ML20141G844: Difference between revisions
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| number = ML20141G844 | | number = ML20141G844 | ||
| issue date = 04/22/1986 | | issue date = 04/22/1986 | ||
| title = Forwards Initial Comments on Draft Proof & Review Tech Specs Provided by NRC | | title = Forwards Initial Comments on Draft Proof & Review Tech Specs Provided by NRC .Justification for Most Comments in Encl Review Submitted in Util | ||
| author name = Thomas G | | author name = Thomas G | ||
| author affiliation = PUBLIC SERVICE CO. OF NEW HAMPSHIRE | | author affiliation = PUBLIC SERVICE CO. OF NEW HAMPSHIRE | ||
| Line 11: | Line 11: | ||
| contact person = | | contact person = | ||
| document report number = SBN-1012, NUDOCS 8604240006 | | document report number = SBN-1012, NUDOCS 8604240006 | ||
| title reference date = 03-13-1986 | |||
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC | | document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC | ||
| page count = 391 | | page count = 391 | ||
Revision as of 10:48, 12 December 2021
| ML20141G844 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 04/22/1986 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Noonan V Office of Nuclear Reactor Regulation |
| References | |
| SBN-1012, NUDOCS 8604240006 | |
| Download: ML20141G844 (391) | |
Text
{{#Wiki_filter:, . _ _ _ _ . George S. Thomas he Presidera RC, eor hM1uct.on UI Pub 5c Service of New Hampshire New Hampshire Yankee Division April 22, 1986 SBN-1012 T.F. B7.1.2 l United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Vincent S. Noonan, Project Director PWR Project Directorate No. 5
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNH Letter SBN-846 dated July 26, 1985, " Technical Specifications for Seabrook Station", G. S. Thomas to G. W. Knighton (c) USNRC Letter dated March 13, 1986, "Seabrook Technical Specifications', V. S. Noonan to R. J. Harrison Subj ect : Seabrook Proof and Review Technical Specifications
Dear Sir:
Enclosed please find our initial comments on the Proof and Review Technical Specifications provided by the staff in Reference (c). Please be advised that further comments may be provided as a result of our continuing review effort. Additionally, justification for most, if not all of the comments provided in this review has previously been submitted by Reference (b). I Please address any questions to Warren IIall at (603) 474-9574, extension 4046. Very truly yours, 8604240006 860422 ./ \, A PDR ADOCK 05000443 A PDR
]f(
George S. Thomas Enclosure 00 cc: ASLB Service List l P.O. Box 300 Seabrook, NH 03874 Telephone (603) 474-9521
1.0 CEFINITICNS o) . The defined terms of this section appear in capitalized type and are applicable. throughout these Technical Specifications. i
. ~
ACTION 1.1 ACTION shall be that part of a Technical Specif.icatic.n which prescribes - remedial measures required under designated conditions. ACTUATION LOGIC TEST " c 1. 2 An ACTUATION LOGIC TEST shall be the application of various simulated in'put combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices. ANAT.0G CHANNEL OPERATIONAL TEST L3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated
- signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL lock and/or Trip TEST Set shall include adjustments, as necessary, of the alam, inter-range and accuracy. points such that the Setpoints are within the required l
s jAXIALFLUXOIFFERENCE - 1.4 AXIAL FLUX DIFFERENCE shall be the difference in nomalized flux signals
- between the top and bottom halves of a two section excore neutron detector.
=- CHANNEL CALIBRATION p --
~
- 1. 5
~
A CHANNEL CALIBRATION shall be the adjustment, as necessary, o'f the channel such that it responds within the required range and accuracy to known . values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alam, interlock and/or trip functions and may be perfor.ed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
, CHANNEL CHECX 1.6 A CHANNEL CHECX shall be the qualitative assessment of channel behavior during operation by observation. This detemination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
o DRIFT SEABROOK - UNIT 1 1-1 M 32 1
l DEFINITIONS
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~
CCNTAINMENT INTEGRITY
!i 1.7 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required to be closed during accident conditions are either:
- 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, of
- 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-2 of Specification 3.6.3.
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of Specification t
3.6.1.3, 1
- d. The containment leakage rates are within the limits o;* Specification 3.6.1.2, and
- e. The sealing mechanism associated with each penetration (e.g., welds, .
bellows, or 0 rings) is OPERABLE.
- l f')
v CONTROLLED LEAKAGE
- 1. 8 CONTROLLED LEAKAGE shall be that seal water flow n,,.M M the reactor coolant pump seals. fro.m CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe ennservative position.
DOSE EQUIVALENT I-131 See IA5&S 1. DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109, Revision 1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I." E - AVERAGE DISINTEGRATION ENERGY
- 1. E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma.
energies per O ,sa!5 h t4 '*,disinte
- e'eJe- * 'o - fpr ration (MeV/d) *" the
- radionuclides in the sample,un M --
-)e-D. .. ,
H;I 1
O INSE1T I DIGITAL CHANNEL OPERATIONAL TEST . 1.10 A DICITAL C'dANNEL OPEP.ATIONAL TEST s' hall consist of exercising the digital co=puter hard are using database c:anipulation and injecting si=ulated process data to verify GPERASILITY of alarm and/or trip functions. e c 6-O i
. 4 I
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I DEFINITIONS ,
. ENGINEERED SAFETY FEATURES RESPONSE TIME ,~
(] '
- 1. The ENGDfEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time .
interval from shen the monitored parameter exceeds its (SF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e. , the valves travel to their required positions, pump oischarge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where appitcable. FREQUENCY NOTATION
- 1. The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
- IDENTIFIED LEAKAGE
- 1. IDENTIFIED LEAKAGE shall be:
4
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere from sources that a,re both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY
- LEAKAGE, or
- c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.
MASTER RELAY TEST
- 1. A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of ea.h relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.
MEMBER (S) 0F THE PUBLIC
- 1. MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-
~
tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include person who use portions of the site for recre-ational, occupational, or other purposes not associated with the plant. OFFSITE 00SE CALCULATION MANUAL
- 1. The OFFSITE COSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.
SEABROOK - UNIT 1 1-3 5 ua 1 t MARi31986
DEFINITIONS i
. l OPERABLE - OPERABILITY
- 1. A system subsystem, train, component or device shall be OPE'RABLE or l have OPERABILITY when it is capable of performing its specified function (s), '
and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL MODE - MODE
- 1. An OPERATIONAL MODE (i e. , MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.
PHYSICS TESTS 11 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental I nuclear characteristics of the reactor core and related instrumentation:
, (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
l PRESSURE BOUNDARY LEAKAGE 12 , j 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l p leakage) through a nonisolable fault in a Reactor Coolant System component
\
body, pipe wall, or vessel wall. ( PROCESS CONTROL PROGRAM
- 1. The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, l sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State Regulations, burial ground requirements, and other require-sents governing the disposal of radioactive waste.
PURGE - PURGING "N
- 1. 2:1 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
5 6 T.7 Z .L -{i ,D L'. O SEABROOK - UNIT 1 1-4
DEFINITIONS
~
QUAORANT POWER TILT RATIO
- 1. , QUADRANTPhERTILTRATIOshallbetheratioofthemaximumupperexcore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever .
is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RATED THERMAL POWER n 1 25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt. REACTOR TRIP SYSTEM RESPONSE TIME j 1. h The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval f
, when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.
REPORTA2LE EVENT 1.f7 A REPORTABLE EVENT shall be any of those conditi6ns specified in ! Section 50.73 of 10 CFR Part 50. f CCNTAINMENT ENCLOSURE INTEGRITY
- 1. . CONTAINMENT ENCLOSURE INTEGRITY shall exist when:
- a. Each door -in-eacn-access-opening is closed except when the access
? -.
opening is being used for normal transit entry and exit then-at:- - 4ea s t-one -doo r-s hal 1-be-clo s ed , -
- b. The Containment Enclosure Filtration System is OPERABLE, and .
- c. The sealing mechanism associated with each penetration (e.g., welds, 1 bellows, or 0-rings) is OPERABLE.
SHUTOCWN MARGIN
- 1. SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. ,
3 SITE BOUNDARY 3F 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither . owned, nor leased, nor otherwise controlled by the licensee. any area wiM the Si be 1 hadryu,cd &v reventional parftses 6 m e mb ers of W o. o bjy__S t alLk ~ 'hd fo ie hegod W O hoWr7 for c r ees e ain Nar 4 43 W % Wha f ""u M'W c ocu f C&ers sLvtll he Saff lE t' -k locafwo & 1 da c4 cwidio 4 5) . - M [i 3 SEABROOK - UNIT 1 1-5 - ggu4986- -
p ~ DEFINITIONS , V , . SLAVE RELAY TEST _ , , g . - . 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices., SOLIDIFICATION , , Si 1 72 SOLIDIFICATION shall be the conversich of wet wastes into a form that meets shipping and burial ground requirements. SOURCE CHECK 44 . 1 33 A SOURCE CHECK shall be the qualitative assessement of channel response when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS
.D 1.34 A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains, or other l desigt.ated components obtained by dividing the specified test interval into n equal subintervals, and
- b. The testing of one system, subsystem, train, or other designated i component at the beginning of each subintervs1.
THERwAL POWER 94 1.35 THER."AL POWER shall be the total reactor core heat transfer rate to the
- 9 reactor coolant. .
_ TRIP ACTUATING DEVICE OPERATIONAL TEST 47 - 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the recuired Setpoint within the required accuracy. UNIDENTIFIED LEAKAGE M 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. 7 o i
- DRAFT SEABROOK - UNIT 1 1-6 j3gggg
DEFINITIONS o- UNRESTRICTED AREA See hed 8
- r . - t g
1.38 An-UNRESTRICTED- AREA-shall-be any area'at or-beyond_thelSITE_ BOUNDARY-access -to which-is-not-controlled-by-the-licensee-for-purposes-of- protection - of individuals _from_ exposure to-radiation and radioactive-materials,-or-any-area-within-the-SITE-BOUNDARY used'for res1Teitial~ quarters or-for-industrial,- commercial, institutional,-and/or recreational purposes.. , - VENTILATION EXHAUST TREATMENT SYSTEM ur) 1.19 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine or radioactive saterial in particulate fo'rm in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing todines or particu-lates from the gaseous exhaust stream prior to the release to the environment.
. Such a system is not considered to have any effect on noble gas effluents.
Englneered Safety Features Atmoscheric Cleanup Systems are not considered to b,e VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING
~
UI 1.40 VENTING shall be the controlled process of discharging air or gas from a . confinement to maintain temperature, pressure, humidity, concentration, or
, other operating condition, in such a manner that replacement air or gas is not T provided or required during VENTING. Vent, used in system names, does not imply
.- a VENTING process. .
TRUnr. EAT
- VASTE GAS: HOLDUP SYSTEM (645E445) uL GMetus wcwASTE rCGnrtwsT 1.41 A WASTE-GAS-HOLDUP SYSTEM shall be any system designed and installed to *
~
reduce. radioactive gaseous ef fluents by collecting Reactor Coolant System . of fgases from the Reactor Coolant System and providing for delay or holdup
-- for the purpose of reducing the total radioactivity prior to release to the environment. .
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SEABRCOX UNIT 1 1-7 ~ MAR 131988 i i
O INSERT III r - UNRESTRICTED AREA (for liquid effluents only)
~
1.39 The UNRESTRICTED AREA for liquid effluents, for the dose calculations, shall be defined at or beyond the surface edge of the initial mixing
- ene where effluents from the subcerged cultiport diffuser discharge have undergone proept dilutica (assu=ed to be 10:1).
s' t. 6' t I i t
. l l
l l
ll TAttE 1.1 g_ -FREQUENCY NOTATION
~
NOTATION -- FREQUENCY 5 At least once per 12 hours. ., . O At least once per-24 hours. ~ W At least once per 7 days. M
' At least once per 31 days.
Q At least once per 92 days. SA
- - At least once per 184 days.
R At least once per 18 months. S/U
- Prior to each reactor startup.
N.A. Not applicable. P 1 Completed prior to each release. O -
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r h., s %. M ; 3 I 51 .i .L A O SEABROOK - UNIT I *** MAR 13 Iggg
. TABLE 1.2 O ,'
O- OPERATIONAL MODES t REACTIVITY % RATED AVERAGE COOLANT
- MODE CCNDITION, K THERMAL POWER
- TEMPERATURE 77 .
- 1. POWER OPERATION -
> 0.99 > 5% , .
> 350*F .
- 2. STARTUP > 0.99
- , . <_ 5% _> 350*F
- 3. HOT STAN0BY < 0.99 0 -
> 350*F t.
4 .- HOT SHUTDOWN < 0.99 0 350*F
> 200*F> T"'9
- 5. COLD SHUTDOWN < 0.99 0 $ 200*F
- 6. REFUELING" . 1 0.95 0 < 140*F
- Excluding decay heat. .
) ** Fuel in the reactor vessel with the vessel head closure bolts less than fully i tensioned or with the head removed.
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O . Atia1 3-SEABROOK - UNIT 1 1-9 Mggj$lggg l
- l
,y . T e
SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS O f o .x. m m _. MAR 131986
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS I () - 2.1 SAFETY LIMLTS , j REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-1 for four loop operation. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL PCWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour; and comply with the require-r ments of Specification 6.6.1. ;L REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, J, 4, and 5. ACTION: []) MODES 1 and 2: Whenever the Reactor Ccolant System pressure nas exceeded 2735 psig, be in HOT STANOBY with the Reactor Coolant System pressure within its limit within )L hour 4 and comply with the requirements of Specifica. tion 6.6.1. dL N00ES 3, 4 and 5: l l Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.6.1. I o SEABROOK - UNIT 1 2-1 TR-FT MAR 131986 r
i , 680 l . l, [ UNACCIPIAld O. . , i j i oneures 440 rNI < c . .,, . c. i
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c.o 0.20 Cao cao cao em IE (FRACTICN Cf RATED THERMAL power) FIGURE 2.1-1 1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION l 1 .
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A d l SEABROOK - UNIT 1 2-2 MAR 131986
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS - '
REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS , 2.2.1 The Reactor Trip System Instrumentation and Inter 1cck Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: 5'
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column
' but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.
- b. With the Reactor Trip System Instru.eentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or
- 2. Declare the channel inoperable and apply the applicable ACTICN statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 I + R + 5 1 TA Where: Z = The value from Column I of Table 2.2-1 for the af fected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel.
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O SEABROOK - UNIT 1 2-3
O O O
. TABLE 2.2-1 REACTOR TRIP SYSTIM INSTRUMINTATION TRIP SETPOINTS l
5 SENSOR E TOTAL ERROR e FUNCTIONAL UNIT AtLOWANCE (TA) Z (5) TRIP SETPOINT Alt 0WARLE VALUE E 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.
~
- 2. Power Range, Neutron Flux 'I
- a. High Setpoint 7. 5 4.56 0 1109% of RTP** 1111.1% of RTP**
- b. Low Setpoint 8.3 4.56 0 125% of RTP** $27.1% of RTP**
- 3. Power Range, Neutron Flux, 1. 6 0.5 0 15% of RIPa* with $6.3% of RTPa* with i High Positive Rate a time constant; a time constant 32 seconds 32 seconds
- 4. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP** with <6.3% of RTP** with
. 7* High Negative Rate a time constant a time constant l >2 seconds >2 seconds
- 5. Intermediate Range, 17.0 8.41 0 $25% of RTP** 131.1% of RTP**
Neutron Flux F 6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps 11.6 x 105cps
& 7. Overtemperature AT 6.5 3 31
-3773'
- I 0*/
1-06 See Note 1 See Note 2
- J
+0r50 #-
- f. g 09'1 ,
- 8. Overpower AT 4.8 1-42 0.12 See Note 3 See Note 4 h Pressurizer Pressure-tow 3.1 0.71 1.69 11945 psig a 9. 31935 psig 4*8 k -
i
- 10. Pressurizer Pressure-High 3.1 0.71 1.69 12385 psig ~<2395 psig i es % 11. Pressurizer Water Level-High 8.0 2.18 1.82 $92% of instrument 193.8% of instrument span span l 97 b] 12. Reactor Coolant Flow-tow 2.5 1,82 0.6 190% of loop design flow"
>89.4% of loop design ficw**
" Loop design flow = 95,700 gpa
** RIP = RATED THERMAL POWER l
-4j- $ e c Af d C S ~
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l - - - - - - _ _
l O O O t
, TABLE 2.2-1 (Continued) ,
[> REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS 8 o SENSOR TOTAL ERROR [ FU'?tTIONAL UNIT Att0WANCE (TA) Z (S) TRIP SETPOINT ALLOWARLE VALUE h 13. Steam Generator Water 17.0 15.28 1,76 117.0% of narrow 115.9% of narrow Level Low-Low range instrument range instroment
' span span
- 14. Undervoltage - Reactor 15.0 1.39 0 310200 volts 19822 volts !
Coolant Pumps t
- 15. Underfrequency - Reactor 2.9 0 0 355.5 Hz 155.3 Hz Coolant Pumps
- 16. Turbine Trip Sto itso ,
- a. Low Fluid 011 Pressure N.A. N.A. N.A. 3355 psig 3336 psig i
- b. Turbine Stop Valve M.A. N.A. N.A. 31% open
- 11% open 4 Closure ,
'l7. Safety Injection input N.A. N.A. M.A. N.A. N.A.
from ESF i i a . i g r'
.. m . R m o m - t - R - l u l- '
t . J .
.w
.- 3
- - _ _ - - - - - - _ . . . .- - .- . - - - _ . - . - = . - - - . . - . . - . -. - . - . - ._. =. .- .. ..
O O O
, TARLE 2.2-1 (Continued) i E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E
O
- SENSOR TOTAL ERROR
[ FUNCTIONAL UNIT Att0WANCE (TAl Z (5) TRIP SETPOINT ALLOWABLE VALC l h 18. Reactor Trip System l _ Interlocks 'l l a. Intermediate Range N.A. N.A. N.A. ll x 10 80 amp 16 x 10 81 amp Neutron Flux, P-6
- b. Low Power Reactor Trips Block P-7
- 1) P-10 input N.A. N.A. N.A. $10% of RTP** <12.1% of RTP**
7 2) P-13 input N.A. N.A. N.A. $10% RTP** Turofrie $12.3% RTP** Turbine e Impulse Pressure Impulse Pressure Equivalent Equivalent
- c. Power Range Neutron M.A. N.A. N.A. <50% of RTP** <52.1% o f RTP**
I Flux, P-8 ' t=- j ! d. Power Range Neutron N.A. N.A. N.A. <20% of RTPa* <22.1% of RTP** N Fluu. P-9 1.---- 9 P-? : e. Power Range Neutron M.A. N.A. M.A 110% of RTP** 17.9% of RTP**
@ l Flux. P-10 e i
- f. Turbine I g ulse Chamber N.A. N.A. N.A. <10% RTP** Turbine <12.3% RTP** Turbine
, Pressure. P-13 -
Impulse Pressure Impulse Pressure Equivalent Equivalent
~ I.,: '
- 19. Reactor Trip Breakers M.A. N.A. N.A E e N.A. N.A.
'm 20. Automatic Trip and Interlock N.A. N.A. M.A. N.A. N.A.
D '-- Logic d' .
**RTP = RATED THERMAL POWER 'I.
l - _ - _
! O O O I TABLE 2.2-1 (Continued)
$ TABLE NOTATIONS 3;
S o NOTE 1: OVERTEMPERATURE AT ' AT y{ (3 f Ts g) $ AT, M - K g{s T ( 3 , y } - T ' + K3 (g - P ) - f, (@ I 5
-4
- Where: AT = Measured AT by RTD Manifold Instrumentation; "I f{ $
3
= Lead-lag compensator en measured AT; v
Ti. T2 = Time constants utilized in lead-lag compensator for AT, ti f( 8 s , 12 4 w 3 s; I y.
= Lag compensator on measured AT; 3
n O T3
= Time constants utilized in the lag compensator for AT, T3 = '2, s ;
AT = Indicated AT at RATED THERMAL POWER; . , o l 1. 0195~ K, = -13 9; o -o e i z. l K2 = 0,0138/*F; I
* = The function generated by the lead-lag compensator for T**U I*'5 dynamic compensation; i 2.
hr - ' t., is = Time constants utilized in the lead-lag compensator for T,yg,et, V 33 s, Ys { 4 5; h
.:ar kN T = Average temperature. *F; u
d% 1 + 1.5
=
Lag compensator on measured Tavg: l
.o l ;;;ij g @j 1 -
Time constant uttiired in the meu ured T,y tag compensator, 1. - 2. s; l . w . t
-4 1 -) .
e em
, __,.._m ._ , . _ . . - _ . _y.- _ ___..v.--, - .
, _ _ _ . , -_--c ..~z -
O
' ~
O O O
,y TABLE 2.2-1 (Continued) y TABLE NOTATIONS (Continued) a
$ NOTE 1: (Continued) x e T' $ 588.5*F (Nominal T**U at RATED Ti!ERMAL POWER);
e c).aoaf/9 { K3 = 000671/psig; s P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure); 5 = Laplace transform operator, s 3; and f (al) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: E (1) For q - g be ween - M and + b
, f (A D = 0, dere q and o are percent M D M R m t b POWER in the top and bottoe halves of the core respectively, and q t
- 4; s t
al THUMAL POWER in percent of RATED THERMAL POWER; l q exceeds - 35%, the AT Trip Setpoint shall (2) For esch percent that the magnitude of q beautomaticallyreducedby.II2t(%ofitsvalueatRATEDTHERMALPOWER;and l (3) For each percent that the magnitude of q q exceeds + , the AT Trip Setpoint shall MI - beautomaticallyreducedbyk % of its value at RATED THERMAL POWER. (-Q l ' 30 2 h NOTE 2: The channel's maximum Trip 5etpoint shall not exceed its computed Trip 5etpoint by more than 2r5% y of AT span. 5 kri , 1 i . 9
o
~
i O O TABLE 2.2-1 (Continued) { TABLE NOTATIONS (Continued) 5 O 7 NOTE 3: DVERPOWER AT E 1 ) ( I'$ } I I I I q AT (I1 1+T* 5) 2s) (1I + v35) < ATo (K* - K 5 (1 + t,5) (1+t.5)T-K[(T (1 + T.5)I- T" - fZQI)] e-
't Where: AT = As defined in Note 1,
= As defined in Note 1, y 3 Yi, 12 = As defined in Note 1, I
! = As defined in Note 1, l 7 1+1 35 l e ! = 13 As defined in Note 1,
= As defined in Note 1, AT,
= 1.09, K.
1 K3 = 0.02/*F for increasing average temperature and 0 for decreasing average temperature, [' 3 = The function generated by the rate-lag compensator for T dynamic compensation, 2.
- lgj ry =
Time constants utilized in the rate-lag compensator for T,yg, t, a 10 s, e U 1 l
= As defined in Note 1, 1 + ts5 1
43 b ^-
,# T. = As defined in Note 1 W .,
y . W' e
=%
O O O
, TARLE 2.2-1 (Continued) h TABLE NOTATIONS (Continued)
B O 7 HOTE 3: (Continued) .
$ K. =
0.00128/*FforT>T"andK.=0 fort <T", H .g
=
s T As defined in Note 1, T" = Indicated T, at RATED THERMAL POWER (Calibration temperature for AT instrumentation, < 588.5'F). S = As defined in Note 1, and f (a!) = 0 for all af, m h NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than LC% of AT. span. , 3'i l yg; I.o4 is % senso~ e*
- o " 4%3 **d 0- 47
- s 'tk e 5enso" cc'er -Co- l'ressa r eb['ressu e- e e
. - - - , rt<se utus ea s " as inaasu,.cr sens - w o- o< ur a-k hs., .mus/ Je aA J k J,1.c tp .6.,
. I L.< $ co, 2 2.- 6 -the ex4ue omte , ~ 4,e d r Jo e /,
I l; . . ., I g; ' l .
=!
- x 3*
g . y y-..t c3 ( , t 1 1
.f -
/ *g d 6
1
O- . e - i 4 EASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l
~
. ~ . . . . , . ~ . - - l O . L9.:.'t._t.A.i i MAR 131986
1 1 !O. i e . ' ~ - 5 i 4 i 4 t j P60TE i i The BASES contained in succeeding pages summarize ! the reasons for the Specifications in Section 2.0, 2 - but in accordance with 10 CFR 50.36 are not part
- of these Technical Specifications.
l I 8 em G an. me . D l O MAR 131986
a e 2.1 SAFETY LIMITS , BASES 2.1.1 REACTOR CORE I The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented i by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure fro;n nucleate boiling (DNB) and the resultant sharp reduction in heat transfer
, coefficient. DNB is not a directly measurable parameter during operation and I therefore THERMAL POWER and reactor coolant temperature and pressure have been related to ONS through the W-3 (R-Grid) correlation. The W-3 ONS correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonunifor1n heat flux distributions. The local DN8 heat flux ratio (DNBR) is defined as the ratio of'the heat flux that would cause DN8 at a particular core location te the local heat flux and is indicative of the targin to DNB.
The minimum value of the ONBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95% pecbability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curves of Figures 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. N These curves are based on an enthalpy hot channel factor, F of 1.55 and a reference cosine with a peak of 1.55 for axial power shape. Abg,llowanceis a included for an increase in F" at reduced power based on the expression: F" = 1.55 [1+ 0.2 (1-P)] Where P is the fraction of RATED THERMAL POWER. These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f g(AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance ef fect on the Over-i temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits. i l 1 r3 4, smf SEABRC0K - UNIT 1 B 2-1 sai s L g-
. a.
M1TlH6 i
SAFETY LIMITS BASES - i, 2.1.2 REACTOR COOLANT SYSTEM PRESSURE . The restriction of this Safety Limit protects the integrity of the Reictor Coolant System (RCS) from overpressurization and thereby prevents the release of radionucildes contained in the reactor coolant from reaching the containment atmosphere. c The reactor vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements. The entire RCS is hydrotested at 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation. O l . l m.
\1 ,
*')
. L -
SEABROOK - UNIT 1 8 2-2 l, y
.s . A l
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l 1 MAR 13 IS86
2.2 LIMITING SAFETY SYSTEM SETTINGS
~
BASES . 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure tha't the core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated ope
- rational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents. The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, 1 Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with Setpoints less conservative than the Trip Set-point b.ut within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration O point for rack and sensor components in conjunction with a statistical combin-ation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equa-tion 2.2-1, Z + R + 5 1 TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with tile sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as measured" devia-tion, in percent span, for the af fected channel from the specified Trip Set-point. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drif t factor, and provides a threshold value for REPORTABLE EVENTS. s en i Y,J watke Me unce plf t w= 4 M7 de o m 1h c j The methodology to derive the~ ~ Trip Set ~p~ornti s based upon combining all of the uncertainties in the channels:/ Inherent to the determination of the Trip Setpoints are the magnitudes of these channel encertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. SEABROOK - UNIT 1 B 2-3 nJ/n% '
'1 1 0
p* g'm
;; d Vs . f ' '
b
LIMITING SAFETY SYSTEM SETTINGS , BASES I
- I REACTOR TRIP SYSTEM INSTRUMEN(ATION SETPOINTS (Continued) ~
l The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the I design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The l Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is i initiated. This prevents the reactivity insertion that would otherwise result l from excessive Reactor Coolant System cooldown and thus avoids unnecessary I actuation of the Engineered Safety Features Actuation System. Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability. Power Range, Neutron Flux In each of the Power Range Neutron Flux channels t5ere are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low
-power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. Power Rance, Neutron Flux, Hich Rates The Power Range Positive Rate trip provides protection against rapid flux ) increases wh.fch are characteristic of a rupture of a control rod drive housing. 1 Specifically, this trip complements the Power Range Neutron Flux High and Low ! trips to ensure that the criteria are met for rod ejection from mid power. The Power Range Negative Rate trip provides pretection for control rod dbr.p accidents. At high power a single or multiple rod drop accident could cause ! local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevetet this from occurring by tripping the
\ reactor. No credit is taken for operation of the Power Range _ Negative Rete trip ,
for those control rod drop accidents for which DNBRs will be greater than 1.30. SEABROOK - UNIT 1 h R 'iu 8 2-4 ., a a r o W e, 1 eN
l l LIMITING SAFETY SYSTEM SETTINGS
~
BASES - Intermediate and Source Range, Neutron Flux . The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consecuences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 beenmes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER nnless manually blocked when P-10 becomes active. Overtercerature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow with respect to piping transit delays j from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips. The Set-point is automatically varied with: (1) coolant temperature to correct for j temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1. . Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g. , no fuel pellet melting and less than 1% cladding strain) under all possible l overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat genera-tion rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases." Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressJre range in which reactor operatinn 4e p r ;itt W .
,n ,f m
a e O 7.g 4 a i SEABROOK - UNIT 1 8 2-5 %u i A6 - h- - m 1 0886
LIMITING SAFETY SYSTEM SETTINGS , BASES I *
~
Pressurizer Pressure (Continued) The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure. On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7. The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure. Pressurizer Water Level The Pre:surizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, auto-s satically reinstated by P-7. Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps. - On increasing power above P-7-(a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 50% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on lcw reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked. Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The specified Setpoint provides allowances for starting delays of the Emergency Feedwater System. SEABROOK - UNIT 1 B 2-6 - 3a p9<
# Sk g.4 it .$.
i [7 2 MAR-131986
LIMITING SAFETY SYSTEM SETTINGS , BASES I
- Undervoltace and Underfrecuency - Reactor Coolant pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorpor ated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips frca momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 seconds.
On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7. Turbine Trio A Turbine trip initiates a Reactor trip. On decreasing power the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 20% of 8tATED THERMAL POWER); and on increasing power, reinstated automatically by P-9. Safety Infection Inout from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3. Reactor Trip System Interlocks The Reactor Trip System interlocks perform the following functions: P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e. , prevents premature block of Source Range trip)-and-de- -)[-
-energizes-the-high--voltage-to-the-detectors ~ On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.
P-7 On increasing power P-7 automatically enables Reactor trips on low flow in_more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. On decreasing power, tne above listed trips O re eute etic iiv biecxed-SEABRC0K - UNIT 1 B 2-7
.o a . 2 ;
~
WWW_
() , LIMITING SAFETY SYSTEM SETTINGS , BASES Reactor Trip System Interlocks (Continued) P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the above trip. e P-9 On increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip. P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Lew Setpoint Power Range trip; and automatically blocks the Source Range trip ar.d deenergizes the Source Range high voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated. 1 Provides input to P-7. P-13 Provides input to P-7. i 0 d J
/
SEABROOK - UNIT 1 8 2-8 DRLFT hutR la 1886
. - . . - . _ . . _ - . . . ... . _ . - -. - - ~.
\
.O .
g * ., c SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION ng SURVEILLANCE- REQUIREMENTS S m e P w e D 9 e e a _a e
' R y O .
4
; l;e
..x .p n. A.H s
- i 1
MAR 131986 , I
g 3/* LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS V; . 3/4.0 APPLICABILITY 5 LIMITINGCCNDITICSFOROPERATION 3.0.1 Cc=pliance with the Limiting Conditions fer Operation co'ntained in the succeeding specificatiens is required during the OPERATIONAL MODES cr other .
~
conditiens specified therein; except that ucen failure to meet the limiting Conditions for Operation, the associated AC. TION requirements shall be cet. P 3.0.2 Nonce pliance with a specification shall exist when the require ents of the Limiting Condition for Operation and associated ACTICN requirements are net met within the specified time intervals. If the Limiting Conditien fer Operatien is restered prict to expiration of the specified time intervals, cc:pletion cf the ACTION require:ents is not requirec'.
/ .
3.0:3 When a Limiting Cenditica for Operatien is not met, except as previded in the asscciated ACTION requirements, within I hour actien shall te initiated te place the unit in a H00E in which the specification does net apply by placicg it, as applicable, in: -
- a. At least HOT'57ANCEY within the next 6 hours,
- b. At least HOT SHUTDOWN within the follcwing 6 hours, and
- c. At leas! COLD SHUIDOVN.within the subsequent 24 hours.
Where corrective measures are ccepleted that per=it cperatien under the ACTION requirements, the actic.' may be taken in accordance with the specified time limits as measured frc= the time of failure te =eet the Limiting Cenditien for
- Operation. Exceptiens to these requirements are stated in the individual
, specificaticns. '
This specification is net applicar.le in MODE 5 or 6. , 3.0.4 Entry into an CPERATIONAL MODE or other specified conditien shall not be made unless the conditions for the Limiting Conditien for Operation are cet without reliance en provisions contained in the ACTION recuire=ents. This prevision shall net prevent passage through cr to OPERATIONAL MODES as required te cc: ply with ACTICN requirements. Exceptiens to these requirements are - stated in the individual spec (fications.
? .
~
~.-
f- ',"i .T bh n Is
k,N.,, m ., 4 SEAERCCK - UNIT 1 3/4 0-1 2 h.)
A'N i 1
- h Mk
APPLICA3ILITY ( SURVEILLANCE REQUIREMENTS l 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES - cr other conditions specified for individual Limiting Conditi6ns for Operation unless otherwise stated in an individual Surveillance Requirement. ' 4.0.2 Each Surveillance Requirement shall he performed'within the speciffed time interval with: e
- a. A maximum allewable extension not to exceed 25% of the surveillance .
interval, but
- b. The ccchined time interval for any three censecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval. .
4.0.3 Failure to perfer= a Surveillance Requirement within the specified time interval shall censtitute a failure to meet the OPERABILITY requirements fer a Li=iting Conditien for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Require:ents de not have to be perfer:ed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be sade unless the Surveillance Requirement (s) associated with the Limiting
~
Cenditica for Operatien has been performed within the stated surveillance interval cr as ctherwise specified. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME, Ccde Class 1, 2, and 3 ccepenents shall be applicable as fc11cws:
- a. Inservice inspection of ASME Code Class 1, 2, and 3 cc penents and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves ,
shall be perfer ed in acccrdance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Co.:sission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(1); O SEABRCOX - UNIT 1 3/4 0-2 n .v,.%.2 9 / n ' ."A 3. m$p} f MAOil880 ! l
O. ^>>ttcast'tTv ( SURVEILLANCEREOdiREMENTS(Centinued)
- b. Surveillance intervals specified in Section XI of the ASME Eciler and Pressure Yessel Code and applicable Addenda fo'r the inservice inspection and testing activities required by the ASME foiler and Pressure Vessel Code and applicable Addenda shall be applicable as . -
fc11cws in these Technical Specifications:- - ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inseection and testino activities activities Weekly . At least once per 7 days
- Monthly At least cnce per 31 days Quarterly or every 3 senths At least once per 92 days Se:iannually or every 6 months At least once per 184 days Every 9 months At least once pe 275 days Yearly or annually At least once pr - 355 days
- c. The previsiens of Specification 4.0.2 are applicable to the a..'ve required frequencies for perfor=ing inservice inspection and testing activities;
- d. Performance of the above inservice inspection and testing activities 3 - # shall be in additien to other specified Surveillance Requirements; and
_ e. Ncthing in the ASME Beiler and Pressure Vessel Code shall be c:nstrued ' T to supersede the require =ents of any Technical Specification. 9 l o O P ., Y* i "T% f"% f(\ .
~
l ?* ,f [O . . n b .r, ~L ..L A . .1 SEABROOK - UNIT 1 3/4 0-3 MAR 131886
~
3/4.1 REACTIVITY CONTROL SYSTFMS 3/4.1.1 BORATION CONTROL - r - SHUTDCVN MARGIN - T GREATER THAN 200'F LIMITING CONDITION FOR OPERATION j 4 3.1.1.1 The SHUTDOWN MARGIN shall be grea'ter than or equal to 1.3% ak/k i p for four loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4.
. ACTION:
With the SHUTDOWN MARGIN less than 1.3% ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm baron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REOUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal A to 1.3% ak/k:
- a. Within I hour af ter detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or tntrippable control rod (s);
- b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;
- c. When in MODE 2 with K,ff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum inser-tion limit of Specification 3.1.3.6; and
*See Special Test Exceptions Specification 3.10.1. ~ ~ ~ - - - - - - - - - - -
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SEABROOK - UNIT 1 3/4 1-1 . ! MAR 131988
REACTIVITY CONTROL SYSTEMS . SURVEILLANCE RE00IREMENTS (Continued)
~
- e. When in MODE 3 or 4, at least once per 24 hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration,
- 2) Control rod position,W
- 3) Reactor Coolant System average temperature,
- 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% Ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Speci .'ication 4.1.1.1.le. , above. The predicted reactivity values shall be adjusted -(normalizedh to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. 8
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_ _ _ - ~_. -_____ _ _ REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,yg LESS THAN OR EQUAL TO 200*F . i , LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.2% k/k. APPLICABILITY: MODE 5. - ACTION: , 5 With the SHUTDOWN MARGIN less than 1. a /k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REOUIREMENTS 4.1.1.2s The SHUTOOWN MARGIN sha" be determined to be greater than or equal to1.2"Jak/k: W
- a. Within 1 hour after cetection of an inoperable control rod (s) and at O- least once per 12 hours thereaf ter while the rod (s) is inoperable.
If the in9perable control red is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and At least once per 24 hours by consideration of the following factors: b. i
- 1) Reactor Coolant System boron concentration,
- 2) Control rod position, * *
- 3) Reactor Coolant System average temperature,
- 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration.
- The Reactor Coolant System boron concentration shall be > 2000 ppm when coolant loops are in a drained condition.
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REACTIVITY CONTROL SYSTEMS O- MODERATOR TEMPERATURE COEFFICIENT .- 7 - LIMITING CONDITION FOR OPERATION ., 3.1.1.3 The moderator temperature coefficient (MTC) shall be:
- a. Less positive than 0 ak/k/*F for' the all rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition; and c
- b. Less negative than -4.2 x 10 4 ok/k/*F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.
APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only**. Specification 3.1.1.3b. -- MODES 1, 2, and 3 only**. ACTION:
- a. With the MTC more positive than the limit of Specification 3.1.1.3a.
above, operation in MODES 1 and 2 may proceed provided:
- 1. Control rod withdrawal limits are established and maintained suf ficient to' restore the MTC to less positive than 0 ak/k/*F within 24 hours or be in HOT STANDBY within the next 6 hours.
O Tae witserawai iimits shaii de <= aeeft<=# to tae ia ertio" limits of Specification 3.1.3.6;
- 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and .
- 3. A Special Report is prepare ,and submitted to the Commission, pursuant to Specification 6.9,.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withcrawn condition.
- b. With the MTC more negative than the limit of Specification 3.1.1.3b.
above, be in HOT SHUTOCWN within 12 hours.
*With K g7 greater than or equal to 1. **See Special Test Exceptions Specification 3.10.3.
