 49., FCV-77-17 RCDT to Gas Analyzer 10
- 50. FCV-77-18 RCDT and PRT to V li 10
- 51. FCV-77-19 RCDT and PRI to V 11 Floor Sump Pump Disch 10
- 52. FCV-77-127 10
- 53. FCV-77-128 Floor Sump Pump Disch Primary Water Makeup 10
- 54. FCV-81-12 ~ 10
- 55. FCV-87-7 UHI Test Line 10
- 56. FCV-87-8 UHI Test Line
- 10
- 57. FCV-87-9 UHI Te'st Line 10
- 58. FCV-87-10 UHI Test Line 10
- 59. FCV-87 UHI Test Line Fire Protection Isol. 20
- 60. FCV-26-240 20
- 61. FCV-26-243 Fire Protection Iso'1.
TVA-SQN-TS35 Change No. 2 JUSTIFICATION FOR CHANGE IN MAXIMUM ISOLATION TIME FOR CONTAINMENT ISOLATION VALVE During a phase A isolation, the outboard containment isolation valve (FCV-62-77) usually isolates before the inboard isolation valve closes (FCV-62-72, 73, or 74); see attachment 1 for the measured stroke times of FCV-62-72, 73, 74, and 77. This will result in overpressurizing the letdown line between the containment isolation valves and lifting the relief valve 62-662. This specification change will allow Sequoyah to adjust the stroke time of FCV-62-77 to approximately 10 - 20 seconds, thereby allowing the inboard isolation valve to close before FCV-62-77 closing. Additional justification for this proposed change is documented in the NRC May 5,1981 letter from Darrel G. Eisenhut to All Licensees of Operating Plants regarding the " Engineering Evaluation of the H. B. Robinson Reactor Coolant System Leak on January 29, 1981." The engineering evaluation was prepared by Wayne D. Lanning of the NRC and is dated March 23, 1981. This evaluation discusses the RCS leak at H. B. Robinson and one of the problems experienced during the event was the rupture of the CVCS relief valve bellows. The discussion on pages 10 ar.d 12 of the repor t was:
. . .from an operational consideration, overpressurizing the CVCS could be prevented, provided the orifice isolation valves were closed before the outboard isolation valves. Correcting the valve closing sequence for isolation would also reduce the challenge to the relief valve.
The closing sequence for the isolt. tion valves appears to cause part of i the CVCS to be pressurized to the setpoint of the relief valve and may i be contributing to the failure of the relief valve bellows whenever the system is isolated. Therefore, based on this event at H. B. Robinson, we request that the maximum allowable stroke time of FCV-62-77 be changed to 20 seconds. 1 i t 1
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- cl STROKE VI"lis I" SECnNJS R ECOSD D : C _ __ _ _ _R E3 0 T_E . .L O C A L _ . _. __ - _ .NYMBE.R_,_ . . _
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__J-FCV NA 1344 1-FCV- 73 C1/26/B2 166.1 4.0 ___ ___ ATTACHMENT 1 . __. __ . (Page _2 _ o f_4) __ _ ___. ._ _ . .._ _... _. __
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1 -F C *l 77 05/22/or 166.1 2.' NA 1365 l 1-FCV 77 C6/16/30 166.5 1.7 NA 1364 FCV 77 0 8 /21/a ? 166 1 24 NA 1367 1-FCV- 62- 77 11/E 2 / c 0 16;.1 1.9 NA 136C 1 rCV- 62- 77 ~1/04/c1 166.6 1.C 2.C 136' 1 cCV- 62- 77 22/19/A1 166 1 2.0 NA 137F l 1 cCV- 62- 77 03/23/a1 166.6 2.C 19 1371 l 1-FCV- 62- 77 93/?!/61 166.6 21 18 1372 1-FCV- 62- 77 05/21/;1 166 1 21 NA 1373 1 CCV- 62- 77 17/?3/c1 - 166.6 21 30 1374 1 rCV- 62- 77 V/23/c1 166.1 2.4 NA 1375 1 rCV- 62- 77 11 /19/5.1 166.1 5.8 NA 1376 1-FCV- 62- 77 01/28/d2 166 1 26 NA 1377 ! 1 cCV- 62- 77 32/23/c2 16%.6 3.3 NA 1373 __ 1 ATTACHMENT. 1 __ _(Page _.4. of _4)_ _ _ _ _ . . . _ _ . . _ _ _ __ __ _ _ __,__. _ _ .__ _ i l I
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k o . I 1 4 ENCLOSURE 3 ! i SEQUOYAH NUCLEAR PLANT TVA-SQN-TS-35 4 I CHANGE NO. 3 2 i ) WINTER FLOOD LEVEL FOR FLOOD PROTECTION AND SURVEILLANCE REQUIRDENTS i. i 1 l l l 4 l f I
s t Enclosure 3 (Cont.) PROPOSED TECHNICAL SPECIFICATIONS TVA-SQN-TS-35 CHANGE NO. 3 1 A I d i. s t. . ( a # i s . k l v .,
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PLANT SYSTEM _5 , 3/4.7.6 FLOOD PROTECTION LIMITING CONDIfl0N FOR OPERATION 3.7.6 The flood protection plan shall be ready for implementation to maintain the plant in a safe condition. APPLICABILITY: When one or more of the following conditions exist:
- a. heavy rainfall conditions in the east Tennessee watershed,
- b. an early warning or alert that a critical combination of flood and/or high headwater levels may or have developed,
- c. an early warning or alert involving Fontana Dam, or
- d. recognizable seismic activity in the east Tennessee region.