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i REACTIVITY CONTROL SYSTEMS l _ I SURVEILLANCE REQUIREMENTS , y 4.1.1. 3 The MTC shall be determined to be within its limits during each fuel f - cycle as follows: j l a. The MTC shall be measured and ccmpared to the BOL limit of Specifi-cation 3.1.1.3a. , above, prior t.o initial operation above 5% of I RATED THERMAL PCWER,,,after each fuel loading; and
- b. The MTC shall be measured at any THERMAL POWER and compared to f -3.3 x 10
- Ak/k/*F (all rods withdrawn, RATED THERMAL POWER I
! condition) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.3 x 10 4 ok/k/*F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specification l 3.1.1.3b. , at least once per 14 EFPD during the remainder of the ' i fuel cycle. l 1 l r l i l l l l
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- REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY F -
LIMITING CONDITION FOR OPERATION I - I 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (T**9) shall be greater than or equal to 551*F. . APPLICABILITY: MODES 1 and 2* **. ACTION:
] With a Reactor Coolant System operating loop temperature (T,y ) less than 551*F, restore T avg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.
i SURVEILLANCE REQUIREMENTS i 4.1.1.4 The Reactor Coolant System tencerature (T'V9) shall be determined to be greater than or equal to 551*F: , j O a. Within 15 minutes prior to achieving reactor criticality, and
- b. At least once per 30 minutes when the reactor is critical and the I
Reactor Coolant System T,yg is less than 561*F with the T,yg-Tref Deviation Alarm not reset. l
*With K,7f greater than or equal to 1.
**See Special Test Exceptions Specification 3.10.3.
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4 REACTIVITY CONTROL SYSTEMS
. 3/4.1.2 BORATION SYSTEMS ,
FLOW PATH - SHUTUUWN t l LIMITING CONDITION FOR OPERATION . 3.1.2.1 As a minimum, one of the f:llowing boron infection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:
- a. A flow path from the boric acid tanks via either a boric acid l transfer pump or a gravity feed connection and a charging pump to I the Reactor Coolant System if the baric acid storage tank in Specification 3.1.2.5a. is OPERABLE, or
- b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.5b. is OPERABLE.
APPLICABILITY: MODES 4, 5 and 6. ACTION: With nr.ne of the above flow paths OPERABLE or capable of being powered from an j i O OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS l 1 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. i i
- SEABROOK - UNIT 1 3/4 1-7 [/ A l M laless
REACTIVITY CONTROL SYSTEMS FLCW PATHS - OPERATING LIMITING CONDITION FOR OPERATION j - l 3.1.2.2 At least two of the following three boron Injection flow paths shall be OPERABLE: .
- a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS), and
- b. Two flow paths from the refueling water storage tank via charging pumps to the RCS.
APPLICABILITY: MODES 1, 2 and 3" j ACTION: With only one of the above required baron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated i to a SHUTDOWN MARGIN equivalent to at least y akA at 200*F within the a next 6 hours; restore at least two flow paths'to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. l-SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated l OPERABLE: . i a. At least once per 31 days by verifying that each valve (manual,
- power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;
- b. At least once per 18 months during shutdown by verifying that each i
automatic valve in the flow path actuates to its correct position on ] a safety injection test signal; and
- c. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.
*The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry i into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours or prior to the temperature of one or more of th,e RCS cold legs exceeding 375*F, whichever comes first.
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REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specifi"ation 3.1.2.1 shall be OPERABLE and capable of being powered from an OPEF RE emergency power source. , APPLICABILITY: MODES 4b5and6. b ,. ACTION: With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REOUIREMENTS O 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across the pump of greater than or equal to 7495 psid is developed when tested pursuant to Specification 4.0.5. AWU 4.1.2.3.2 All charging pumps, exc,luding the above required OPERABLE pump, shall be demonstrated inoperable
- gat-least once-per-31-days r-except-when-the-reactor-vessel-head is removedr-by-verifying-that-the-motor-circuit-breakers-
-a re -s e cure d-i n-the-op e n -po s i ti o n.
X -A-maximum-of_one charging _ pump _shall_be OPERABLErand-that-pump-shall-be-a-centrifugal-pump,-whenever the-temperature-of-one-or-more-of-the-RCS-cold-
-l e g s -i s-l e s s -th a n -o r-eq ua l-to -305' F .
, ' 'An
, inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with ,
power removed from the valve operator, or by a manual isolation valve ' secured in the closed position. m & R & M & ca) CJiruAuY U 0 - pp &k q kcum o}ih atirb7 MDE 4 /*n 11W 3 "' batwuuazu g m e ~~ 44 Ac5 co d ^ # T f s l O 32S*F
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f REACTIVITY CONTROL SYSTEMS CHARGING PL'MPS - OPERATING r . i LIMITING CONDITION FOR OPERATION 3.1.2.4 At least tw charging pumps shall be OPERABLE. V APPLICABILITY: MODES 1, 2 and 3.* ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be 'n at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least y ak/k at 200*F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours, l.2 SURVEILLANCE REOUIREMENTS O 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2;495 psid is developed when tested pursuant to Specification 4.0.5. ;2.60 t j i
*The provisions of Specifications 3.0.4 and 4.0.4 are not applic'ble for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored ,
to OPERABLE status within 4 hours or prior to the temperature of one or more I of the RCS cold legs exceeding 375'F, whichever comes first. I SEABROOK - UNIT 1 3/4 1-10 , mQ j f-}. g
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REACTIVITY CONTROL SYSTEMS
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BORATED WATER SOURCE - SHUTDOWN - i LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE: ,
- a. A B,qric Acid Storage System with:
- 1) A minimum contained borated water volume of 4,800 gallons,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A minimum solution temperature of 65'F. *
- b. The refueling water storage tank (RWST) with:
- 1) A minimum contained borated water volume of 24,500 gallons,
- 2) A minimum boren concentration of 2000 ppm, and
- 3) A minimum solution temperature of 50*F O aaa'tcaet'trv: acoes s na e.
ACTION: With no borated water sourcq OPERABLE, suspend all operations involving CORE ALTERATIONS or positive rear'civity change . SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1) Verifying the boren concentration of the water,
- 2) Verifying the contained borated water volume, and
- 3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST temperature.
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-REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING .
t. LIMITINGCONDIT5ONFOROPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:
- a. A Boric Acid Storage System with:
?
- 1) A minimum contained borated water volume of 20,200 gallons,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A minimum solution temperature of 65'F.
- b. The refueling water storage tank (RWST) with:
t/'l'/,000
- 1) A minimr.u contained borated water volume of 479;000-ga11ons,
- 2) A minimum baron concentration of 2000 ppm,
- 3) A minimum solution temperature of 50*F, and
- 4) A maximum solution temperature of 86*F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restor'e the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equisalent to at least $ ak/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within Inext 30 hours.
- b. With the RWST inopera e, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
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SEABROOK - UNIT 1 3/4 1-12 - MAR 131986
J . I i 1 ; REACTIVITY CONTROL SYSTEMS , SURVEILLANCE REQt)IREMENTS - 4.1.2.6 Each borated water source shall be demonstrated OPERABLE: i i j a. At least once per 7 days by: i 1) Verifying the boron concentration in the water, g 2) Verifying the contained borated water volume of the water source, and , 3) Verifying the Boric Acid Storage System solution temperature j when it is the source of borated water. I b. At least once per 24 hours by verifying the RWST temperature. 4 O l 4 i 4 l
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REACTIVITY CONTROL SYSTEMS UNB0 RATED WATER SOURCES - SHUTDOWN . 0 LIMITING CONDITION FOR OPERATION 3.1.2.7 The Boron Thermal Regeneration System shall be rendered incapable of performing its dilution function by: a.
% CS-t!3oteBTRS M & ^'*-o o
-Removing. power- from-the-Chiller _ Compressor-(CS-E-18) -and, b.
f km CS- V o S' r B TR U1 Positioning-the-t}ree-way-vafve-(HCV-387)-to-bypass all flow
.the-The rma l- Rege ne ra ti v e- Demi ne ra l i z e rsr--
APPLICABILITY: HODES 4, 5, and 6 CS-V30L n CS - U 305~
- a. With either an-OPERABLE-Chiller-Compressor-or-f-low-path-alignment-to-
.t he-The rma l- Re ge ne rativ e__D enin e r a14 z e r vi mme d i a te ly- remo v e- powe r-f rom the-Chi 44er-Compressor-and/oe align the flow path to bypass the demineralizers.
\Qn-the-event-that-testing-is required-during or after m4Tntenancey ' ~
operatJon of the Chiller Compressor or the three-way byp, ass valve is permittTeas-lor.g as the two components are not out-of'5pecification 3.1.2.7 simultaneo'us1 b During such ,t_esting'the unaffected component shall be verified to be in compliance with this specification through-out the test. Jen-the' test is completed'all-conditions for Specifi-ca, tion-3.t 2'7 shall_be_re-established-h SURVEILLANCE REQUIREMENTS 4.1.2.7 The above required condi'tions shall be verified at least once per 31 days. - 1 A ..rj7 1.r i SEABROOK - UNIT 1 3/4 1-14
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REACTIVITY CONTROL SYSTEMS UNB0 RATED WATER SOURCES - SHUTDOWN _ LIMITING CONDITION FOR OPERATION 3.1.2.7 The Boron Thermal Regeneration System shall be rendered incapable of performing its dilution function by: C f U +h a M
- a. -Re% cs - v 3o moving-power % BTRs m~
from-the-Chiller - _ Compressor-(CS-L-18),-and-. (J<.m CS- v 3 cS' r STRS &% knat c&p nibl dN"
- b. Positioning-the-three-way-valve-(HCV-387)-to-bypass all ' flow around-
.the--Thermal-Regenerative-Demineralizers --
APPLICABILITY: MODES 4, S, and 6 ACTION: CS-V302 n CS-U365^^ ms?cl'**Y
- a. With either an-OPERABLF-Chiller-{ompressor-or-flow-path-alignment-to-
-th e-Th e rma l-- Re g e ne rativ e._D e ci n e ra l i z e r sri mm ed i a te ly- remo v e- p owe r- f ro m the-Chiller-Compressor-and/oe align the flow path to bypass the demineralizers.
N' ~
)Qn-the-event-that-testing-is opera 'on of the Chiller Compressor required ~ during or the three way bypassor af ter ma~Tntenance.
valve'is permitte as-long as the two components are not out-of' Specification ~
. 3.1.2.7 simultaneoD 1y s Dur__ing i such testing the unaffected component shall be verified to be in compliance _with this specification through-out the test. When-the~tEt is complet'e'd'all-conditions for Specifi-cation 3.-l.277 shall be_re-established-h SURVEILLANCE REOUIREMENTS 4.1.2.7 The above required conditions shall be verified at least once per 31 days. -
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REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES
- 1 GROUP HEIGHT LIMITING CONDITION FOR OPERATION _
3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indicated po,sition) of their group step counter demand position. 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 one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position). POWER OPERATION may continue provided that within 1 hour:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figure 3.1-1. 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 MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
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;
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
O _ 3/4 1-15
^ ,,, *m SEABROOK - UNIT 1 3 3 4.. t g,
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REACTIVITY CONTROL SYSTEMS LIMITING CONDITION-FOR OPERATION
- 1 ACTION (Continued) c) A power distribution map is obtained from the movable
.incore detectors andq F (Z) and F" are verified to be within their limits within 72 hours; and e
d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux , Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.P m h NMd 16l4cc,epdU d $+v 41y
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- c. With more than one rod trippable but inoperable due to causes other al a 1 m than addressed by ACTION a. above, POWER OPERATION may continue provided that:
- 1. Within 1 hour, the remainder of the rods in the bank (s) with the inoperable rods are aligned to within 1 12 steps of the inoperable ro,ds while maintaining the rod sequence and insertion s
limits of Figure 3.1-1. The THERMAL POWER level shall be s restricted pursuant to Specification 3.1.3.6 during subsequent operation, and
- 2. The inoperable rods are restored to OPERABLE status within 72 hours.
~ ~
- d. With more than one rod misaligned from its group step counter demand height by more than 1 12 steps (indicated position), be in HOT STANDSY within 6 hours.
SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals 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 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. O l SEABROOK - UNIT 1 3/4 1-16
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1 g j MAR 171986---
TABLE 3.1-1 _ ACCIDENT ANALYSES REQUIRING REEVALUATION , IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment - Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in c Large Pipes Which Actuates the Emergency Core Cooling System ;j{ Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident) Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) O , ( ( O _. SEABRC0K - UNIT 1 3/4 1-17
*[3 13 ]Ls & .a a MAR 131986
- _- i
REACTIVITY CONTROL SYSTEMS
~
POSITION INDICATION SYSTEMS - OPERATING . LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within i 12 steps. APPLICABILITY: MODES 1 and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoperable ei ther:
- 1. Determine the position of the non cati rod (s) indirec ,
by the movable incore detectors at leasboh,ce-per 8 hour era FA 2- b immediately after any motion of the nonindicating rod which Me ~ exceeds 24 steps in one direction since the last determinatf ort-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 demand position indicator per bank inoperable either:
- 1. Verify that all digital red 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 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.
SURVEILLANCE REQUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours except during time intervals when the rod position deviation monitor is i inoperable, then compare the Demand Position Indication System and the } Digital Rod Position Indication System at least once per 4 hours. l O , _ _ _ SEABROOK - UNIT 1 3/4 1-18 p -) 5 ya
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REACTIVITY CONTROL SYSTEMS
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POSITION INDICATION SYSTEM - SHUTDOW'l . 1, LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod , position within i 12 steps for each shttdown or control rod not fully ) inserted. ,l APPLICABILITY: MODES 3* **, 4* **, and 5" **. ACTION: , With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers. - SURVEILLANCE REOUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (:-) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 s'.eps when exercised over the full-range of rod travel at least once per 18 mor,Uts. .~ _ s T
- *With the Reactor Trip System breakers in the closed position.
**See Special Test Exceptions Specification 3.10.5.
l l O _ SEABROOK - UNIT 1 3/4 1-19 t 9; f\ i 2; i 1.- A l-c. p<-
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M 131986- ' ~{
REACTIVITY CONTROL SYSTEMS
- R00 DROP TIME r .
LIMITING CONDITION FOR OPERATION 2Q 3.1.3.4 The individual full-length (shutdown and control) ro drop time from
- the fully withdrawn position shall be less than or equal to-3 -seconds from t beginning of decay of stationary gripper coil voltage to dashpot entry with
, a. T,yg greater than or equal to 551*F and k kLs. ActulLI ef1 Of 6
f)
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2. ACTION: With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2. SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:
- a. For all rods following each removal of the reactor vessel head,
- b. For specifically affected individual rods follow 1,1g any maintenance '
on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and l
- c. At least once per 18 months. i l
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REACTIVITY CONTROL SYSTEMS , f SHUTDOWN ROD INSERTION LIMIT , LIMITING CONDITION FOR OPERATION 3.1.3.5 A1.1 shutdown rods shall be fully withdrawn. APPLICABILITY: MODES 1* and 2* **. e ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour 'ther:
- a. Fully withdraw the rod, or l
- b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
O SURVEILLANCE REOUIREMENTS
~
4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:
- a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
l *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
**With K,77 greater than or equal to 1.
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SEABROOK - UNIT 1 3/4 1-21 ne td a" I pI\ M _. A , ^ A .:f ir,:7l5 X NMll) 66__
i REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS r . ; LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures 3.1-1. APPLICABILITY: MODES la and 2* **. e ACTION: With the control banks inserted beyond the above insertion limits, except for
~
surveillance testing pursuant to Specification 4.1.3.1.2:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the above figure, or
- c. Be in at least HOT STANDBY within 6 hours.
O suavetuance atoutasse"Ts 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 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 Specifications 3.10.2 and 3.10.3.
**With K g7 greater than or equal to 1.
O SEABROOK - UNIT 1 3/4 1-22 r 9s n rm mwnisas ;
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, e 3 1' 228 220 / (0 30 ' 22e> '
1Ao.s 4 4.2 d a) 200 / f 2 .$ANK B N 180 t' - - 160 E m 140 / (l'U' 146) y' / BANK C
$ 120 5
p 100 > B f O o
- 80 / -
5 BANK D
$ 60 a ~
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0.;2 . 0.4 0.l6 0.8 10 FRACTION OF RATED THERMAL POWER FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP OPERATION s ueR0 u - UNIT 1 au 1-u p.agpj aon 4 77:p ,. hR~0-1986
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE ," LIMITING CONDITION FOR OPERATION . i 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target bands (flux difference units) about the target flux difference:
- a. t 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU;
- b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTU; and
- c. +3%, -12% for each subsequent cycle.
The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed 1 hour during the previous 24 hours. ( $ ec 6' O.rurn 0 l a ouvu. </. .t.1. 2.& The indicated AFD may deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours. 4 (,L_a_ Lumuham % ~2. /. 2.() APPLICABILITY: MODE 1, above 15% of RATED THERMAL DOWER.* ACTION: O a. With the indicated AFD outside of the sbove required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:
- 1. Restore the indicated AFD to within the target band limits, or
- 2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
- b. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:
- 1. THERMAL PO4ER to less than 50% of RATED THERMAL POWER within 30 minutes, and
- 2. The Power Range Neutron Flux * ** - High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
*See Special Test Exceptions Specification 3.10.2. I en Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.
SEABROCK - UNIT 1 3/4 2-1 7 ,a, a, 7 q i % '.' !
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POWER DISTRIBUTION LIMITS , LIMITING CONDITTUN FOR OPERATION
- ACTION (Continued) -
- c. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER ' the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above required target band.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
- b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed O to exi t eeries the intervai Precee<#2 eaca iossi#s.
4.2.1.7 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:
- a. One minute penalty deviation for each I minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and
- b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERFAL POWER levels between 15% and 50% of RATED THERMAL POWER.
4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least n me per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable. 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provi-sions of Specification 4.0.4 are not applicable. O n bvs a am 3/4 2-2
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i l i UNACCEPTABLE (-11,90) (11,90) UNACCEPTABLE OPERATION - OPERATION f 9 . ACCEPTABLE OPE R ATION 50 e l l (-31,50) (31,501 40 20 0
- 50 -40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE f.a.8) %
FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUf4CTION OF O a^tco '"raxa' eowta
"', o SEABRC0K - UNIT 1 3/4 2-3 "%
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l POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F i LIMITING CONDITION FOR OPERATION 3.2.2 Fq (Z) shall be limited by the following relationships: Fq (Z) 1 2.32 K(Z) for P > 0.5 P F9 (Z) 1 (4.64) K(Z) for P $ 0.5
, and Where* P = THERMAL POWER RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY: MODE 1. ACTION: With F (Z) exceeding its limit: q
- a. Reduce THERMAL POWER at least 3 for each 3 Fq (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range
- Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Set-points have-tieen reduced at least M for each 3 F (Z) q exceeds the limit
- b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a. , above; THERMAL POWER may then be increased provided Fq(Z) is demonstrated through incore mapping to be within its limit.
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f (12.0.0.65) I I I N g 0.4 0.2 0.0 6 8 10 12 0 2 4 CORE HEIGHT FT FIGURE 3.2-2 K(Z) - NORMALIZED F n(Z) A5 A FUNCTION OF CORE HEIGHT sEAsac a - uN!r 1 3u 2.s n. m3 p. ., h, ,., . ,
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POWER DISTRIBUTION LIMITS SURVEILLANCE REGDIREMENTS ! 4.2.2.1 The previsions of Specification 4.0.4 are not applicable. 4.2.2.2 F,y shall be evaluated to determine if qF (Z) is within its limit by:
- a. Using the movable incore detectors to obtain a power distribution
,, map at any THERMAL POWER greater than 5% of RATED THERMAL POWER,
- b. Increasing the measured F,y component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties,
- c. Comparing the F,y computed (F, ) obtained in Specification 4.2.2.2b.,
above to:
- 1) The F,y limits for RATED THERMAL POWER (F, ) for the appropriate measured core planes given in Specification 4.2.2.2e. and f.,
below, and
- 2) The relationship: -
F,y' = F,R [l+0.2(1-P)], Where F ' is the limit fo
- fractional THERMAL POWER operation P
express as a function of F and P is the fraction of RATED THERMAL POWER at which F,y was measured.
- d. Remeasuring F,y according to the following schedule:
- 1) When F,C is greater than the F,RTP limit for the appropriate measured core plane but less than the F relationship, additional C RTP power pistribution maps shall be taken a d F "Y com ared to F "Y and F, either:
a) Within 24 hours af ter exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which F C ,,, j,,g determined, or ## b) At least once per 31 Ef fective Full Power Days (EFPD), whichever occurs first. O - SEABROOK - UNIT 1 3/4 2-6 np'.Mgn u j fj
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.E MAtla l068
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
. t' C P
- 2) When the F is less than or equal to the F limit for the appropriate measured core plane, additional power distribution maps shall be taken and F compared to F and F, at least once per 31 EFPD.
- e. The F,y limits for RATED THERMAL POWER (F,RTP) shall be provided for all core planes containing Bank "0" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specifica-t'on 6.9.1.6; f.
The F,y limits of Specification 4.2.2.2e. , above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:
- 1) Lower core region from 0 to 15%, inclusive,
- 2) Upper core region from 85 to 100%, inclusive,
- 3) Grid plane regions at 17.8 2 2%, 32.1 1 2%, 46.4 2 2%, 60.6 2 2%,
and 74.9 1 2%, inclusive, and O 4) core piae re2 ions ithin 1 2% of core seioht ct 2.88 inches) about the bank demand position of the Bank "0" control rods.
- g. With F, exceeding F, , the effects of F,y on Fq(Z) shall be evaluated to determine if F q (Z) is within its limits.
4.2.2.3 When qF (Z) is measured for other than F,y determinations, an overall measured qF (Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, i SEABROOK - UNIT 1 3/4 2-7 D' T.) A
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Malt 131966
l POWER DISTRIBUTION LIMITS 1 ] J/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION ' t I' 3.2.3 F g shall be iess than 1.49 [1.0 + 0.2 (1-P)]. .l APPLICABILITY: MODE 1. 4 ACTION: With F exceeding its limit.
- a. Within 2 hours reduce the THERMAL POWER to the level where the
) LIMITING CONDITION FOR OPERATION is satisfied. a b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required by ACTION a. , l above; THERMAL POWER may then be increased provided FN 93 i demonstrated through incore mapping to be within its limit. j J SURVEILLANCE REQUIREMENTS . ) 4.2.3.1 The provisions of Specification 4.0.4 are not appitcable. I i O 4.2.3.2 rL shaii be demoastratee to be withia its iimit prior to operation above 75% RATED THERMAL POWER af ter each fuel loading and at least once per 31 EFF0 thereafter by: , a. Using the moveable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER. ~
- b. Using the measured valve of Fh which does not include an allowance 1 for measurement uncertainty.
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i 1 1 3 POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO 1 . !
- LIMITING CONDITION FOR OPERATION
} { 3.2.4 The QUADRANT POWER TILT RATIO shall not' exceed 1.02. i j APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *. ACTION: ! j a. With the QUADRANT POWER TILT RATI0' determined to exceed 1.02: ) i
- 1. Within 2 hours reduce THERMAL POWER at least 3% from RATED <
i THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO i in excess of I and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.
- 2. WiththeQUADRANTPOWERTILTRATIOdeterminedtoexceed1.02jx j within 24 hours and every 7 days thereafter, verify tha{'F,y j and Fg are within their Ilmits by performing Surveillanc j Requirements 4.2.2.2 and 4.2.3.2. THERMAL POWER and setpoint reductions shall,be in accordance with the ACTION statements of i Specifications 3'2.2 and 3.2.3.
O m i i
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4 i, l I i I 1 i a 1
*5ee Special Test Exceptions Specification 3.10.2. j 1
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! MM 151886 , i i !
.---_ _ _ _ _ _ _ _ _ - . - - -a
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATIO L 1 SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and
- b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable.
~
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm indicated QUADRANT POWER TILT RATIO at least ance per 12 hours by either:
- a. Using the four pairs of symmetric thimble locations or
- b. Using the moveable incore detection system to monitor the QUADRANT POWER TILT RATIC subject to the requirements of Specification 3.3.3.2.
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d 3 f POWER DISTRIBUTION LIMITS Q- 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 5 l 1 3.2.5 The following DNS-related parameters shall be maintained within the 1 the following limits: 1 514
- a. Reactor Coolant System T,yg, <_ 581*F
- b. Pressurizer Pressure > 2205 psig*
I *
- c. Reactor Coolant System Flow > 391,000 gpm**
) APPLICABILITY: MODE 1. " 1 ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of ! RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS i I 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours. . J 4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months. ! 4.2.5.3 The RCS total flow rate shall be determined by precision heat l balance measurements at least once per 18 months. - l 4 2S.H ~jk
& kg m% inn,p QUEdevp L4.0 3 r+-<lll 0Si AM +k of kf yN N S & RCS S. M ad wk nl4 7 $n Ac5 m & avd m & Tlififb AC5 fourAdh.$+d4 rad 4J.
I
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%
,' of RATED THERMAL PNER. l ** Includes a 2.1% flow measurement uncertainty. . . m ;tir. p>,>;uiv., 4mQ&n%4 3 01 <ta +4otylYnLQ n L), loJ ataclrt AlIA$"_f JrJ 3 7L ud, g,d 2nd &ts&Q '1/CS ,Mi wa w Q m pg,.uftz ' l i tBAMW$ Genk 4% b YL $h OW y$* L <*t} .e ,** *~+s SEABROOK - UNIT 1 3 4 2-11 ..
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3/4.3 INSTRUMENTATION
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3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION
= . ,
LIMITING CONDITION FOR OPERATION i -
! 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2. -
i APPLICABILITY: As shown in Table 3.3-1. (* i ACTION: 1 As shown in Table 3.3-1. 1 i SURVEILLANCE REOUIREMENTS i j 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and I_ the automatic trip icgic shall be demonstrated OPERABLE by the performance of i the Reactor Trip System Instrumentation Surveillance Requirements specified in l Table 4.3-1. j 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function i shall be demonstrated to be within its limit at least once per 18 months. 4 Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number I of redundant channels in a specific Reactor trip function as shown in the "To'.al No. of Channels" column of Table 3.3-1. i 1 i - h l 1 i 4 i ) 1O SEABROCK - UNIT 1 3/4 3-1 Pl s 1
- ,t. t <3 - - 3 k
- AWt 1
- 11Hs
. . . . . . . . - - _ - - - . , - - - . . . . . , , , . _ _ _ , , . ,--,-,.-,,,,,__,.,,.,-,~,,,=,.-,,.n. . , - - - - _ , _ , , _ . . , -
. _ _ _ _ _ _ _ _ _ _ . _.-_____.._-_-_-_m. _ _ . _ _ _ __ . . . _ , _ _ _ _ .. _ _ _ ._____ _ _ _ . __ _ _ _ _ _ _ _ _
O O O TABLE 3.3-1 m
> REACTOR TRIP SYSTEM INSTRUMENTATION E
o MINIMUM R TOTAL NO. CHANNELS CilANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION c- FUNCTIONAL UNIT , h-i 2 1, 2 1
- 1. Manual Reactor Trip 2 1
~ 2 1 2 3*, 4*, 5* 10.g
- 2. Power Range, Neutron Flux 2 3 1, 2 2#
High Setpoint 4 a. l Low Setpoint 4 2 3 1###, 2 2# b. 4 2 3 1, 2 2#
- 3. Power Range, Neutron Flux High Positive Rate 2 3 1, 2 2#
l m 4. Power Range, Neutron Flu <, 4
} High Negative Rate i
w 2 1###, 2 3 l 4 5. Intermediate Range, Neutron Flux 2 1
- 6. Source Range, Neutron Flux 4 2 1 2 2##
- a. Startup
- b. Shutdown ,2'Y 0 1 3,4,5 5
- c. Shutdown 2 1 2 3*, 4*, 5* 10 ;
4 2 3 1, 2 6#
- 7. Overtemperature AT
{w- '.
- 8. Overpower AT 4 2 3 -1, 2 6# .
.i .' 2 3 1*# 6# (1)
- 9. Pressurizer Pressure--Low 4 j q)
Pressurizer Pressure--High 4 2 3 1, 2 6# (1) 10. 4
- 11. Pressurizer Water Level--High 3 2 2 1** 7#
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O O O TABLE 3.3-1 (Continued) 3; REACTOR TRIP SYSTEM INSTRUMENTATION 8 R MINIMUM
. TOTAL NO. CHANNELS CHANNELS APPLICABLE c- FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERARLE MODES ACTION 5
- 12. Reactor Coolant Flow--Low
- a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 . j7# ,
any oper- each oper-ating loop ating loop
~
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 7#
below P-8) - two oper- each oper-ating loops ating loop
- 13. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 6# (1) w Level--Low-Low in any oper- each oper-1 ating stm. ating stm.
u, gen. gen. de \
- 14. Undervoltage--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus 1** 6#' % {
Pumps i 1
- 15. Underfrequency--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus 1** 6#
Pumps ) I b. , 16. Turbine Trip i
- a. Low Fluid Oil Pressure 3 2 2 1*** 7#
I 2
' b. Turbine Stop Valve Closure 4 4 "L;Y 1*** 11#
IN 17. Safety injection Input p.-~ d from ESF . 2 1 2 1, 2 9 hw X. *.;M
- 18. Reactor Trip System Interlocks
- a. Intermediate Range )
k^ Neutron Flux, P-6 2 1 2 2## 8 l M- -t l {3f] sn .
i i O O O 1 . l TABLE 3.3-1 (Continneg i vs - I j $ REACTOR TRIP SYSTEM INSTRUMENTATION i 8 o MINIMUM i j TOTAL NO. CHANNELS CHANNELS APPLICABLE 1 FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION -1 z ! U 18. b. Low Power Reactor 1 i w , Trips 8 lock. P-7 "I . P-10 Input 4 2 3 1 8 or ' P-13 Input 2 1 2 1 8
- c. Power Range Neutron .
Flux, P-8 4 2 3 1 8 , i i d. Power Range Neutron 4 2 3 1 8 i Flux, P-9 I i D e. Power Range Neutron i i ' i Flux, P-10 4 2' 3 IM 8
- f. Turbine Impulse Chamber l Pressure, P-13 2 1 2 1 8
~'
l l 19. Reactor Trip Breakers 2 1 2 1, 2 9, 12
- 2 1 2 3* , 4* , 5* 10 l}- _-
! g
- ' 20. Automatic Trip and Interlock 2 1 2 1, 2 9
]
, ,. Logic 2 1 2 3*,4*,q* 10 g'
w. e se i, g 6 a i l - l. i ,_ _ _ _ _ _ _ _ . f I l
i i i i O rast3 2.2-1 < cent ee) TABLE NOTATIONS -
, Win i "Only-if the Reactor Trip System breakers -happen-to-be in the closed position and the Control Rod Drive System -is capable of rod withdrawal.
! ** Trip function automatically blocked or bypassed below the P-7 (At Power) j Setpoint. ,
*** Trip function automatically blocked below the P-9 (Reactor Trip /. Turbine Trip i Interlock) Setpoint.
i i #The provisions of Specification 3.0.4 are not applicable. i B elow
## Trip-functior% p-6t-automaticaHy-blockeMbove-the-P-7 (Intermediate Range Neutron j Flux Interlock) Setpoint. l i TSejo uJ l
### Trip-function-automaticaHy-blocked-above the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
i i (1) These channels also provide inputs to ESFAS. Comply with applicable
- MODES and ACTION statements of Specification 3.3.2 for any portion J
of the channel required to be OPERABLE by Specification 3.3.2. 4 O i ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum i Channels CPERABLE requirement, restore the inoperable channel i to OPERABLE status within 48 hours or be in HOT STANDBY within i the next 6 hours. . i i ACTION 2 - With the number of OPERABLE channels one less than the Total j Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: i a. The inoperable channel is placed in the tripped condition r
- within',1 hourf
.l b. The Minimum Channels OPERABLE requirement is met; however, i the inoperable channel may be bypassed for up to hours for surveillance testing of other channels per Specification 4.3.1.1, and t c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron l Flux Trip Setpoint is reduced to less than or equal to 4
' J85)% of RATED THERMAL POWER within 4 hours; or, the
, QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2. i . l l SEABROOK - UNIT 1 3/4 3-5 q, q 4 ' h)\lHG-
_ __ _. _ - _ _ _ _ _ . _ _ _ _ _ - _ _ _ _ __ _ ..__.____.m _ _ _ _ _ _ _ 4 . TABLE 3.3-1 (Continued)
- ACTION STATEMENTS (Continued) ,
l l ACTION 3 - With the number of channels CPERABLE one less than the Minimum
- Channels OPERABLE -quirement and with the THERMAL POWER level: i i
l a. Below the P-6 (Intermediate Range Neutron Flux Interlock) l Setpoint, restore the inoperable channel to OPERABLE status prior to increising THERMAL POWER above the P-6 Setpoint, and g. J ! b. Above the P-6 (Intermediate Range Neutron Flux Interlock) j Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing i THERMAL POWER above 10% of RATED THERMAL POWER. ACTION 4 - With the number of OPERABLE channels one less than the Minimum ! Channels OPERABLE requirement, suspend all operations involving j positive reactivity changes. I 1 ACTION 5 - With the number of OPERABLE channels one less than the Minimum 1 Channels OPERABLE requirement, restore the inoperable channel l j to OPERABLE status within 48 hours or open the Reactor Trip t System breakers, suspend'all operations involving positive j reactivity changes.and-verify-Valves are-closed-and - l
. secured-in position within-the-next-hour--
~
ACTION 6 - With the number of OPERABLE channels one less than the Total , i Number of Channels, STARTUP and/or POWER OPERATION may proceed i l provided the following conditions are satisfied: 4 . r The inoperable channel is placed in the tripped ccndition
~
l a. l i within A houri and , h j b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to hours { for surveillance testing of other channels per ) Specification 4.3.1.1. j ACTION 7 - With the number of OPERABLE channels one less than the Total J Number of Channels, STARTUP and/or POWER OPERATION may proceed }' until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped I condition within 1 hours lo ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3. ; i !o
, SEABROOK - UNIT 1 3/4 3-6 7 y: -
i Mg)e s .]d < rc 11
! h dd 1
- A lbasa --
1 TABLE 3.3-1 (Continued)
. ACTION STATEMENTS (Continued) i . i.
1 ACTION 9 - With the number of CPERABLE channels one lest than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to l 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.
~
- ACTIDN 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour.
1 ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the in perable channels are placed in the tripped condition within hour 5 1 6
' ACTION 12 - With one o" the diverse trip features (undervoltage or shunt trip attactment) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply ACTION 9. The breaker shall not be bypassed while one of the ,
diverse trip features is inoperable except for the time required
- for performing maintenance to restore the breaker to OPERABLE status.
4 l 3 l i I l O - SEABROOK - UNIT 1 3/4 3-7 - -
,- ;, ,p - g l
.b 1S L hithlea6
O O . O TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRtMENTATION RESPONSE TIMES E O
~
7 FtmCTIONAL UNIT RESPONSE TIME
^
E 1. Manual Reactor Trip N.A.
- 2. Power Range, Neutron Flux $ 0.5 second* l
- 3. Power Range, Neutron Flux, High Positive Rate N.A.
- 4. Power Range, Neutron Flux, High Negative Rate $ 0.5 second*
- 5. Intermediate Range, Neutron Flux N.A.
{ 6. Source Range, Neutron Flux N.A.
~
- 7. Overtemperature AT $ 4 seconds * ,
- 8. Overpower AT $ 4 seconds
- I
- 9. Pressurizer Pressure--Low < 2 seconds t
- 10. Pressurizer Pressure--High 1 2 seconds
) 11. Pressurizer Water Level--High N.A.
i Ed L s vwsf g_ ,,v ,o-! M .
"Neutroc detectors are exempt from response time testing. Response time of the neutron flux signal portion g of the channel shall be measured from detector output or input of first electronic component in channel.
.g.{. UK
! ~W I , q IkA l
L
- - - - - - - - - - - - _ . - ,- , - , e ,
O O O TABLE 3.3-2 (Continted) REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E , ?> ' l
^ FUNCTIONAL UNIT RESPONSE TINE 1
E
~
- 12. Reactor Coolant Flow--Low
-4
- a. Single Loop (Above P-8) $ 1 second <g
- b. Two Loops (Above P-7 and below P-8) $ 1 second
- 13. Steam Generator Water Level--Low-Low $ 2 seconds l
t
- 14. Undervoltage - Reactor Coolant Pimaps 5 1.5 seconds
- 15. Underfrequency - Reactor Coolant Pumps $ 0.6 second
- 16. Turbine Trip 4
- a. Low Fluid Oil Pressure
- N.A.
Y b. Turbine Stop Valve Closure N.A. e 3
- 17. Safety Injection Input from ESF N.A.
- 18. Reactor Trip System Interlocks N.A.
! 19. Reactor Trip Breakers N.A. L {- 20. Automatic Trip and Interlock Logic N.A. D f- P 1 ) C :~- 'j Pe F '
,. .e y- , 5_ ., ,.
W
-.__ , _ - _ _ - - . - .- . , - ,-. - - - . ~ . . . - . . . - . - - _ . - .
l O O . O TARLE 4.3-1 . m 9 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS l E 8 TRIP ~ ANALOG ACTUATING MODES FOR CilANNEL DEVICE WHICH E CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE l Z FUNCTIONAL UNIT CllECK CALIBRATION TEST TEST O LOGIC TEST IS REQUIRED H 'l l /3
- 1. Manual Reactor Trip N.A. N.A. N.A. R(14) N.A. 1, 2, 3*, 4*, 5*
- 2. Power Range, Neutron Flux M60) l l a. High Setpoint S D(2, 4), N.A. N.A. 1, 2
![ ( M(3, Q(4, R(4, 4), 6), 5) ( b. Low Setpoint S R(4) K5/4 b) N. A. N.A. 1***, 2 M 3. Power Range, Neutron Flux, N.A. R(4) 'MA(j6) N.A. N.A. 1, 2 [ High Positive Rate , j b 4. Power Range, Neutron Flux, N.A. R(4) 'M,dh6 N.A. N.A. 1, 2 High Negative Rate S. Intermediate Range, S R(4, 5) S/U(1) N.A. N.A. 1***, 2 Neutron Flux i
/f
- 6. Source Range, Neutron Flux 5 R(4, 5) S/U(1),Q(9,17.) N.A. N.A. 2**, 3, 4, 5
. , 7. Overtemperature AT S R(12) N.$h6) N.A. N.A. 1, 2 e
u .q 4 8. Overpower AT S R KO U6) N.A. N.A. 1, 2 w.ci. 9. Pre-:urizer Pressure--Low 5 R %db3)(U) N.A. N.A. 1 i
-Q
- 10. Pressurizer Pressure--High 5 R X dblf)bh N.A. N.A. 1, 2
'M, d b(,)
C 11. Pressurizer Water Level--High 5 R N.A. N.A. I 88 12. Reactor Coolant Flow--Low 5 R V.dO6) N.A. N.A. 1
$~[d -
~.1
f % A' % 2"*+ A , N' TABLE 4.3-1 (Continued) m 9 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILtANCE REjulREMENTS E 8
~*
TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH E CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE Q FUNCTIONAL UNIT CHECK CALIllRATION TEST TEST LOGIC TEST IS REOUIRED w t
- 13. Steam Generator Water Level-- S R 'MG.Q6)b7) N.A. N.A. 1, 2 Low-Low 14 Undervoltage - Reactor Coolant N.A. R N.A. h$h6) N.A. 1 Pumps
- 15. Underfrequency - Reactor N.A. R N.A. k$hh N.A. 1 Coolant Pumps M
a
- 16. Turbine Trip y a. Low Fluid Oil Pressure N.A. R . N.A. S/U(1, 10) N.A. 1 Z b. Turbine Stop Valve N.A. R N.A. S/U(1, 10) N.A. 1 Closure
- 17. Safety injection Input from N.A. N.A. N.A. R N.A. 1, 2 ESF
- 18. Reactor Trip System Interlocks
._ e a. Intermediate Range
' "] Neutron Flux, P-6 N.A. R(4) 'H.R N.A. N.A. , 2**
- b. Low Power Reactor m, l Trips Block, P-7 N.A. R(4) -H(8 )-f' N.A. N.A. 1
,, cj c. Power Range Neutron l . a- Flux, P-8 N.A. R(4) M(8)- k N.A. N.A. I j Power Range Neutron .)D>
4- _ v .> d. Fluu, P-9 N.A. R(4) M(8}k N.A. N.A. I <r I CC y _. . - J c3 ;. t . I [ :-'1
O O , O TABLE 4.3-1 (Continued) REACTOR TRIP SYSTEM INSTRt1 MENTATION St!RVElll ANCE REQtlIREMENTS 8 TRIF l ANAt0G ACTUATING MODES FOR l CHANNEL DEVICE
- 1tICH E CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE l 0 Ft'NCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED J5. Reactor Trip System Interlocks (Continued)
- e. Power Range Neutron flux, P-10 N.A. R(4) -M(8)( N.A. N.A. 1, 2
- f. Turbine Impulse Chamber Pressure, P-13 N.A. R M(8) f\ N.A. N.A. 1
[j ?O. Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A. 1, 2, 3*, 4*, 5* s 2021, Automatic Trip and Interlock N.A. N.A. N.A. w a. M(7) I , 2, 3 * , 4 * , 5
- Logic
$ 72. Reactor Trip Bypass Breaker N.A. N.A. N.A. M(14),R(15) N.A. 1, 2, 3*, 4*, S* %r/ -E. ..; l '
- = . . 4 N rQ l!
m sa CID " LD f2- j
.3 l I
9 .
TABLE 4.3-1 (Continued) TABLE NOTATIONS , g "Only if the Reactor Trip System breakers happen to be closed and the Control
~
Rod Drive System is capable of rod withdrawal.
**Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
***Below P-10 (Low Sctpoint Power Range Ne'utron Flux Interlock) Setpoint.
7 # (1) If not performed in previous fdays. (2) Ccmparison of calorimetric to excore power indication above 15'. of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1. (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATCC THERMAL POWER. Recalib n te if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for /ntry into MODE 2 or 1. l (4) Neutron detectors may be excluded from CHANNil CALIBRATION. 3 (5) Initial plateau curves shall be measured for each detector. Subsequent (V plateau curves shall be obtained, evaluated and coccared to the initial curves. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into
~'
MODE 2 or 1. (6) Incore - Excore Calibration,-above -75%-of-RATED-THERMAL-POWERr The provisions of Spccification 4.0.4 are not applicable for entry into , MODE 2 or 1. . (7) Each train shall be tested at least every 62 days on a STAGGERED
~ TEST BASIS. .
NOT (ASE.J) (8) -With-power-greater _than or-equal-to-the-Inteclock. Setpoint the-required
-ANALOG-CHANNEL-OPERATIONAL TEST-shall-consist _of vecifying-that-the .
Mnterlock is~in the required-state-by-observing-the-permissive-annuns ciator-wi ndow: 5 QuoMty (9)b Surveillance in MODES 3*, 4", and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the perr.issive annunciator window. (10) Setpoint verification is not applicable. (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments. of the Reactor Trip Breakers. O
~
SEABROOK - UNIT 1 3/4 3-13 %
\ M.3) s.j .AlAg ]j h[>bm h . Dll
~ . _
TABLE 4.3-1 (Continued)
. TABLE NOTATIONS (Continued)
(12) Verify the RTD bypass inops flow rate. (13) The TRIP ACTUATIhG DEVIli OPERATIONAL TEST shall indepe_ndently verify the OPERABILITY 'of the u dervoltage and shunt trip circuits for the Manual Reactor Trip Function. 15e test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s). (14) Local manual shunt trip prior to placing breaker in service. (Or for s' plants that do not actuate the shunt trip attachment of the bypass breakers on a manual reactor trip): Remote manual undervoltage trip when breaker placed in service. (15) Automatic undervoltage trip. (/(,) M k Q M k ' N na s r n c c. e t> r e s r s a w s (rz) Cn p M b 7 A "l A 0 " " A T q.5.2.\ gx y @ Apm ,QM 3 3z. I \ f O c SEABROCK - UNIT 1 3/4 3-14 f!l~'f? Qj l 444noa[ . 7l"1 L i1
INSTRUMENTATION 3/4.3.2 ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION , g LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3. ACTION:
- a. With an ESFAS Instrumentation or Interlock Trip Setpoint trir less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table 3.3-4, either: ,
- 1. Adjust the Setpoint censistent with the Trip Setpoint value of O T "'e 3 3-4. "a eeter='#e it"'" 12 a #rs t" t ea tio" e-2 was satisfied for the affected channel, or ~3 3 - l
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
33-EquationJr2-1{ Z + R + 5 h TA Where: Z = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error for the l affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for l the affected channel.
- c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
O SEABROOK - UNIT 1 3/4 3-15 1 I, ,,-
-s g y t.
,s 1
Mr(allag:r ! gj a
4 1 () INSTRUMENTATION
~
SURVEILLANCEREdUIREMENTS !
'4...3 2 1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.
l 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function ! shall be demonstrated to be within the limit at least once per 18 months. l Each test shall include at least one train such that both trains are tested at t. least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number j of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" column of Table 3.3-3. o - 1 l l $ I { i l i SEABROOK - UNIT 1 3/4 3-16 . ., i
~
).I-r - Saa ,n . a>
____ $h,.11.1986 g*m+m--ee--p- -,y-y ,-p y,_,,+,g,e ,w=,-, ,, -, ,-. - y.m, , . - - - - -,-my-,.y.v
O O , O TABLE 3.3-3 1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I E 8 MINIMUM + TOTAL NO. CilANNELS CilANNELS APPLICABLE E FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION q e-. 1. Safety injection (Reactor 'l Trip, feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation,and Emer0 enc Feedwater),9cevvvedt>s <A/J Sccid b0 a. C6n WM -k/-7.</&delimbml) Manual Initiation 2
%1 2 1,2,3,4 17 4* b. Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation '
T Relays *
- c. Containment 3 2 2 1,2,3 14*
Pressure--Hi-1 .
- d. Pressurizer 4 2 3 1, 2, 3# 18*
Pressure--Low P 7 14* e e. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# Pressure--Low any. steam line (7 n
' h. 3* .m, -
6- i g% c3gy , c. CD'
, i g .
l s I
~
(- ( L..) L L,. TABLE 3.3-3 (Continued) i ENGINEERED SAFETY ffATURES ACTUATION SYSTEM TNSTRUMENTATION
$o
- MINIMUM TOTAL NO. CllANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Z
s 2. Containment Spray . g
- a. Manual Initiation 2 1 with 2 1,2,3,4 17 2 coincident switches
- b. Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays R c. Containment Pressure-- 4 2 3 1,2,3 15 Hi-3 Y
5 3. Containment Isolation -
- a. Phase "A" Isolation
- 1) Manual Initiation 2 1 2 1,2,3,4 17
- 2) Automatic Actuation 2 1 2 1,2,3,4 13 4
f j Logic and Actuation Qf Relays See item 1. above for all Safety injection initiating functions and
,l h e]. 3) Safety injection r,equirements.
j pqp I
- b. Phase "B" Isolation t ~,, ,
ff [ a
- 1) Manual Initiation 2 1 with 2 coincident 2 1,2,3,4 17 l switches
^ M .
l ) 00 f J.
.ee.
t O O , O l TABLE 3.3-3 (Continued) h ENGINEERED SAFETY FEATtJRES ACTtJATION SYSTEM INSTRtJMENTATION E 8 MINIMUM
' CHANNELS APPLICABLE TOTAL NO. CllANNELS E FUNCTIONAL llNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION w '
l s 3. Containment Isolation (continued)
- b. Phase "B" Isolation (continued)
- 2) Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays
- 3) Containment 4 2 3 1,2,3 15 Pressure--lii-3 T *
- c. Containment Ventilation .
T Isolation ,
"e
- 1) Manual Initiation 2 1 2 1,2,3,4 16
- 2) Automatic Actuation 2 1 - 2 1,2,3,4 16 Logic and Actuation Relays
("_-- i.g'",o 3) Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements. ,
- 4) Containment On Line 1 2 1,2,3,4 16 l lt.# ' Radioactivity-High /*
i Steam Line Isolation 4. j:# M
- a. Manual Initiation (System) 2 1 2 1,2,3 21 5[h lN - ~s O$ -
%E A 9
.M
O O , O TABLE 3.3-3 (Continued) N
$; ENGINEERED SAFETY FEATURES ACTilATION SYSTEM INSTRUMENTATTON
^
MINIMUM TOTAL NO. CilANNELS CllANNELS APPLICABLE
' filNCTIONAl llNIT Of CllANNfl5 10 TRIP, OPfRAnlf , MODFS , ACTION
'z y 4. Steam Line Isolation (continued) l
- b. Automatic Actuation 2 1 2 1, 2, 3 20 Lo0ic and Actuation Relays
- c. Containment Pressure-- 3 2 2 1,2,3 14*
lli-2
- d. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 14*
Pressure-Low any steam y line y e. Steam Generator . g Pressure - Negative 3/ steam line 2/ steam line 2/ steam line 3**** , 14* Rate--Hi0h any steam line
- 5. Turbine Trip .
- a. Automatic Actuation 2 1 2 1, 2 22 l Logic and Actuation
, 3 , 1 Relays W"J4 - b. Steam Generator 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 18*
._ Water Level--
y.j j j High-High (P-14) . r c s 6. Feedwater Isolation QN,, a. Steam Generator Water 4/sta. gen. 2/stm. gen. 3/stm. gen. 1, 2 18*
.g;M Level--Hi0 h-High (P-14)
- b. 4 2 3 1, 2 18
= Low RCS T,yg Coincident
~ with Reactor Trip 4# 'd c. Safety Injection See Item 1. above for all Safety Injection initiating functions
'9 5 and requirements.
hI' . Co m . d-3e
O O , O TABLE 3.3-3 (Continued) m
$s ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION o"
R MINIMUM .
. TOTAL NO. CilANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION g
Z 7. Emergency Feedwater
- a. Manual Initiation 'l s
(1) Motor driven pump 1 1 1 1,2,3 21 (2) Turbine driven pump 2 1 2 1,2,3 21 ,
- b. Automatic Actuation Logic 2 1 2 1,2,3 20 and Actuation Relays
- c. Stm. Gen. Water Level--
Low-Low . Start Motor-Driven Pump 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 18* t' and Start Turbine - .
- Driven Pump w
c'3 d. Safety Inlection
~ Start Motor-Driven Pump See~ Item 1. above for all Safety injection initiating functions and and Turbine-Driven Pump requirements.
rt o -rv Pump and Turbine- d"-
" 'f M W "[ [N -
" P
- db/ h
}? l _l Driven Pump -2 1 -2 172r3 -17
, 8. Automatic Switchover to
, I , Containment Sump ,
I '.[i.~. ) a. Automatic Actuation 2 1 2 1,2,3,4 13 g Logic and Actuation
; V !
y . m Relays
?> C b. RWST Level--Low-Low 4 2 3 1,2,3,4 18*
3 CoincidentWithh %
! [np om SafetyInjection) < See Item 1. above for all Safety Injection initiating functions
--4 and requirements.
, e-i ~. ; ,
1
O O . O - TABLE 3.3-3 (Continued) m i!; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION a g n 4 MINIMUM . TOTAL NO. CilANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION c 5 i H 9. Loss of Power 18* 1
- a, 4.16 kV Bus E5 and E6- 2/ bus 2/ bus 1/ bus 1, 2, 3, 4 Loss of Voltage
- b. 4.16 kV Bus E5 and E6-Degraded Voltage 2/ bus 2/ bus 1/ bus 1, 2, 3, 4 18*
i Coincident with SI . See Item 1. above*for all Safety Injection initiating functions and requirements.
- 10. Engineered Safety Features
] y Actuation System Interlocks [ a. Pressurizer Pressure, 3 2 2 1,2,3 19
. 4 P-11 ,
- b. Reactor Trip, P-4 2 2 2 1, 2,/ 21
- c. Steam Generator Water 4/stm. gen. 2/stm. gen. y//stm. gen. 1, 2, 3 18*
i Level, P-14 . i t-a 3 O e 1 t ~
- ftV m .
N # sem na v u , .
~
g .NV
,2 .
M .* i e 4 - 1 4 so.
TABLE 3.3-3 (Continued) O ' TABLE NOTATIONS ,
.- t
*The provisions of Specification 3.0.4 are not appitcable.
# Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.
** Trip-function-marbe blocked-in-this-MODE-below-the P-12-(Low-Low _.T- -
-Interlock)-Setpoint. **9
**** Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.
J ACTION STATEMENTS ACTION 13 - With the number of OPERABLE channels one less than the Minimum Ch_annels OPERABLE requirement, be in at least HOT STANDBY (2- wit Dh3 hours and in COLD SHUTDOWN within the following i 30 hours; however, one channel may be bypassed for up to X hours for surveillance testing per Specification 4.3.2.1, provided i the other channel is OPERABLE. ACTION 14 - With the number of OPERA 3LE channels one less than the Total
, Number of Channels, operation may proceed until perfomance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within
{pY. hour 3 ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to (hours for surveillance testing per Specification 4.3.2.1. q ACTION 16 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed. ACTION 17 - With the number of OPERABLE channels one less than the Minimum 9hannels OPERABLE requirement, restore the inoperable channel
.o OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
ACTION 18 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
/
SEABROOK - UNIT 1 3/4 3-23 )! b 6 f i a J ) ~r .1 [PL,! '1 'h
. v, e ~_ .
/~
/
TABLE 3.3-3 (Continued) ~ ACTION STATEMENTS (Continued)
- a. The inop,erable channel is placed in the tripped condition within'l.hourf and
- b. TheMinimumChannelsOPERABLErequirementismet;hyev.tr.J one additional channel may be bypassed for up to T hours for surveillance testing of other channels per Specification 4.3.2.1.
ACTION 19 - With less than the Minimum Number of Channels OPERABLE, within I hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3. ACTION 20 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANOBY within 6 hours and in at least HOT SHUTDOWN within the foi owingours'N 6 hours; however, one channel may be bypassed for up to )2 for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. (~ ACTION 21 - With the number of OPERABLE channels one less than the Total u/ Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STAND 8Y within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. Tolal Almber of ACTION 22 - With the number of OPERABLE channels one less than the -Minimum ChannelsOPERABLErequirement,beinatleastHOTSTANDBYgthin,Y 6 hours; however, one channel may be bypassed for 'up to 7,, hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. y 3/4 3-24 l Q ~ ~ e a p
% g --
pfl m f SEABROOK - UNIT 1 '_
..G.
A JM 131886
O O . O TABLE 3.3-4 , vi 0 ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS S l E SENSOR c- TOTAL ERROR 2 FUNCTIONAL ~ UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE
- 1. Safety injection (Reactor Trip, 'l Feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, l
Containment. and Emergency Ventilation Isolation Feedwater). See P3e 3b 3-0
- a. Nanual Initiation N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
$ c. Containment Pressure--Hi-1 4.2 0.71 1.67 5 4.3 psig $ 5.3 psig u . 7 y- , 4 d. Pressurizer Pressure--Low 13.1 10.71 1.69 's1850psig ,
3 1840 psig
"' 7.- ?
l
- e. Steam :.ine Pressure--Low 13.1 . 10.71 1.63 3 585 psig T 568 psig*
- 2. Containment Spray 1
- a. Nanual Initiation N.A. N.A. N.A. N.A. N.A.
4
,T. ._ ,, b. Autcmatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
i and Actuation Relays WA 30 is'.7 .
- c. Containment Pressure--Hi-3 4:0' O.71 1.67 $ 18.0 psig 5 1971 psig q
1.gt= fU , ss f.! 4 ,
' I,, l7-. .
p'l e .
O O , O TABLE 3.3-4 (Continued) M j ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR O TOTAL ERROR Z (S) TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT ALLOWANCE (TA) c-
- 3. Containment Isolation "I
a .' Phase "A" Isolation N.A.
- 8. A. N.A. N.A. H.A.
- 1) Nanual Initiation
- 2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- 3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values, ,
$ b. Phase "B" Isolation N.A. N.At N.A. N.A. N.A.
- 1) Nanual Initiation N.A. N.A. N.A. N.A.
- N.A.
- 2) Automatic Actuation Logic and Actuation '
P:1ays y,o (g 7
-4:0- 0.71 1.67 5 18.0 psig $ 19 2 psig
- 3) Containment Pressure--
Hi-3
- c. Containment Ventilation Isolation sV" x- 1) Nanual Initiation N.A. N.A. N.A. N.A. N.A. ,
l N.A. N.A. N.A.
' 2) Automatic Actuation N.A. N.A.
Logic and Actuation - g %[ 7 j . -) Relays See Item 1. above for all Safety Injection Trip Setpoints and
.2' 3) Safety injection Allowable Values.
g.
- m. 6. ,:'d i l
N.A. N.A. N.A. <2x N.A.
- 4) Containment On Line Purge Uackground Radioactivity-High g
cri I [aTf . f N
- 4. ,r J
.I .
u -
, ,n LJ G w TABLE 3.3-4 (Continued) m 3; ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION TRIP SETPOINTS '
SENSOR . x TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE c-E H
- 4. Steam Line Isolation 'I
- a. Manual Initiation (System) N.A. N.A. N.A. N. A. N.A.
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- c. Containment Pressure--Ni-2 5. 2 0.71 1.67 $4.3 psig $5.3 psig 2-. 2.
- d. Steam Line Pressure--Low 13.1 10.71 1.63 3585 psig 2568 psig*
w
- e. Steam Generator Pressure - 3.0 0.5 0 $100 psi $ 123 psi **
}
w Negative Rate--Nigh A . N 5. Turbir.e Trip
- a. Automatic Actuation Logic H.A. N.A. N.A. H.A. N.A.
Actuation Relays
- b. Steam Generator Water 4.0 2.18 1.76 <86.0% of <87.2% of narrow t . Level--High-High (P-14) narrow range range instrument 7 "__ instrument span,
) span.
(39 . i f f 6. Feedwater Isolation g'pdle . '
- a. Steam Generator Water 4.0 2.18 1.76 <86.0% of narrow range
<87.2% of narrow range instrument n.g r Level--Hi-Hi-(P-14)
J.c l instrument span. Lj
. span.
- b. 4.6 1.12 1.35 7._564
- F ? 561. 2'F Qt 3- g . Low RCS T,yg Coincident
~^ gi ,
with Reactor Trip , Safety injection N.A. N.A. N.A. N.A.
'g *
- c. N.A. .
WC l J'
7 m . y/ j \ / TABLE 3.3-4 (Continued) m
$; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS o"
SENSOR n - TOTAL ERROR ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE c- FUNCTIONAL UNIT
?!
- 7. Emergency Feedwater
* .g a'. Manual Initiation (1) Motor driven pump N.A. N.A. N.A. N.A. N.A.
(2) Turbine driven pump N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. 'N.A. N.A. N.A. N.A.
and Actuation Relays
- c. Steam Generator Water 17.0 15.28 1.76 > 17.0% of > 15.9% of narrow Level--Lew-Low iiarrow range range instrument t'
- Start Motor-Driven Pump instrument span.
';" and Start Turbine-Driven span.
y Pump ,
- d. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Start Motor-Driven Pump Allowable Values.
and Turbine-Driven Pump
- e. Loss-of-Of fsite Power NrA. - N , A ,- - - N .-A . >-4800V >-4692V' b~ Start Motor-Driven Pump and Turbine-Driven Pump 5'u ,0L.,9. fqq %d pp g Ug y j
- 8. Automatic Switchover to '
f- Containment Sump b ?IQ a. Automatic Actuation Logic .N.A. N.A. N.A. N.A. N.A. and Actuation Relays
- 2. /23,77t,
)) j>.h' 2.'7 o.gl /.8 Z /27,94g _C116;449- gal s. q 118,771: gal s. [
.:::7
- b. RWST Level--Low-Low Coincident With NrA: -N,A. -N:A.
See Item 1. above for all Safety injection Trip Setpoints and b y m. c.o Safety Injection Allowable Values. M; E . CV) -
O G i LJ TABLE 3.3-4 (Continued) m 3; ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION TRIP SETPOINTS 8 o SENSOR 7 10TAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE e fi
- 9. Loss of Power , 7
- a. 4.16 kV Bus E5 and E6 N.A. N.A. N.A. o2975 4 2908 volts!
Loss of Voltage $oltswith with a $ 1735 / 3/f a < 1.20 second time second time d e l ay. -g-k -)<- de l ay. -k M '*-
- b. 4.16 kV B n E5 and E6 N.A. N.A. N.A. k3933 volts 3902 volts Degraded Volta 0e with a $ 10 with a i 10.96 second time second time de l ay. -A- y-y- delay. K y(-%-
m
} Coincident with:
w Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and . J, Allowable Values. w .
- 10. Engineered Safety Features Actuation System Interlocks
- a. Pressurizer Pressure, P-11 N.A. N.A. N.A. $ 1950 psig i 1960 psig .
. b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.
- c. Steam Generator Water Level, See Item 5. above for dTT Steam Generator Water Level Trip P-14 Setpoints and Allowable Values. i WW 5aT U -
3L m
- Ptw c.
C3 f -W.E I 1 4
O v HBLE 3.3-4 (Continued) T .[TABLENOTATIONS - -
- Time constants utilized in t*.e lead-lag controller for Steam Line Pressure-Low are r y > 50 seconds and 12 (5 seconds. CHANNEL CALIBRATION shall ensure that these time constants art adjusted to these values.
**The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is less than or equal to 50 seconds. CHANNEL CALIBRATION shall ensure that,this time constant is adjusted to this value.
ft.% & $wq j Y $0 f 3 Y k
-th AJ % Ak"J k O
S e l 7 , - -- _ SEABROOK - UNIT 1 3/4 3-30 o
'l'a --
.O f-Er- Fi O*a [.)L $s [3, i$f I
. DiARih 1406--- 1
( i TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES .~ ! INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 1. Manual Initiation
- a. Safety Injection (ECCS) N.A.
- b. Containment Spray -
N.A. e c. Phase "A" Isolation N.A.
- d. Phase "B" Isolation N.A.
- e. Containment Ventilation Isolation N.A.
- f. Steam Line Isolation N.A.
- g. Feedwater Isolation N.A.
E,m:evo-
- h. .: Auxiliary-Feedwater N.A.
- i. Essential Service Water 4oGCCid M N.A.
- j. Containment Cooling Fans N.A.
- k. fonfr Emls'o [t on- N.A.
- 1. Reactor Trip N.A.
,- m. Start Diesel Generator N.A. \
- 2. Containment Pressure--Hi-1
- a. Safety Injection (ECCS) < 27(1)/12(5)
- 1) Reactor Trip <2
- 2) Feedwater Isolation < 7(3)
- 3) Phase "A" Isolation 17(2)/27bMb
- 4) Containment Vent Isolation 25I1)/10(2)
- 5) Emergency Feedwater < 60 ;
- 6) Service Water System f2 /47 I2)
- 7) Start Diesel Generator < 10 E) CSA 43y3 %/ gg v
s, ,q,' gni SEABROOK - UNIT 1 3/4 3-31 '
/ - -
a' . a . .uN.a [ .h
TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES - -
!j INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
' 3. Pressurizer Pressure--Low -
- a. Safety Injection (ECCS) 1 25(1)/12(5)
- 1) Reactor Trip '
<2
- 2) Feedwater Isolation 7(3)
- 3) Phase "A" Isolation -<-17b/27(1)N4
- 4) Containment Ventilation Isolation h25(1)/10(2)
- 5) Emergency Feedwater < 60
- 6) Service Water System 47(1)/12(2)
- -77~~-P r i ma ry-Compo ne n t-Coo li n g -Wa t e r- 47 I1)/(({2) 7 2-) Start Diesel Generators -< 10 F) <t3 A W y 9 d -7 M D 4 M d.
- 4. Steam Line Pressure--Low
. a. Safety Injection (ECCS) 1 12(5)/22I4)
- 1) Reactor Trip <2
- 2) Feedwater Isolation 7(3)
- 3) Phase "A" Isolation e-17 b23/27(1) M
- 4) Containment Ventilation Isolation 25I1)/10(2)
- 5) Emergency Feedwater < 60
- 6) Service Water System ~
g2)/47II) 7 Start Diesel g tors 10
- b. Steam L N IsY ation 9(3) .
- 5. Containment Pressure--Hi-3 27
- a. Containment Spray < 2)/5 (1)
- b. Phase "B" Isolation . -65II)/75(2 W
- 6. Containment Pressure--Hi-2 Steam Line Isolation I3) 19
- 7. Steam Generator Pressure - Negative Rate--High Steam Line Isolation 19 I3)
- 8. Steam Generator Water Level--High-High (P-14)
- a. Turbine Trip i 2.5
- b. Feedwater Isolation 1 7( )
O 4 SEABROOK - UNIT 1 3/4 3-32 793FT Ihh. 24aism
O TABLE 3.3-5 (Continued)
- . t-ENGINEERED SAFETY FEATURES RESPONSE TIMES
. INITIATING SIGNAL AND FUNCTION RESPONFE TIME IN SECONDS
- 9. Steam Generator Water Level--Low-Low
- a. Motor-Driven Jmergency Feedwater Pu 1 60
- b. Turbine-Driven Emergency Feedwater Pump 1 60
- 10. Loss-of-Offsite Power
- a. Motor-Driven Emergency Feedwater Pump < 60
- b. Turbine-Driven Emergency Feedwater Pump 5 60
-11. Trip-of-AM-Main-Feedwater-Pumps AM-Emergency-Feedwater-Pumps -N r A:-
f l12'. RWST Level--Low-Low Coincident with Safety Injection - G Automatic Switchover to Containment 250 D Sump <100'f24N [2_13. Loss of Power
- a. 4.16 kV Bus E5 and E6 < Id / 36
~
JLossofVoltage)
- b. 4.16 kV Bus E5 and E6 < 10' // ClO Degraded Voltage 4oincident -
with-Safety-Injection-is hl T1 h a. I SEABROOK - UNIT 1 3/4 3-33 g
- l. ,
i V TABLE 3.3-5 (Continued)
~
F TABLE NOTATIONS (1) Diesel generator starting and sequence loading delays included. (2) Diesel generator st.srting and sequence loading delay M included. Offsite power available. (3) Air-operated valves. e (4) Diesel generator starting and sequence loading delay included. RHR pumps not included. (5) Diesel generator starting and sequence loading delays not included. RHR pumps not included. O ~ l l O _ - SEABROOX - UNIT 1 3/4 3-34 .' [ , [11 , M4filiisas
O O , 0 TABLE 4.3-2 ta , 9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION SURVEILLANCE REQUIREMENTS h O ,
. TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH
- U CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
- FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIREtt t ,
- 1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel 1 Generators, Phase "A" Isolation, Containment Ventilation Isolation, .
and Emergency feedwater). See lluje MV 3'I
- a. Manual Initiation N.A. N.A. N.A. R N.A. O N.A. N.A. 1, 2, 3, 4
, b. Automatic Actuation N.A. N.A. N.A. N.A. 5A M(1) 54 M(1) Q(2) 1, 2, 3, 4 g Logic and Actuation Relays ,
O c. Containment Pressure- S R 'MGC2d N.A. N.A. N.A. N.A. 1, 2, 3 Hi-1
- d. Pressurizer Pressure S R 'MOl.1) N.A. N.A. N.A. N.A. 1, 2, 3 Low
~
- e. Steam Line S R 'Md[1) N.A. H.A. N.A. N.A. 1,2,3 P---- . Pressure-Low
- 2. Containment Spray
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. NsA. 1, 2, 3, 4 g{,
? i b. Automatic Actuation Logic and Actuation N.A. N.A.
N.A. N.A. f/)h(1) ggM(1) Q 8.) 1, 2, 3, 4 Relays g'5 ,J' I C. Containment Pressure- S Hi-3 R KQ(.d N.A. N.A. N.A. N.A. 1, 2, 3
.y
' y z:;-t
'm as -u4 j NT 080
O O , O TABLE 4.3-2 (Continued) m S ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION n
! SURVEILLANCE REQUIREMENTS o
* ~
TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICil U CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE e FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST -[ 15 REQUIRED
- 3. Containment Isolation
- a. Phase "A" Isolation
- 1) Manual Initiation d.A. N.A. N.A. R N.A. N.A. N.A 1, 2, 3, 4
- 2) Automatic Actuation N.A. N.A. .N.A. N.A. SAH(1) $ M H(1) 1,2,3,4 Logic and Actuation Q(2}
Relays 1 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. w h b. Phase "B" Isolation .
- 1) Manual Initiation N.A. N.A. N.A. R N.A. N. A. N.A 1, 2, 3, 4
- 2) Automatic Actuation N.A. N.A. N.A. N.A. $FM(1) 54-H(1) Q dL) 1, 2, 3, 4 Logic Actuation
_ Relays
, , 3) Containment Pressure-Hi-3 S R Y d(L) N.A. N.A. M.A. N.A. 1, 2, 3
-R
- c. Containment Ventilation Isolation N "
g Q 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 2 2) Automatic Actuation Logic and Actuation N.A. N. A. N.A. N.A. ffM(1) % H(1) Q(t)1,2,3,4 L N, M Relays 87 3. -f I g 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. " 4) Containment On Line S R M(2) N.A. N.A. N.A. N. A. . 1, 2, 3, 4
,f"" i Purge Radioactivity-High j .
'j. .
t= .l s
f O i O , O TARLE 4.3-2 (Continued) sn
- 7) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g? S_URVEILLANCE REQUIREMENTS o
^ TRIP ..
' ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH
@ OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE Q CHANNEL CHANNEL CHANNEL CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST, IS REQUIRED
~ FUNCTIONAL UNIT
, 4. Steam Line Isolation ,
N.A. N.A. N.A. n N.A. N.A. 1, 2, 3
- a. Manual Initiation N.A. R (System)
N.A N.A N.A. 6 # N(1) 5 # M(1) QC2) 1, 2, 3 i
- b. Automatic Actuation N.A.
Logic and Actuation Relays
- c. Containment Pressure- S R 'Md(t) .
N.A. N.A. N.A. N.A. 1, 2, 3 Hi-2
'M GD-) N.A. N.A. N.A. N.A. 1, 2, 3 1:' d. Steam Line S R ,
- Pressure-Low
';' 'MD('Lh- N.A. N.A. N.A. N.A. 3
- e. Steam Line Pressure- S R
$ Negative Rate-High ,
- 5. Turbine Trip
- a. Automatic Actuation N.A. N.A. N.A. N.A. 64'M(1) 54M(1) QGt) 1, 2 Logic and Actuation
, Relays
' N.A. N.A. N.A. N.A. 1, 2
- b. Steam Generator Water S R T.d(1-)
[%
~
Level-High-High (P-14)
- 6. Feedwater Isolation
- a. Steam Generator Water S R H N.A. N.A. N.A. ' N.A. 1, 2 i D' ' Level--High-High (P-14)
M. N.A. N.A. N.A. 1, 2
, g,' S A M N.A.
- b. with LowReac RCS TTrip@ Coincident See Item 1. above for all Safety Injection Surveillance Requirements.
- >4.h)w~ c. Safety Injection se -
- 7. Emergency Feedwater
~
ess 'I '
- a. Manual Initiation N.A. N.A. N.A. 1, 2, 3
'$ D (1) Motor driven pump N.A. N.A. N.A. R (2) Turbine driven pump N.A. N.A. N.A. R N.A. N.A. N. A.. 1, 2, 3 b,
- b. Automatic Actuation N.A. N.A N.A. N.A. $4M(1) f4M(1) QC2J-1, 2, 3 and Actuation Relays A
O h o n o LJ TABLE 4.3-2 (Continued) i $ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS y . i E TRIP
- ACTUATING MODES ANALOG CilANNEL DEVICE MASTER SLAVE FOR WilICH E RELAY RELAY SURVEILLANCE CHANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION U CHANNEL TEST LOGIC TEST TEST TEST., IS REQUIRED
- FUNCTIONAL UNIT CllECK CALIBRATION TEST
- 7. Emergency feedwater (Continued)
N.A. N.A N.A 1, 2, 3 R 'Md.(.~t) N.A.
- c. Steam Generator Water S
- Level-Low-Low, Start - Motor-Driven Pump and Turbine-Driven Pump I d. Safety Injection, Start See Item 1. above for all Safety Injection Surveillance Requirements, w Motor-Driven Pump and 5 Turbine-Oriven Pump N.A. M N.A. N.A. N.A 1, 2, 3 I w e. Loss-of-Offsite Power N.A. R Start Motor-Driven , Pump and Turbine-Driven Pump
~
- 8. Automatic Switchover to
)
Containment Sump 3 L-- ni a. Automatic Actuation N.A. N.A. N.A. N.A. M M(1,) Q 1, 2, 3, 4 j Logic and Actuation , D.1 - g , Relays l *:f H N.A. H.A. N.A. N.A 1,2,3,4
- b. RWST Level-Low-Low 5 R
' l'E 3
^4.2./ Coincident With ik[g I' Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
l Cis.77r i l 4:e a_ e
- s.
%a e
m) 'Ot G TABLE 4.3-2 (Continued) NI E; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS E
, TRIP
- c. ANALOG ACTUATING MODES
$ CHANNEL DEVICE MASTER SLAVE FOR WHICH H CHANNEL CHANNEL CNANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
- CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED FUNCTIONAL UNIT
- 9. Loss of Power
- a. 4.16 kV Bus E5 and N.A. R N.A H N.A. N.A. N.A. 1, 2, 3, 4 E6 Loss of Voltage ,
- b. 4.16 kV Bus E5 and N.A. R. N.A. H N.A. N.A. N.A. 1, 2, 3, 4 E6 Degraded Voltage Coincident With g Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements y 10. Engineered Safety O Features Actuation -
System Interlocks
- a. Pressurizer N.A. R h.G.CO N.A. N.A. N.A. N.A. 1, 2, 3 Pressure, P-11 .
- b. Reactor Trip, P-4 N.A. N.A N.A. N.A. Kf4 6),M/ N.A. N.A. 1, 2, 3 Dl ., c. Steam Generator 5 R
'M()CL) N.A. M(1) .M(-1) -Q- 1,2,3 l Water Level, P-14 N+ #4' NM-
[. . 9
,)
EE f TABLE NOTATION
'D .
(1) Each train shall be tested at least every 2 days on a STAGGERED TEST BASIS. ,, [. - , 4 (Z) A DIGITAL CHANNEL OPERATIONAL TEST will be performed on this instrumentation. . Q &,k. e.& M \ssMS a$ ). lend y % dag gn a S T*WM WORS ym a m m " ' * '" cm act n. a.at MuhAa1 h~t w"t a b '. p_ _g; .
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION ___ l i . 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within e the specified limits.
. APPLICABILITY: As shown in Table 3.3-6.
ACTION:
- a. With a radiation monitoring channel Alarm / Trip Setpoint for plant l
operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours or declare the channel inoperable,
- b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.
- c. The provisions of' Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE-RIOUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3. O -
*4 "6$
I 3/4 3-40 lj
) --
f u .: SEABROOK - UNIT 1 mE1rless
O o , 0 TABLE 3.3-6 M s RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 8 9 MINIMUM . CHANNELS CHANNELS APPLICABLE ALARM / TRIP c H0 DES SETPOINT ACTION
- TO TRIP / ALARM OPERABLE FUNCTIONAL UNIT
~ 1. Containment f jogg, ,
- a. Containment-Atmosphere-N 4 -2 / -A11/ 2,19 $ 2-mR/h 2 ~/
-Radioactivi ty-High-dan M at
- b. RCS Leakage Detection N.A. 26
- 1) Particulate Radioactivity H.A. 1 . 1, 2, 3, 4 26
- 2) Gaseous Radioactivity N.A. I 1, 2, 3, 4 N.A.
- 2. Containment Ventilation Isolation
- a. Ya'rhiNde"ki[dlo'[cki by- 1 '12. -AH /,2,3, y 23 w
b A11 4
- 4 02 5 b. N$s N R dN a M M ly *- 1 1. 2.-
Main StesT. Line 1/ Valve 1/ Valve 1, 2, 3, 4 H . 'A ~27 1 3.
-4. Fuel-S torage-Pool -A rea s--
ar-Radi oactivi ty-Nigh-- " Gaseous-Radf oactivity- 2 5-2-mR/hr 24 -
" * * - 25---
b Criticality-Radiatfon-Level- --I 2 - ----1-15-mR/h r
!,_ q 3,. Control Room Isolation
/do C#4 n 'lxluulgadu/*4 5
'h a. Air Intake-Radiation Level
- 1) East Air Intake 1/ intake 1/ intake 1/ intake All 5 mR/hr 24' 24 ga
- 2) West Air Intake 1/ intake All 5-2-%R/he h.
1-1 3-2 mR/hr- 94 - ['4'd -b-Control-Room-Atmosphere- A l g -Radiation =High-3 -Primary-Component-Cooling Water-- a Loop _A _1 1 - Al <_2-x- 28 - L J.' -Haekground-Q._, E,/ b-Loop 8 - 1_- 1 Al1 3-2 a -- ---. *
-Bac kground -
,i pJ l
TABLE 3.3-6 (Continued) ,
~ ~
i TABLE NOTATIONS
* .Must-satisfy-Specification 3c11 _W,2cl-requirements.M hyC A. & &l MovemudCl n w 0
** -Wi th-irrad i ated -f uel-i n-the-f uel-storage -poal-are as-
-* *
- 71 th-f uel-i n -the -f ue l-s to ra g e -pool-a reh s .
ACTiONSTATEMENTS ACTION 23 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment ventilation isolation valves are maintained closed. ACTION 24 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation. NOT~ USED
-W i th-l e s s-th a n-the -M i n i mum -C hanne l s-O P E RAB LE-re qu i reme n tMp e ra -
ACTION 25 - ti o n -may-co n t i n u e -f o r-up-to d ay s- p ro v i ded -a n-a pp ro p r i a t e -
-portable-continuous monitor-with the-same-Alarm-Setpoint_is _
-provided-in-the-fuel storage pool-ar,ea:---Restore-the-inoperable-
/]
V ,-monitors-to-OPERABLE-status within 30-days-or-suspend-a11~
--operations involving fuel-sovement-in-the-fuel-storage-pool
--a re a s ..
ACTION 26 - Must satisfy the ACTION requirement for Specification 3.4.6.1.