ACTION:
- a. With a Stage I flood warning issued initiate and complete within 10 hours the Stage I flood protection procedure which shall include being in at least HOT STANDBY within 6 hours, with a SHUTDOWN MARGIN of at least 5% delta k/k and T less than or equal to 350 F within the following 4 hours. If wit 850 10 hours following the issuance of a Stage I flood warning communications between the TVA Division of Water Resources and the Sequoyah Nuclear Plant cannot be verified, initiate and complete the Stage II flood protection procedure within the following 17 hours. With a Stage II flood warning issued initiate the Stage II flood protection plan in time to ensure completion before the predicted flooding of the site and no later than 17 hours prior to the predicted arrival time of the initial critical flood level (703 ft ms1 winter and summer). '
- b. With a seismic event occurring after a critical combination of flood and/or headwater alerts are issued verify and maintain communications between TVA Power Control Center and the Sequoyah Nuclear Plant ,
. within 6 hours or initiate and complete the Stage I flood protection plan within the following 10 hours. If communications have not been established upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours.
- c. With a Fontana Dam Alert issued verify and maintain communications between Fontana Dam and the Sequoyah Nuclear Plant with 1 hour or initiate and complete the Stage I flood protection plan within 10 hours. If communications have not.been established upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours.
l SEQUOYAH - UNIT 1 3/4 7-15 _i
PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION , LIMITING CONDITION FOR OPERATION (Continued)
- d. With either the Norris, Cherokee, Douglas, Fort Loudon, Fontana, Hiwassee, Apalachia, Blue Ridge or Tellico dam failed seismically, after a critical combination of flood and/or headwater alerts is
- issued initiate and complete the Stage I flood protection plan
- within 10 hours. Upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours. Both the Stage I and the Stage II flood protection plans will be terminated if it is determined that the potential for flooding the site does not exist.
SURVEILLANCE REQUIREMENTS 4.7.6.1 The water level in the forebay shall be determined at least once per 8 hours when the water. level __i.s less than or equ_al to_703. feet _Mean_ Sea Level USGS datum _and at least once per 2 hours when the water level is above these limits. 4.7.6.2 Communications between Sequoyah Nuclear Plant:
- a. and TVA Division of Water Resources shall be maintained every 4 hours during heavy rainfall condition in the east Tennessee watershed.
- b. and TVA Power Control Center shall be maintained every 3 hours following a-recognizable seismic event that has occurred when a critical combination of flood and/or headwater alert is issued.
Communications shall be maintained until it has been determined i that the potential for flooding the site does not exist.
- c. and Fontana Dam shall be maintained every hour.when an alert involving Fontana Dam has been issued by TVA Division of Water Resources.
SEQUOYAH - UNIT 1 , 3/4 7-16
PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION PLAN LIMITING CONDITION FOR OPERATION 3.7.6 The flood protection plan shall be ready for implementation to maintain the plant in a safe condition. APPLICABILITY: When one or more of the.following conditions exist:
- a. heavy rainfall conditions in the east Tennessee watershed,
- b. an early warning or alert that a critical combination of flood and/or high headwater levels may or have developed,
- c. an early warning or alert involving Fontana Dam, or
- d. recognizable seismic activity in the east Tennessee region.
ACTION:
- a. With a Stage I flood warning issued initiate and complete within 10 hours the Stage I flood protection plan which shall include being in at least HOT STANDBY within 6 hours, with a SHUTDOWN MARGIN of at least 5% delta k/k and T/ avg less than or equal to 350*F within the following 4 hours. If within 10 hours following the issuance of a Stage.I flood warning communications between the TVA Division of Water Resources and the Sequoyah Nuclear Plant cannet be verified, initiate and complete the Stage II flood protection procedure within the following 17 hours. With a Stage II flood warning issued initiate the Stage II flood protection plan in time to ensure completion before the predicted flooding of the site and no later than 17 hours pr.ior to the predicted arrival time of the initial critical flood-level (703 f t ms1 winter and summer).