~
ACTION 27 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours, and:
- 1) either restore the inoperable Channel (s) to OPERABLE status or within 7 days of the event, 8
- 2) prepareandsubmitaSpecial[ReporttotheCommission pursuant to Specification 6.'$.2 within 14 days following the event outlining the actions taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
AldT~ U SE D ACTION 2B - -Wi th-the -numb e r-o f- O P ERAB LE- C ha n n e l s-l e s s -th a n -- the - M i n i mum -
-Channels-OPERABLE-teqtiffement, cliT1Fct grab Tamples-daily-from---
-the-Primary Component-Cooling-Water-System-and-the-Service l
' ~ Water-System and-analyze-the-radioactivity-until-the-inoperable
-Channel (s)~is restored _to_0PERABLE-status.
O - pph . ,i
. ~..
SEABROOK - UNIT 1 3/4 3-42 L ' d 's. ~ , .
)
Nb
TABLE 4.3-3 ui RADIATION MONITORING INSTRUMENTATION FOR PLANT 9 OPERATIONS SURVEILLANCE REQUIREMENTS
, E S 8
DIGITAL
- CHANNEL MODES FOR MilCil
' SURVEILLANCE CHANNEL CHANNEL OPERATIONAL l
E FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED
- 1. Containment
[ ., a '. Containment -Atmosphere
-Radioactivi ty-High-dha # S R H -Al-1-/i 2,3 '/
- b. RCS LeakaDe Detection
- 1) Particulate Radio- S R H 1, 2, 3, 4 activity
- 2) Gaseous Radioactivity S R M 1,2,3,4 l
l
- 2. Containment Ventilation isolation 1 a. P rt atNad[oIcivIty S R H -A11-42,3,Y R b. S R H -Al l- 4 *-
* -Gaseous-RadioactivitEs
%&ilm.Cunn daan i$t.S .M 1,2,3,4 Main Steam Line
.i w 3. R ! b -Fuel-Storage-Pool-Areas > a Radioactivity-HiDh---
**~
1 -Gaseous-Radioactivity-- -5 -R H j b Criticality-Radiation-Level -S R H i1 Control Room Isolation ) a. Air Intake Radiation Level
- 1) East Air Intake S R H All i I 2) West Air Intake S R H All ,
l i M -b--Control-Room-Atmosphere---
-Radiation-High - -S R -M- - Al gC 'J i
i 6----g r i mgCgepo p ne n t-C ool i ng-Wa te r s R -- - M All-
-b r Loop B- 3 R -M All -
3
$V g %c_-l ditceddt%rd;%T2::"f 2"'Ed'" "'s "With-irradiated fuel _in the fuel-storage-pool-areas- .
- a m
- i se
>:::i' J
INSTRUMENTATION - MOVABLE INCORE DETECTORS T- .- !. LIMITING CONDITION FOR OPERATICN 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:
- a. At least 75% of the detector thimbles,
- b. A minimum of two detector thimbles per core quadrant, and
- c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.
APPLICABILITY: When the Movable Incore Detection System is used for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoring the QUADRANT POWER TILT RATIO, or N
- c. Measurement of F pq(Z) and F ,
ACTION: , With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS
-4 r3 r3-2-Th e -Nova b l e-I nc o re - De te c t i o n -Sy s tem -s h a l l -b e-demo n s tra ted - O P ERAB L E-a t
.__.least-once-per-24-hours-by-normalizing-each detector _ output when-required-for:
-acRecalibration of-_the_Excore Neutron Elux-Detection-System; or----
-b .- Mo n i to ri ng -th e -QU AD R ANT- POWE R -TI LT- RATI 0 ro r N
_c._._ Measurement-of-F pz(Z-)-and-F , l
/
Dd a
~;?'h-t .-
5; w }?d, ,W'*{ W t : SEABROOK - UNIT 1 3/4 3-44 N 50 l000
INSTRUMENTATION SEISMIC INSTRUMENTATION
. LIMITING CONDITION FOR OPERATION 3.3.3.3 The seis.mic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.
APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
p/ w SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECX, CHANNEL CALI-BRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4. - skaj t b < 4.3.3.3.2 Each of the above required seismic monitoring f nstruments actuated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours andfa CHANNEL CALIBRATION',$erformed within 3010 days following the seismic event 9 Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 14 days describing the magnitude, frequency spectrum, and resultant effect upon facility features important to safety, f & h3 u YA A N SZjh}&S 210 -l779
/
O Ph D g t m SEABROOK - UNIT 1 3/4 3-45 -]J , h.1 A gy j31986 1
TABLE 3.3-7
~
SEISMIC MONITORING INSTRUMENTATION ,,
- MNIMUM MEASUREMENT INSTRUMENTS RANGE OPERABLE INSTRUMENTS AND SENSOR LOCATIONS
- 1. Triaxial Time-History Accelerographs
- ' t Ig 1*
- a. 1-SM-XT-6700 Free Field East Cont.
Room Air Intake c 2 Ig 1*
- b. 1-SM-XT-6701 Containment Foundation
- c. 1-SM-XT-6710 Cont. Opr. Floor t Ig 1*
- 2. Triaxial Peak Accelerographs a.1-SM-XR-6702 Reactor Vessel Support 0-20 Hz. I
- b. 1-SM-XR-6703 Reactor Cool. Piping 0-20 Hz. I 0-20 Hz. 1
- c. 1-SM-XR-6704 PCCW Piping
- 3. Triaxial Seismic Switches N.A. 1*
l
- a. 1-SM-XS-6700 Free Field N.A. la
- b. 1-SM-XS-6701 Containment Foundation
-NrArd o2fgdo 0 253 la
- c. 1-SM-XS-6709 Containment Foundation d.1-SM-XS-6710 Cont. Opr. Floor N.A. ,
1*
- 4. Triaxial Response-Spectrum Recorders ~
1-30 Hz. 1*
- a. 1-SM-XR-6705 Containment Foundation Il
- b. 1-SM-XR-6706 SG JIB Support 1-30 Hz. I
- c. 1-SM-XR-6707 Prim. Aux. Bldg. 1-30 Hz. 1 d.1-SM-XR-6708 Service Water Pump House 1-30 Hz. 1
*With reactor control room indication O i e SEABROOK - UNIT 1 3/4 3-46 ,
.] ,y
.1.i I.. s 1-4. ,t a
~ W 1[\hg
) TABLE 4.3-4
~
SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS .- ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. Triaxial Time-History Accelerographs .
- a. 1-SM-XT-6700 Free Field East Cont. ,, M* R SA Room Air Intake
- b. 1-SM-XT-6701 Containment Foundation Ma R N.A.
c.1-SH-XT-6710 Cont. Opr. Floor M* R H.A.
- 2. Triaxial Peak Accelerographs
- a. 1-SM-XR-6702 Reactor Vessel Support N.A. R 'N.A.
- b. 1-SM-XR-6703 Reactor Cool. Piping N.A. R N.A.
- c. 1-SM-XR-6704 PCCW Piping N.A. R N.A.
- 3. Triaxial Seismic Switches
- a. 1-SM-XS-6700 Free Field H R SA (N b. 1-SM-XS-6701 Containment Foundation ** M R N.A.
- c. 1-SM-XS-6709 Containment Foundation ** M R H.A.
- d. 1-SM-XS-6710 Cont. Opr. Floor *' H R N.A.
- 4. Triaxial Response-Spectrum Recorders
- a. 1-SM-XR-6705 Containment Foundation ** M# R N.A.
Il b.1-SM-XR-6706 SG HB Support N.A. R N.A.
- c. 1-SM-XR-6707 Prim. Aux. Bldg. N.A. R N.A.
- d. 1-SM-XR-6708 Service Water Pump House N.A. R N.A.
*Except seismic trigger
**With reactor control room indications.
# CHANNEL CHECK to consist of turning the test / reset switch and verify all lamps illuminate on 1-SM-XR-6705.
N. L ,] SEABROOK - UNIT 1 3/4 3-47 E ". i l} l - c- . l -
.u %s I ( h s sia ':: .
INSTRUMENTATION
. METEOROLOGICAL INSTRUMENTATION i
, LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE. ,
APPLICABILITY: At all times. , ACTION:
- a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
O SuavEILtANCE aE0ulaE8ENTS 4.3.3.4 Each of the meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be demonstrated OPERABLE by the performance of:
- a. A Daily CHANNEL CHECK, and ,
- b. A Semiannual CHANNEL CALIBRATION
~
SEABROOK - UNIT 1 3/4 3-48 I
- i AD A . i
- - - 1
~~
.- . _ - . = _ _ _ _
4 TABLE 3.3-8 METECROLOGICAL MONITORING INSTRUMENTATION . g 4
~
MINIMUM INSTRUMENT LOCATION OPERABLE i
- 1. Wind Speed
- a. Lower Level Nominal Elev. 43 ft. 1 i e
! b. Upper Level Nominal Elev. 209 ft. 1
- 2. Wind Direction
^
- a. Lower Level Nominal Elev. 43 ft. 1
- b. Upper Level Nominal Elev. 209 ft. 1 1 3. Air Temperature - AT I
- a. Lower Level Nominal Elev. 43-150 ft. I
- b. Upper Level Nominal Elev. 43-209 ft. 1 O
i 1 , J 1 4 I i l s i O o ~ 1- g ~- % j SEABROOK - UNIT 1 3/4 3-49 3ff h 74 f -4
.. - - _ _ _ . _ . - _ ___m_ l 98f
/
INSTRUMENTATION REMOTE SHUTDOWNTSYSTEM ! LIMITING CONDITION FOR OPERATION - Sde 3.3.3.5 The Remote E Shutdown System trans.fer switches, power, controls and monitoring instrurrentation channels shown in Table 3.3-9 shall be OPERABLE. c APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours.
Safe
- b. With one or more Remote > Shutdown System-transfer switches, power, or control circuits inoperable, restore the inoperable switch (s)/
circuit (s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours. ,m U c. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REOUIREMENTS sde 4.3.3.5.1 Each remote
- shutdown monitoring instrumentation chtnnel in Table 3.3-9 shall be demonstrated OPERASLE:
by
- a. Every 31 days -De performance of a CHANNEL CHECK, and
- b. Every 18 months by performance of a CHANNEL CALIBRATION
- except-the-
-CHANNEL CALIBRATI0tLfor_ power-range-neutron 41ux-shall-be-performed-
-every quarter.~
%f e
. . 4.3.3.5.2 Each Remote + Shutdown-System-transfer switch, power and control lide) to circuit 3 including the actuated components, shall be demonstrated OPERABLE gg,33-j ~ at least on:e per 18 months.
" Intermediate and Source Range neutron flux instrumentation -andtReactor-trip-
-breaker-indication-instrumentation are not required to have CHANNEL CALIBRATION operations.
("; &.~..~,, .
, p 3/4 3-50 # e d., d SEABROOK - UNIT 1 j
%h Lb Ji
~
k l
(% , i
\. s ,
TABLE 3.3-9 m REMOTE SHUTDOWN SYSTEM
~
HINIMUM
$ TOTAL NO.
g READOUT OF . CHANNELS X INSTRUMENT LOCATION CHANNELS OPERABLE e c -Power-Range-Neutron-Flux- -2 2 -
= 1 '2. Intermediate Range Neutron Flux CP-108 A and B 2 2 H t.'3. Source Range f;eutron Flux CP-108 A and B 2 2
- -4h R e a c t o r-T r i p-B rea k e r-I nd i ca ti on -1/ t r i p-b rea k er--1/ t r i p-b re a k e r- 'l 3 '5. Reactor Coolant Temperature -
Average klid,* LN
- f=4..J T'c '* ta'r$ CP-108 A and B 2 -2j
-6 Reactor-Coolant-F-low-Rate- -2 "2-9 7. Pressurizer Pressure CP-108 A and B 2 2i SB. Pressurizer Level CP-108 A and B 2 1 CP-108 A and B 9 2/stm. gen. si-i-2/stm. gen.
g g 7 f_J. Steam _ Generator Pressure 110. SteamCP-108 GeneratoMWater A and B Level gen. 4-l-2/s tm. gen. q-1-2/stm.
-Control-Rod Position-Limit Switches- -1/ insertion 1/i nserti on-l imi t--
m limit-switch / switch / rod-
} -rod-
, --R H R-F l ow-R a te--- - 4 -13.--RHR Temperatu- e i ,
- 314. Steam Generator-Emergency feedwater Flow Rate CP-108 A and B --2'/-'/A6or -2"/ -lh/+' k'3 9 15. Boric Acid Tank Level CP-108 A and B 2.-//T4,i A 2 -1 / Ge /t so u ri s.pr.m*M TRANSFER SWITCHES W SWITCH 2.- I LOCATION H _a I N uxiliary Feedwater Control
{d , i 2. Safe. Shutdown Equipment Power
': g,q up' Feed *'ter see atheLJ M ,
- c. Pressuriz
@;>jf r'
- d. Valves /\
eaters
- 3. CVCS Makeup 11ow Control -
h- g 4. Diesel-Generator Control , . '*T > 5., Ele ~ctrical Distribution System Control ' CONTROL CIRCUITS SWITCH LOCATION [_r, , f.
'lG Co
. Iw Auxiliary Feedwater,f16w yg 2. Pre's'surizer Heaters
- 3. CVCS Makeup-F. low gpg h-[- .
%mr,j , ~
- 4. Diesel Generator N
,5. -Electrical Distribution System l
TRANSFER SWITCHES LOCATION r - Emergency Feedwater Pump Steam Supply Valves CP-108 A MS-V-393/127 . Emergency Feedwater Pump Steam Supply Valves CP-108 B MS-V-394/128 Emergency Feedwater Pump Steam Supply Valves CP-108 A and B > MS-V-395 c Emergency Feedwater Pump FW-P-37 B BUS 6 SWCR Emergency Feedwater Recire Valve FW-V-346 CP-108 A Emergency Feedwater Recirc Valve FW-V-347 CP-108 B SG A EFW Control Valve FW-FV-4214 A CP-108 A SG A EFW Control Valve FW-FV-4214 B CP-108 B SG B EFW Control Valve FW-FV-4224 A CP-108 A SG B EFW Control Valve FW-FV-4224 B CP-108 B SG C EFW Control Valve FW-FV-4234 A , CP-108 A SG C EFW Control Valve FW-FV-4234 B CP-108 B SG D EFW Control Valve FW-FV-4244 A CP-108 A SG D EFW Control Valve FW-FV-4244 B CP-108 B SG A EFW Atmos Relief Valve MS-PV-3001 CP-108 A l SG B EFW Atmos Relief Valve MS-PV-3002 CP-108 B SG C EFW Atmos Relief Valve MS-PV-3003 CP-108 A l SG D EFW Atmos Relief Valve MS-PV-3004 CP-108 B MS Isol Valves MS-V-86/88/90/92 CP-108 A MS Isol Valves MS-V-86/88/90/92 CP-108 B Pressurizer Heaters Group A CP-108 A Pressurizer Heaters Group B CP-108 B () Charging Pump CS-P-2A BUS 5 SVCR Charging Pump CS-P-2B BUS 6 SWCR l
O TRANSFER SWITCHES I.0 CATION 1 Charging Pump Suction f rom R'JST CS-LCV-Il2D CP-108 A
~
Charging Pump Suction from R'4ST CS-LCV-ll2E CP-108 B Pressurizer Relief Valve (PORV) RC-PCV-456A CP-108 A Pressurizer Relief Valve (PORV) RC-PCV-4563 CP-108 B PORV Block Valve RC-V-122 c CP-108 A PORV Block Valve RC-V-124 CP-108 B I Righ Pressure Injection SI-V-138 CP-108 A High Pressure Injection SI-V-139 CP-108 B Charging Pump Miniflow CS-V-197 MCC-E612 VCT Discharge Isol Valve CS-LCV-Il2B CP-108 A VCT Discharge Isol Valve CS-LCV-Il2C CP-108 B O
~
CONTROL CIRCUITS _ LOCATION SG A Atmos Relief Valve MS-PV-3001 CP-108 A SG B Atmos Relief Valve MS-PV-3002 CP-108 B SG C At=os Relief Valve MS-PV-3003 CP-108 A SG D Atmos Relief Valve MS-PV-3004 CP-108 B a O 't
\~/ INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION
. LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.
c APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels except the containment atmosphere-high range radiation monitor, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status (J) within 48 hours or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- c. With the number of OPERABLE channels for the containment atmosphere-high range radiation monitor less than required by the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s), within 72 hours, and either restore the inoperable channel (s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specifi-cation 6.8.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS X 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE:
- a. Every 31 days by performance of a CHANNEL CHECK, and
- b. Every 18 months by performance of a CHANNEL CALIBRATION.
o O +: HelsTwn N Tesr af wi m ps 3t 6,y ,ad of,Jiut m p n Jmy .;r Test ansts y puj _ j a gja c4 a wu r ' , _ _
~
SEABROOK - UNIT 1 3/4 3-52 D) [*) .
- y~se,7,"{
d' r j
- s i_Ms. .ah>y.
( .) . M4911la r l.._
'v l J i ( %a l' <
,. J w' TABLE 3.3-10
$ ACCIDENT MONITORING INSTRUMENTATION g;
8 TOTAL MINIMUM
- Q NO. OF CilANNELS CilANNELS OPERABLE c INSTRUMENT
- 1. Containment Pressure
- 1 'l 2
al Normal Range 1 Extended Range 2 b. 4 2
- 2. Reactor Coolant System llot Leg-Water-Ti10T ( ide Range)
Temperature Acq 2 4
- 3. Reactor Coolant System Cold-Water - TCOLD (Wide Range)
Temperature 2 1
- 4. Reactor Coolant Pressure - Wide Range g
2 I y S. Pressurizer Water Level . w 2/ steam generator 1/s' team generator
- 6. Steam Generator Pressure Steam Generator Water Level - Harrow Range 1/ steam generator -1/s$eam generator-7 7.
W-Steam Generator Water Level - Wide Range 1/ steam generator -1/ steam-genera tor j 8. I 9. Refueling Water Storage Tank Water Level 2 1
/ valve -1/ valve,-
-Ma i n-S te am-I s ol a t i on-Va l ve-Pos i t i on
-2/va l ve- -1/ valve-
---Feedwater-Isolation Valve-Position ,
.. h 10 Reactor Coolant System Subcooling Margin Monitor 2 1
'12.
2 1 J N b. Containment Building Water Level
~u y h)J=$
- Edle s I may4w age m i Ed q pr Diem gauutt. .
J
LJ LJ Lj TABLE 3.3-10 (Continued) m 3; ACCIDENT MONITORING INSTRUMENTATION 8 HINIMUM ji! TOTAL NO. OF CilANNELS INSTRUf!ENT CilANNELS OPERABLE c-5 r2. 2/ core quadrant
- 14. Core Exit Thermocouples 4/ core quadrant
" *l hiik Kin m .Ye~t Plant Vent -Gaseousoactivi _Radbe &H L315. t -Monifor-- N.A. NA
-16:-Co n t a i nmen t- Enc l o s u re- B u i l d i ng - E f f l ue n t-Ra d i oa c t i v i ty- - 1 -N A-1417. Containment - r Ma'ti$- d M 2 1 IM8, Reactor Coolant system Inventory (RVLIS) 2 I w IU19. Containment liydro0en Concentration 2 1 s
[ )~170. Intermediate Range Neutron Flux 2 1
&* 1871. Intermediate Range Neutron Flux Rate 2 l'
-22.-Source-Range-Neutron-Fl ux - ~ 2- -
.23 Power-Range-Neutron-Flus 2 --I
} li M. Containment Isolation Valve Position
- 2/ Penetration 1/ Penetration Containment Enclosure Negative Pressure 2 1 h 10 25.. _
-Co nde n s a t e-S t o ra ge-Ta n k-Wa t e r- L e v e l 2 1 ---
f -y4 i (?%Y l g
- Applies to penetrations with 2 active valves ir. series. These valves are moved to the closed a g position by automatic signals.
t = s Y
,- ,ea
.'.G .
y CQ ' k.
- e. ?
4 O ' INSTRUMENTATION i FIRE DETECTION INSTRUMENTATION - 1 LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-11 shall be OPERASLE. , APPLICABILITY: Whenever equipment protected by the fire detection instrument ' is required to be OPERABLE. ACTION:
- a. With any, but not more than one-half the total in any fire zone, Type. X --Function-A-fire detection instruments shown in Table 3.3-11 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next 1 hour establish a fire watch patrol to inspect the zone (s) with the incperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours (or monitor the containment air temperature at least once per hour at thelocationslistedinSpecification4.6.1.p.
i
- b. With more than one-half of the Function A fire detection instruments O in any fire zone shown in Table 3.3-11 inoperable, or with any Type y -Function-8 fire detection instruments shown in Table 3.3-11 inoperable, or with any two or more adjacent fire detection instruments shown in Table 3.3-11 inoperable, within 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the contain-ment, then inspect that containment zone at least once per 8 hours i
(or monitor the containment air temperature at least once per hour atthelocationslistedinSpecification4.6.1.j).
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.3.7.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated " OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months. 4.3.3.7.2 The NFPA Standard 720 supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. 4.3.3.7.3 The nonsupervised circuits, associated with detector alarms, between O the instrument and the control room shall be demonstrated OPERABLE at least once per 31 days. MAfL13JS8.6 SEABROOK - UNIT 1 3/4 3-55 , aa.n1 _ 1 _
- - - - --w,-. , ,,,.,y,.yv,.-w - - _ , . .._,r. n,.4 . _ - , , - - + - ---.,w4, ,.- .--.--y-y#y.r.__- ,, y-w - ...--e e--- -
--w
TABLE 3.3-11 FIRE DETECTION INSTRUMENTS - F ' TOTAL NUM8ER INSTRUMENT LOCATION OF INSTRUMENTS
- HEAT FLAME SM0KE (x/y) (x/y) (r/y)
- 1. CONTAINMENT **
i Control Panel #376 E1. O'0" Zone #1 - 16/0 Zone #2 El. O'0" 19/0 Zone #3 El 0'0" 12/0 Zone #4 E1. O'0" 16/0 Zone #5 El. (-)26' 0" 23/0 Zone #6 E1. (-)26' 0" 8/0 Zone #7 El. (-)26' 0" 12/0 Zone #8 El. (-)26' 0" 20/0 Zone #9 El. (-)26' 0" 11/0
- 2. CONTROL BUILDING Control Panel #377
. Zone #1 El. 75' 0" 17/0
) Zone #2 El. 75' 0" ide- 10/0 Zone #3 El. 75' 0" 1/o ' 18/0 Zone #4 El. 75' 0" 9/0 Zone #5 El. 21' 6" 12/0 O. Zone #6 Zone #7 El. 21' 6" 12/0 El. 21' 6" 3/0 Zone #8 El. 21' 6" 3/0 Zone #9 E), 21' 6" 3/0 Zone #10 El. 21' 6" 3/0
. Zone #11 El. 21' 6" 14/0 Zone #12 El. 21' 6" 1/0 .
Zone #13 El. 21' 6" 1/0 Zone #17 El. 50' 0" 14/0 Zona #18 E1. 50' 0" 12/0 4 Zone #19 E1. 75' 0" 2/0 Zone #20 El. 21' 6" 9/0 Zone #21 E1. 21' 6" 9/0 Zone #22 El. 21' 6" 30/0 Zone #23 E1. 75' 0" 13/0
- Zone #24 E1. 21' 6" 10/0
*(x/y): x is number of-function-A-fearly warning fire detection and notification only) instruments.
y is number of -Function-8-(actuation of Fire Suppression Systemsandearlywarningandnotification} instruments.
**The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A containment
^
leakage rate tests.
' ~ ~ -
~MM l$ l$g '
SEABROOK - UNIT 1 3/4 3-56 ,s ; f %. .
TOTAL NUMBER
\ INSTRUMENT LOCATION OF INSTRUMENTS
- HEAT FLAME SMOKE (x/y) (x/y) (x/y) -
, 3. PRIMARY AUXILIARY BUILDING
~
Control Panel #378 Zone #1 El. 7' 0" 2/0 , Zone #2 El. 7' 0" . 2/0 Zone #3 El. 53' 0" 9/0 e Zone #4 El. 81' 0" 12/0 Zone #5 El. 7' 0" 10/0 Zone #6 El. 7' 0" 2/0 Zone #7 El. 53' 0" 14/0 Zone #8 El. 53' 0" 18/0 Zone #9 El. 7' 0" 4/0 Zone #10 El. 7' 0" 8/0 Zone #11 El. 7' 0" 17/0 Zone #12 El. 53' 0" 2/0 Zone #13 El. 53' 0" 2/0
- 4. SERVICE WATER PUMPHOUSE Control Panel #380 O Zone #1 El. 21' 6" 14/0 V Zone #2 El. 21' 6" 9/0 i 5. SERVICE WATER COOLING TOWER Control Panel #381 Zone #3 E1. 22' 0" 3/0 Zone #4 El. 22' 0" 3/0 .
Zone #6 El. 46' 0" 19/0 )
- 6. ELECTRICAL TUNNELS (A & 8)
Control Panel #409 Zone #1 El. (-)26' 0" 0/28 Zone #2 El. (-)26' 0" 0/28 Zone #3 El. O' 0" 0/28 2J Zone #4 El. O' 0" 0/25 Zone #7 El. 50' to (-)2' 0/7 Zone #8 El. 50' to O' 0/2
- 7. DIESEL GENERATOR BUILDING "A" )
Control Panel #412 Zone #1 El. 25' 0" 0/10 Zone #2 El. 25' 0" 8/0 Zone #3 El. (-)16' 0" 0/7 ! SEABROOK - UNIT 1 3/4 3-57 3 4 4 . r emuum. a .L p] k
TOTAL NUMBER O'- INSTRUMENT LOCATION OF INSTRUMENTS * . HEAT FLAME SMOKE . 7 (x/y) (x/y) (x/y) -
- 7. DIESEL GENERATOR BUILDING "A" (Continued)
~
Zone #4 El. (-)16' 0" 0/7 Zone #5 El. 51' 0" 1/0 Zone #6 El. 51' 0" 1/0 Zone #11 El. 51' 0" - 10/0 Zone #12 El. 25' 0" 27/0
- 8. DIESEL GENERATOR BUILDING "B" Control Panel #413 Zone #1 El. 25' 0" 0/10 Zone #2 El. 25' 0" 8/0 t Zone #3 El. (-)16' 0" 0/7 Zone #4 El. (-)16' 0" 0/7 Zone #5 El. 51' 0" 1/0 Zone #6 El. 51' 0" 1/0 Zone #11 El. 51' 0" 10/0 Zone #12 El. 25' 0" 27/0
- 9. CABLE SPREADING ROCM i
' Control Panel #414 Zone #1 El. 50' 0" 0/21 Zone #2 El. 50' 0" 0/15 Zone #3 El. 50' 0" 0/6 Zone #4 El. 50' 0" 0/6 Zone #5 El. 50' 0" 0/6 .
, Zone #6 El. 50' 0" 0/6 Zone #7 El. 50' 0" 0/6
- 10. CABLE TUNNEL C - (PA8 TO CONTROL BUILDING ABOVE RHR VAULT) f Control Panel #421 .
Zone #1 El. 30' 0" 0/8 Zone #2 E2. 30' 0" 0/9 , 11. CONTAINMENT FAN ENCLOSURE Control Panel #444 1 Zone #1 El. 25' 0" 9/0 Zone #2 El. 25' 0" If10/0 1 e
..m , ,_
SEABROOK - UNIT 1 3/4 3-58 [) ,.{' }("; y z j V; AY
. . ~2 d [ }
TOTAL NUMBER
\ INSTRUMENT LOCATION OF INSTRUMENTS * .
HEAT FLAME SMOKE . I (x/y) 6/y'} (x/y)
- 12. EMERGENCY FEEDWATER PUMP BUILDING Control Panel #445 Zone #1 El. 27' 0" 11/0
- 13. MECHANICAL PENETRATION AREA e
Control Panel #446 Zone #1 E1. (-)34' and (-)20" 8/0 Zone #2 El. (-)34' and (-)20" 19/0
- 14. EAST - MS AND FW PIPECHASE Control Panel #451 Zone #1 El. 12' 0" 9/0 Zone #2 El. 8' 0" 4/0 3/0 Zone #3 El. 28' 0" '
2/0 Zone #4 El. 28' 0" 2/0 Q 15. WEST - MS AND FW PIPECHASE Control Panel #452 Zone #1 El. 12' 0" 6/0 Zone #2 El. 8' 0" 6/0 Zone #3 El. 25' 0" 1/0 Zone #4 El. 28' 0" 2/0 -
- 16. PRIMARY AUXILIARY BUILDING Control Panel #453 Zone #1 El. (-)6' 0" 29/0 Zone #3 El. 25' 0" 12/0 Zone #4 El. 25' 0" 18/0
- 17. FUEL STORAGE BUILDING r Control Panel #454 I
Zone #1 E1. 7' 0" 7/0 Zone #2 E1. 21' 0" 9/0 Zone #3 El. 64' 0" 30/0 Zone #4 . E1. 21' 0" 11/0 Zone #5 Elf 64' 0" 13/0 0 MAR 131988 SEABROOK - UNIT 1 3/4 3-59
\\
I u ua -y
~~_~_-;___----
ue-- a
q TOTAL NUMBER V INSTRUMENT LOCATION OF INSTRUMENTS
- HEAT FLAME SMOKE '
.-- (x/y) (x/y) (x/y) - t
,18. PRIMAR AUXILIARY BUILDING
~
Control Panel #471 Zone #1 El, 25' 0" 0/30 Zone #2 El. 25' 0" . 0/30.29 Zone #3 El. 25' 0" 0/17
- Zone #4 El. 25' 0" c 0/21 Zone #5 El. 25' 0" 0/13
- 19. SERVICE WATER PUMPHOUS (PUMP AREA)
Control Panel #474 Zone #1 El. 21' 6" 21/0 Zone #2 E1. 21' 6" 26/0
- 20. RHR VAULTS Control Panel #475 ,
Zone #1 E1. (-)61' to O' 5/0 O Zone #2 El. (-)61' to O' 5/0 V Zone #3 E1. (-)61' 0" 2/0 Zone #4 El. (-)61' 0" 2/0 Zone #5 EI. (-)S0' 0" 2/0 Zone #6 El. (-)50' 0" 2/0 Zone #7 El. (-)31' to O' 3/0 Zone #8 El. (-)31' to O' 3/0 Zone #9 El. (-)61' to 20' 7/0 . Zone #10 El. (-)61' to 20' 7/0
- 21. FIRE PUMP HOUSE Control Pane #176 Zone #1 El. 21' 0" 4/0 Zone #2 El. 21' 0" 4/0 Zone #3 El. 21' 0" 1/0 O ....-. MAR 131986 I -.. .
SEABROOK - UNIT 1 3/4 3-60 y -- ,, 1 ., L' 1 - ./ .. . ,[
, O ' tasTausta'attoa LOOSE-PART DETECTION SYSTEM , LIMITING CONDITION FOR OPERATION 3.3.3.8 The Loose-Part Detection System shall be OPERABLE. APDLICABILITY: MODES 1 and 2. - ACTION: P i
- a. With one or more Loose-Part Detection System channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, i
SURVEILLANCE REQUIREMENTS 4.3.3.8 Each channel of the Loose-Part Detection Systems shall be demonstrated OPERABLE by performance of:
- a. A CHANNEL CHECK at least once per 24 hours,
- b. An ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and
- c. A CHANNEL CALIBRATION at least once per 18 months.
3/4 3-61 p .,,,) '
~.
*r,;
+f SEABROOK - UNIT 1 r
; 4: , .f, , ;
J-j A G. J.~.4 j '
.d.
*N* =.. .
INSTRUMENTATION
~
. RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION
. t LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm /
Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL P (ODCM). APPLICABILITY: At all times. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable,
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown l in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the-time-specified-in-the-ACTION;-or explain in the O so aD __ next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.8.1.4 why this inoperability was not corrected within the time specified.
- c. The provisions of Specifications 3.0.3 and 3.0.4, are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECX, CHANNEL CALIBRATION, and HANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3.7. g Di (TA L MAR 131986 O y ) v ;; n r, SEABROOK - UNIT 1 3/4 3-62 h/ J 3h, A y 1
,.3 o
O O . O TABLE 3.3-12 . M
$; RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 5 -
S
. MINIMUM CHANNELS c
Z INSTRUMENT OPERABLE ACTION
- Radioactivity Monitors Providing Alarm and *!
1. Automatic Termination of Release
- a. Liquid Radwaste Test Tank Discharge 1 29
- b. Steam Generator 810wdown Flash Tank Drain 1 30
- c. Turbine Building Sumps Effluent Line 1 30 w 2. Flow Rate Measurement Devices 1 31 m a. Liquid Radwaste Test Tank Discharge . 1 6 .
31
" b. Steam Generator Blevdown Flash Tank Drain
- 1
- c. Circulating Water Discharge w *- -31 ^^ d -
' t .
t s
! r y, 4O M ina e,0 a.o d Md M nP m w,a m utilft/ kddtW' g.. . c.,
s a m
- CD Lw ,
.. I
~ .~
TABLE 3.3-12 (Continued) . ACTION STATEMENTS ,
. ACTION 29 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue -for-up-to-14. days provided that prior to initiating a release:
- a. At least two independer'it samples are analyzed in accordance with Specification 4.11.1.1.1, and ,,
- b. At least two technically qualified members of the station .
staff independently verify the release rate calculations and discharge line valving. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 30 - With the number of channels OPERABLE less than the Minimum Chan :Is OPERABLE requirement, effluent releases via this pathway 4y continue provided grab samples are analyzed for radioactivity fer-up-to-30-days at a lower limit of detection of no more than 10 7 microcurie /ml: O a. At least once per 12 hours when the specific activity of V the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or
- b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
ACTION 31 - With the number of channels OPERA 8LE less than the' Minimum Chane.els CPERABLE requirement, effluent releases via this pathway may continue for-up-to-30-days provided the flow rate is estimated at least once per 4 hours during actual releases. Pump perfor-mance curves genersted in place may be used to estimate flow, s
---- MAR 131986
. O - . , .
! Yi fi ,,) l D* ,'*]h U
} Lq d > ( )((.j g.
I SEABROOK - UNIT 1 3/4 3-64 ,
-+ . ;
i ~~ ~ -
- . ~ _
e
O O , O TABLE 4.3-5
% RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
o
- n
- ANAL h CHANNEL c-CHANNEL SOURCE CHANNEL OPERATIONAL 5* CALIBRATION TEST INSTRUMENT CHECK CHECK
~ ,
'l
- 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release l'
- a. Liquid Radwaste Test Tank Discharge D . P R(3) %(1)
- b. Steam Generator Blowdown Flash Tank Drain D M R(3) Q(1) m c. Turbine Building Sumps 1 Effluent Line D M R(3) Q(1)
J, 2. Flow Rate Measurement Devices ,
- a. Liquid Radwaste Test Tank Discharge D(4) N.A. R Q(2)
- b. Steam Generator Blowdown Flash Tank Drain D(4) N.A. R N.A. *
~ .--.
- N.A. N.A.
l, c. Circulating Water Discharge N.A. lf7 j
-e
, Pump curves may be used to estimate flow.
j
. . , n
.. .= 2 a
.d h l
g' 5
,e 4
[ 2: i ' f t 5
- 2
TABLE 4.3-5 (Continued)
- TABLE NOTATIONS
- - .~ s (1) The DIGITAL CHANNEL OPERATIONAL TEST shall be performed.
' ~
(2) The ANALOG CHANNEL OPERATIONAL TEST shall be performed. (3) The inicial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by.the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with N85. These standards shall permit calibrating the system over its intended range of energy and measurement 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 on which continuous, periodic, or batch releases are made. O I l l l l l l l l l O r7ix
. MAR 131986 ,
SEABROOK - UNIT 1 3/4 3-66 h S *' b' 1 d 7 1 A a
INSTRUMENTATION RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
. t LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to 4 ensurethatthelimitsofSpecifications3.11.2.1and3.11.2.3Xrenotexceeded.
The Alarm / Trip Setpoints of these channels meeting Specification 3.11.2.1 shal) be determined and adjusted in accordance with the methodology and parameters in the ODCM. APPLICABILITY: As shown in Table 3.3-13 ACTION:
- a. With a radioactive gaseous ef fluent monitcring instrumentation l channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.
- b. With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table 3.3-13. Restore r _the._inoperabl_e so clup O sastru=eatatior to OeERABLE stetus withia<the-time-epecirree-ia-tne.
-AC-TION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.8.1.4 why this inoperability was not corrected within the time specified.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK,' CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-3. pha 6 O p 1,. . ,.. , .-_ c . - SEABROOK - UNIT 1 3/4 3-67 s A J MAR 108a8 il
O O , O
^
TABLE 3.3-13 3; RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION a R .
. MINIMUM CilANNELS
- c. INSTRUMENT OPERABLE APPLICABILITY ACTION 5
- 1. RADI0 ACTIVE GAS WASTE SYSTEM EXPLOSIVE GAS
- MONITORING SYSTEM 'I s
0xygen Monitor (Process) 1 34
- 2. PLANT VENT
- a. Noble Gas Activity Monitor 1 33
- b. Iodine Sampler 1 35 w
- 1 c. Particulate Sampler 1 35
- 32
- d. Flow Rate Monitor 1 ,
co
- e. Sampler Flow Rate Monitor 1 32
-3. FU E L-STO R AG E- A R E A-VENT I L AT I ON-SYST EM- l
- Noble-Ga s-Activi ty-Moni tor--- -*- -3 3-
-a. *
! -b. Iodine-Sampler- L""' / ,c. Particulate Sampler- -1 ~ -*- '
g: h e .' Flow-Rate-Monitor- *-- p, i 1 l
-d .
w Sampler-Flow Rate _ Mon { tor _ _ 1- '" e. CD k...e-er> Y- , J. , N- I . J
O O . O TABLE 3.3-13 (Continued) v, k RADIDACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION 8 9 - MINIMUM CHANNELS c INSTRUMENT OPERABLE APPLICABILITY ACTION
. TURBINE GLAND SEAL CONDENSER EXHAUST
~
*I
- a. Iodine,'j fitetscda.f Sampler 1
- 35
-b. Particulate-Sampler--- -1 35 h h. Sampler Flow Rate-Monitor 1 .
*** A/4 R.
w-r: , l m b4 -
> pr#
l .2P
- f.-- **
u I
~ = hes gli pa-
- a. .
s* ,
%ee 0 e n
- s. ,
em
TABLE 3.3-13 (Continued)
. TABLE NOTATIONS i
- At all times. -
** During RADI0ACTTVE GAS WASTE ." STEM one'ation.
W [10lb511fm4 A jg{nffg yf ACTION 32 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue -for-up-to-30-days provided the flow rate ('is estimated at least once per 4 hours. ACTION 33- With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for-up-to-30-days provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. ACTION 34 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, operation of this RADIOACTIVE GAS WASTE SYSTEM may continue provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours. ACTION 35 - With the numb,er of channels OPERABLE less than the Minimum (",) Channels OPERABLE requirement, effluent releases via the affected pathway may continue for-up-to-30-days provided samples are con-tinuously collected with auxiliary sampling equipment as required in Table-4.-11 % ODCM s *
/
O 5 4, nesb 7. ci fI
,a W
SEABROOK - UNIT 1 3/4 3-70 ,___
, ~g+ +/ ,)b ,1 1
-MAR l3-1986 mwee
O - O , O TABLE 4.3-6 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8 -
% " ANALOG =
CHANNEL MODES FOR WHICH c CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE 5
- INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED w I
- 1. 'RADI0 ACTIVE GAS WASTE SYSTEM EXPLOSIVE GAS MGNITORING SYSTEM 0xygen Monitor D N.A. @-3) or QC M .
(Process)
- 2. PLANT VENT w a. Noble Gas Activity Monitor 'D M R(2) Q(1) s
- b. Iodine Sampler W N.A. N.A. N. A.
U c. Particulate Sampler ~W N.A. N.A. N.A. *
- d. Flow Rate Monitor D N.A. R 4A/J- /
- e. Sampler Flow Rate Monitor D N.A. R -Q#A
- N' -3 -FUEt-STORAGE-AREA-VENTILATION-Q SYSTEM
' *~
p".s-j -a:-Noble-Gas-Activity-Moni tor 1: M -R(-2) Q
,f j '
[ -b-Iodine-Sampler W- NrA. NrA. -NrA. *
^
2 5 - c-Particulate-Sampler W N.A. N.A. NrA. " p. a tfrflow-Rate-Moniter 0 NrA. -R y *- &&'^.. , 4 - e -SampTer Flow-Rate-Monitor D NrA. R Q
*~
hei?;
. ~ .
O O , O TABLE 4.3-6 (Cont.inued) RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTRirdENTATION SURVEILLANCE REQUIREMENTS o" . O *
~
7 ANALOG c- CHANNEL MODES FOR Ml!CH 5 CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK . CHECK CALIBRATION TEST IS REQUIRED , 3% TURBINE GLAND SEAL CONDENSER EXHAUST ' 1LT M
- a. Iodinef Sampler W N.A. N.A. N.A
- M*-
3
-b-Part-iculate sampler . W- N-A. N. ^ _ -
_N,A. *-
. b t. Sampler Flow Rate D N A. R -Q-A/A. *D R
Y M e 1 1 ,m Yh b- J
, Nb!
ya A wV I 2 ); - , e 8' g %Tyl g CO Y ' N))
.M
J TABLE 4.3-6 (Continued) TABLE NOTATIONS , At all times.
** During RADI0 ACTIVE WASTE GAS SYSTEM operation.
M HOOEs g %Q u>dkm Ns.w rt FtSzy bjM Mb & (1) The DIGITAL CHANNEL OPERATIONAL TEST shall be performed (2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified bf the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initi(1 calibratirn shall be used.
-f3 )----Th e-CH ANN E L-CA LI B RAT ION - s ha lbi n c l ude -the -u s e-o f-s ta nda rd -ga s -s amp l e ss
-containing-a-nominal.
---a .-One -vol ume -pe rc e n t-hyd roge nr balance nitrogen, and-
-b:---Four volume-percent-hydrogen;--balance-nitrogen.
3141 The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: Q a. One volume percent oxygen, balance nitrogan, and
- b. Four volume percent oxygen, balance nitrogen.
O rs n p nq r n SEABROOK - UNIT 1 3/4 3-73 j h { :' ,i;\
.n J' .C n i~ii .d. .h.
Mg la 1886- , 1
l l l i INSTRUMENTATION - 3/4.3.4 TURBINE OVERSPEED PROTECTION , LIMITING CONDITIDN FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE. APPICABILITY: MODES 1, 2, and 3g d%/6 de/,) stw3/- ACTION:
- a. With one stop valve or one control valve per high pressure turbine steam line inoperable and/or with one intermediate stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours.
- b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours isolate the turbine from the steam supply.
' SURVEILLANCE REOUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.
4.3.4.2 The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE: 3
- a. At least once per)), days by cycling each of the following valves through at least one complete cycle from the running position:
- 1) Four high pressure turbine stop valves,
- 2) Four high pressure turbine governor valves,
- 3) Four low pressure turbine reheat stop valves, and
- 4) Four low pressure turbine reheat intercept valves.
- b. At least once per 31 days by direct observation of the movement of each of the above valves through one complete cycle from the running position,
- c. At least once per 18 months by performance of a CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems, and
- d. t.t least once per 40 months by disassembling at least one of each of tne above valves and performing a visual and surface inspection of valve seats, disks, and stems and verifying no unacceptable flaws or excessive corrosion. If unacceptable flaws or excessive corrosion are found, sll other valves of that type shall be inspected.
nn c, p i 3/4 3-74 U' -c [; 4 SEABROOK - UNIT 1 L Y b h. D N;** y!j E Y 41886
3/4.4 REACTOR COOLANT SYSTEM -
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- 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCUf.~.TTON r -
STARTUP AND POWER OPERATION
~
LIMITING CONDITION FOR OPERATION i 3.4.1.1 All reactor coolant loops shall be in operation. APPLICABILITY: MODES 1 and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours. O SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. _ -l
*See Special Test Exceptions Specification 3.10.4.
SEABROOK - UNIT 1 3/4 4-1 I 5. .' C /Q k ( I[4 L AA
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REACTOR COOLANT SYSTEM 5 - HOT STANDBY f, LIMITING CONDITI@ FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listeo below shall be OPERABLE with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when the Reactor Trip System breakers are open:*
- a. R2 actor Coolant Loop A and its associated steam generator and-reactor coolant pump,
- b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
APPLICABILITY: MODE 3.*
- ACTION:
- a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be-in HOT SHUTDOWN within the next 12 hours.
b b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within I hour open the Reactor Trip System breakers.
- c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the
- required reactor coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17% at least once per 12 hours. 4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours.
" All reactor coolant pumps may be deenergized for up to I hour provided: (1) no coerations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
O *
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REACTOR COOLANT SYSTEM HOT SHUTDOWN - LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:*
- a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,**
c
- b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,"*
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,**
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump,**
- e. RHR Loop A, and-
- f. RHR Loop B. ,
{} APPLICABILITY: MODE 4. ACTION:
- a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE
- status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 tours. .
- b. With no loop in operation, suspend all operations involving a reduc-tion in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop to operation.
*All reactor coolant pumps and RHR pumps may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
**A reactor coolant pump shall not be started unless the secondary water.
temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures. w & Q TAA C4arkn 310 4 , i O -MAR 131986
- e. .j M Dr h.;
/.A 3/4 4-3 b,. c' l, .
SEABROOK - UNIT 1 . . .n r. i as .4. .s s d A ( I e-, - - - - , .-- -- ,- . , - - - - - - - - - , - ,,
O acacroa coo'a~r svsres HOT SHUTDOWN -. SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor ecolant pump (s), if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. , 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17% at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours. 4 e O a s v 'E T f O y 9my l SEABROOK - UNIT 1 3/4 4-4 .m *;m ., .
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I 1 l i REACTOR COOLANT SYSTEM COLD SHUTDOWN - ROOPS FILLED . LIMITING CONDITICN FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:
- a. One additional RHR loop shall be OPERABLE **, or
- b. The secondary side water level of at least two f' team generators shall be greater than 17%.
APPLICABILITY: MODE 5 with reactor coolant loops filled ***. N ACTION:
- a. With one of the RHR loops inoperable and with less than the required steam generator water level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR h loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
*The RHR pump may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at lcast 10*F below saturation temperature.
**0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
***A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures,
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COLD SHUTDOWN - IDOPS NOT FILLED - l 4 LIMITING CONDITION FOP. OPERATION _ 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE
- and at least one RHR loop shall be in operation.**
APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION:
- a. With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE REOUIREMENTS
~
4.4.1.4.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. .
*0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
**The RHR pump may be deenergized for up to I hour provided: (1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
l l
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O MAR 131986 n, n.9 [, d y,., 3/4 4-6 ' SEABROOK - UNIT 1 ? A\
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REACTOR COOLANT SYSTEM
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3/4.4.2 SAFETY VALVES - - SHUTDOWN _ LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig i 1%.* APPLICABILITY: MODES 4 and 5. ACTION: With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. O SURVEILLANCE REOUIREMENTS 4.4.2.1 No additional requirements other than those required by Specification 4.0.5.
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
O -MAR -131986 pn UU n[1g kri rp LY fi
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- OPERATING -- , ,
LIMITING CONDITION FOR OPERATION _ 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig 1% * - APPLICABILITY: MODES 1, 2, and 3. ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. ) SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specification 4.0.5.
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
O v4aja issa mn a nn IW h I SEABROOK - UNIT 1 3/4 4-8 c 2; M u .E. 3.
REACTOR COOLANT SYSTEM
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3/4.4.3 PRESSURPZER . t LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 92% of pressurizer level (1656 cubic feet), and at least two groups of pressurizer heaters each having a capacity of at least 150 kW. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours.
O SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by . energizing-the-heaters-from-the-emergency-power ~
--supply-and measuring circuit current at least once per 62-days--
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3/4.4.4 RELIEF YALVES . . t i LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise, be in at least d
HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one PORV inoperable due to causes other than excessive seat leakage, within I hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY utihi sthe next 6 hours and in COLD O Sauto0Wu within the foiiewins 30 houb <su
- c. With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour either restore each of the PORV(s) to OPERABLE l
status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. -
- d. With one or more block valve (s) inoperable, within I hour:
(1) restore the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s), or close the PORV and remove power from its assoicated solenoid valve; and (2) apply the ACTION b. or c. above, as appropriate, for the isolated PORV(s).
- e. The provisions of Specification 3.0.4 are not applicable.
O YD SEABROOK - UNIT 1 3/4 4-10 [ Q _m .1 #( l1U ' ( T r __ _ _ _ _ _ _-- __-____m.___ - _ - .
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3/4.4.4 RELIEF 5ALVES
. SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:
e
- a. Performance of a CHANNEL CALIBRATION, and
-b. Ope ra t i ng -t he - v a l v e -t h ro ug h_c ne - c omp l e t e - cy c l e-o f-f ul l-tra v e h 4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.
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3/4.4.5 STEAM GENERATORS , . g LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. c ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200'F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. , 4.4.5.1 Steam Generator Sample Selection and Insocction - Each steam generator b^ shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Samole Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in 3pecification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except;
- a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;
- b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
O Mall 131986
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SEABROOK - UNIT 1 3/4 4-12 .; , tj ? . ~. 1 j . a u. 2
REACTOR COOLANT SYSTEM STEAM GENERATORSr- . g SURVEILLANCE REOUIREMENTS (Continued)
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection (pursuant to Specificaticn 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
- 2) The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories: Catecorv Inspection Results , C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. O MAR 131986 SEABROOK - UNIT 1 3/4 4-13
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SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.
Subsequent iriservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preser-vice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.'5.3a.; the interval may then be extended to a maximum of once per 40 months; and
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1) Reactor-to-secondary tubes leak (not including leaks originating from tube-to-tube' sheet welds) in excess of the limits of Specification 3.4.6.2, or .
- 2) A. seismic occurrence greater than the Operating Basis Earthquake, or WH
- 3) Ak loss-of-coolant accident requiring actuation of the Engineered Safety Features, or 64.t m TIT
- 4) Ak main steam line or feedwater line break.
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O MAR 131986
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SEABROOK - UNIT 1 3/4 4-14 "T'hi6'.e 7?] d b.\
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SURVEILLANCE REOUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
- a. As used in this specification: ,
- g. 1) Imperfection means an exception to the dimensions, finish, or -
contour of a tube from that required by fabrication drawings or"~ specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;
- 2) Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube;
- 3) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
- 4) % Degradation means the percentage of the tube wall thickness affected or removed by degradation;
~~)
V Defect means an imperfection of such severity that it exceeds 5) the plugging limit. A tube containing a defect is defective;
- 6) Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40%*
of t!e nominal tube wall thickness;
- 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above;
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the tcp support of the cold leg; and
*Value to be determined in accordance with the recommendations of Regulatory Guide 1.121, August 1976.
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n v MAR 131986 SEABROOK - UNIT 1 3/4 4-15 71 Vh 1 w.
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STEAM GENERATOR -- , SURVEILLANCE REQUIREMENTS (Continued)
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Recorts
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.8.2; O a- Tae c =aiete re=#it er tae = tee = semeretor t#ee '# ervice '# a ct<="
shall be submitted to the Commission in a Special Report pursuant to V Specification 6.8.2 within 12 months following the completion of the inspection. This Special Report shall include:
- 1) Number and extent of tubes inspected,
- 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and 4
- 3) Identification of tubes plugged.
- c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.8.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
O $.' 99 %v a.Mq SEABROOK - UNIT 1 3/4 4-16 ] . , - j f; [. / i d d" d .L[2,\::. 1
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_ TABLE 4.4-1 g
. MIMIMUM NUMBER OF STEAM GENERATORS TO BE '
INSPECTED DURING INSERVICE INSPECTION ! Preservice Inspection ~ Yes No. of Steam Generators per Unit Four First Inservice Inspection
- Two Second & Subsequent Inservice Inspections 1 One
_ TABLE NOTATION 1. The inservice inspection may be limited to one steam generator on a ro-tating schedule encompassing 12% of the tubes if the results of the first or aprevious in inspections indicate that all steam generators are performing like manner. ditions in one or more steam generators may be found to be than those in other steam generators. sequence shall be modified to inspect the most severe conditions.Under s Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections The scribedfourth and subsequent inspections shall follow the instructions de-above. 9
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O O l n O M TABLE 4.4-2 3; E R STEAM GENER ATOR TUBE INSPECTION E IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION l Z Sample Size Result Action Required Result Action Required Result Action Required j g A minimum of C-1 None N. A. N. A. N. A. N. A. S Tubes per .c,e i S. G. j C-2 Plug defective tubes C-1 None N. A. N. A. I and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G. ) Perform action for C-3 C-3 result of first I y sample
? Perform action for ,
g C-3 C-3 result of first N. A. , N. A. .I samp!e C-3 Inspect all tubes in All other this S. G., plug de- S. G.s are - None N. A. N. A. fective tubes and C-1 inspect 2S tubes in S me S. G.s Perform action for '4 N. A. N. A. each other S. G. C-2 but no C-2 result of second i l additional g,
. A Notificatiort to NRC S. G. are pursuant to $50.72 C-3 i
< (bil2) of 10 CFR Additional inspect all tubes in Part 50 S. G. is C-3 each S. G. and plug p'g-)jg- { defective tubes. Notification to NRC N. A. N. A. m pursuant to 550.72 w (bil2) of 10 CFR lj@,*'. g Part 50
! Where N is-the' number-of-steem generators-in-the-un!t,-and n is the number of steam generators inspected .
gq j n during an inspection
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't rN REACTOR CG0LANT SYSTEM V ' 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE
~
,- . s LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant Sistem Leakage Detection Systems shall be OPERABLE:
- a. The Containment Atmosphere Particulate Radioactivity Monitoring System,
- b. The Containment Drainage Sump Level Monitoring System, and
- c. Containment Radioactive Ga: Monitor APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: With only two of the above requireu .ukage Detection Systems OPC.RABLE, operation may continue for up to 30 cays provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours when the O' re9=<ree cesec== or eertic iete aeetoective saattoriass v te <= <aoneredie: otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD w following 30 hours. ;ble S D e/ @-/- &M 3 #*3 SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Gaseous and Particulate Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-37DIC.tTAL
- b. Containment Drainage Sump Level Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.
\
/
l O . MAR 1a isse l SEABRC0K - UNIT 1 3/4 4-19 l p j
! 3 9 I A N N] , P '*g I L _.
f- REACTOR COOLANT SYSTEM
. OPERATIONAL LEAKAGE -
LIMITING CONDITION FOR OPERATION 3.4.6 2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE, -
- b. I gpm UNIDENTIFIED LEAKAGE,
- c. 1 gpm total reactor-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 52 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig, and f.
0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the -
above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following l 30 hours.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of
- the affected system from the low pressure portfon within 4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
O -- MAR 131986 l SEABROOK - UNIT 1 3/4 4-20 g
. gnW L x
q
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE 7 . ! SURVEILLANCE REQUIREMENTS 4.'4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
- a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours; g,
- b. ~
Monitoring the containment drainage sump inventory and discharge at least once per 12 hours; c. km Measurement of the CONTROLLED LEAKAGE to-the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days .with-the-modulating-valve-ful4y-operr. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;
- d. e.of a Reactor Coolant System water inventory balance M ' within-12 #hours after achieving steady state operation
- and at least once per 72 hours thereafter during steady state operation except that not more than 96 hours shall elapse between any two successive inventory balances; I O e. Monitoring the Reactor Head Flange Leakoff System.at least once per 24 hours.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying 1.eakage to be within its limit:
- a. At least once per 18 months,
- b. Prior to entering MODE 2 whenever the plant has been in COLD 1 l
SHUTDOWN for 72 hours or more and if leakage testing has not been l performed in the previous 9 months,
- c. Prior to returning the valve to service following maintenance, 1 l
repair or replacement work on the valve, and l l
- d. Within 24 hours following valve actuation due to automatic or manual I action or flow through the valve.
l The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
*Tavg being changed ]y 8.9 '
an 5 F/ hour o MAR 131986 SEABROOK - UNIT 1 3/4 4-21 n dD' a bJ ' m rn N b .). ) l l
f] v TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES . 7 - t
. VALVE VALVE MAX. ALLOWABLE NUMBER SIZE FUNCTION -
LEAKAGE (GPM) SI-V144 1-1/2" SI to RCS Loop 1 Cold Leg Injection 0.75 SI-V148 1-1/2" SI to RCS Loop 2 Cold Leg Injection 0.75 SI-V152 1-1/2" SI to RCS Loop 3 Cold Leg Injection 0.75 SI-V156 1-1/2" SI to RCS Loop 4 Cold Leg Injection 0.75 e i SI-V81 2" SI to RCS Loop 3 Hot Leg Injection 1.0 SI-V86 2" SI to RCS Loop 2 Hot Leg Injection 1. 0 SI-V106 2" SI to RCS Loop 4 Not Leg Injection 1. 0 SI-V110 2" SI to RCS Loop 1 Hot Leg Injection 1.0 SI-V118 2" SI to RCS Loop 1 Cold Leg Injection 1. 0 SI-V122 2" SI to RCS Loop 2 Cold Leg Injection 1.0 SI-V126 2" SI to RCS Loop 3 Cold Leg Injection 1.0 SI-V130 2" SI to RCS Loop 4 Cold Leg Injection 1.0 SI-V140 3" SI to RCS Cold Leg Injection 1.5 SI-V82 6" SI to RCS Loop 3 Hot leg Injection 3.0 SI-V87 6" SI to RCS Loop 2 Hot Leg Injection 3.0 RH-V15 6" RHR to SI Loop 1 Cold Leg Injection 3.0 RH-V29 6" RHR to SI Loop 3 Cold Leg Injection 3.0 RH-V30 6" l I RHR to SI Loop 4 Cold Leg Injection 3.0 l RH-V31 6" RHR to SI Loop 2 Cold Leg Injection 3.0 RH-V52 6" ( SI to RCS Loop 1 Hot Leg Injection 3.0 ' RH-V53 6" SI to RCS Loop 4 Hot Leg Injection 3.0 RH-V50 8" RHR to RCS Loop 4 Hot Leg Injection 4.0 RH-VS1 8" RHR to RCS Loop 1 Hot Leg Injection - 4.0 SI-V5 10" SI to RCS Loop 1 Cold Leg Injection 5.0 SI-V6 10" SI Tank 9A Discharge Isolation 5.0 SI-V20 10" SI to RCS Loop 2 Cold Leg Injection 5.0 i SI-V21 10" SI Tank 9B Discharge Isolation 5.0 SI-V35 10" SI to RCS Loop 3 Cold Leg Injection 5.0 SI-V36 10" SI Tank 9C Discharge Isolation 5.0 SI-V50 10" SI to RCS Loop 4 Cold Leg Injection 5.0 SI-VS1 10" SI Tank 90 Discharge Isolation 5.0 RC-V22 12" RHR Pump 8A Suction Isolation 5.0 RC-V23 12" RHR Pump 8A Suction Isolation 5.0 RC-V87 12" RHR Pump 88 Suction Isolation 5.0 RC-V88 12" RHR Pump 88 Suction Isolation 5.0 SEABROOK - UNIT 1 3/4 4-22 ) ) / 13 S. .L L L
REACTOR COOLANT SYSTEM
. 3/4.4.7 CHEMISTRY -
LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2. APPLICABILITY: At ill times. AbTION: MODES 1, 2, 3, and 4:
- a. With any one or more chemistry parameter in excess of its Steady-State Limit but within its Transie.nt Limit, restore the parameter to within its Steady-State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and
- b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
At All Other Times: V With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to l 1ess than or equal to 500 psig, if applicable, and perform an engineering l evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remair.s acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4. SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters specified in Table 3.4-2 at least once per 72 hours.*
- Sample and analysis for dissolved oxygen is not required with T <, 50'F.
gg SEABROOX - UNIT 1 3/4 4-23 MAIL 13J866 A3.n 1. A 1
d. i TABLE 3.4-2
-0 -
REACTOR COOLANT SYSTEM CHEMISTRY LIMITS 1 STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen * < 0.10 ppm , 5 1.00 ppm Chloride < 0.1$ ppm $ 1.50 ppm Fluoride $ 0.15 ppm 1 1.50 ppm I .I
- O-It?O Limit not applicable with T, g less than or equal to 250*F.
d E 1 1 i l O M a taIsas l c_ _ - SEAEROOK - UNIT 1 3/4 4-24 l g 'th - . b . L);b! [g2a. d, E gr.w--m,,e-y < m ___..,,.,..,,m _
, , , , , _ , , _m._,,_,,, ,.,_ , . , _ _ _ _ . , , _ _ , ,,,, , ,_, _ , , , ._._, _ _y_, _ , _ _, ,.____ _ _ ,___ , , , _ , . . , _ _ , , _ _ _ , . y ..,_ ,
REACTOR COOLANT SYSTEM fS L) - 3/4.4.8 SPECIFIC ACTIVITY T . i LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shal,1 be limited to:
- a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and
- b. Less than or equal to 100/E microCuries per gram of gross radioactivity.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*:
- a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T**9 less than 500*F within 6 hours; and O b. wita_the ePecific ctivity or the reector cooi ot are ter ta -
100/E microcuries per gram, be in at least HOT STANDBY with T less than 500*F within 6 hours. **9 MCDES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than _ - 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro-Curies per g \, perform the sampling and analysis requirements of Item 4.a) of Table 4.4 until the specific activity of the reactor coolant is restored to hin its limits. SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the-limits by performance of the sampling and analysis program of Table 4 44 3
/
f" *With T,y greater than or equal to 500*F. ' ggjajl86-SEABROOK - UNIT 1 3/4 4-25 k D T
?
w\ A .. ._h w >
' . . i' 300 E . : .
N,, . . u - w x
- l U .
250- -** - - --- * - **
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.4 c.
3 o . . . o 200 .............. ...... ............ ................ ........;. .. -
- 6. .
o" . . . r* . ! UNACCEPTABLE
- i. .. OPERATION Z . . .
4 . . .
,3 330 ........,..... ..,........,.....
- . t........,.........,.................
C . . . . e . : : : o . . . . . . g . . . . . . Q =
,00 ..........................;........ . . . . :. .
ACCEPTABLE ! . : E OPERATION : w . . a . . . .
< s0 ........................... .......................
3, . . . .
- s- . : : : : : :
C - * * *
- y . . . . . .
m . . . . . m . C * - * *
- Q .
C. . , , . . , 20 30 40 SC 40 70 to 90 100 PERCENT OF RATED THERW AL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC
, ACTIVITY >l pC1/ gram DOSE EQUIVALENT I-131
.. MAR 131986 . _ . .
SEABROOK - UNIT 1 3/4 4-26 d
] t7].
> i'7y]j tti
.r d<i, ,
/ . + k. .i .1 d
9 _ _ . , _ _ , , - - _,,-%w f- p,,, , , , , , , , , . ,7- , - ,7 ,, -, , _ _ , , , , , . , . , _ , - . - - - , - , , - ,,-,
O O o-t . . TABLE 4.4-3
$ REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE E AND ANALYSIS PROGRAM 8
S SAMPLE AND ANALYSIS MODES IN MIICH SAMPLE TYPE OF MEASUREMENT ~ e AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED E At least once per 72 hours. 1, 2, 3, 4 k
- 1. Gross Radioactivity Z Oetermination g
" 1 per 14 days. 1 i 2. Isotopic Analysis for DOSE EQUIVA-LENT I-131 Concentration .
- 3. Radiochemical for E Determination
- 1 per 6 months ** 1
! 4. Isotopic Analysis for Iodine a) Once per 4 hours, 1#, 2#, 3#, 4#, S# j Including I-131, I-133, and I-13S , whenever the specific i activity exceeds 1 pCf/ gram DOSE EQUIVALENT I-131 7 or 100/E pC1/ gram of g gross radioactivity, and b) One sample between 2 1,2,3 f $ and 6 hours following r a THERMAL POWER change exceeding 15% of the RATED THERMAL ! POWER within a 1-hour I period. a M:Eq r ! . pm , p-d g
- ar i
v e. - I g ee.
TABLE 4.4-3 (Continued) - .O U . TABLE NOTATIONS 7
*A radiochemical analysis for 5 shall consist of the quantitative measurement of the specific activity for esch radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in the reactor coolant. The specific activities for these individual radio-nuclides shall be used in the determination of E for the reactor coolant sample. Determination of the contributors to 5 shall be based upon those energy peaks identifiable with a 95% confidence level.
**Samp'le to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.
#Until the specific activity of the Reactor Coolant System is restored within its limits.
O V og
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O - ma la isse SEABROOK - UNIT 1 3/4 4-28 ,- { .- ..) Ir', ,, s D y EM..
.14'.s.sd. .u. A j(,
REACTOR COOLANT SYSTEM O-- - 3/4.4.9 PRESSURE / TEMPERATURE LIMITS . r - REACTOR COOLANT SYSTEM
~
LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (exc'ept the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 100 F in any 1-hour period,
- b. A maximum cooldown of 100*F in any 1-hour period, and
- c. A maximum temperature change of less than or equal to 10*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T,yg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours. . SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required y*10 CFR Part 50, Appendix H, in accordance with the schedule an)d 3.4-3.in Figures 3.4- Table 4.4 (d The results of these examinations shall be used to update O _. MAa la issa
~
SEABROOK - UNIT 1 3/4 4-29
l i 1 O. COPPER CONTENT : CONS ER V ATIVELY A S S UM E D TO B E
~
' RT NDTINITI A L 0.10 W T % ( A CTU A L C O NTE NT= 0. 0 8 W T*4 )
- 40 F
~ !'I RTNOTA F T E R 18 E F P Y :1/4T.110 F 3/4T.57 F .
CUR VE APPLIC A B LE FO R H E ATUP R ATE S UP TO 8 0 F/HR FOR TH E SERVICE PEntOO UP TO 16 EFPY AND CONTAINS MARGINS OF to F Of wu AND 80 PSIG FOR POSSIB LE INSTRUMENT ERRORS l - k g LE A K'TE ST LI M IT * *
- l C. *
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- TEST O TEMPER ATU AE (2 5 5 F1 FOR H - -
TH E S E AVICE g *
. PEAICOUP 4: . : .
70 15 E P P T, W . C; ; h $ h h h 0.0 100.0 200.0 300.0 400.0 500.0 INDICATED TEMPERATURE (DEG. F) FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY O MAR 131986 SEABROOK - UNIT 1 3/4 4-30 yb $ 9
]Df ' Q N) fN L
. u x..u 2 a 4. .t.
i O v MATERIAL PROPERTY B ASIS .
-. CO PPER CONTENT
- CONSERVATIVELY ASSUMEO TO B E .
t* O 10 WT% (ACTU AL CONTENT =0 06 WT%) RTNOTIN ITI A L .40 F , RTNOTA FTER 16 EFPY .1/4T.110 F ~ 3/4T.87 F . CURVE APPLICABLE FOR COOLOOWN M ATES UP TO 100 F/bR FOP THE SERVICE PERICO UP TO 16 EFPY AND CONTAINC M ARGINS OF 10 F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS
. : : P 4 . .
o 1000 m
.............b. .
a. w w - l : gg . . . 2 : . m m ; - f' w .........................,.. ................................... u : o : .. : W
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- Da- .
100 . l i : : : : 4 . . , , 0 900 200 300 400 300 lNDICATED TEMPER ATURE (DEG. r) FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY MAR 131986 SEABROOK - UNIT 1 3/4 4-31 ;
- f'
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- 3 "" MAR 131986 O i
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SEABROOK - UNIT 1 3/4 4-32 : <
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REACTOR COOLANT SYSTEM
~
PRESSURIZER r - 1 LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup of 100*F in any 1-hour period,
- b. A maximum cooldown of 200"F in any 1-hour period,,and
- c. A maximum spray water temperature differential of 320*F.
APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. O SURVEILLANCE REOUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. SEABROOK - UNIT 1 3/4 4-33 . MAR 131986
REACTOR COOLANT SYSTEM 4 - OVERPRESSURE PROTECTION SYSTEMS .' r- -
, LIMITING CONDITION FOR OPERATION 3.4.9.3 At leas t one of the following Overpressure Protection Systems shall be OPERABLE: -
- a. Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig +_1%, or
- b. Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or
- c. The Reactor Coolant System (RCS) depressurized with an RCS vent of
; greater than or equal to M . square inches, j /58 APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than i or equal to 329*F, MODE 5 and MODE 6 with the reactor vessel head on.
ACTION:
- a. With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves or to OPERABLE status within 7 days or depressurize and vent the RCS through at
, least a M square inch vent within the next 8 hours. 189
- b. With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 'h2 square inch vent within 8 hours. /66
- c. In the event the PORVs, or the RHR suction relief valves, or the RCS
} vent (s) are used to mitigate an RCS pressure transient, a Special ~ Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 30 days. The report shall describe the l circumstances initiating the transient, the effect of the PORVs, or l the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable.
/
SEABROOK - UNIT 1 3/4 4-34 L, ,4 \ e H MAR 13 taas
; O ,
o , O e i N
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g 2500 , , , l x * ' '1 . s , VALID FOR THE FIRST 86 EFPY. SETPOINT c 5 CONTAINS MARGIN: OF ,50'F FOR
--e *g TRAN5IENT EFFECTS.
- g
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- RCS T EMPERATURE (*F )
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SURVEILLANCE REQUIREMENTS 4.4.9.3.1 'Each PORV shall be demonstrated OPERABLE by:
- a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required
. OPERABLE;
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and
- c. Verifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.
4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:
- a. For RHR suction relief valve RC-V89 O 1) By verifying at least once per 31 days that RHR RCS Suction Isolation Valve RC-V88 is open with power to the valve operator removed, and
- 2) By verifying at least once per 12 hours that RC-V87 is open.
- b. For RHR suction relief valve RC-V24
- 1) By verifying at least once p?r 31 days that RC-V22 is open with power to the valve operator removed, and
- 2) By verifying at least once per 12 hours that RC-V23 is open.
- c. Testing pursuant to Specification 4.0.5.
4.4.9.3.3 The RCS vent (s) shall be verifled to be open at least once per 12 hours
- when the vent (s) is being used for overpressure protection.
*Except when the v'ent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at O i *" car 318r-SEABROOK - UNIT 1 3/4 4-36 g i ,
y
! i;n 0 r- \ 3j
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i REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY . f r - 1, LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components i shall be maintained in accordance with Specification 4.4.10. APPLICABILITY: All MODES. ACTION:
- a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant j System temperature more than 50*F above the minimum temperature i required by NOT considerations.
- b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
- c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service,
- d. The provisions of Specification 3.0.4 are not applicabl,e. -
l l i SURVEILLANCE REQUIREMENTS l 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. O _ . 3/4 4-37 m
- p h Ks ,W%
SEABROOK - UNIT 1 ' 1l /7g h 19 m A . A $L A . l
f~ t REACTOR COOLANT SYSTEM O. , l 3/4.4.11 REACTOR COOLANT SYSTEM VENTS .- t l LIMITING CONDITION FOR OPERATION 3.4.11 At least one Reactor Coolant System vent path consisting of one vent . valve and one block valve powered from emergency busses shall be OPERABLE l and , closed at each of the following locations: ihe geht udve T
- a. Reactor vessel head, and ',
i b. Pressurizer steam space, 4nd. f l APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves and block valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN with'in the following 30 hours.
O b. With both Reactor Coolant System vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuators of all the vent valves and block valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours or be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. , SURVEILLANCE REQUIREMENTS 4.4.11.1 Each Reactor Coolant System vent path block valve not required to be closed by ACTION a. or b., above, shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel from the control room. 4.4.11.2 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying all manual isolation valves in each vent path are locked in the open position,
- b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and l c. Verifying flow through the Reactor Coolant System vent paths during
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SEABROOK - UNIT 1 3/4 4-38 "h Lb)~7) . ABm p.
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3/4.5 EMERGENCY CORE COOLING SYSTEMS O. 2'4.s.1 Acco u'Aroas HOT STANDBY, STARTUP AND POWER OPERATION . LIMITING CONDITION FOR OPERATION 3.5.1.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:
- a. The isolation valve open, '
4/CO ddOO
- b. A contained borated water v@lume of between.8640 and.9306 gallons, 19tc 2/00
- c. A boron concentration of between 2000-and 2600 ppm, 477
- d. A nitrogen cover-pressure of between 600 and:680 psig,and:
APPLICABILITY: MODES 1, 2, and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a closed isola-tion valve, p store the inoperable accumulator to OPERABLE status g within-7,houg or be in at least HOT STANDBY within the next 6 hours
, and reduce pressurizer pressure to less than 1000 psig within the following 6 hours. O b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least H01 STANOBY within 6 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per hours by:
- 1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and
- 2) Verifying that each accumulator isolation valve is open.
a b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution; and
" Pressurizer pressure above 1000 psig.
SEABRC0K - UNIT 1 3/4 5-1 h 3
- a o I
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i MAR 131986 I
p EMERGENCY CORE COOLING SYSTEMS f SURVEILLANCE REQUIREMENTS (Continued) - -
~
cf At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator.is disconnected, dy-removah.of-the-breaker--from-the circuitpande Y. At least once per 18 months by verifying that each accumulator isola-c tion valve opens autcoatically under each of the following conditions:
- 1) When an actual or a simulated RCS pressure' signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint, and
- 2) Upon receipt of a Safety Injection test signal.
4.5.1.2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE:
- a. At least once per 31 days by the performance of an ANALOC CilANNEL OPERATIONAL TEST, and
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
({\ -
- O
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O ' ( SEABROOK - UNIT 1 3/4 5-2 1
EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS .~ r -
. LIMITING CONDITION FOR OPERATION 3.5.1.2 Each reactor coolant system accumulator isolation valve shall be shut with power removed from the valve operator.
APPLICABILITY: MODES 4* and 5 ACTION:
- a. With one or more accumulator isolation valve (s) open and/or power available to the valve operator (s), immediately close the accumulator isolation valves and/or remove power from the valve operator (s).
- b. The provisions of Specification 3.0.4 are not applicable for entry into MODE 4 from MODE 3.
SURVEILLANCE REQUIREMENTS 4.5.1.2 Each accumulator isolation valves will be verified shut with power removed from the valve operator at least once per 31 days. O
*Within 12 hours prior to entry into MODE 3 from MODE 4 and if pressurizer pressure >1000 psig each accumulator isolation valve shall be open as required by Specification 3.5.1.1.a.
m.. .. NR 13Isog O , SEABROOK - UNIT 1 3/45'z) .
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EMERGENCY CORE COOLING SYSTEMS O- 3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350*F . 1 lSl40TDOWNM LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERA 8LE Safety Injection pump,
- c. One OPERABLE RHR heat exchanger,
- d. One OPERABLE RHR pump, and
- e. An OPERABLE flow path
- capable of taking suction from the refueling ~
water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation. APPLICABILITY: MODES 1, 2, and 3 P ACTION:
- a. With one ECCS subsystem inoperablo restore the inoperable subsystem to OPERABLE status within 72-hourn or be in at least HOT STANDBY within the next 6 hours and in HOT SHUT 00WN within the following 6 hours. *
- b. In the event the ECCS is actuated and injects Coolant System, a Special Report shall be submitted ared and pare) water to into the R the Commission pursuant to Specification 6.E 2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
*During MODE 3, the discharge paths of both Safety Injection pumps may be isolated by closing for a period of up to 2 hours to perform surveillance testing as required by Specification 4.4.6.2.2.
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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS i 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per JS hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position 6-SI-V-3 Accumulator Isolation Open SI-V-17 Accumulator Isolation Open SI-V-32 Accumulator Isolation Open SI-V-47 Accumulator Isolation Open CBS-V-47 RWST-/SI-Pump-Isolat-ion Open-CBS-V-51 -RWST/SI-PumpJ sola Qon Open
~
SI-V-114 SI Pump to Cold Leg Isolation Open RH-V-14 RHR Pump to Cold Leg Isolation Open RH-V-26 RHR Pump to Cold Leg Isolation Open RH-V-32 RHR to Hot Leg Isolation Closed RH-V-70 RHR to Hot leg Isolation Closed SI-V-77 SI to Hot Leg Isolation Closed SI-V-102 SI to Hot Leg Isolation Closed
- b. At least once per 31 days by:
- 1) Verifying that the ECCS piping is full nf water by venting the ECCS pump casings and accessible discharge piping high points, and
- 2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall l be performed:
- 1) For all accessible areas of the containment prior to establish-ing PRIMARY CONTAINMENT INTEGRITY, and
- 2) Of.. the areas af fected within containment at the completion of each containment entry when PRIMARY CONTAINMENT INTEGRITY is established.
Q SEABROOK - UNIT 1 3/4 5-5 {P
.L . 4 i
.gg
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) - r - i
- d. At least once per 18 months by:
- 1) Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by ensuring that; a) With a simulated or a'ctual Reactor Coolant System pressure signal greater th prevent the valve'an or equal s from to 365 psig being opened, and the interlocks b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 660 psig the interlocks will cause the valves to automatically close.
- 2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion,
- e. At least once per.18 months, during shutdown, by:
- 1) Verifying tha*t each automatic valve in the flow path actuates to its correct position on (Safety Injection actuation and O ^#ta tic s itc"ever te c=#t i eet se a) test is" i #d
- 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:
a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.
- f. By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:
- 1) Centrifugal charging pump 1 2480 psid,
- 2) Safety Injection pump 1 1445 psid, and
- 3) RHR pump 2.183 psid.
1%
- g. By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:
- 1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and SEABROOK - UNIT 1 3/4 5-6 (* ?
s .. . . a. . :. 1 A MMlaIH6 -
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) . t
- 2) At least once per 18 months. -
High Head SI System Intermediate Head SI System Valve Number Valve Number SI-V-143 SI-V-80 P SI-V-147 SI-V-85 SI-V-151 SI-V-104 SI-V-155 SI-V-109 SI-V-117 SI-V-121 SI-V-125 SI-V-129
- h. By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
- 1) For centrifugal charging pump lines, with a single pump running:
a) The sum of the injection Ifne flow rates, excluding the highest flow rate, is greater than or equal to 337 gpm, O and b) The total pump flow rate is less than or equal to 550 gpm.