- b. With a seismic event occurring after a critical combination of flood and/or headwater alerts are issued verify and maintain communications between TVA Power Control Center and the Sequoyah Nuclear Plant I - within 6 hours or initiate and complete the Stage I flood protection plan within the following 10 hours. If communications have not been established upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours.
SEQUOYAH - UNIT 2 3/4 7-15
l PLANT SYSTEMS - ACTION: (Continued) I c. With a Fontana Dam Alert issued verify and maintain communications between Fontana Dam and the Sequoyah Nuclear Plant with I hour or initiate and complete the Stage I flood protection plan within 10 hours. If communications have not been established upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection. plan within the following 17 hours,
- d. With either the Norris, Cherokee, Douglas, Fort Loudon, Fontana, Hiwassee, Apalachia, Blue Ridge or Tellico dam failed seismically after a critical combination of flood and/or headwater alerts is issued initiate and complete the Stage I flood protection plan within 10 hours. Upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours. Both the Stage I and the Stage II flood protection plans will be terminted if it is determined that the potential for flooding the site does not exist. -
SURVEILLANCE REQUIREMENTS 4.7.6.1 The water level in the forebay shall be determined at least once per l 8 hours when the water _ level is less than or_. equal _to_703 feet Mean Sea Level USGS datum and at least once per 2 hours when the water level is above these limits. 4.7.6.2 Communications between Sequoyah Nuclear Plant: .
- a. and TVA Division of Water Resources shall be maintained every '
4 hours during heavy rainfall condition in the east Tennessee
- watershed.
- b. and TVA Power Control Center shall be maintained every 3 hours following a recognizable seismic event that has occurred when a '
critical combination of flood and/or headwater alert is issued. Communications shall be maintained until it has been determined that the potential for flooding the site does not exist,
- c. and Fontana Dam shall be maintained every hour when an alert involving Fontana Dam has been issued by TVA Division of Water Resources.
SEQUOYAH - UNIT 2 3/4 7-16 l
TVA-SQN-TS35 Change No. 3 JUSTIFICATION FOR CHANGES IN FLOOD PROTECTION TECHNICAL SPECIFICATIONS This request changes the winter flood level in Technical Specification 3.7.6 from 697 feet mean sea level to 703 feet mean sea level and the surveillance requirement, 4.7.6.2.a, to require communication be maintained every four hours in lieu of three hours. The reason these changes are being requested is to allow TVA the option to postpone initiating stage II shutdown after a stage II flood warning as long as possible and still ensure a minimum of 17 hours to complete the stage II flood protection plan before flooding of the site. The flood level for the summer months is 703 feet mean sea level, and therefore has already been accepted for the technical specifications. Also, maintaining communication between Sequoyah Nuclear Plant and the Division of Water Resources every four hours provides a minimum of twenty-seven hours for safe plant shutdown in all flood producing conditions.
-- _4- 2 - a ._--h, 'l e 6 ENCLOSURE 4 SEQUOYAH NUCLEAR PLANT TVA-SQN-TS-35 CHANGE NO. 4
.l MODIFICATION OF SURVEILLANCE REQUIREMENTS FOR TESTING OF CONTAINENT PROTECTIVE FUSES l l l l ' a s T l _ e
ELECTRICAL POWER SYSTEMS SURVEILLANCE RE0UIREMENTS (Continued) (c) For each circuit breaker found inoperable during these functional tests, an additional representaive sample of at least 1 of the . curcuit breakers of'the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
- 2. By selecting and functionally testing a representative sample of at least 10% nf each type of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating basis. For the lower voltage circuit breakers the nominal trip setpoint and short-circuit response time are-listed in Table 3.8-1.
Testing of these circuit breakers will consist of injecting a current in excess of the breakers nominal setpoint and measuring'the response
. time. The measured response time will be compared to the manufacturer's data to ensure that_it is less than or equal to a value specified by the manufacturer. Circuit breakers found inoperable'during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found. inoperable during these-functional tests, an additional representative- sample.of at least .
10% of all the circuit breakers of the inoperable. type shall also be functionally tested until no more failures are found or all. circuit breakers of that type have been functionally tested. I I
\;
l
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l -b. At least once per 60 months'by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures : prepared ir. conjunction with its manufacturer'sL recommendations. ! c. A fuse ins.pection.and maintenance' program will be maintained to. ensure that: l 1. the; proper size'and type of fuse is installed,
- 2. the fuse shows no' sign of deterioration,'and L- 3. the' fuse connections are tight and. clean.
l I ^ 7 SEQUOYAH - UNIT l' 3/4 8-16. l l l.
ELECTRICAL POWER SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) (c) For each circuit breaker found inoperable during these functional tests, an additional representaive sample of at least 1 of the curcuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
- 2. By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers. Circuit' breakers selected for functional testing shall be selected on a rotating basis. For the lower voltage circuit breakers the nominal trip setpoint and short-circuit response time are listed in Table 3.8-1.
Testing of these circuit breakers will consist of injecting a current in excess of the breakers nominal setpoint and measuring the response time. The measured response time will be compared to the manufacturer's data to ensure that it is less than or equal to a value specified by the manufacturer. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sainple'.of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more-failures are found or all~ circuit breakers of that type have been functionally tested. 1
- b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in. conjunction with its manufacturer's recommendations.
- c. A fuse inspection and maintenance program will be maintained to ensure that:
- 1. the proper size and type of fuse is installed,
- 2. the fuse shows no sign of deterioration, and
- 3. the fuse connections are tight and clean.
SEQUOYAH - UNIT 1 3/4 8-17 et
TVA-SQN-TS35 Change No. 4
>USTIFICATION FOR MODIFICATION OF SURVEILLANCE REQUIREMENTS FOR TESTING OF CONTAINMENT PROTECTIVE FUSES
.Sequoyah Nuclear Plant Technical Specification 4.8.3.1.a.3 requires that at
- least 10 percent of the fuses used for containment penetration conductor overcurrent protection be tested every 18 months. TVA believes that this requirement is costly and unnecessary because it does not enhance the safety or reliability of the plant. In fact, the testing may be detrimental to safety because excessive removal and insertion of fuses in i the fuse holders may damage the contact points and the removal and replacement of in-line current limiters may compromise cable integrity.
Both of these conditions can lead to an unwanted deenergization of equipment. Gould Shawmut, a major fuse manufacturer, has provided TVA information to the effect that "(U)nder no condition can a current limiting fuse ever become less protective over life." It indicates that high temperatures, current surges, or unusual cycling conditions can reduce the life of a fuse "but this simply means it becomes more protective. Fuse resistance will only begin to change increase when unusual loading, cycling or indeed a short circuit occurs." In addition, it provided the following information on cable protecting fuses. Under no circumstances can cable protecting fuses be removed from cables without physically destroying them because of the crimped joint. Of all fuses, cable protector fuses are extra heavy so are (sic) the least susceptible to deterioration of all types. They are designed for cable isolation only under extreme overcurrent condition (sic). Utilities typically install them permanently with~no intention of ever disturbing them. We have not seen a cable protecting fuse fail for any reason in nearly thirty years of sales.. TVA has 24 of these large cable protecting fuses installed for each unit on the reactor coolant pump motor feeds. In addition to the crimped joints, these fuses are wrapped with heat shrink insulation material which must be cut off in order to remove the fuse. The fuse and wrap are destroyed for each test. Could Shawmut uses resistance measurements for production quality control. However, these " resistances are not published because construction changes can occur at any time as designs change or materials are improved." The fuse resistance is a good measure of a fuse's rating, but it is not necessary to periodically remeasure the resistance because the fuse degradation mode does not decrease resistance.
4 In sussuary, TVA believes that a fuse inspection and maintenance program to verify that the proper size and type of fuse is installed, the fuse shows no signs of deterioration, and the fuse connections are tight and clean would ensure that the necessary level of containment penetration conductor overcurrent protection is maintained. The present technical specification requires testing that is unnecessary and costly because fuse resistance does not decrease under degrading conditions, it only will increase (and become more protective). In the case of large cable protecting fuses, the testing is destructive. TVA met with NRC's Instrumentation and Controls System Branch on January 26, 1980 to discuss the electrical system technical specifications
,for Sequoyah unit 1. At that time, TVA stated its objections to the technical specification requirement to measure ~ fuse resistances. However, the staff indicated that TVA did not have enough evidence to support the request. We trust that the experience of Gould Shawmut is sufficient evidence to support our claim that fuse resistance is unnecessary to ensure that containment penetrations are protected from conductor overcurrents.
1 j * . ! l t i i i i
)
ENCLOSURE 5 i l , t i SEQUOYAH NUCLEAR PLANT-4 i 1 TVA-SQN-TS-35 i CHANGE NO. 5 i I CHANGES IN ROD POSITION INDICATION e ) 1 SYSTEM REQUIRIDENTS o a d I. ( i I 1 h [ l r t i. ? i l I l e i _- . . . _ _ . . , . . _ . , . . _ . , , , , ,._.u.._ . . _ , . . . . , _ _ __ . _ . . _ . . _ . _ _ . - -. .u _
Enclosure 5 (Cont.) PROPOSED TECHNICAL SPECIFICATIONS TVA-SQN-TS-35 CHANGE NO. 5 l
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUT 00WN LIMITING CONDITION FOR OPERATION 3.1.3.3 The group demand position indicator shall be OPERABLE and capable of determting within + 2 steps, the demand position for each shutdown or control rod not fully inserted. APPLICABILITY: MODES 3*#, 4*# and 5*#. ACTION: With less than the above required group demand position indicator (s) OPERABLE, immediately ope the reactor trip system breakers. SUR'/EILLANCE REOUIREMENTS
- 4.1.3.3 Each of the above required group demand position indicator (s) shall_ '
be determined to be OPERABLE by movement:of the associated control rod at
.least 10 steps in any one direction at least once per 31 days.