- 2) For Safety Injection pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 445 gpm, and b) The total pump flow rate is less than or equal to 660 gpm.
- 3) For RHR pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 2828 gpm.
O SEA 8 ROOK - UNIT 1 3/4 5-7 D [9 - i O 2.t h .l h 4 Isn
4 1 EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS. SUBSYSTEMS - T,yg LESS THAN 350*F .
' LIMITING CONDITION FOR OPERATION i
3.5.3 1As a minimum, one ECCS subsystem comprised of the following shall be OP ERABLE: - l a. One OPERABLE centrifugal charging pumph
- b. One OPERABLE RHR heat exchanger, l c. One OPERABLE RHR pump, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4. . ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of
. O ~ either the centrifugal charging pump or the flow path from the ! refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours.
i b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR' pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reac-tor Coolant System T**9 1ess than 350*F by use of alternate heat 4 removal methods.
- c. In the event the ECCS is actuated and injects water into the Reactor l Coolant System, a Special Report shall be prepared and submitted to a the Commission pursuant to Specification 6.8.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for
- each affected Safety Injection nozzle shall be provided in this i
Special Report whenever its value exceeds 0.70. i
- A maximum.of- one. centrifugal-charging pump.and one-Safety-Injection-pump-
.shall-be-OPERABLE whenever the' temperature-of-one or-more.of the RCS cold-legs-is-less-than or-. equal _to.305?F,-
m.g . j SEABROOK - UNIT 1 3/4 5-8 h I'
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EMERGENCY CORE COOLING SYSTEMS ~ O.' SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE.per the applicable I requirements of Specification 4.5.2.
-t-5.-3r2-All char ingyumps -an gretyynje,cj; on pupp ,-gr3ept- he abov,e -
ALIWe'd Q2(RABL'E umpshshall ' demo'nspr.ated nopera)birby v r fpigg'yhah-
--1;hepo r-circdi bre,akeps e- -ar/cgred b, yn th p
--d hdu f whene've h'd_ temp'e'ra'turbof4ne/or-more,en of-the RCS positforCat cold Megseast onSe is less ~ per -
-than ' dual-to-305*F.
4.5.3.2 All centrifugal charging pumps and Safety Injection pumps, except the above allowed OPERABLE pumps, shall be demonstrated inoperable
- by verifying that the motor circuit breakers are secured in the open position within 4 hours af ter entering MODE 4 from MODE 3 or prior to the temperature of one or more of N
the RCS cold legs decreasing below 325 F whichever comes first, and at least once per 31 days therecfter. O
*An inoperable pump may be energized for testing or for filling accumulators provided the discharge of the pump has been isolated from the RCS by a closed b
isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position. O SEA 8 ROOK - UNIT 1 3/4 5-9 9
g 2 Cf' lLL edbN EMERGENCY CORE COOLING SYSTEMS n [e_o_bu. -
- ECCS SUBSYSTEMS - T,yg <,200 F LIMITING CONDITION FOR OPERATION
- 3. 1T~ afety Injection pumps shall e inoperable. ,,
APPLICABILITY: MODE 5 and MODE 6 with the reactor vessel head on. ACTION: With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours, s , SURVEILLANCE REQUIREMENTS 9 A)
.5.4 All Safety Injection pumps shall be demonstrated inoperable
- by O ve Hying that the motor circuit breakers are secured in the open position at least once per 31 days.
*An inoperable pump may be energized for testing or for filling accumulators O provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
L 'dhreo Q - UNIT 1 3/4 5-% 10
BORON INJECTION SYSTEM O. - 3/4.5.4 REFUELING WATER STORAGE TANK - I LIMITING CONDITION FOR OPERATION l 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:
*/77, ed \
l a. A minimum contained borated, water volume of 479,000 gallons, I
- b. A minimum boren concentration of 2000 ppm of boron, 60
- c. A minimum solution temperature of 35'F, and
- d. A maximum solution temperature of *F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. O SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:
- a. At least ence per 7 days by:
- 1) Verifying the contained borated water volume in the tank, and
- 2) Verifying the baron concentration of the water,
- b. At least once per 24 hours by verifying the RWST tempeiature.
W - - - _
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O r., 4 h SEABR00Y, - UNIT 1 3/45-}0, - je p ~ ~ - MM 13 \888
p 3/4.6 CONTAINMENT SYSTEMS V - 3/4.6.1 PRIMARY CONTAINMENT
. s CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3, and 4.
- ACTIONf' g Without(primary > CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour or be in at least NOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
- a. At least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves l d i secured in their posi j Specification 3.6.3),tionsexceptasprovidedinTable3.6-2of
- b. By verifying that each containment air lock is in compliance with
, the requirements of Specification 3.6.1.3; and
- c. After each closing of each penetration subject to Type B testing,
, except the containment air locks, if opened following a' Type A or B test, by leak rate testing the seal with gas at a pressure not less qvl thanP,,j6r8psig,andverifyingthatwhenthemeasuredleakagerate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L,.
*Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.
SEABROOK - UNIT 1 3/4 6-1 1 k MAR 131986
O. costatusen' svsress CONTAINMENT LEAKAGE -
!l l
8 LIMITING CONDITION FOR OPERATION _ [ 3. 6.1. 2 Containment leakage rates shall be limited to: l l a. An overall integrated leakage rate of: Less than or equal to L,, 0.15% by weight of the containment air per 24 hours at P,,-46-8 psig; il$.9 b. A combined leakage rate of less than 0.60 L, for all penetrations I and valves subject to Type 8 and C tests, when pressurized to P,. No individual penetration will be allowed to exceed 5% of the total allowed (0.05 L,); and c. A combined leakage rate of less than or equal to 0.60 L, for all penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P,.
]
/ APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: With (a) the measured overall integrated containment leakage rate exceeding 0.75L,,or(b)themeasuredcombinedleakagerategrallpenetrationsand valvessubjecttoTypesBandCtestsexceedingh.-75L,,or(c)thecombined bypass leakage rate exceeding 0.60 L,, restore the overall integrated leakage rate to less than or equal to 0.75 L, as-applicablea the combined leakage rate for all penetrations and valves subject to Type B and C tests to less
- 0. foo than &75' L,, and the combined bypass leakage rate to less than 0.60 L, prior to increasing the Reactor Coolant System temperature above 200 F.
{ l SEABROOK - UNIT 1 3/4 6-2 1 1 1 l __ MAR 14Igas
CONTAINMENT SYSTEMS SURVEILLANCERE3UIREMENTS ! 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI / N45.4 1972[: .
- a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 + 10 month intervals during shutdown at a pressure not less than-either P,, 46:8 psig, during each 10 year service period. The third test of e c set shall be conducted during the shutdown for the 10 year plant inservice inspection;
- b. If any periodic Type A fails to meet 0.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet either-0.75 L,, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either-0.75 L, at which time the above test schedule may be resumed; O c. The ecc=recx er eec" Trae a te t "eii ee ver<ried bv test which:
>#aaie #tei
- 1) Confirms the accuracy of the test by verifying that the supple-mental test result, L , cminus the sum of the Type A and the super imposed leak, L,, is equal 4;nto-or-les tr-than- 0. 25 L,;
- 2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and
- 3) Requires that the rate at which gas is injected into the contaln-ment or bled from the containment during the supplemental test is between 0.75 L, and 1.25 L,.
- d. Type B and C tests shall be conducted with gas at a pressure not less than P,,46g psig, at intervals no greater than 24 months except for tesN involving:
- 1) Air locks,
- 2) Purge supply and exhaust isolation valves with resilient material seals, G) - Penetrations-using-continuous-Leakage-Monitoring-Systems, -
--a nd -
SEABROOK - UNIT 1 3/4 6-3 l[ hbl /7 L' dr dt- ) 1 .1- i e
~ ~ ~
1
~ Mall 131986"
3 (V - CONTAINMENT SYSTEMS ' r - SURVEILLANCE REQUIREMENTS (Continued) 4)---Valves pre's'surized with-fluid-from-a-seal system.
- e. The combined bypass leakage rate shall be determined to be less than or equal to 0.60 L, by applicable Type B and C tests at least o,nce per 24 months,except-for penetrations which-are-not-Individually-testable; peitetrations not individually-testable-shall-be determined
-to have-no-detectable -leakage-when-tested-with-soap-bubbles while the-containment-is pressurized-to-P,;-46:8 psig;--during-each-Type-A-test;
- f. Purge supply and exhaust isolation valves with resilent material seals shall be tested and demonstrated OPERABLE by the requirements of Speci-fication 4.6.1.7.3 or 4.6.1.7.4, as applicable;
- g. Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3; 4i. Leakage-f rom-isolation valves-that-are-sealed-with-fluid ~from a seal-
-system-may-be-excluded,- subject _to_the_ provision of-Appendix,.1, Section-III-Cr3,-when determining-the-combined-leakage-rate-provided-s __the-seal-system and-valves-are-pressurized to at- least-1 10-P
-50r7-psig;--and-the seal-system-capacity-is-adequate-to maintaiIi-system --
pressure for-at-least-30-days ;~
-1. Type-B-tests-for-penetrations-employing-a-continuous-Leakage M6Htoring System-shall-be con' ducted at-Pv46:8 psigT at intervais no-greater than-once per-3 yearsr and-The provisions of Specifications 4.0.2'are not applicable.
o , SEABROOK - UNIT 1 3/4 6-4 I af()1dy y l
}
MAR 131986 1
j
,_ TABLE 3.6-1 , ,
SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS PENETRATION - NO. FUNCTION RELEASE LOCATION
)(-16 Containment On-Line Purge '
Primary Auxiliary Building (Exhaust) Equipment Vent (RCDT) Waste Processing Building '
%- 17
%-18 Containment On-Line Purge Primary Auxiliary Building (Supply)
[-19 Post Accident Monitoring Primary Auxiliary Building , Sample [-20 PCCW Loop A (Supply) Primary Auxiliary Building [- 21 PCCW Loop A (Return) Primary Auxiliary Building j-22 PCCW Loop B (Return) Primary Auxiliary Building [-'23 PCCW Loop B (Supply) Primary Auxiliary Building l O.
-32 Equipment and Floor Drainage Waste Processing Building (RCDT)
[- 34 Equipment and Floor Drainage Waste Processing Building i (RC Sump) l In')e cfIcr.
~
i K-35A Safety injector (Test Line) Waste Process,ing Building
%-358 Reactor Coolant (Pressurizer Primary Auxiliary Building Steam / Liquid Sample)
)
)(-35C Reactor Coolant Primary Auxiliary Building (RC Sa:nple Loop I) <
%-35D Reactor Coolant Primary Auxiliary Building (RC Sample Loop III) h36A Demineralized Water Demineralized Water Storage Tank (Outside)
N-36B Nitrogen Gas (High Pressure) Primary Auxiliary Building p36C Reactor Makeup Water Waste Processing Building (Tank Farm)
/
[-37A Chemical and Volume Control Primary Auxiliary Building (Letdown) SEABROOK - UNIT 1 3/4 6-5 E.
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MM 131986
O - TABLE 3.6-1 - T -
~
SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS (continued)
~
PENETRATION NO. FUNCTION RELEASE LOCATION
)(-37B Chemical and Volume Control Primary Auxiliary Building (Excess Letdown)
X-3876A Fire Protection Fire Water umphouse/ Fire Water Tanks
)(-38/768 Combustible Gas Control Bldn 6dMm a.ad Feedvder Pip e C k se-K-39 Spent Fuel Pool Cooling and Fuel Storage Building Cleanup
[-40A Nitrogen Gas (Low Pressure) Primary Auxiliary Butiding K- 408 PRT Sample Primary Auxiliary Building A ~ (o.2.
%-67 puel Trxnsfer Tak Service Air n.I % sse y eu;) din s Main Steam'and Feedwster Pipe Chase f-710/74D Lead Detection /Yld et Stea M lLA Y F'*d"A e r Pi p e C.ha.Sc.
HVAC-1 Containment Air Purge Primary Auxiliary Building HVAC-2 Containment Air Purge Primary Auxiliary Building Efa:pment tideh Ohfside
- Pe,g ,,, nc j t/ del; ,%: n Steam u d }1edader P;te CJase O _- MAR 13 Isss ---
f\ SEABROOK - UNIT 1 3/4 6-6 . ( _'
'l" Y
g
. CONTAINMENT SYSTEMS .
1, CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with: e
- a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 L
- at P,, 46-8 psig.
VS1 APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one containment air lock door inoperable:
- 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed,
- 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days, .
- 3. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hcurs, and
- 4. The provisions of Specification 3.0.4 are not applicable.
- b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
an ; n oferable V Except d&;9 entry to refu r L in n er door- MAR 13 lg88 SEASROOK - UNIT 1 3/4 6-7 Da k f4) i i N LA t,1 E
- /n .. .s. 1 : :. >t
O . CONTAINMENT SYSTEMS - r - l' SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
- a. Within 72 hours following each' closing, except when the air lock is
- g. being used for multiple entries, then at least once per 72 hours, by verifying that the seal leakage is less than 0.01 L, as determined by precision flow measurements when measured for at least 30 seconds within the volume between the seal at a constant pressure of 46;8~psig; y3 ?
- b. By conducting overall air lock leakage tests at not less than P ,
psig, and verifying the overall air lock leakage rate is w thin its limit:
- 1)
- and Atleastonceper6 months,7 2)
FqaR Prior to establishing CONTAINMENT INTEGRITY when maintenance has been perfortned on the air lock that could affect the air lo-k sealing, capability.**
- c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
*The provisions of Specification 4.0.2 are not applicable.
**ThisrepresentsanexemptiontoAppendixJ,paragraphIII.D.W(j' of 10 CFR Part 50. lh86 SEABRDOK - UNIT 1 3/4 6-8 g-4: pq7q
.7- ...
,k a' . --- .- u
l o CONTAINMENT SYSTEMS
. 1 INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between 14.6Tand 16.2 psia.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within I hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. O SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours. 1 y ylla ;s no lowa I:m;E v% Ue e,nhin., ant MAR 13 1986 O- u -1:n s u m ,,- > - ,. ,
,] ,U ' - .
Accordnce < p,,ith w $ jee;p,. e4ys,5,6,f,7,f,' l '. j} f k{ SEABROOK - UNIT 1 3/4 6-9 .d m a [- ! N
.. . - . _ _ _ . . _ - - u
l l
)
O CONTAINMENT SYSTEMS
- V .
AIR TEMPERATURE ~ 5 LIMITING CONDITION FOR OPERATION - 3.6.1.5 Primary containment average air temperature shall not exceed 120 *F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment average air temperature greater than 120 F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ( SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at the following locations and shall be determined at least once per 24 hours: . , Location *
- a. Elevation 45 feet
- b. Elevation 71 feet '
- c. Elevation 110 feet
- d. Elevation 130 feet
- . MAR 13.1986 O D '."') ;\ ri4 U. .n. i i i fLn!!-I A t4 SEABROOK - UNIT 1 3/4 6-10 - _ _ . _ . ._____
CONTAINMENT SYSTEMS - r - - ; CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specific 5 tion 4.6.1.6. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined p/ s during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degrada-tion of the containment vessel detected during the above required inspections shall be reported to the Commission in a Special Report pursuant to Specifi-cation 6.C.2 within 15 days. - MAR 131986 7'"I 6 i /O 3/4 6-11 D'n" s. JA SEABROOK - UNIT 1 M a,g k A4 A w
O , CONTAINMENT SYSTEMS - e
- l CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and: 6'
- a. Each 36-inch containment shutdown purge supply and exhaust isolation valve shall be closed and locked closed, and
- b. The 8-inch containment purge supply and exhaust isolation valve (s) may-be-open for up to-.1000 hours-duri6,g-a calendar _ year _provided-no-mo re-tha n-one -pa i r-(o n e-s up p ly-a nd-one-ex ha u s t-)-a re -op e n - a t-one -
-t-ime; S}pJJ be Sexle d clos e.d io 6 he- ina t} o'n am o tenf yr.te f.*ce h/e bu t u.y be open f r pu ro #e sys k ogemt;o" for ^;" con t roi, fo r APPLICABILITY: MODES 1, 2, 3, and 4. Fo r Me rso ##^ A/* f r ess a r<.
" #en#85/'
a l es fry e,, g p,,f'4garye;jja
/M)' C* an G idt fier.
e ,_ ACTIGN: b*s s M t regulrs l-h e ulve. to be open.
- a. With a 36-inch containment purge supply and/or exhaust isolation p/ valve open or not locked closed, close and/or lock close that valve
- s. or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With the 8-inch).$ containment purge supply and/or exhaust isolation y g/gr g j n within4 hours,otherwisebe'i}natleastHOTSTANDB close the open 8-inch *. valve'(s or isolate the penetration (s) 4 6 hours, and in COLD SHUTDOWN within the following 30 hours.
- c. With a contair. ment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifica-tions 4.6.1.7.3 and/or 4.6.1.7.4, restore the inoperable valve (s) to OPERABLE status \within 24 hours, otherwise be in at least HOT STANDBY within the nex 6 hours, and in COLD 5HUTDOWN within the following 30 hours.
,,. ;s, Me g-he. puated'<n6d suh thd ik incasa. red lea)<'y e ca.de cloes no f e y ee e.d 8 h e- lo'NIis ICf C E R e a b s'e n t/. 6. l. 9. 3 a.ndl'sr t/.6.l.2.hfyj3lggg f
- M L SEABROOK - UNIT 1 3/4 6-12 J d A .L ,
)R 1 W g
i . l l 7_s
, CONTAINMENT SYSTEMS -
{ SURVEILLANCE REQUIREMENTS 4.6.1.7.1 The 36-inch containment purge supply and exhaust isolation valves shall be verified to be locked closed and, closed at least once per 31 days.
'4.6.1.7.2fhe-cumulative-time-that-the 8-inch purge = supply:and: exhaust-isola--
1on-valves *haVFbeen'open-during a calendar-year-shall=be determined-at-least~ once-per-7-days: 4.6.1.7.3 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard valves with resilient material seals in each locked closed 36-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L
. when pressurized to
- P,.
92 days 4.6.1.7.4 At least once per months each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to
.D<01 L when pressurized to P,.
0.05 , U,, fach S-inc it codaiaen t f u y supply u d ex W' t isold so n vxtre shtl be. v e r; fr e d i s se ejosed ud /ocA'ed Clc5ed in &cco rduc e u; H, S p ; p ,. g .,.o , 3 , g , ,, y, g yd once. fer si da.yg, 1
$h80 SEABROOK - UNIT 1 3/4 6-13 .M
~ n3. hx .a\ if"i
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a G x
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS O-
- CONTAINMENT SPRAY SYSTEM r .
LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and transferring suction to the containment sump. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: q g yg With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72-hours--or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to OPERA 3LE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in-position is in its correct position;
- b. By verifying, that on recirculation flow, each pump ddvelops a dicmed;4! discharge pressure of greater than or equal to M B psig'when tested pursuant to Specification 4.0.5; d
- c. At least once per 18 months during shutdown, by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal, and
- 2) Verifying that each spray pump starts automatically on a Containment Pressure-Hi-3 test signal.
- d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
O * '3 1986
;1 9,k t 'q rn SEABROOK - UNIT 1 3/4 6-14 1 N ,
~x CONTAINMENT SYSTEMS (O
SPRAY ADDITIVE SV5 TEM ! LIMITING CONDITION FOR OPERATION - 3.6.2.2 The Spray Additive System shall be OPERABLE with:
- a. A spray additive tank containing a volume of between 9500 and 6 9750 gallons of between 19 and 21% by weight NaOH solution, and
- b. Two gravity feed paths each capable of adding NaOH solution from the chemical additive tank to the Refueling Water Storage Tank.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the Spray Additive System inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours s or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position;
- b. At least once per 6 months by:
- 1) Verifying the contained solution volume in the tank, and
- 2) Verifying the concentration of the NaOH solution by chemical analysis.
- c. At least once per 18 months, during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal lyapd MAR 131986 O _ _. -
SEABROOK - UNIT 1 3/4 6-15 E D *R AA F m J
CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves specified in Table 3.6-2 shall be OPERABLE with isolation times as shown in Table 3.6-2. P APPLICABILITY: MODES 1, 2, 3, and 4. , ACTION: With one or more of the isolation valve (s) specified in Table 3.6-2 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
- b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or
- c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange; or
- d. Be in at least HOT STAN35Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS 4.6.3.1 The isolation valves specified in Table 3.6-2 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair, or replacement work is performed on the valve or its associated actuator, control, or power circuit by performance of a cycling test and verification of isolation time. MAR 131986 0 . _ .. Tr% D % in ry A. 1 k SEABROOK - UNIT 1 3/4 6-16
CONTAINMENT SYSTEMS F - i SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTOOWN or REFUELING MODE at least once per 18 months by: .
- a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position,
- b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position, and
- c. Verifying that on a Containment Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation position.
4.6.3.3 The isolation time of each power-operated or automatic valve of Table 3.6-2 shall be determined to be within its limit when tested pursuant to Specification 4.0.5. O I
- O ___hfAL13.Iggg n v,m e- ra.,.s rj n
< n D d>. ir r,. l i li' A
-~
SEABROOK - UNIT 1 3/4 6-17 ---
7 x TA8LE 3.6-2 L/ - CONTAINMENT ISOLATION VALVES , A. PHASE "A" ISOLATION MAXIMUM ISOLATION TIME VALVE NUM8ER FUNCTION (Seconds) CAH-FV6572 Radiation Monitorjng Skid 60 Inlet 1 CAH-FV6573 Radiation Monitoring Skid 60 Inlet 1 CAH-FV6574 Radiation Monitoring Skid 60 Outlet 1 [nules W 9 Exhaust Filter Isolation CGC-V14*m Containment 8
... Er.ch4?
"c CGC-V28*f Containment Exhaust Filter Isolation 8 CS-V149 Reactor Coolant Letdown 10 CS-V150 Reactor Coolant Letdown 10 CS-V167 Gxcess Leid* fd""
RCP Seal Water Return - 10
&$ces s [e fa r o' CS-V168 10 RCP Seal WaterJReturn- f. e E d" '~
NG-FV4609 Nitrogen Gas Supply 1 NG-FV4610 Nitrogen Gas Supply 1 NG-V13 Accumulator Nitrogen Supply 10 NG-V14 Accumulator Nitrogen Supply 10 RC-FV2830 PZR Steam Scr.ple 1 RC-FV2831 PZR Liquid Sample 1 RC-FV2832f RCS Loop 1 Sample 1 i RC-FV2833) RCS Loop 3 Sample 1 RC-FV2836 PZR Relief Tank Gas Sample 1 RC-FV2837 PZR Relief Tank Gas Sample 1 RC-FV2840 PZR Steam / Liquid Sample 1 RC-FV2874,f ' RCS Loop 1 Sample 1 si ( RC-FV2876* RCS Loop 3 Sample Qfgj3I880 1 g luhP1 a 1
"N
(,) TABLE 3.6-2
-- CONTAINMENT ISOLATION VALVES .
(continued) A. PHASE "A" ISOLATION , MAXIHUM ISOLATION TIME VALVE NUMBER FUNCTION (Seconds) RH-V27# RHR Test Line 10 c RH-V28# RHR Test Line 10 RH-V49# RHR Test Line 10 RMW-V30 Reactor Makeup Water 10 SB-V9# SG Blowdown 10 SB-V10# SG Blowdown 10 SB-V11# SG Blowdown 10 SB-V12# SG Blo.wdown 10 O', SI-V62 Accumulator Fill and Test Line 10 SI-V70 Accumulator Fill and Test Line 10 SI-V131# SI Test Line 10 SI-V134# SI Test Line . 10 SI-V157g/ Accumulator Fill and Test Line 10 SI-V158# SI Test Line 10 SI-V160# SI Test Line 10 SS-FV2857 P+sst SampleAccident Flush Tank Orain 1 VG-FV1661 Hydrogenated Equipment Vent Header 1 VG-FV1712 Hydrogenated Equipment Vent Header 1 WLD-FV8330 Containment Floor Drains 1 WLD-FV8331 Containment Floor Drains 1 WLD-V81 Reactor Coolant Drain Tank 10 WLD-V82 Reactor Coolant Drain Tank 10 MAR 131986 - [h [9 ,b Ei ,9 SEABROOK - UNIT 1 3/4 6-19 8 :
.O E't .L li, b i_b iff j
. Li,
TABLE 3.6-2 )
- CONTAINMENT ISOLATION VALVES '
(continued)
- 8. PHASE "B" ISOLATICN -
MAXIMUM ISOLATION TIME VALVE NUMBER FUNCTION (Seconds) CC-V57 PCCW Loop A Supply 10 e CC-V121 PCCW Loop A Return 10 CC-V122 PCCW Loop A Return 10 CC-V168 PCCW Loop A Supply 10 CC-V175 PCCW Loop B Supply 10 CC-V176 PCCW Loop B Supply 10 CC-V256 PCCW Loop B Return 10 CC-V257 PCCW Loop B Return 10 L'~ GC-V1092# RCP-Thermal-Barrier-Cooling . 3 0____ RCP Thermal Barrier Co~oling--
-CC-V1095# --- - 2 0___._
CC-V1101# - RCP-Thermal-Barri e r-Cool i ng -30 tC-V1109# 7CP TheFm~al Barrier-Cooling 30 - C. CONTAINMENT PURGE AND EXHAUST MAXIMUM ISOLATION TIME VALVE NUMBER FUNCTION (Seconds) CAP-VI Containment Refueling Purge & Exhaust 5 CAP-V2 Containment Refueling Purge & Exhaust 5 CAP-V3 Containment Refueling Purge & Exhaust 5 CAP-V4 Containment Refueling Purge & Exhaust 5 COP-V1 Containment On-Line Purge ,2'8 COP-V2 Containment On-Line Purge ,2' COP-V3 Containment On-Line Purge ,25 COP-V4 Containment On-Line Purge ~M$k }3l$gg # 5 l ._..____ SEABR00K - UNIT 1 3/4 6-20 " ' I- '^
; f 7,~i bi i
TABLE 3.6-2 (J3 CONTAINMENT ISOLATION VALVES I (continued) D. MANUAL MAXIMUM ISOLATION TIME VALVE NUMBER FUNCTION (Seconds) CGC-V3#* Hydrogen Analyzer 00tlet NA e CGC-V10#* Hydrogen Analyzer Inlet NA CGC-V15* Containment Exhaust Filter ORC Isolation NA CGC-V24#* Hydrogen Analyzer Outlet NA CGC-V32#* Hydrogen Analyzer Inlet NA CGC-V36* Containment Exhaust Filter ORC Isolation NA CGC-V43* Compressed Air Supply to Containment NA CGC-V44* Compressed Air Supply to Containment NA q b CGC-V45 Portable Air Compressor Connection NA DM-V4 Demineralized Water Supply NA DM-V5 Demineralized Water Supply NA FP-V5927 Containment Fire Protection Feader , NA SA-V229 Containment Service Air NA SA-V1042 Containment Service Air NA SF-V86 Refueling Cavity Cleanup NA SF-V87 Refueling Cavity Cleanup NA LD-V1 Leak Detection NA LD-V2 Leak Detection NA E. OTHER MAXIMUM ISOLATION TIME VALVE NUMBER , FUNCTION (Seconds) O CAN-v12 Redietion Monitoring Sxid 60 IRC Caecx NA l CC-V410 PCCW Loop A Return Relief NA SEABROOK - UNIT 1 3/4 6-21 MAR.131ssa
'i j /k Qi - [
d Its. fA 11 i
TABLE 3.6-2
. CONTAINMENT ISOLATION VALVES F (continued)
E. OTHER (CONT'D)
- MAXIMUM ISOLATION TIME VALVE NUMBER FUNCTION (Seconds)
CC-V474 PCCW Loop B Return R'elief NA CC-V840 PCCW Loop B Supply Relief NA CC-V845 PCCW Loop A Supply Relief NA CGC-V4# Hydrogen Analyzer Outlet IRC Check NA CGC-V25# Hydrogen Analyzer Outlet IRC Check NA CGC-V46 Compres Air Supply IRC Check NA CS-V794 /E xcess Letdown Return Check RCP Seal'# NA DM-V18 Containment Demineralized Water Supply NA Relief O FF-V588 Containment Fire Protection Header NA IRC Check RC-V312 Pressurizer Sample Relief NA RC-V314 RCS Loop 1 Sample Relief NA RC-V337 RCS Loop 3 Sample Relief NA RMW-V29 Reactor Makeup Water IRC Check NA SF-V101 Refueling Cavity Cleanup Relief NA SI-V247 Accumulator Fill / Test Header Relief NA ht Ace.id eaf SS-V273 '4 Sample Flush Tank Drain IRC Check NA WLD-V209 Sump "B" to FDT Relief NA WLD-V213 PDT to RC Drain Tank Relief NA RC-FV2894* RCS Loop 1 Sample NA RC-FV2896* RCS Loop 3 Sample NA
# Not subject to Type C leakage test
- May be opened on an intermittent basis under administrative.contr_o]__
SEABROOK - UNIT 1 3/4 6-22 h lM MM AIMOA >['h 1 .1
C' d M 74~.~6J4'~ COMBUSTIBLE GAS CONTROL _ _ __..-~~f
/ - . N '
HYOROGEN MONITORS. , [
/
LIMITING CONDITION FOR OPERATION _ - 3.6.4.1 Two independent containment hydrogen monitors, hall be OPERABLE. APPLICABILITY: MOD and 2.
/
[ ACTION: /
- a. With one hydrogen monitor inoperab e, restore the inoperable monitor to s or be in at least HOT STANDBY within the OPERABLE next 6 hours. status within DO day /
/
- b. With both hydrogen monitors inop'erable, restore at least one monitor to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours'.
/
/
SURVEILLANCE REOUIREMENTS
/
Each hydrogen monitor shall be demonstrated OPERABLEs by the 4.6.4.1 performance of a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 9,2 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing:
- a. One volume percent hydrogen, balance nitrogen, and Fou g olume percent hydrogen,~ balance ~ nitrogen. ~ ~ ~ - _
,jb .
Q, ? e (;hls sfeclfierlion - The eg a ymed is coterad by 5fecWicaJi a n s. 3. 3. b '/$ce: dad Mo r,llor;y 3. n % ,gg;, SEABROOK - UNIT 1 3/4 6-23 J 'R
- -f$ -N d$$
A Pio : j
CONTAINMENT SYSTEMS f-) I .
~
ELECTRIC HYDROGEN WECOMBINERS ! LIMITING CONDITION FOR OPERATION - 3.6.4.2 Two independent Hydrogen Reconibiner Systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE: () a. At least once per ' months by verifying during a Hydrogen Recombiner System functional test that the minimum heater sheath temperature increases to greater than or equal 700*F within 90 minutes. 1 Upon reaching 700*F, increase the power setting to maximum power l for 2 minutes and verify that the power meter reads greater than or equal to 60 kW; and . i
- b. At least once per 18 months by:
- 1) Performing a CHANNEL CALIBRATION of all recombiner l instrumentation and control circuits, j
- 2) Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e. , loose wiring or structural connections, deposits of foreign materials, etc.), and
- 3) Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.
/
O V U$ 3 D'
.fJv f) ^I -
ri \ M [ []% f SEABROOK - UNIT 1 3/4 6-24 w
CONTAINMENT SYSTEMS [] .
~
[onh:nmenf Straep re fe circ a ltNn - r HYDROGEN = MIXING-SYSTEM
~
LIMITING CONDITION FOR OPERATION Oaf 1],p.ex f $$rxc2 fare 0CWC" 3.6.4.3 Two independent Hydrogen:MfXing=SIystems-shall be OPERABLE. APPLICARILITY: MODES 1 and 2. I g ;n n e n } S t ra c i k t e t*E C
- l# N'* " y With one Hydrogen: Mixing-System inoperable, restore the inoperable system-to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.
SURVEILLANCE REQUIREMENTS cobl^m**$ S!ruiurc c'cI'c"k W L 4.6.4.3 Each Hydrogen: Mixing-System-shall be demonstrated OPERABLE:
- a. At least once per 92 days on a STAGGERED TEST BASIS by starting each system from the control room and verifying that the system operates for at least 15 minutes, and
- b. At leas > once per 18 months by verifying a system flow rate of at least 4 LOO cfm. .
/
" " ~ ~ ~ ~ . -
e
)m\}*2r-) /% hy1'*RrII
-- A h. P A 1 ,
SEABROOK - UNIT 1 3/4 6-25 [ j MAR 131986
CONTAINMENT SYSTEMS 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING - CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent Containment Enclosure Emergency Air Cleanup Systems shall be OPERABLE. c APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one Containment Enclosure Emergency Air Cleanup System inoperable, re-store the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. - SURVEILLANCE REQUIREMENTS O 46s1 e ca ce t immeat e#cie #re emerseacx ^<r cie #ua sx tem seii ne demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal ,
adsorbers and verifying that the system operates for at least 10' 15 minale 5. continuous-hoursy .
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi- l cating with the system by: j
- 1) Verifying that the cleanup system satisfies the in place pene-
/,0% pt tiog and bypass leakage testing acceptance criteria of less than 4:05% and uses the test procedure guidance in Regulatory Positions C.5.a. C.S.c, and C.5.d of Regulatory Guida 1.52, Revision 2, March 1978, and the system flow rate is 2275 cfm i 10%; ,
- 2) Verifying, within 31 days after removal,'that a laboratory analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978*, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2 March 1978*, for a methyl iodide penetration of less than O 2.14%; and n, MAR131986 p. m nb
- t G I, lc \
- w1
'j 7 3/4 6-26 h *' i-i
" ^q\
r
*- i 4
SEABROOK - UNIT 1 -.
f . f~% d CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) - 1 C t o ?.115 ( > n3
- 3) VerifyingasystemflowrateYof2025[cfmi10%duringsystem operation when tested in accordance with ANSI N510-1980.
- c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory.
l: Position C.6 b of Regulatory Guide 1.52, Revision 2, March 1978[ meets the laboratory testing criteria of Regulatory Position C.6.a ofRegulatoryGuide1.52, Revision 2, March 19787foramethyl iodide penetration of less than 2.14%:
- d. At least once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA l filters and charcoal adsorber banks is less than flinches 4*5 )
Water Gauge while operating the system at a flow rate,yof 2100 cfm + 10%, ro n.F
$ h fMs
- 2) Verifying that the system starts on a Safety Injection test signal,
' c ra S S ca n c<f
- 3) Verifying that the filter cooling-bypass-valvef can be manually G cpened, V
- 4) Verifying that each system produces a negative pressure of greater than or equal to 0.25 inch Water Gauge in the annulus l within 1 minutefafter a start signalfand - I S.V
- e. After each complete or partial replacement of a HEPA filter bank, by l verifying that the cleanup system satisfies the in place penetratjion /. 6 l and bypass leakage testing acceptance criteria of less than Ord5% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 2275 cfm i 10%; and
- f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less j,o than-0,.-05% in accordance with ANSI H510-1980 for a halogenated hydro-Mcar on refrigerant test gas while operating the system at a flow rate of 2275 cfm i 10%.
- ANSI 510-1980 shall be used in place of ANSI 10-1975 referenced in Regulatory Guide 1.52, Rev. 2, March 1978.
/
O MAR 131988 iPb y n> ' .c
/J nP8 ;qd J-" :Ul ";s .\W ;-)1 J}N SEABROOK - UNIT 1 3/4 6-27 - ....
C-N
CONTAINMENT SYSTEMS O.
- CONTAINMENT ENCLO5URE BUILDING INTEGRITY -
LIMITING CONDITION FOR OPERATION - 3.6.5.2 CONTAINMENT ENCLOSURE BUILDING INTEGRITY shall be maintained. APPLICABILITY: MODES 1, E, 3, and 4. c-ACTION: 5 l Without CONT?.INMENT ENCLOSURE UILDING INTEGRITY, restore CONTAINMENT ENCLOSURE BUILDING INTEGRITY within 24 ours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS O 4es2 co"T^1NMENT ENCLOSURE BUILDING INTEGRITY sha11 de demonstrated at least once per 31 days by verifying that each door-inzeachraccess-opening. is closed except when the access opening is being used for normal tran:it entry and exit,-then-at"least=one xfcor3hallrbe-c4osed, MAR 131986 O - nn R
/ YI s
3'.t rs. I
.a,I~1 ll
- .2 . a SEABROOK - UNIT 1 3/4 6-28 . ____ _ _ .
I i O
~
CONTA1NsENT Sv51Ess 7 . 5 CONTAINMENT ENCLOSURE BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.3 The structural integrity of the containment enclosure building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.5.3. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the structural integrity of the containment enclosure building not con-forming to the above requirements, restore the structural integrity to within the limits within 24 $ours or be la at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. 7 J's O - SURVEILLANCE REQUIREMENTS , 4.6.5.3 The structural integrity of the containment enclosure butiding shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed acces-sible interior and exterior surfaces of the containment enclosure building and l verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation. Any abnormal degradation of the containment enclosure building detected during the above required inspections shall be reported to theCommissioninaSpecialReportpursuanttoSpecification6.h2within 15 days. g l l 9 MAR 131986 SEABROOK - UNIT 1 3/4 6-29 LH LL - - . - - . . . _ L$ k
l l l p 3/4.7 PLANT SYSTEMS U 3/4.7.1 TURBINE CYCLE ,
~
SAFETY VALVES LIMITING CONDITION FOR OPERATION All main steamiline Code safety valves associated with each steam f 3.7.1.1 generator shall be OPERABLE with lift settings as specified in Table 3.7-2. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that f within 4 hours, either the inoperable valve 9 s restored to OPERABLE (
. status or the Power Range Neutron Flux High Trip Setpoint is reduced
> per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. O ks b. The provisions of Specification 3.0.4 are not applicable.
~ -
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. l l
/
O
. MAR 131986 n
j II .) P 9 p' g SEABROOK - UNIT 1 3/4 7-1 {] ,)j {bf, i 4 .1 d .
TABLE 3.7-1 --
. MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH ' '
INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP OPERATION ! I MAXIMUM NUMBER OF INOPERABL2 MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) p 1 87 2 65 3 43 6 0 - - l l 4 c - MARL 31986 O Tr3 q ' SEABROOK - UNIT 1 3/4 7-2 1
~.
DAs.1[*A)
O Ta8'e 3. 7->= STEAM LINE SAFETY VALVES PER LOOP - - 1' VALVE NUMBER LIFT SETTING (i 1%)* _ ORIFICE SIZE Loop 1 Loop 2 Loop 3 Loop 4 V6 V22 V36 V50 1185 psig 16.0 sq. in. V7 V23 V37 V51 1203 psig 16.0 sq. in. V8 V24 V38 VS2 1220 psig 16.0 sq. in. V9 V25 V39 V53 1238 psig 16.0 sq. in. V10 V26 V40 V54 1255 psig 16.0 sq. in. O
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
O MARL 3.1988.
/h I,t? T SEABROOK - UNIT 1 3/4 7-3 d} 3aDra n a
, , _ _ _ _ , , j
PLANT SYSTEMS f) ht e rgencJ' 'd . AUXILIARY FEEDWATER SYSTEM . r . 1 LIMITING CONDITION FOR OPERATION 6is o e emern ac 3.7.1.2 At least -three independent steam generator auxiliary)'feedwater pumps and associated flow paths shall be 0PERABLE with:
- a. One motor-driven emergency feedwater pump and-one-startup ',
feedwater-pump,--each capable of being powered from separate GLf1 emergency buss &ri, and
- b. One steam turbine-driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1, 2, and 3. . ACTION: emer3cnQ
- a. With one auxiliary feedwater pump inoperable, restore the required emergenc-7 auxiliary feedwater pumph to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
O
- b. With two ugnency auxiliary feedwater pumps inoperable >f be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
-c . With=three-auxiliary feedwater pumpsdnoperable mmediately initiate orrective action to restore atleast one auxiliary feedwater pump to OPERABLE status as soon as possible,or C '" "D * "I' SURVEILLANCE REOUIREMENTS eatm ercy 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
- a. At least once per 31 <.lys on a STAGGERED TEST BASIS by:
f h e.
- 1) Verifying that aach-motor-driven pump develops a discharge pressure of greater than or equal to 1450 psig at a flow of greater than or equal to 222 gpm;
- 2) Verifying that the sc6sa m ine-driven pump develops a discharge pressure of greater than or equal to 1450 psig at a flow of greater than or equal to 222 gpm when the secondary O steam suPnix pressure is areater thaa 22 ofSpecification4.0.4arenotapplicab[.09sio.
The provisieas e for entry into MODE 3; soo MARi3-1986 SEABROOK - UNIT 1 3/4 7-4 f hI 1./ n ha x -- i
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) . i
- 3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured-in position is in its correct position; and
- 4) Verifying that each automatic va ve q,t e flow path is in the fully open position whenever the AuxtItary Feedwater System is placed in automatic control or whert.above 10% RATED THERMAL POWER.
- b. At least once per 18 months during shutdown by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater System Actuation test signal, end-
- 2) Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal.
-3) Varifying-that-under-manual-control-the startup-feidwat~er~ pump s ta rts -a nd -th a t-a ll-re q u i re d -va l v e-re a l-i g nme n ts -c a n- b e -ma d e -
v w i t h i nathenti mezal.l owe d -i n :-the -de sri g n fmerAene-y 4.7.1.2.2 Auxi4tary feedwater flow paths to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying design flow to each steam generator.
'from:- -
-1) Each emergehcTfeedwatir~pEmp--- +
- 2) -The-startup-feedwater-pump-via-the-main-feedwater flow path ~ !
a nd -v i a-the -eme rge ncy-fe e dwa te r - he ade r.
]
c . At lead once fer, F nmed:C; monk &c: cal;or s f,y s kwn, o r Ql{ouing com f)IeBok o 6 he GFG) Spfe m D hlc h allers lhe Sysbo's FW G hkr" der ls t; cs, b y verify;ny flow from the CGT 6o ac h s'leam yncedo r-O ( MAR +31986-pai
-: . a
/'
n yme' , I A (e'
- r:!e l' SEABROOK - UNIT 1 3/4 7-5
] lv l
"" " ' " ^ -
(5
-.. i
f) v PLANT SYSTEMS
- CONDENSATE STORAGE TANK -
- ^ .
s. LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) system shall be OPERABLE with
- a. A volume of 210,000 gallons of water contained in the condensate storage tanudnd-bc -A-concrete-CST-enclosure-that-is capable of-retaining-210,000-gal ~
,l o n s -o f-wate r.-
- c. The-condenser-hot well and the~d&miiferalizedwater storage-tank DPERABLE-as-backup supply of-210,000-gallons-of water-APPLICABILITY: MODES 1, 2, and 3.
ACTION: With the CST crathe= CST enclosure inoperable, within 4 feithere
\ Restore the CST and-the= CST-enc-losure .to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,.or*
-b- Demonstrate-the-OPERABILITY-of-the-condenser-hot-well an~d the- demin -
.eralized-water-storage-tank-as-a-backup-supply to the auxiliary-feed-wa te r-pump s-a n d -re s to re -th e -C ST-to-O P ERAB LE-s tatu s71 thi n d ay s-
.o r-be --i n-a t-l e a s t-HOT-STANDBY-w i t h i n-t h e -n e x t-6-ho u rs -a n d -i n - HOT-4 HUTDOWN vi th i n-th e -f o i-l owi n g-6 -h o u rs e SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The CST and-the-CST enclosure shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume is within its limits and-the-CST = enclosure = integrity-is: maintained- when the tank is the ar feedwater pumps.
supply source for the gx 4r7rir3.-2-The-condenser-hot-wel1 and-the-demineralized-water-storage tank ~ sha1Lbe demonstrated-OPERABLE-at-least once per-12-hours-by-verifying-the-avai-lability-of-210,000 -gallons of-water-whenever-the-condenser-hot-well-and-
-the-demineralized-water storage-tank is-the-supply source-for-the-auxiliary -
-feedwater-pumps.- ,
MAR 131986 O -.-- 8 SEABROOK - UNIT 1 3/4 7-6 - -- - .
PLANT SYSTEMS
. SPECIFIC ACTIVITY -
r - LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurie / gram DOSE , EQUIVALENT I-131. p APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the specific activity of the Secondary Coolant System greater than 0.1 microcurie / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. O Suave 1 uasCe accuraEsE"TS 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. ~ ' ~ ~ e MAR 131986 [
~ * ' * * - _.. ,_
y \p L a /-$1 111 i SEABROOK - UNIT 1 3/4 7-7
TABLE 4.7-1 b'
~
SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY 7 .- SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY
- 1. Gross Radioactivity -
At least once per 72 hours. Determination
- c
- 2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 10% of the allowable Ifmit for radioiodines.
b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10%
/~S -
of the allowable limit kl for radiofodines.
*A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the secondary coolant except for radio-nuclides with half-lives less than 10 minutes. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level, i
MAR 131986
- (2)
Tb F? 3. p m SEABROOK - UNIT 1 3/4 7-8 S 'J- Ad e--
PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES
~. !
LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ' ACTION: MODE 1: With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY within the next 6 hours and in HOT SHUTOOWN within the following 6 hours. MODES 2 and 3: With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following () 6 hours. t/ SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5.0 seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
/
U< , MAR 131986 I . h Sug 6 -* g{ g4 I-l k il /, [.) , rj 3/4 7-9
.iJ.J ga.
SEABROOK - UNIT 1 o ! w >
" - Q.
PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION .
]
~. ~
LIMITING CONDITION FOR OPERATION 3.7.2 The temperatures of both the reactor and secondary coolants in the steam generators shall be greater than 70*,F when the pressure of either coolant in the steam generator is greater than 200 psig. c APPLICABILITY: At all times. ACTION: With the requirements of the above specification not satisfied:
- a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and
- b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator. Determine that the' steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F.
SURVEILLANCE REQUIREMENTS 4.7.2 The pressure in each side of the steam generator shall be determined to . be less than 200 psig at least once per hour when the temperature of either the reactor or secondary coolant is less than 70*F. s O -MAR 13 l886-T T[,:) p.,
.U. 4 .l-h A%~kks iQ O t.f la,
. ., 6 SEABROOK - UNIT 1 3/4 7-10 --
PLANT SYSTEMS 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM r - ' LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent primary component cooling water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. c ACTION: W ith only one primary component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.7.3 At least two primary component cooling water loops shall be demonstrated
~
OPERABLE: O - at ie st e#ce Per 31 d vs hv verifyias taat ch vaive cmenoai. power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and
- b. At least once per 18 months during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related eq0fpment actuates to its correct position on its' associated Engineered Safety Fea-ture actuation signal, and
- 2) Each primary Component Cooling Water System pump starts auto-matically upon loss of or-failure ~to' start of the~ridundant pump'
.w i th i n -the _.l oop .-
4 W e,r tesf fjg h /. O MAa la 1988 -- SEABROOK - UNIT 1 3/4 7-11 ,_A , . , y, 2Q i, ' }
- . - - . . =
-s PLANT SYSTEMS fv)
, 3/4.7.4 SERVICE WATER SYSTEM -
7' . - t LIMITING CONDITION FOR OPERATION losPs 3.7.4 At least two independent service water systems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, 4. c ACTION: With only one service water loop OPERABLE restore two loops to OPERABLE status within 72 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS l lbo Ps 4.7.4 At least two Station Service Water $ystems- shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual,
({]) power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and
- b. Et least once per 18 months during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related equipment actuates L: its correct position on its associated Engineered Safety Fea-ture actuation test signal, and
- 2) Each Station Service Water System pump starts automatically upon loss of or-failure-to-start-of-the-redundant pump-within -the--
, cop.. yo w er +es t sy I.
r O n- ~
-MAR 131986
~~
} 3, ,
). y h.
SEABROOK - UNIT 1 3/4 7-12 ,s . . s .. . .... ...
p G PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with:
- a. A service water pumphouse water level at or above minus 37'0" Mean P Sea level, USGS datum, and ;
/
- b. A mechanical draft cooling tower comprised of two cooling tower fans W and a contained basin water voU me of equal to or greater than 35M M4X105 ga11ons at an average water temperature of less than or equal to 70*F, and gfg 47.3
.c --A portable-tower makeup-pump-system-stored-to-be-OPERABLE for-30-days following a-Safe-Shutdown-Earthquake.,
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With the service water pumphouse inoperable, restore the service water pumphouse to OPERABLE status within 72 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With the mechanical draft cooling tower inoperable, restore the cooling tower to OPERABLE status within 72 hours, or be in at least NOT STANDBY within the next 6 hours and in COLD SHUTD WN within the following 30 hours.
.cr - -With-the-portable-tower makeup-pump-system-inoperable; continue-operation 7nd notify-the NRC within_L.hourJfL-accor_ dance with:the
-procedure-of-10-CFR -50. 72-o f-acti ons-or-conti ngencies-to -en s ure-an -
adeq uate -s upply-o f-ma ke up-wate r-to -the - mecha n i cal-dra f t-cool i ng -
-towe r-fo r-a -mi n i mum - o f day s .
SURVEILLANCE REQUIREM*yTS 4.7.5 The ultimate t.24t sink shall be determined OPERABLE:at-least once=per:- g.. M led nc e ger M hours by -
,A,! /).24-hours by h,erifying the water level in the service water than:orcequalsto 37'- 0" Mean Sea Level, pumphouse to bea.c greaterbon or A. Ai Aqs
'V,,d24-hours-byYerifyingthewaterinthemechanjcaldraftcooling j FMT tower basin to be greater than or equal to a v$II;me of.4X105- 35.4
. gallons-and-at-a-temperature Jess than-or_. equal to-70*F ,
- b. At just*ber- once pr adays From Inn I la Sq t. go e e s p ac,by vec n i3 MAR-WIS86
i ? 'y ; {
y( p)
^
Mat coo ky iow,,e bxsin balR verr os br ~ . egff,rg 6e is ,j u l 6, or /e:s $ etha n , ;, ; .j-SEABROOK - UNIT 1 3/4 7-13
- k:*
#^ "\ T"' p * [ji
+ weu rA.,%h w.uA,p>su ,vwa su-v4_,u , , , , *: m - .-. - . - .
PLANT SYSTEMS LIMITING CONDITIOM.FOR OPERATION CONTINUED C. $1Everf 31d.tvs bvp g;, 31: days-by< starting from the control room each UHS cooling tower fan that is required to be operable and operating each of those fans for at least 15 minutes,
-dM1-~ days'by Verifying-that-the portable-tower-makeup-pump-system-is-
. stored-in-its-design operationaf-readiness-state, d; ert
%. 18 months by verifying automatic actuation of each cooling tower
#' fan on a Tower Actuation test signal.
l 1 I I c O MAR 133985 ' 1 3y) a
?,- VM
- r. ,
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"^L eL. \ .
SEABROOK - UNIT 1 3/4 7-14 -
PLANT SYSTEMS
) 3/4.7.6 SNUBBERS LIMITING CONDITI0ttFOR OPERATION ,
3.7.'6 All lubbers shall be OPERA 8LE. The only snubbers excluded from the requirement are t1ose installed on nonsafety-related systems and then only iftheirfailureo)f,failureofthesystemonwhichtheyarieInstalledwould have no adverse effect on any safety related system. APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.' ACTION: c-With one or more snubbers inoperable on any system, within 72 hours replace or re-store the inoperable snubber (s);to OPERABLE status and perform an engineering eval-uation per Specification 4.7y9'g. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS 4.7.6 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. In heu af'
- a. Insoection Types As used in this specification, type of snubber shall mean snubbers
(] of the same design and manufacturer, irrespective of capacity.
\"-
- b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be inspected independently according to the schedule below. The first inservice visual inspection of each type of snubber shall be performed after 4 months but within 10 months of commencing POWER OPERATION and shall include all snubbers. If all snubbers of each type [on any system] are found OPERABLE during the first inservice visual inspection, the second inservice visual inspection [of that system] shall be performed at the first refueling outage. Otherwise, subsequent visual inspections [of a given system] shall be performed in accordance with the following schedule:
No. of Inoperable Snubbers of Each Type Subsequent Visual [on Any System] per Inscection Period Inspection Period * ** 0 18 months 2 25% 1 . 12 months i 25% 2 6 months i 25% 3,4 124 days 25% 5,6,7 62 days i 25% 8 or more 31 days i 25%
*The inspection interval for each type of snubber [on a given system] shall not n be lengthened more than one step at a time unless a generic problem has been V identified and corrected; in that event the inspection interval may be lengthened one step the first time and two steps thereafter if no inoperable snubbers of that type are found [on that system]. _ . . ._ _ g - -
**The provisions of Specification 4.0.2 are not appl icabg. m fSgg , 3 SEABROOK - UNIT 1 3/4 7-15 H h , ,,/ t i C- j ?
O_ _-L . hh. h h.
1 q D PLANT SYSTEMS l l SURVEILLANCE REQUIREMENTS (Continued)
. I
- c. Visual Inspection Acceptance Criteria Visual inspections shall verify that: (1) there_are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are functional, and (3) fasten-ers for attachment of the snubber to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the pur-pose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespec-tive of type [on that system] that may be generically susceptible; and (2) the affected snubber is functionally tested in th q s-found condition and determined OPERABLE per Specification 4.7.)f. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers. [For those sr."bbers common to more than one system, the OPERABILITY of such snubbers shall be considered in assessing the surveillance schedule for each of the related systems.]
- d. Transient Event Inspection h' An inspection shall be performed of all snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual inspection of the systems within 6 months following such an event. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified '
using at least one of the followingi (1) manually induced snubber movement; or (2) evaluation of in place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel,
- e. Functional Tests During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers of each type shall be tested using one of the following sample plans.
The sample plan for each type shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan selected for each snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:
- 1) At least 10% of the total of each type of snubber shall be functionally tested either in place or in a bench test. For each snubber of a type that does not meet A the functional test acceptance criteria of Specification 4.7.ff. , an additional 10%
of that type of snubber shall be functionally tested until no O ore faiieres are feund or untii aii nubbere or that tvPe have been functionally tested; or 8 SEABROOK - UNIT 1 3/4 7-16 :: b[ a . n.k. .h. ~. l -.
PLANT SYSTEMS y I J SURVEILLANCE REQUIREMENTS (Continued) T i
- e. Functional Tests (Continued)
- 2) A representative sample of each type of snubber shall be func-tionally tested in accordance with Figure 4.7-1. "C" is the total number of snubbers of a type fount'not meeting the accept-ancerequirementsofSpecification4.7.ff. The cumulative number of snubbers of a type tested is denoted by "N". At the end of e,ach day's testing, the new values of "N" and "C" (pre-vious day's total plus current day's increments) shall be plotted on Figure 4.7-1. If at any time the point plotted falls in the " Reject" region, all snubbers of that type shall be functionally tested. If at any time the point plotted falls in the " Accept" region, testing of snubbers of that type may be terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the
" Reject" region, or all the snubbers of that type have been tested; or
- 3) An initial representative sample of 55 snubbers shall be func-tionally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until
('"') the total number tested is equal to the initial sample size multiplied by the factor, 1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. The results frora this sample plan shall be plotted using an " Accept" line which follows the equation N = 55(1
+ C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the
" Accept" line, testing of that type of snubber may be terminated.
If the point plotted falls above the " Accept" line, testing must continue until the point falls in the " Accept" region or all the snubbers of that type have been tested. Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time provided all snubbers tested with the failed equipment during the day of equipment failure are retested. The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are represen-tative of the various configurations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same location as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of (l V snubber, the functional test results shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the functional tR531ng. _ _ _ _ 3/4 7-17 SEABROOK - UNIT 1
~l,_.s1.u.a
i
. . i
)
l l l PLANT SYSTEMS O . SURVEILLANCE REQUIREMENTS (Continued) ,
- f. Function 1 Test Acceptance Criteria The snubber functional test shall verify that: -
1
- 1) Ac'tivation (restraining actit..i) is achieved within the specified range in both tension and compression;
- 2) Snubber bleed, or release rate where required, is present in both tension and compression, within the specified range; e
- 3) For mechanical snubbers, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel; and
- 4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement.
Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.
- g. Functional Test Failure Analysis
- An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be
' subject to the same failure mode. . _ ,
For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service. If any snubber selected for functional testing either fails to lock up or fails to move, i.e. , frozen-in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be func-tionally tested. This testing requirement shall be independent of ' the requirements stated in Specification 4.7.!)'e. for snubbers not meeting the functional test acceptance criteria. uAn ta issa O
- SEABROOK - UNIT 1 3/4 7-18 7+ D .w Ja -
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PLANT SYSTEMS _ O , SURVEILLANCE REQUIREMENTS (Continued)
- h. Functional Testino of Repaired and Replaced Snubbers Snubbers whi'.:h fail the visual inspection or the-functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and' snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance c'riteria subsequent,to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.
- 1. Snubber Service Life Procram The service life of hydraulic and mechanical snubbers shall be moditored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for various seals, springs, and other critical parts shall be deter-mined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history. Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be docu-O mented and the documentation shall be retained in accordance with Specification 6.10.3.
O [ m% I:,2) m7 J ,
.j SEABROOK - UNIT 1 3/4 7-19 '
b l -- C "_k , f k d _ _ _ _ _ _ N
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*J CONTINUE
' TESTING _ 2 2 ogM s, ACCEPT e 1 J / 60 70 80 90 100 U 10 '20 30 40 50 N l FIGURE 4.7-1 MAR 131986 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST , R '[5) }\ T]'. !]'4 3/4 7-20 $_,M ,'A ,,, 4
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--, __SEABRM} - UNIT 1- - - - . _ _ _ _ __ _
PLANT SYSTEMS im C 3/4.7.7 SEALED SOURCE CONTAMINATION 7 .. i LIMITING CONDITION FOR OPERATION 3.7.7 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination. c APPLICABILITY: At all times. ACTION:
- a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and either:
- 1. Decontami' ate and repair the sealed source, or
- 2. Dispose of the sealed source in accordance with Commission Regulations.
3 b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. (G SURVEILLANCE REOUIREMENTS _ 4.7.7.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by: -
- a. The licensee, or
- b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample. 4.7.7.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.
- a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:
- 1) With' a half-life greater than 30 days (excluding Hydrogen 3),
Il G and
- 2) In any form other than gas. MAR 1U988 b% *i} ! } %
SEABROOK - UNIT 1 3/4 7-21 ~%j
.u j) M fY jii I-m[
_ _ _ _ _ _ _ _ _ _ _ _ _ - - - - :_ = __-- __- _ -_____ _ ___.
p PLANT SYSTEMS U .
~ SURVEILLANCE REQUIREMENTS (Continued)
. i
- b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use; and
- c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.
4.7.7.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination. O l
- r
. O ' 3/4 7-22 i e AaN d.!.18 [ D.;*Ansg'ro SEABROOK - UNIT 1 - ~ . . . - y- - _,
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PLANT SYSTEMS O 3/4.7.8 FIRE SUPPRESSION SYSTEMS . 7 .. !* FIRE SUPPRESSION WATER SYSTEM
~
LIMITING CONDITION FOR OPERATION 3.7.8.1 The Fire Suppression Water Systerii shall be OPERABLE with: two At least-three fire suppression pumps, each with a capacity of g. a. 1500 gpm, with their discharge aligned to the fire suppression header,
- b. p#aSeparate water supplies, each with a minimum contained volume of 300.000 gallons, and ,
- c. An OPERABLE flow path capable of taking suction from the fire water tank and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3. 7 10'. 2, 3. 7.10',. 5, and 3. 7.10.~ 6.
W '$ '8 APPLICABILITY: At all times. ACTION:
- a. With one pump and/or one water supply inoperable, restore the inoper-able equipment to OPERABLE status within 7 days or provide an alter-nate backup pump or supply. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
- b. With the Fire Suppression Water System otherwise inoperable, establish a backup Fire Suppression Water System within 24 hours.
O WR 13 !988 _ __ _
'O A 7-SEABROOK - UNIT 1 3/4 7-23 ;
Y h ~[b.Y,rn
-'1
PLANT SYSTEMS
. SURVEILLANCE REQUIREMENTS
' 1 4.7.8.1.1 The Fire Suppression Water System shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying the contained water supply volume,
- b. At leas
- once per 31 days by starting the electric motor-driven pump and oper ating it for at least 15 minutes on, recirculation flow,
- c. At least once per 31 days by verifying that each valve (menual, power-operated, or automayic) in the flow path is in its correct position, I;~
- d. At least once per'6 months by performance of a system flush,
- e. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
- 1) Verifying that each automatic valve in the flow path actuates O to its correct position,
- 2) Verifying that the pump develops at least 1500 gpm at a system head of 250 feet, c230
- 3) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and
- 4) Verifying that each fire suppression pump starts sequentially:
to maintain the Fire Suppression Water System pressure greater than or equal to 120 psig,
- g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.
l l l O MAR 13 Iggg I .'"') M T '; ' U's 3/4 7-24 ./ .' \ [l SEABROOK - UNIT 1 - g ., ,,1 A . A A. h. 51 o_
i PLANT SYSTEMS SURVEILLA1TCE REQUIREMENTS (Continued) , The. \ 4.7.8.1.2 Each* fire pump diesel engine shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying:
869
- 1) The fuel storage tank contains at least 200 gallons of fuel, and
- 2) The diesel itarts from ambient conditions and operates for at least 30 ininutes on recirculation flow,
- b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270 -1975-is-within-the-acceptable-limits-specified in Table _1_of ASTM _D975-1977-when-checked-for vis~c~osity andMter~and sediment;-and ~ 6a.c- afdae / icd in s e r f-I
- c. At least once per 18 months, during shutdown, by subjecting the diesel to an inspection in accordance with procedures prepared in conjunction I with its manufacturer's ret amendations for the class of service.
1 4.7.8.1.3 The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE: .
- a. At least once per 7 days by verifying that:
- 1) The electrolyte level of each battery is above the plates, and
- 2) The overall battery voltage is greater than or equal to 24 volts.
- b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and
- c. At least once per 18 months by verifying that:
- 1) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and
- 2) The battery-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.
'me, O ,
MAR 1.yggg_ SEABROOK - UNIT 1 3/4 7-25 fl R k ,ji' {
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PLANT SYSTEMS SPRAY AND/OR SPRINKLER SYSTEMS - LIMITING CONDITION FOR OPERATION 3.7.8.2 The following Spray and/or Sprinkler Systems shall be OPERABLE:
-a. Cable-Spreading-Room -
-b. Computer-Room-
-c. Electrical-Tunnels.
-1. Control-Bui-1 ding-to-Containmenk Se _ e DseA~
.2-Control-Building-to-PAB-
.d. Diesel-Generator-Bui-1 ding-
- 1. Fuel-O f-1-Storage-Tanks-
- 2 .- Fuel-011 Pip ~e Trencher"
-3. Fuel-011-Day-Tanks-
_e. Primary-Auxiliary-Building- _1 Electr_ical Chases - I APPLICABILITY: Whenever equipment protected by the Spray / Sprinkler System is required to be OPERABLE. ACTION:
- a. With one or more of the above required Spray and/or Sprinkler Systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. j SURVEILLANCE REQUIREMENTS 4.7.8.2 Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE:
- a. At least 'once per 31 days by verifying that each valve (manual, power-operated, oc automatic) in the flow path is in its correct position, ER_13198L _
\c \
SEABROOK - UNIT 1 3/4 7-26 b# -. b A N. i
l The following spray and/or sprinkler systems shall be operable: 3.7.8.2
~
. t*
A. Cable Spreading Room
- 1. System 1 ,
- 2. System 2
- 3. System 3
- 4. System 4
- 5. System 5 -
c B. Diesel Generator Building - Train A C
, 1. Fuel Oil Storage' Tank System
- 2. Redundant Fuel Oil Storage Tank System
- 3. Fuel Oil Day Tank System 4 Fuel Oil Pipe Trench System
- 5. Diesel Generator Room System l 6. Fuel Oil Storage Tank Room Sump System C. Diesel Generator Building - Train B
- 1. Fuel Oil Storage Tank System
- 2. Redundant Fuel Oil Storage Tank System I
O 3. 4. Fuel Oil Day Tank System Fuel Oil Pipe Trench System
~
- 5. Diesel Generator Room System
- 6. Fuel Oil Storage Tank Room Sump System D. Electrical Tunnel - Train A -- - - - - - - - - - - - - - - -
E. Electrical Tunnel - Train B F. Primary Auxiliary Building
- 1. Electrical Chase
- a. Vertical Portion of Fire Area PAS-F-1G-A
/
- b. Horizontal Portion of Fire Area PAB-F-1G-A
- 2. Elevation 25' Area System O
q,
- b. At least once per 12 months by cycling each testable valve in the O fie aath throue8 at iea t ome cemniete excie of f#ii travei.
~
- c. At least-once per 18 months: . . g
- 1) By performing a system functional test which includes simulated automatic actuation of the system, and: ,
a) Verifying that the automatic valves in the flow path actuate to their correct positions on a simulated test signal, and b) Cycling each valve in the flow path that 15 not testable during plant operation through at least one complete cycle of full travel.
- 2) By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity; and
- 3) By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed. j l
- d. At least once per 3 years by performing an air flow test through l each open head spray / sprinkler header and verifying each open head spray / sprinkler nozzle is unobstructed.
O - l l O MAR 131986
., 'A SEABROOK - UNIT 1 3/4 7-27 y ;, f .hA a a+ .na
.L d '
l PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION glve 3.7.8.3 The fire hose stations /r in Ta ble 3.7-3 shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. .. ACTION: e
- a. With one or more of the fire hose stations inoperable, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the
- wye shall be connected to the standard length of hose provided for the hose station. The second outlet of the wye shall be connected to a length of hose sufficient to provide coverage for the area lef t unprotected by the inoperable hose station. Where it can be demon-strated that the physical routing of the fire hose would result in a recognizable hazard to operating technicians, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the OPERABLE hose station. Signs shall be mounted above the gated wye (s) to identify the proper hose to use. The above ACTION require-ment shall be accomplished within 1 hour if the inoperable fire hose is the primary means of fire suppression; otherwise route the addi-p V tional hose within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not appitcable.
SURVEILLANCE REQUIREMENTS 4.7.8.3 The fire hose stations'shall be demonstrated OPERABLE:
- a. Atleastonceper31 days,byavisualinspectionofi.hefirehose stations accessible during plant operations to assure all required equipment is at the station,
- b. At least once per 18 months, by:
- 1) Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station,
- 2) Removing the hose for inspection and re-racking, and
- 3) Inspecting all gaskets and replacing kny degraded gaskets in the couplings.
- c. At least once per 3 years, by:
- 1) Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and
- 2) Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, O 8$caever 4 are ter-MAR 131986
.) DM P
3/4 7-28 5 -- SEABROOK - UNIT 1 .L // z_ n 1. 1
.I ._) L
i i i TABLE 3.7-3 FIRE HOSE STATIONS LOCATION ELEVATION - HOSE REEL NUMBER i ENCR. W BLDG. - i West end opposite door 27' p 71 [4)tsf C.Ad } n SbOVR-f yQ E [R EQUIPME?-T VAULTS North Vault ,
-50' 27 i
South Vault -50' 26 North Vault -31' 25 South Vault -31' 24 North Vault 3' 23 . O seuth vau1t s' 22 CONTROL BUILDING + Stairway - 21' ._. , 30 ,
~ ~ '
j Turbine Bldg. by door to Essential Swg. Rs. A 21' 8A e 2 Stairway 50' 29 Turbine Bldg. by door to Cable Spd. Rs. '50' 15A Stairway 75' 28 Turbine Bldg. by tornado door , 75' 20A i i i O SEABROOK - UNIT 1 3/4 7-3
m TABLE 3.7-3 - V FIRE HOSE STATIONS r - LOCATION ELEVATION . HOSE REEL NUMBER a DIESEL CEN. BLDC. l "A" Train in stairway ,
-8' 67 1
"B" Train in stairway -8' 70
. . "A" Train in stairway. 21' 66 "B" Train in stairway 21' 69 "A" Train in stairway 51' 65 4 "B" Train in stairway 51' 68 ELECTRICAL TUNNELS l Q "A" Train - West stairway O' 63A "A" Train - East stairway O' 63 "B" Train - West end of Tunnel -20' 64A j
' ~ ~ ~ ~
l "B", Train - East stairway
-26' '64 ~~ j 9
O SEABROOK - UNIT 1 3/4 7- 74
^
/~ TABLE 3.7-3 FIRE HOSE STATIONS LOCATION ELEVATION -
HOSE REEL NUMBER PRI. AUX. BLDC. (PAB) ,. p Piping Penetration A.ea 373 pipin3 Peonder n Acer -a c,'-26 ' 2 , C. e North stairway -6' 37 Outside Demineralizer Access Room -6' ; 37A North stairway 7' - 36 South stairway 7' 38 North stairway 25' 34 South stairway 25' 35 North stairway 53' 33
~
South stairway 53' 32 Outside HVAC Eq. Room 81' 31 ~ - ~ FUEI. STORAGE BLDG. t Outside SF pump area 7' ; 49 By West doorway 21' 48 By West stairway 64' 47 MAIN STM - FW PIPECHASE South stairway 12' , 22A South stairway 21' , 22B SEABRC0K - UNIT 1
/
3/47-)(
TABLE 3.7-3 FIRE HOSE STATIONS r - LOCATION ELEVATION - HOSE REEL NUMBER CONTAINMEYr ge . Approx. 40# on outside wall -26' 53 19 ' e Approx.140*' by "C" accumulator -26' 51 Approx. -1709/d -26' 60 Approx. 320* by stairway -26'- 57 Approx. 55' 0' 54 Approx.120' by equip. hatch O' 52 Approx. 220* opposite inst. rack O' 61 l Approx. 310* o as* Hz Approx. 30# behind lh; recom. . O' 25' 58 55 ! 135* t Approx. 140**by equip. hatch 25' 50 ;
. Approx. 225* 25' 59 .
g,g * . .. . - - - Approx. 320" by personnel hatch "~ ~ ~ ~ '25' ' ~ ~ 56 i O / SEABROOK - UNIT 1 3/47-l r
PLANT SYSTEMS () , YARD FIRE HYDRANTS AND HYDRANT HOSE HOUSES . LIMITING CONDITIO$~FOR OPERATION Sueg in Tih/e 3. 7- 4' 3.7.8.4 The yard fire hydrants and associated hydrant hose houses /shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE. - s' ACTION:
- a. With one or more of the yard fire hydrants or associated hydrant hose houses inoperable, within 1 hour have sufficient additional lengths of 2 1/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area (s) if
, the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression; otherwise, provide the additional hose within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
S*JRVEILLANCE REQUIREMENTS O's 4.7.8.4 The yard fire hydrants and associated hydrant hose houses shall be demonstrated OPERABLE:
- a. At least once per 31 days, by visual inspection of the hydrant hose house to assure all required equipment is at the hose house, _
- b. At least once per 6 months (once during March, April . or May and once during September, October, or November), by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged,and-
- c. At least once per 12 months by:
- 1) Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater,
- 2) Inspecting all the gaskets and replacing any degraded gaskets ,
in the couplings, and
- 3) Performing a flow check of each hydrant to verify its OPERABILITY.
/
ALAR 131986 O - DD #
$5 g in 1 t' SEABROOK - UNIT 1 3/4 7-29 *
,,- -.y
s TABLE 3.7-4 ~ YAID FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES
- LOCATION HYDRANT Nb. HOSE HOUSE NUMBER CONTROL BUILDING Outside "B" Diesel doorway 17A HH9A Opposite stairway door 16 i
DIESEL GEN. BLDC. Outside "B" Diesel doorway 17A HH9A j gaathacf oP. k syI Gen.Eldj - l4 FUEL STORAGE BLDG. South of Pri. Aux. Bldg. (PA3) 6 HH4 Opposite Fuel Storage Bldg. 7 LMou%sl o P YueI St'org 2lh. 9 Hg5 PRI. AUX. BL')G. (PAB) South of Pri. Aux. Bldg. _6 HH4 Outside "B" Diesel doorway 17A HH9A Nor th ves f oP'Pri. Awy. Bld - dou&husf of Pri. Au.v. F/ SERV. WTR. PUMP HOUSE Ih\ Southwest of Serv. Wtr. Pump House 8 HH5 2 ~
-No r th-o f-condensate- S to rage _Tanic_. 10 EH6' S o w t h a P G r e. h 4 r. P u Hf o a n pg SERV. WTR. COOLING TOWER East end of Cooling Tower 5A HH10
-a O SEABROOK - UNIT 1 3/47 0
/
(') V TABLE 3.7-5 YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES LOCATION HYDRANT NO. ROSE HOUSE NUMBER MAIN ST:t - W PIPECHASE Nears'outhentrancItopipechase 9 HH6 FIRETPUMP HOIJSE l -/ - \ Egs/ to \(Jump Ho se 1 HH1 q1 EstERcExcr F6EMTER PwyPNour 2 - M e.r r S oc d h E d r u e< t o M x t n 9 yy& f[w, peedu.tfer p;pe Cla.s el $xd) O .
-. p . mo e & e6 -m W 4 O
i s SEABROOK - UNIT 1 3/4 7-}(
PLANT SYSTEMS D 3/4.7.9 FIRE RATED ASSEMBLIES
~
LIMITING CONDITION FOR OPERATION ! 3.7.9 All fire rated assemblies (walls, floor / ceilings, ca'ble tray enclosures, and other fire barriers) separating safety-related fire areas or separating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire rated assembly penetrations (fire doors, fire windows,firedampers, cable, piping,andventilationductpenetrationseals)P shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within 1 hour either establish a continuous fire watch on at least one side of the affected assembly, or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly and establish an hourly fire watch patrol,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
'] SURVEILLANCE REQUIREMENTS 4.7.9.1 At least once per 18 months the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by performing a visual inspection of: - -
- a. The e;:::d :frfaces of each fire rated assembly,
- b. Each fire window / fire damper and associated hardware, and
- c. At least 10% of each type of sealed penetration. If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10% of each type of sealed penetration shall be made. This inspection process shall continue until a 10%
sample with no apparent changes in appearance or abnormal degradation is found. Samples shall be selected such that each penetration will be inspected every 15 years. O 3/4 7-30 IlE? i /1 WT 2L.F L A .L L .d. L i SEABROOK - UNIT 1 _ _ _ _ 1
PLANT SYSTEMS O . SURVEILLANCE REQUIREMENTS (Continued) ,
. 1 4.7.9.2 Each of the above required fire doors shall be verified OPERABLE by inspecting the automatic hold-open, release and closing mechanism and latches at least once per 6 months, and by verifying:
- a. The OPERABILITY of the fire door supervision system for each electrically supervised fire doqr by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days, c
- b. That each locked closed fire door is closed at least once per 7 days,
- c. That doors with automatic hold-open and release mechanisins are free of obstructions at least once per 24 hours, and a functional test is performed at least once per 18 months, and
- d. That each unlocked fire door without electrical supervision is closed at least once per 24 hours.
O O MAR 13198g SEABROOK - UNIT 1 3/4 7-31 .LLe D [ li])i,h,[/\T(pm
3 PLANT SYSTEMS 3/4.7.10 AREA TEMPERATURE MONITORING
~. - 1 LIMITING CONDITION FOR OPERATION 3.7.10 The temperature of each area shown in Table 3.7-3 shall not be exceeded for more than 8 hours or by more than 30*F, or those. v e.ts ub;ch are m:jj en v i ro n h en f s s h.x lf n o t ey c e e d 13 o *F.
APPLICABILITY: Whenever the equipment in an affected area is required to be 6' OPERABLE. ACTION:
- a. With one or more areas exceeding the temperature limit (s) shown in Table 3.7-5 for more than 8 hours, prepare and submitjto the Commissionwithin30 days,pursuanttoSpecification6.J.2,a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specifi-cations 3.0.3 and 3.0.4 are not applicable.
- b. With one or more areas exceeding the temperature limit (s) shown in Table 3.7-5bymorethan30'p.prepareandsubmitaSpecialReport
- orgi anymmeb g y as required by ACTION'a M bove and within 4 hours either restore (Lrc A 4# ^j f he t area (s) to within the temperature limit (s) or declare the equip-ment in the affected area (s) inoperable.
~~
SURVEILLANCE REQUIREMENTS 4.7.10 The temperature in each of the areas shown in Table 3.7-5 shall be determined to be within its limit at least once per 12 hours. O man 13agsg SEABROOK - UNIT 1 3/4 7-32 D?] $ A _._ _[ $__
Table 3.7-r - Area Temperature Monitoring Area Temperature Limit (*F) Control Room
- 1. -
85
- 2. Cable Spreading Room
- 99.5 P 3. Switchgear Room - Train A* 99.5
- 4. Switchgear Room - Train B* 99.5
- 5. Battery Rooms - Train A 90.5
- 6. Battery Rooms - Train B 90.5
- 7. ECCS Equipment Vault - Train A 99.5
- 8. ECCS Equipment Vault - Train B 99.5
- 9. Centrifugal Charging Pump Room - Train A 99.5
- 10. Centrifugal Charging Pump Room - Train B 99.5
- 11. ECCS Equipment Vault Stairwell - Train A
- 12. ECCS Equipment Vault Stairwell - Train B
- 13. PCCW Pump Area 99.5
- 14. Cooling Tower Switchgear Room - Train A* 99.5
- 15. Cooling Tower Switchgear Room - Train B* 99.5
- 16. Cooling Tower SW Pump Area
- 122.5
- 17. SW Pumphouse Electrical Room - Train A* 99.5 O' 18. SW Pumphouse Electrical Room - Train B*
~
99.5
- 19. Sw Pump Area
- 99.5
- 20. Diesel Generator Room - Train A* 115.5
- 21. Diesel Generator Room - Train B* 115.5
- 22. EFW Pumphouse .