4 "With tne reactor trip system breakers in the closed position.
#$ee Special Test Exception 3.10.-5.
SEQUOYAH'- UNIT 1 3/.4~1 -
/
4 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUTDOWN LIMIT:NG CONDITION FOR OPERATION 3.1.3.3 The groep demand position indicator shall be OPERABLE and capable of deterating within + 2 steps, the demand position for each shutdown.or control rod not fully inserted. APPLICABILITY: MODES 3*#, 4*# and 5*#. ACTION: With less than the above required group demand position indicator (s) OPERABLE, immediately open the reactor trip system breakers. SURVEILLANCE REQUIREMENTS-4.1.3.3 Each of the above required group demand position indicator (s) shall be determined to be OPERABLE by movement'of the associated control rod at < least 10 steps in any one direction at least once per 31 days.
"With the reactor trip system _ breakers in the closed position.
#$ee Special Test Exception 3.10.5.
4 SEQUOYAH - UNIT 2 3/.4. 1-18
TVA-SQN-TS35 Change No. 5 JUSTIFICATION FOR CHANGES IN ROD POSITION INDICATION SYSTEM REQUIREMENTS
- Westinghouse Electric Corporation recommended several changes to the standard technical specifications for rod position indication in their letter from J. C. Miller to H. J. Green dated July 23, 1981 (LOO 810803 075). A copy of this letter is attached. There were three basic changes that were recommended to improve the accuracy and usefulness a of the analog rod position indicator ( ARPIs).
- 1. A reference position would be defined to recognize that the positions of control and shutdown banks need to be known accurately only in specified regions of the core. Westinghouse recommended that in the regions where less accuracy was acceptable, the ARPIs need only agree within 12 steps of the reference rather than 212 steps of the demand counter. The reference positions would take into account any nonlinearities associated with the ARPIs. Detection of a misaligned rod is primarily limited to control banks C and D.
- 2. Immediate verification of position after rod movement would be shifted from the ARPIs to the group step counters with subsequent verification by the ARPIs after temperature equilibration. This change was made to account for the temperature sensitivity of the ARPIs.
3 Rod position indication requirements would be shifted from the ARPIs to the group demand counters when the reactor is not critical. Westinghouse indicated that this approach had been informally reviewed by M. Duenfeld of NRC's Core Performance Branch and found to be acceptable. Westinghouse believes that these changes would decrease the number of uninformative licensing event reports filed and alleviate some of the problems associated with the ARPI system. We have reviewed the recommended changes and recommend that. item 3 be implemented. Sequoyah has not experienced extreme non-linearity in the ARPIs; therefore, item 1 was rejected because it provided no benefit. Sequoyah has not experienced. extreme temperature sensitivity in the ARPIs except at . . relatively cold temperatures; therefore, item 2 was rejected because it provided no benefits. . Item 3 was accepted because it would provide benefits during operation with temperatures less .than no-load temperatures. In the past Sequoyah has experienced significant disagreement between the
. demand counters and ARPIs during hot and cold shutdown conditions.
LOO 810803o E wa smu own Westinghouse Water Reactor Electric Corporation Divisions t ,, 37m
! ; nmwenrewymo n2y) j PRn) /
. , July 23, 1981 TVA-81-28 Mr. H. J. Green, Director Thi l! . ' ,
l of Nuclear Power ; o ' Tennessee Valley Authority
. ~-
I D
*hg7 1750 Chesnut Street Tower II Chattanooga, Tennessee 37401 d 5 /.c i f ,
8A - RMS.A. 3 7
Dear Mr. Green:
Tennessee Valley Authority Sequoyah Unit 1 REVISED ANALOG R0D POSITION INDICATION SYSTEM TECHNICAL SPECIFIC Atta'ched for your information is the Standardized Technical Specifications revised by Westinghouse to allow continued operation by plants with the These specs were 75 Vised'To7 Analog Red Position Indication System (ARPI). TUOM3p NRC and plant T.dncerns .ibbut'th'e' accuracy'a:ah Tech Spec required minimums. Specifically, work was performed allowing an increase in the inaccuracy of the APRI and an increase in the indicated misalignment allowed. However, this work was very plant and cycle specific and is heavily dependent on the size of the DN8R margin present in the cycle design. Recognizing that this method tould not allow a generic solution to the problem of large ARPI inaccuracies, work was performed to determine what changes, if any, could be made to the plant. Tech Specs to reflect the real safety requirements of the ARPI. Afterliscussio'n witFthe'NRC1h'e~att'
~
~ ~~ ^ ~ ached ~~ ~'~ ~3 "s79 Vested r vis~i.6iislseldevelopid..)
Page 1 of the attachment provides a new defined term to be. included in the Tcch Spec Definitions 'section. The term defines what will be considered the ARPI's Reference Position. The revised specs are based on the following:
- 1. lShutdowTB'aKlis3n~d~CEntF51~A~end;8: positions need y,to be ,known3Mr]iR) win a.-very -limited . range,-near.the top-and. bottom of..the core /
~ ~
~
- 2. (C6htrol Banks"C'an'd D pisi t ons.neidit[b'e 'iEcurately'known'at-.the bottom.6f the ch're and from somewhat below the Full Power
~
'~' ~ ~ inseption
~ ~ "
0.imits-(s.150. steps)tol.the_tiplof.the. core
- U Page 2 July 23, 1981
- 3. CRecognizing.that';the ARPI is'very'; temperature sensitive; immediate~-
t verification of. position after rod movement isishifted from the ARPI3 Ithe group step' counters w thf ub,se_quent-verification s by.the,ARPI.z afte'r
' temperature. equilibration.
- 4. I Detiicti oh?'o f "a 7 mi s al i gned rod '.i s? pFimhi i l[ fi ditid "to :Cohtf611 Banks;C,,g.,
~
.and D and' through,the,.use.of .the. Reference Position.'
Pages 2, 3 and 4 of the attachment present the revised current NRC spec for a misalinned rod. Westinghouse recommends the use of this spec to assure an adequate response to a misaligned rod, including the evaluation of the transients listed on page 4. Pages 5 and 6 list the revised accuracy and Page 7 operability requirements for the ARPI and the group' demand counters. erabi l i ty. .whenTnot 'Er i ffdil! biitTwl th is a revision of the NRC spec on RPI.To~p!bben~Vevisid'toTplaceTril.lan~ce.on X el the trip breakers closed. The spec li'as igio.UpTstTpi.~co~unfeEs'." Rages 8 and 9 are revised irsertion limit specs noting the use of the group step counters with ARPI verification. Pages 10 and 11 provide revised Bases for the attached Tech Specs revisions. Finally, page 12 provides a typical figure defining the Reference Position for a Control C or 0 Bank and is referenced in the attached Bases. The attached have been informally ~ reviewed by-M.-Duenfeld of"the' Cote
.P5Ffd'rmihEe' Bra ~nch of 'the NRC and'hhveibein found1to be7 acceptable ~ The Tech
~
~
Speisi's'h6'ulii be inciUEfed ii1 the plant sp'ecific_ Tech Specs. Itis' believed that the attachedTill' deb'riase thei~hiimbes3f LERs: fi. led and remove _someif (ths~;iroblemsas' soc.iai,edwi.ththi.sisystemDTheserecommendedchangesshould be made to your plant Tech Specs as soon as reasonably possible. If you have any questions please contact the undersigned.' Very truly yours,
. , i( .,~
J.iC. Miller ~,' Manager Operating Plant Service Southern Region Attachment cc: R. U. Mathieson i
i .' I REFERENCE POSITION Analog Rod Position Indication System REFERENCE POSITION is defined as:
- a. For all Shutdown Banks, Control Banks A and B, and the Part length Banks; the group demand counter indicated position between 0 and 30 steps withdrawn inclusive and between 200 and 228 steps withdrawn inclusive.
- b. For Control Banks C and 0; the group demand counter indicated
~
position between 0 and 30 steps withdrawn inclusive and between 150 and 228 steps withdrawn inclusive. For the withdrawal range of 31 to 149 steps inclusive the REFERENCE POSITION shall be the individual rod calibration curve noting indicated analog rod position vs indicated group demand counter position. W 0 4 1 -
o .
* #. l *
,e REACTIVITY CONTRCL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMTIrlG CONDITION FOR OPEPATI0li 3.1.3.1 All full length (shutdown and control) rods, and all part length rods which are inserted in the core, shall be OPERABLE and positioned within i 12 steps (indicated position) of the REFERENCE POSITION corresponding to the group demand counter position within or.e hour after rod motion. APPLICABILITY: MODES ~1* and 2*. ACTION:
- a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours,
- b. With more than one full or part length' rod inoperable or misaligned from the REFERENCE POSITION-by more than i12 steps (indicated position), be in HOT STANDBY within 6 hours.
- c. With one full or part length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its REFERENCE POSITION by more than i 12 steps (indicated position). POWER OPEPATION may continue provided that within one hour ei.ther:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The remainder of the rods. in the bank with the inoperable rod are aligned to within i12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figures (3.1-1) and(3.1-2); the THERMAL POWER level shall be restricted ~
pursuant to Specification (3.1.3.6) during subsequent operation, or
- 3. The rod is declared inoperable and the SHUTDOWN i1ARGIN require-ment of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
*See Special Test Exceptions 3.10.2 and 3.10.3.
2, -
.. v..-
a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed. results of these accidents remain valid for the duration of operation under these conditions. b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours. c) A power distribution map is obtained from the movable incore detectors and FQ (Z) and F!g are verified to be within their limits within 72 hours. d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours the high neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. THERMAL POWER shall be maintained less than or equal to 75% of RATED THERMAL POWER until compliance with ACTIONS 3.1.3.1.c.3.a and 3.1.3.1.c.3.c above are demon-strated. - SURVEILLANCE REQUIREMENTS _ 4.1. 3.1.1 The position of each full and part length rod shall be determined to be within + 12 steps (indicated position) of the REFERENCE POSITION corresponding to the group demand position at least once per 12 hours (allowing for one hour thermal soak after rod motion) except during time intervais when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours. 4.1. 3.1. 2 Each full length rod not fully inserted and each part' length rod which is inserted in the core shall be determined to be OPERABLE by movement , of at least 10 steps in any one direction at least once per 31 days. 3 -
e i
.i TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN fHE EVENT OF AN INOPERABLE FULL OR PART LENGTH R00 Rod Cluster Control Assembly. Insertion Characteristics Rod Cluster Control Assembly Misalignment less Of Reactor Coolant From Small RuptJred Pipes Or From Cra -
Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power
~
Major Reactor Coolant System Pipe Ruptures (Loss Of' Coolant Accident) Major Secondary System Pipe Rupture Rupture of a control Rod Drive Mechanism Housing'(Rod Cluster Control Assembly Ejection) ., 4 e f i - 8 . m O 3/4 1-16 007ffB761 W-STS m:
'/ .
. .- h
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The shutdown, control and part length individual rod position indication system and the demand position indicaticn _ system shall be OPERABLE and capable of determining the actual and demanded control rod positions, respectively, as follows: Analog red position indicators, within one hour after rod motion (allowance for thermal soak); All Shutdown Banks - 112 step: of'the group demand counters for withdrawal ranges of 0-30 steps and 200-228 steps. Control Banks A and B - f;12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-228 steps. Control Banks C and D - 112 steps of the group demand counters for withdrawal ranges of 0-30 steps and 150-228~ steps. i 12 steps of the REFERENCE POSITION for withdrawal range of 31-149 steps. All hart Length Banks - i12 steps of the group demand counters for withdrawal ranges of.0-30 steps and 200-228 steps. Group demand counters; i 2 steps APPLICABILITY: MODES 1 and 2. ACTION:
- a. With a maximum of one analog rod position indicator per bank ir. operable either:
l 1. Determine the position of the non-indicating rod (s) indirectly by the movable incore detectors at least once per 8 hours and - within one hour after any motion of the non-indicating rod which
~
exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours,
- b. With a maximum of one group demand position indicator per bank inoperable
~
either: S - -
. 2 .
- 1. Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps (corrected
~
indicated position) of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER -
within 8 hours. I SURVEILLANCE REQUIREMENTS
- 4.1.3.2.1 Each analog rod position indicator shall tua determined to be OPERABLE by veri f.ying that the demand position indication system and the rod position indica-tion system (by use of the REFERENCE POSITION) agree within 12 steps (allowing for one hour thermal soak after rod motion) at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then. compare the demand position indication system and the rod position indica-tion system (by use of the REFERENCE POSITION) at least once per 4 hcurs. ;
Each of the above required rod position indicator (s) shall be deter-4.1.3.2.2 mined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.
'k I
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- (o . - . - .
- 0, REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 The group demand position indicator shall be OPERABLE and capable of deter-mining within + 2 steps the demand position for each shutdown, control or part length rod not fully inserted.
APPLICABILITY: MODES 3*f, 4*# and 5*# 1 ACTION: With less than the above required group demand position indicators (s) OPERABLE, immediately open the reactor trip system breakers. SURVEILLANCE REOUIREMENTS 4.1.3.3 Each of the above required group demand position indicator (s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 .
. steps in any one direction at least once per 31 days.
- With the reactor trip system breakers.in the' closed position.
# See Special Test Exception 3.10.5 7 -
REACTIVITY CONTROL SYSTEMS SHUTDOWN R0D INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn. APPLICABILITY: MODES 1* and 2*# ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification (4.1.3.1.2), within one hour either:
- a. Fully withdraw the rod, or
.b .
Declare the rod to be inoperable and apply Specification (3.1.3.1).. SURVEI'LLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn by use of-the group demand counters, and verified by the analog rod position indicators within one hour after rod motion.
- a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or 0 during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
*See Special Test Exceptions 3.10.2 and'3.10.3 fWith Keff greater than or equal to 1.0 r
y - - 1
-~~ REACTIVITY C0flTROL SYSTEMS C0tiTROL R00 INSERTION LIMITS LIMITIftG C0flDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures (3.1-1) and(3.1-2). APPLICABILITY: MODES 1* and 2*#. ACTI0ft: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification (4.1.3.1.2), either:
' a. Restore the control banks to within the limits within two hours, or
- b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figures, or
- c. Be in at least HOT STATIDBY within 6 hours.
3URVEILLAf!CE REQUIREMEllTS 4.1. 3. 6 The position of each control bank shall be' determined to be within the insertion limits at least once per 12 hours, by use of the group demand counters and verified by the analog rod position indicators within one hour of rod motion, except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours.
*See Special Test Exceptions 3.10.2 and 3.10.3 ,
#With Keff greater than or equal to 1.0 ,
s. e'
- O 4 3/4.1.3 MOVABLE CONTROL ASSE!!BLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTOOWN !%RGIN is maintained, and (3) limit the potential effects of rod misalignrcent on associated accident analyses.
OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure ccepliance with the control rod alignment OPERABLE condition for the analog rod position indicators and insertion limits. is defined as being capable of indicating rod position to within + 12 steps of , the reference position. For the Shutdown Banks, Control Banks A and-8, and the Part Length Banks the REFERENCE POSITION is defined as the grcup demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between J 200 and 228 steps withdrawn inclusive. This pennits the operator .to verify. that the control rods _in these banks are either fully withdrawn or fully;. inserted.
^
Knowledge of these bank positions the~nonnalipera~ ting mode's for these banks. i6'the'se' .two'. areas: satisfies all accident an'alysis' assumptions concerning'.their..
'pliiBFn) For Control Banks C and D the REFERENCE POSITION is defined as the group demand counter indicated position between 0 and 30 steps withdr For the withdrawal range of 31 and between 150 and 228 steps withdrawn inclusive. --
to 149 steps inclusive the REFERENCE POSITION is defined as the individual rod calibration curve noting indicated analog rod position vs indicated group demand counter position (Figure B 3/4.1-1). C3 hip'irison oflhe"ihdicated anilbijNod; position to:the calibration curve i_s.. suff.icient. to . allowdeterhiination that a; control .rodii s indeed misal.igned from'..its.. bank.~ JComparis66~of.Lthe@rsujCdeidand counters tol.the: bank:. insertion limitis' 'wi~th'_ verification of, rodfposition'.With, the analog ~ rod posi..t_i_on indicators (oafter thermal soak after rod motion)?is7 su.f.._ficient3er.ifidation' that the contro.l'ro.ds.._> a.re. .a..b._ove. t.h_e .insertio_
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e Q 0 1 ENCLOSURE 6 SEQUOYAH NUCLEAR PLANT i TVA-SQN-TS-35 CHANGE NO. 6 i CLARIFICATION OF LOCATION OF CAPSULES T AND S IN TABLE 4.4-5 5 I i j i 4 ( i I t' ( r
,. , . - . - .. , - - _ , -n.. , , . - . , - - , .... . n.- ..-- - , , ,- ,
Enclosure 6 (Cont.) PROPOSED TECHNICAL SPECIFICATIONS TVA-SQN-TS-35 CHANGE NO. 6
TABLE 4.4-5 e
' REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM --WITHDRAWAL SCHEDULE g , '
=
CAPSULE -VE5SEL LEAD
.5 NUMBER LOCATION FACTOR WITH0RAWAL TIME (EFPY)
T
- 40 3.73 10 REFUELING 1
U- 140* 3.73 3 1 X 220* 3.73 5 Y- 320 3.73 9 S 4* - 1.09 EOL V 176* 1.09 STBY w W 184* 1.09 STBY l 1
-Z 356* 1.09 STBY U,
o
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-TABLE 4.4-5 .
j REACTOR VESSEL MATEP.IAL SURVEILLANCE PROGRAM - WITHDRAWL SCHEDULE x CAPSULE VESSEL LEAD g -NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY) T 40*' 3.73 18 REFUELING U 140* 3.73 3
.X 220* 3.73 5
'Y. 320 3.73 9 S- 4* 1.09 E0L V -176* 1.09 STBY s
-W 184" 1.09 STBY
@- Z' 354* 1.09 STBY t
1 TVA-SQN-TS35 Change No. 6 JUSTIFICATION FOR Q.ARIFICATION OF LOCATION OF CAPSULES T AND S IN TABLE 4.4-5 r The changes are to resolve a typographical error in Table 4.4-5. The locations of capsules T and S were inadvertently interchanged. We request this change before the first mfbeling outage because capsule T is scheduled for withdrawal at that time. e i 7 I
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