99.5 23. - ~~ Electrical Penetration Area - Train A 93.5
- 24. Electrical Penetration Area - Train B 80.5
- 25. Fuel Storage Building Spent Fuel Pool 99.5 Cooling Pump Area
- 26. Main Steam and Feedwater Pipe Chase - East 125.5
- 27. Main Steam and Feedwater Pipe Chase'- West 125.5
- Mild Environment Area O
wol yy w e
3/4.8 ELECTRICAL POWER SYSTEMS O) 3/4.8.1 A. C. SOURCES OPERATING , g LIMITING CONDITION FOR OPEAATION
~
3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E' Distribution System, and
- b. Two separate and independent diesel generators, each with:
c
- 1) A separate day fuel tank containing a minimum fuel volume frac-tion of 3/8 (600 gallons),
- 2) A separate Fuel Storage System containing a minimum volume of
-75f)00OO-40:0 gallons of fuel,
- 3) A separate fuel transfer pump,
- 4) Lubricating oil storage containing a minimum total volume of 274~'~ gallons of lubricating oil, and .
- 5) Capability to transfer lubricating oil from storage to the diesel generator unit. ,
APPLICABILITY: MG'ES 1, 2, 3, and 4. O ACT10": ro.a er.c,cc~+cen +s."'2 - 3 " w e w w W
- a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifi-
. cations 4. 8.1.1.la. -and-4.-8rl-172a.-5)- within 1 hour and at least
' once per 8_ hours thereaften; restore at least two offsite circuits ~~-~--
~
and two diesel generators to OPERABLE status within 72-hourN '7 Jag be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica-tions 4.8.1.1.la. .and-4r8rl--Ir2a.-5) within 1 hour and at least once per 8 hour _s_thereaf_tecy, restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two offsite circuits and two diesel generators to OPERABLE status within hoursB %lmp the time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With one diesel generator inoperable in addition to ACTION a. or b.
above, verify that:
- 1. All ' required systems, subsystems, trains components, and devices that depend on the remaining OPERABLE ddsel generator as a source of emergency power are also OPERABLE, g_
1 - 4 .7g f 3 SEABROOK - UNIT 1 3/4 8-1 ){ j' I
!I '
~~
h & .i \ a 1
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ELECTRICAL POWER SYSTEMS LIMITING CONDITION-FOR OPERATION .- ACTION (Continued) -
- 2. Wher in H0DE 1, 2, or 3, the steam-driven emergency f eedwater pump is OPERABLE.
If these conditions are not sat [sfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the c' following 30 hours.
- d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing _ the g requirements of Specification 4.8.1.1.2a.5) within MourJand-at-least once-per-8-hours--thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within houCN#fS time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- e. With two of the above required diesel generators inoperable, demonstrate 3 the OPERABILITY of two offsite A.C., circuits by performing the require-(d ments of Specification 4.8.1.1.la. within 1 hour and at least once per 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE statu:, within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two diesel generators to OPERABLE status within 72 hours from time of initial loss or be in - - -
least HOT STANDBY within the next 6 hours and in COLD, SHUTDOWN within the following 30 hours. SURVEILLANCE RECUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class IE Distribution System shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the normal circuit to the alternat'e circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
- a. In accordance with the frequency specified in Table 4.8-1 on a g/
(, STAGGERED TEST BASIS bj-
- 1) Verifying the fuel level in ;5e day fuel tank, Nh
!m q ,,
qyg SEABROOK - UNIT 1 3/4 8-2 {j1h$3bbbbN d i.- (
ELECTRICAL POWER SYSTEMS l} C _S'JRVEILLANCE REOUIREMENTS (Continued) ,
.- t
- 2) VerYfying the fuel level in the fuel storage tank, *
- 3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank, 4)
Verifying the lubricating oil inventory in storage,
- 5) Verifying the diesel starts from ambient condition and acceler-ates to at least 514 rpm in less than or equal to 10 seconds.*
The generator voltage and frequency shall be 4160 1 420 volts and 6011.2 Hz within 10 seconds
- after the start signal.
The diesel generator shall be started for this test by using one of the following signals: a) Manual, or b) Simulated loss-of-offsite power by itself, or g c) Simulated loss-of-offsite power in conjunction with an -ESF Actuation test signal, or
, d) An SI Actuation test, signal by itself.
- 6) Verifying the generator is synchronized, loaded to greater than or equal to 6083 kW in less than or equal to 120 seconds *,
and operates with a load greater than or equal to 6083 kW for at least 60 minutes, and
- 7) Verifying the diesel generator is aligned to provide standby
. power to the_as_sociated emergency busse _s.
J o'd when - - --
- b. At least once per 31 days and af4ebeach-operat4on-ofMhe-diesel where the period of operation was greater than or equal to I hour by checking for and removing accumulated water from the day -and-engine-
. mounted fuel tank;s;
- c. At least once per 92 days by checking for and remoying accumulated water from the fuel oil storage tanks; Si
- d. By sampling new fuel oil in accordance with ASTM-D4057' prior to addition to storage tanks and:
- 1) By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:
^All diesel generator starts for the purpose of this surveillance test may be preceded by an engine prelube period. Further, all surveillance tests and all other engine starts for the purpose of this surveillance tests, with the g exception of once per 184 days, may also be preceded by warmup procedures (e.g. ,
{V gradual acceleration and/or gradual loading > 60 seconds) as recommended by the manufacturer so that the mechanical stress and wear on the glg eg ne is minimized. H -8 91UC 0 i SEABROOK - UNIT 1 3/4 8-3 f L 1 _ 1 .L 2
.r
l l ELECTRICAL POWER SYSTEMS p V SURVEILLANCE REQUIREMENTS (Continued)
. a) An API Gravity of within 0.3 degrees at 60 F, or a specif'ic !
gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60 F of greater than or equal to 0.81 but less than or equal to 0.89, or an API gravity of greater than or equal to 28 degrees but less than or equal to 42 degrees; b) A kinematic viscosity at 40 C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alternativh'ly, Saybolt viscosity, SUS at 100 F of greater than or equal to 32.6, but not less than or equal to 40.1), if gravity was not determined by comparison with the supplier's certification; c) A flash point equal to or greater than 125*F; and d) A clear and bright appearance with proper color when tested in accordance with ASTM-04176-82. See Inser4 I
- 2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested in accordance with ASTM-0975-81 except that trie
, 1 we v .
analysis for sulfur may be performed in accordance with ASTM-01552-79 or ASTM-D2622-82, and the analysis for carbon (~ 4 residue may be performed in accordance with ASTM-D524-76 or s ASTM-0189-75 and converted by FIG X 3 of ASTH-0524-76 to
- Ramsbottom Carbon Residue. See_Togerr E
- e. At least once every 31 days
,1) By obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying 3that total particulate contamination is less than fwg@g[ lu mg/ liter when checked in accordance with ASTM,-D2276-78,-
Method A, samtiq l 2) By verifying the operability of the air intake prbheaters and their power a3 d c ntrol
, circugtry4 g g ,;),y,, ,af l,f of
- 3) By visually inspecting the flanged joints oni'the diesel exhaust r
system for leakage (also after an extended operation of greater than 24 hours).
- f. At least once per 18 months, during shutdown, by:
- 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service;
- 2) Verifying the generator capability to reject a load of greater than or equal to 671 kW shile maintaining voltage at 4160 1 420 volts and frequency at 601 t/ O
- 3) Verifying the generator capability to reject a load of 6083 kW p without tripping. The generator voltage shall not exceed d 4784 volts during and following the load rejection;
- 4) Simulating a loss-of-offsite power by itself, and:
a) Verifying deenergization of the emergency-busses a p i SEABROOK - UNIT 1 shedding from the emergency busses, and 3/4 8-4 b q I\ d, I,b L .t z m 1 0 886.
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c.u%ia._in_S&l.J.Ld. %. sha(L2egase b f Sdkiay : 1. G/vro}'.* Lds J<4$$ conb51.ZhLst/ sl.eJ/ ke venftid e, ee.sloxel /= meef %_ab.oue_c.nderk taEAa_M
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O_ MAR 311988 en--m k
ELECTRICAL POWER SYSTEMS Y SURVEILLANCE RE0VIREMENTS (Continued)
- b)I Verifying the diesel starts on the auto-start signal,
. energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the emergency pbwer sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After ener-gization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 + 420 volts and 60 1 1.2 Hz during this test.
- 5) Verifying that on an SI actuation test signal, without loss-of-offsite power, the diesel generator starts on the autc-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall bs 4160 + 420 volts and 60 + 1.2 Hz within 10 seconds after the auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test;
- 6) Simulating a loss-of-offsite power in conjunction with an SI
-' n'tractuation test signal, and tower actuation test signal ~(TA),'
and a) Verifying deenergization of the emergency busses and load shedding from the emergency busses; b) Verifying the diesel starts on the auto-start signal,
. energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the emergency power sequencer 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 shall be maintained at 4160 + 420 volts and 60 + 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except engine overspeed, low lube oil pressure, 4160-volt bus fault, and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concur-rent with a Safety Injection actuation signal.
- 7) Verifying the diesel generator operates for at least 24 hours.
During the firrt 2 hours:of this test, the diesel generator shall be loaded to greater than or equal to 6697 kW and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 6083 kW. The gererator ,-3 voltage and frequency shall be 4160 1 420 volts and 60 1 1.2 Hz within 10 seconds after the start signal; the steady-state V generator voltage and MAlu31988 SEABROOK - UNIT 1 3/4 8-5 m mm
) 2 a
,. ELECTRICAL POWER SYSTEMS -
t s V SURVEILLANCE REQUIREMENTS (Continued) , r , i frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, perform-Spec ~Tf-icetion 4 8 Ale 2f 6) bht Sec 3 n5erT.EE
- 8) Verifying that the auto-connected loads to each diesel generator do not exceed thi short time rating of 6697 kW;
- 9) Verify 1. , the diesel generator's capability to: c a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standay status.
- 10) Verifying that with the diesel generator operating in a test
.. .c; ode, connected to its bus, a simulated Safety Injection signal - ---
overrides the test made by: (1) returning the diesel generator
, to standby operation, and (2) automatically energizing the U,. emergency loads with offsite power;
- 11) Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day tank of each diesel via the in-stalled cross-connection lines;
~
- 12) Verifying that the emerge $cy power sequence timer is OhERABLE ----
.with the interval between each load block within + 10% of its ~
design interval;
- 13) Verifying that the following diesel generator lcckout features prevent diesel generator starting:
a) Barring device engaged, or b) Differential lockout relay.
~14}
-Veri 4ying that-wi-H-aH-diesel-generator air _ star.t-receivers
-pressurized-to-less-than-or-equal-to-600 ps-fg and the com--
-pressors securedy-the-diesel generator starts-at-least- -
time s-f rom-a mb i e n t-co n d iti o ns_and _acce l e ra te s-to-a t-l e a st.
-900-rpm -i n-l e s s-t h a n-o r-e q u a l-t o--10-s eco nd s .
7 a,
- 7 % arc.
*If Speci-f4 cat-ton-478-1-1-2f-6-)b-) -ts not satisfactorily completed, it is not O necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated at 6083 kW for 1 hour or until operating temperature has stabilized. 7 gg SEABROOK - UNIT 1 3/4 8-6
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.. signal,.ana reachatyated_volkage., uuLx w w!A, ad Svegu encys 60 rd.LH3 ,_tolfkA /0 x and1.
- b. Verd y %_d t$seL-geneexice_is muually sy ekromjed ,laided A.-
l y e&%.se_epd_s joe.tk w_a kn_naag adA i __ 120 secondsi.and.op erales for_yedec.%_se-.epae/fv f_/m/E4/cf. O. 4
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ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Cortinued) .
. I" p[15) Viirifying that the undervoltage load shed relays are bypassed during diesel generator -load. load sequencing, and that the load shedding bypass is reinstated after the EPS RM0 pushbutton is depressed. ~
1576-) Simulating a Tower Actuation signal (TA) while the diesel generator is loaded with the permanently connected loads and auto-connected emergency (accident) loads, and verifying that the, service water pump automatically trips, the cooling tower pump and fan automatically starts. After energization the steady state voltage and frequency of the emergency buses shall be maintained at 4160 1 420 volts and 60 1.2 Hz.
/6 174 While diesel generator 1A is loaded with the permanently connected loads and auto connected emergency (accident) loads, manually connect the 1500 hp star tup feedwater pump to 4160 volt bus E5. After energization the steady state voltage and frequency of the emergency bus shall be maintained at 4160 1 420 volts and 60 1 1.2 Hz.
,,..o,,y...... g. At least once per 10 years or' after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 514 rpm in less than or equal to 10 seconds; and
- h. At least once per 10 years by:
. _ _ _ 71) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hy'pochiorite solution, and
- 2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure. .
4.8.1.1.3 Reoorts - All diesel gener'ator failures, valid or nonvalid, shall N be reported to the Commission in a Special Report pursuant to Specification 6.,12 within 30 days. Reports of diesel generator failures shall include the informa-tion recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revi-sion 1, August 1977. If the number of failures in the last 20 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. -
/
~
O MAR 131986
' T t\ c b /3]lq$30'q[
(!
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3/4 8-7 SEABROOK - UNIT 1 ,s.3
TABLE 4.8-1 N . DIESEL GENERATOR TEST SCHEDULE . NUMBER OF FAILURESTIN 1 LAST 20 VALID TESTS
- TEST FREQUENCY
<1 At least once per 31 days
>2 At least once per 7 days **
a e e O
. V.
- Criteria for determining number of failures and number of valid tests shall !
be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the number of tests are determined on a per diesel generator basis. For the purpdse of this schedule, only valid tests conducted after the completion of the preoperational test requirements of Regulatcry Guide 1.108, Revision 1, August 1977, shall be included in the computation of the "Last 20 Valid Tests."
**This test frequency shall be maintained until seven consecutive failure free
('~#) demands have been performed and the number of failures in th lid demands has been reduced to one or less. t @l J %6 SEABROOK - UNIT 1 3/4 8-8 . gg i
- 1. - a .A . ..
ELECTRICAL POWER SYSTEMS A.C. SOURCES , E SHUTDOWN --
.' g LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE: ,.
- a. One circuit between the offsite transmission network and the Ons'te 01 ass IE Distribution System, and
- b. One diesel generator with:
- 1) A day fuel tank containing a minimum fuel volume fraction of 3/8 (600 gallons of fuel),
/0 /000
- 2) A fuel storage system containing a minimum volume of -J5 000 7 gallons of fuel,
- 3) A fuel transfer pump, 4[ -Lubricating oil storage containing a minimum total volume of p 275'~ga11ons of lubricating oil, and O
- 5) Capability to transfer lubricating oil from storage to the diesel generator unit.
APPLICABILITY: MODES 5 and 6. ACTION: - With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiatec' fuel, or crane operation with loads over the fuel storage pool, and within 8 hours, depressurize an /.53 vent the Reactor Coolant System through a greater than or equal to 3. st;uare inch vent. In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required , sources to OPERABLE status as soon as possible. l SURVEILLANCE REOUIREMENTS 4.8.1.2 The above te' quired A.C. electrical power sources shall be demonstrated OPERABLE by the performanc.e of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.6)), and 4.8.1.1.3. 4 SEABROOK - UNIT 1 3/4 8-9 MAR 131986r AJ~NEx I..O.A 'l
!.RM11 l- _ _ _ _ _
* ' ~- s
3/4.8.2 D.C. SOURCES C,, , OPERATING . LIMITING CONDITION-FOR OPERATION . t 3.8.2.1 As a mininum, the following D.C. electrical sources shall be OPERABLE: S ee Inse.rt E
-a . 125-voit-Battery-Bank-No.-1ATand-its-associated-full-capacity l -charger r-and- ,
t
- t. 125 wolt Battery Bank-No.-1B ra n d-i t s-a s s oc ia te d -f u ll-ca p a c i.ty_ g G' . charger, and-
- c. 125-volt-Battery-Bank-No:-ICpnd fts associated-ful-1-capacity
.chargerrand < _d. -l-25-vol t-Battery-Bank-No--1Drand-its associate ~d-full-capacity, I i chargerr l APPLICABILITY: MODES 1, 2, 3, and 4. -
1 ACTION: . In herYrain
~
-Se mimmm d.. With one of--the required battery banks and/or-fuM-capac4ty-chargers inoperable, restore the inoperable battery bank and/or-fuM-capac4ty-charger to OPERABLE (q
_/ status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
' " Cf * "AAIS demondvafc.%e
- b. OPEdA61LLTY W M ene d ~b re%" L*c5. -fd\ capachhy Percha"6C"> fora l^j ReF"*'d 8
of LIS AS50Clafe.d b% ha.A 6a*veillanet 4 9 2..\.a..-\ wh one hour , aw6 at ,lsasi ence __ _f;iA.qor y A .hkM in Ta.ble. 4 9 '2. is stoi mc+i cleclaic~th~e ~ - beevyper hour T Inopera6 . SURVEILLANCE REOUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The parameters in Table 4.8-2 meet the Category A limits, and
- 2) The total battery terminal voltage is greater than or equal to 128 volts on float charge.
QQ JAAR a { l956] SEABROCK - UNIT 1 3/4 8-10 %ar -
__Q .Tnscri W -
--. G.-_7kvn A : I._)25 selt__haHery. .had, _No J&.De_/c.
L_6ne.lalLeapacdy. baNety chargeun bus HA.anJ 3.6ne_Ca}Leapacdy_baNcry ckrgean basJic. b Tm.in_B'. I- 1Wvolt-_baHeey hank. NoL161R ID l-2.A n e_ Cal Lcapaciiy_ hah ev y_ckeg eunkdD.18.s.d R. anehlLcapcdy._ha&<yAargrun has LID _ O -
, ~.
e O MAR 311986
* * * ' "- em m _ _ _ _ _ _ _ _
D. C. SOURCES p, V
- SURVEILLANCE REOUIREMENTS (Continued) . .
- b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, of battery overcharge with battery terminal voltage above 150 volts, by verifying that:
- 1) The parameters in Table 4.8-2 meet the Category B limits, l 2) There is no visible corrosion at either terminals or connectors, l c- or the connection resistance of these items is less than 150 x 10 8 ohm,* and (4 cel\s per roN
- 3) The average electrolyte temperature of 16 connected cells +is
. above 65 F.
- c. At least once per 18 months by verifying that 1
- 1) The cells, cell plates, and battery racks show no visual I indication of physical damage or abnormal deterioration, l 2) The cell-to-cell and ter:iiinal connections are clean, tight, and i coated with anticorrosion material, O
i L) 3) The resistance of each cell-to-cell and terminal contaction is l less than or equal to 150 x 10 8 ohm,* and
- 4) Each battery charger will supply at least 150 amperes at a minimum of 132 volts for at least 8 hours.
~ - - - - . - _ . -
- d. At least once per 18 months, during shutdown, by veri'fying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design j duty cycle when the battery is subjected to a bittery service test;
- e. 'At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when l subjected to a performance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and
- f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation:is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.
rm * ()M _0btained by subtracting 1) the cross room rack connector 1400 x 10 8the chm,normal resistance typical) and 2) theof: bi-level rack connector (50 x 10 8 ohm, typical) from the measured cell-to-cell conection resistance. 3/4 8-11 i SEA ROOK - UNIT 1 MA 3 M...E g U .\
~.. <. ) h; 3 h. T:i
. . . ~
TABLE 4.8-2
, BATTERY SURVEILLANCE REOUIREMENTS .
. i' CATEGORY A(1) CATEGORY B(2)
PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level e indication mark, indication mark, plates, and < %" above and < \" above and not maximum level maximum level overflowing indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts (6) > 2.07 volts Not more than
~
0.020 below the ( average of all Specific > 1.200 5) > 1.195 connected cells H _. . ar avity(4) --
.. ;3_ ._
l Average of all Average of all connected cells connected cells (o ,/ 1.205
> 1.195( )
TABLE NOTATIONS (1) For a~ny Category A parameter (s) outside the limit (s) shown, the battery p- may be considered OPERABLE provided that within 24 hours all the Cate-gory B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days. (2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B. parameters are within their allowe.ble values and provided the Category B parameter (s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery. (4) Corrected for electrolyte temperature and level. (5) Or battery charging current is less than 2 amps when on float charge. (6) Corrected for ' average electrolyte temperature. ' A SEABROOK - UNIT 1 3/4 8-12 i h6?
.s
D.C. SOURCES SHUTDOWN
\ .
LIMITING CONDITION FOR OPERATION
~
l b 3.8.2.2 As a minimum, one 125-volt battery bank and itsta_ssociated full-capacitycharge]hallbeOPERABLE. APPLICABILITY: MODES 5 and 6. gTION: With the required battery bank and/or fu11-capacit9 charger inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required battery bank and full-capacity charger to OPERABLE status as soon as possible, and within 8 hours, depressurize and vent the Reactor Coolant System through a 3 d square inch vent.
/SF SURVEILLANCE REOUIREMENTS 4.8.2.2 The above required 125-volt battery bank and full-capacity charge [
O sheii be eemoastreted OeERABLE in accordence with Spectricatioa 4.8.2.1 O e m$
/
O :
]D k$k M3hA8ER tm l SEABROOK - UNIT 1 3/4 8-13 -
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3/4.8.3 ONSITE POWER DISTRIBUTION p Q , OPERATING . LIMITING CONDITION FOR OPERATION , 3.8.3.1 The following electrical busses shall be energized _in the specified manner:
- a. Train A A.C. Emergency Busses consisting of:
- 1) 4160-Volt Emergency Bus #E5 -
- 2) 480-Volt Emergency Bus #E51 N, nd-
- 3) 480-Volt Emergency Bus #E52.4 -
c
- b. Train B A.C. Emergency Busses consisting of:
. 1) 4160-Volt Emergency Bus #E6,
- 2) '480-Volt Emergency Bus #E61 **
- 3) 480-VoltEmergencyBus#E62Mnd-
- 4) 480-Volt Emergency Bus #E64.
- c. 120-Volt A.C. Vital Panel #1A energized from its associated inverter connected to D.C. Bus #11A,*
- d. 120-Volt A.C. Vital Panel #18 energized from its associated inverter connected to D.C. Bus #11B,* - .
- e. 120-Volt A.C. Vital Panel #1C energized from its associated inverter connected to D.C. Bus #11C,*
' f. 120-Volt A.C. Vital Panel #1D energized from its associated inverter connected to D.C. Bus #110,*
- g. 120-volt A.C. Vital Panel #1E energized from its associated inverter connected to D.C. Bus #11A.*
- h. 120-volt A.C. Vital Panel #1F energized f rom its assoc.f ated inverter -
"connicted to D.C. Bus #118.* ~ ~
- 1. Train A 125-volt D.C. Busses consisting of:
- 1. 125-volt D.C. Bus #11A energized from Battery Bank 1A or IC.
- 2. 125-volt D.C. Bus #11C energized from Battery Bank 1A or IC. .
- j. Train B 125-volt D.C. Busses consisting of: ,
- 1. 125-volt D.C. Bus #11B energized from Battery Bank 18 or 10. ,
- 2. 125-volt D.C. Bus #11D energized from Battery Bank 1B or 10.
*Two inverters may be disconnected from their D.C. bus for up to 24 hours as necessary, for the . purpose of performing an equalizing charge on their associated battery bank provided: (1) their vital busses are energized, and (2) the vital busses associated with the other battery bank are energized from their U,3 associated inverters and connected to their associated D.C. bus. ,
CI 565-f% hese buses cw be consQevel opem.ble. @ b %gO vol r- bT6 [m #
}D :} Ti These b% fte3 wiU bc uMev ahnmsf ydbe codvol5 ,
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~ R 3Al9861 SEABROOK - UNIT 1 3/4 8-14 .1 H
@R 131886 - I ' -
O oustTe rowea Dtsrateurio" 7 1 LIMITING CONDITION FOR OPERATION APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the required trains of A.C. emergency busses not fully energized, reenergize the train within 8 hours or be in at c least HOT STANDBY within the next 6 hours and in COLD SH WDOWN within the following 30 hours.
- b. With one A.C. vital panel either not energized from its associated inverter, or with the inverter not connected to its associated D.C.
bus: (1) reenergize the A.C. vital panel within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and (2) reenergize the A.C. vital panel from its associated inverter connected to its associated D.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. one. e
- c. WithoneD.C.busnot'energijedfrom'itsassociatedbatterybanh z
(9 reenergize the D.C. bus from%i t s associated battery banhithin 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses and panels shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses. pa MAR 131886
, S %% * "
h iSLR 3,1198,6 1'? 3 h N. E l 2 !.j'.L' p '
~
SEABRC0K UNIT 1 3/4 8-15 --
3 ONSITE POWER DISTRIBUTION (G SHUTDOWN t
- T . .
LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:
- a. One train of A.C. emergency busses consisting of the 4160-volt and the 480-volt A.C. emergency busses listed in 3.8.3.1(a) and (b)
(excluding 480-volt Emergency Bus #E64) e ' oHhe Out-l y
- b. Two 120-volt A.C. vital Panels 1A, 18, 1C and ID energized from their associated inverters connected to their respective D.C. busses, and
- c. One assocta of.edghe f.wo inverter 120-volt-toA.C.
ccutnected Vital Panels i t r espective DC bu.s.1E or IF <nerkel-from t
- d. One 125-volt D.C. bus energized from4its,associatedbatterybah (TN ofl APPLICABILITY MODES 5 and 6.
ACTION: rO With any of the above required electrical busses and panels not energized in'the
" required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses and panels in the specified manner as soon as possible, and within 8 hours,' depressurize and vent the RCS
~ through at.least a $r2: square inch vent. ---
'),58 '
SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses and panels shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses. MM131886
, n, . - - - , p7 i, i
SEABROOK - UNIT 1 3/4 8-16 k j. h $ l \000 l
'- <t l ys .L 4 .
Tbl5 $f e c- an 4 ' b* ELECTRICAL PO!.T.R SYSTEMS add ed . j ONSITE P0tre.R DISTRIBUTION TRIP CIRCUIT FOR INVERTER I-2A , (- \ r - LIMITINC CONDITION FOR OPERATION 3.8.3.3 The safety related trip circuit which trips the D.C. feed f rom D.C. Bus 11C to inverter I-2A af ter 15 minutes of discharge from the battery shall be OPERABLE. Note - this LIMITING CONDITION FOR OPERATION is applicable only when D'.C. Bus'11C is required to be OPERABLE. APPLICABILITY: tiODEk 1, 2, 3, 4, 5 and 6 ACTION:
.. With this safety related trip circuit inoperable, restore the trip circuit to i OPERABLE status within 7 days or de-enargize the D.C. f eed to inverter I-2A by tripping the D.C. circuit breaker in D.C. Bus llc. Verify that this breaker is i open once per 7 days thereaf ter.
4 5 SURVEILLANCE REQUIREMENTS i 4.8.3.3 The safety related trip circuit shall be demonstrated operable at least i once per 18 months. . O c:(t l 1 I 1 - i I : j MAR 311986 4 3/4 8- 16 t (
! . *~ . .
j
3/4.8.4 ELECTRICAL EOUIPMENT PROTECTIVE DEVICES O-A.C. CIRCUITS INSIDE PRIMARY CONTAINMENT
. 1 LIMITING CONDITION FOR OPERATION 3.8.4.1 The circuit breakers feeding the following loads inside primary con-tainment shall be padlocked in the open position: .
I-ends dircu i f. Fanel Refueling Canal Skimmer Pump - 1-SF-P-272 l- E D-Mcc - Il J Polar Gantry Crane I- MM-CR-3 t -E D-us-ii s C Distribution Panel 1- ED-PP-7A l- E D - "5 - H Distribution Panel I i-Ep PP-78 l - E D. up23 Rod Control Cluster Change Fixture 1-FH-RE 22 1- E P-Ncc-ill 12.
' APPLICABILITY: MODES 1, 2, and 3.
ACTION: With any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel (s) within 1 hour. SURVEILLANCE REOUIREMENTS Ver't0l d It*S+ once Per 3 \ acts -thr tu ct'rewit- b wher3 listeJ above S 4.n tockca th-tb open pa si+1 bo , 4.8.4.1 -Eachaf-the-above-required-A-C- circuits shah-be-determined-to-be-deenergi-zed-at-least-once-per-24 hour-s*--by-verifying-that-the associated e-i rcuit-b rea ke rs- a re-i n-the-tr ipped-co nd i t-i o n-
~.
loads **e we guired fo y ExcEPTrou - T-f aq d -ib abouementioned
- l offrol1C"I s b*ief dedh (.nal- fo exend 72 koes) duri,ng P ant k pehne.& ct#cu.it breder can h e. unlocked awd be closed 0"
% regutred dedian providej -this chamge 0s areaker Pos tba'i keccmes pav4 of tG oppiscab(1 eperaf t'ng procedue med &
-the wor!t l4sth fo rd ai nmedt .
t i O *Except-at-least-once per-31-days-i f-locked,--sealedror-othe rwi se-s ecured-in-
-the-tripped -cond LLion.
MAR-13 I888 3/4 8-17 Qf. MAR <. / 31C &BT-3 j mh SEABROOK - UNIT 1 1 \ d aAxr _____ _ _ _ __ - _____-__ _ _ _ __ r
r" ELECTRICAL EQUIPMENT PROTECTIVE DEVICES (
~
CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES -[ !
~
LIMITING CONDITION FOR OPERATION - 3.8.4.2 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be OPERABLE. ,. APPLICABILITY: MODES 1, 2, 3, and 4. ( ACTION: With one or more of the containment penetration conductor overcurrent protective device (s) given in Table 3.8-1 inoperable:
- a. Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated circuit breaker or racking out or removing the inoperable protective device within 72 hours, .
declare the affected system or component inoperable, and verify the circuit breaker to be tripped or the inoperable protective device to be racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their circuit breakers tripped, or their inoperable protective devices racked out, or removed, or
- b. Be in at least HOT STANDBY within the next 6 hours and in MLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS --,- 4.8.4.2 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERABLE:
- a. At least once per 18 months:
- 1) By verifying that the medium voltage 13.8 kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers, and performing the following:
a) A CHANNEL CALIBRATION of the af sociated protective relays, s ee 7 nser1- y_ b) An integrated system functional test which includes' simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designedp -a nd - ygg 33 ;888 O f+ueswum%s % s ee>.+ m r ( Twc 2.v.-i y___ fbybM S 311985lff
\ b O SEABROOK - UNIT 1 3/4 8-18 4
O [ INSERT V r - (due to the large currents involved, it is impractical to inject pri=ary side signals to current transformers. Therefore, the channel calibration will be performed by injecting a signal on the secondary side of these transfor:ers at their test plug), and O (
**O N eh - . - m S
O ( MAR 311986
() ELECTRICAL EQUIPMENT PROTECTIVE DEVICES
- . r
~
SURVEILLANCE REQUfREMENTS (Continued) c) 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. See InSyt ZE See'Chsoi- VIT
.2y_ selecting and-functionally tes. ting _a_ representative-sample 2) of t least 10% of each type of lower voltage circuit br,eafers.
Circu l breakers selected for functional testing shall-be selectedson a rotating basis. Testing of these circu'it breakers sla,11 consist of injecting a current wi,th'a value equal to 300Lof the pickup of the long-time delay trip of the pickup of the short-t'ime delay trip element and e,lement, and ver150'g'ifying that the circuit breaker operates
~
wjthinthetimedelaybandwidthforthatcurrentspecified the manufacturer htet instantaneous element shall be by\stedbyinfectingacurrentequa1to120%ofthepickup te value of the element and
-trips-instantaneously with verifyin,g,that_the.
n~o~f6tentional time circuit-breaker delay. Molde d
\
ng shall also follow this procedurd cakecircuitbreakertest)morethantwoAripelements, except that generally no time Circuit breakers delayandinstantaneous,willbeinvolved.hhallberestoredto found inoperable dufing functional testing
~
OPERABLE status-p'rior to resuming operation. Yqr each circuit bre'aker foun,d' Inoperable during these functionalMests, an - "- additional-representative sample of at least 10% ofg all the cidcui,tdreakersoftheinoperabletypeshallal,sobegunction- f ally tested until no more failures are found or all circuit b@rs_of--that-type hafe bekn~fu~nEtTHHil19~feMed; and
- b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedure's prepared in conjunction with its manufacturer's recommendations.
MAR 131986 O- . t I.
.,u IM. A.R 3119.86
- ia t SEABROOK - U?llT 1 3/4 8-19 [~_
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[ ' devie shall_be-__veplaceLot rchainha+ed .No anhana.I Seshing - f I ts h 4 td ted . e
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i T r\ s e t Y M l \.
- 2. By selecting and functionally testing a representative sampic of at least 10% of each type of lower voltage circuit breakers and overload devices.
-/ Circuit breakers and overload devices selected for functional testing shall be selected on a_ rotating basis. . I-
~
1 1 Testing of air circuit breakers shall consist of injecting a current with a i value equal to 300% of the pickup of the long-time delay trip clement and
- 150% of the pickup of the short-time delay trip element. 'The instantaneous element shall be tes ted by injecting a current equal to +20% of the pickup
} value of the element. Testing of thermal magnetic molded-case circuit breakers shall consist of
- inecting p current with a value equal to 300% of the circuit breaker trip rating and -25% to +40% of the circuit breaker instantaneous trip range or
, setpoint.
i Testing of combina tion s tarters (a magne tic only molded-case circuit breaker l in series with a motor starter and integral overload device) shall consist of injecting a current with a value equal to -25% to +40% of the circuit
- breaker ins tantaneous trip setpoint and, 200% and 300% of the thermal overload device trip rating to the respective devices. '
The above circuit breakers and overload devices shall be verified operable by tripping at a time value which does not exceed the penetra tion withstand time as shown in Table 3.8-1. i () Circuit breakers and or overload devices found inoperable during functionti tasting shall be restored to OPERABLE status prior to resuming operation, For each circuit breaker and or overload devices found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers and or overload devices of the inoperable type i shall also be functionally tested until no more failures are found or all
, _ circuit breakers and or overload devices of that type have been functionally ! te s ted. If the trip time does not exceed the penetration withstand time but is outside the verification time listed in Table 3.8-1, the device shall be replaced or recalibrated. No additional testing is required.
- 3. Corrective actions for any generic degradation of overcurrent protective
, devices, such as setpoint drif t, manufacturing deficiencies, ma terial 2
defects, e tc. , shall be applicable to all (Class 1E and non-Class 1E) pro-tective devices of identical design. i e . l I ) 1 i 1 O . MAR 311936 l
a= A8sa -s9 4 a4 M-es e- w+ 3aAJu-- s--44w,, a~n-- - e4w,_.m, _sn a -- _s, ,,e_ _ , , ,., ,_ ,u_ _ _ , , _ _ , _ _ _ . _ _ _ _ _ _ _ _ , _ __ ___ _ _ __ _ i
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! ), , .h.. l N 1 e < ... !O. i 1 l l MAR 131986 3/4-20 SEA O'* "' M ~ U N ' l
~
' yeuleo) aad bl<ms e. AofC N tYcrm5 ln th)'s tim e are Mclev*
Wi!.l be Ck Wl)l Ec -fcrwed ed A6 bY Ef **Cs a ccLU a. M e % 1 % . Up & d c5 (' TABLE 3.8-1 d CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT -
. PROTECTIVE DEVICES (CONTINUED) .
CT Secondary Verification Penetration Device Number Test Current Time Withstand Time System and Location (amos) (seconds) - (Seconds) Powered Instant Instant Instant I Lonotime Lonotime Lonotime I. 13.8 kV l e e
- a. Circuit Breakers
- 1. Bus 1 - Overcurrent Trio Devices - IAC Relays Incoming Line 40* .37 .43 8 RC Breaker UAT-X2A (S) 12 3.7-4.3 90 Incoming Line 40* .37 .43 8 RC Breaker RAT-X3A (S) 12 3.7-4.3 90
._ ... Reactor. Coolant Pump 28.5-31.5 0 .047 __600 ___ _ _. RC RC-P-1A Feeder .. 15 26-30 1000 Breaker (P)
Reactor Coolant Pump 28.5-31.5 0 .047 600 RC RC-P-18 Feeder 15 26-30 1000 Breaker (P) .
- 2. Bus 2 - Overcurrent Trio Devices - IAC Relays
~
Incoming Line 40* .37 .4'3 8 liC' Breaker UAT-X2B (S) 12 3.7-4.3 90 Incoming Line 40* .37 .43 8 RC Breaker RAT-X38 (S) 12 3.7-4.3 90
~
Reactor Coolant Pump 28.5-31.5 0 .047 600 RC RC-P-IC Feeder 15 26-30 1000
, Breaker (P)
Reactor Coolant Pump 28.5-31.5 0 .047 600 RC RC-P-10 Feeder 15 26-30 1000 Breaker (P)
- 3. The opening response time to a drip signal for all 13.8 kV circuit breakers should be less than 0.042 seconds .for the verification time purposes. The sum of the overcurrent device response time and the circuit breaker opening response time must be less than the penetration withstand time.
ote: hlh0I (P) - Primary ~ _ _ _ _ - m (S) - Backup / Secondary 1- _.__, (*) - Short-Time Value i p
- 6 _ 4 1 h ._ 3, SEABROOK - UNIT 1 3/4 8-21 _fjD [.i'1ARE a n. 19% .L. t:w --
. , , p y_ -- ,- , ,pp - , - , - . . - . . . , - - - ,. --..i-Ty*-r
l i l fs TABLE 3.8-1 (_,), CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) -
. . I I. 13.8 kV (Continued)
- b. Fuses Resistance System (Micro-ohms) Powered Bus 1 c Reactor Coolant Pump 48-64 RC RC-P-1A Fuses Reactor Coolant Pump 48-64 RC RC-P-18 Fuses Bus 2 Reactor Coolant Pump 48-64 RC ""' ~~
RC-P-1C Fuses ~ ~ ~ - Reactor Coolant Pump 48-64 RC (}) RC-P-ID Fuses Test Verification Penetration Setpoint Time Withstand Time System
. (amos) (seconds) (Seconds) Powered Instant Instant Instant -.
Longtime
~
Longtime Longtime II. 480 V
- a. Unit Substations Bus E53 Secondary 9600* .48 .84 20 CAH Breaker (S) 4800 10-28 90 Containment Structure 3000-4500 0 .070 100 CAH Cooling Fan CAH-FN-1C (P) 990 10-28 1000 Containment Structure 3000-4500 0 .070 100 CAH Cooling Fan CAH-FN-1E (P) 990 -
10-28 1000 Containment Structure 2160-3240 0 .080 250 CAH Cooling Fan CAH-FN-IF (P) 990 10-28 1000 O MAR 131086 SEABROOK - UNIT 1 3/4 8-22 \ -
O T^8'e 3 8-1 CONTAINMENT PENETRATION CDNDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) , ,_- Test Verification ~ Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant Longtime - Longtime Longtime c II. 480 V (Continued)
- a. Unit Substations (Continued)
Bus E63 Seccndary 9600* .48 .84 20 CAH Breaker (S) 4800 10-28 90 Containment Structure 3000-4500 0 .070 100 CAH Cooling Fan CAH-FN-1A (P) 990 .10-28 2000 Containment Structure 2160-3240 0 .080 ~ 250 CAH
'C661ing Fan CAH-FN-1B (P) 900 . 10-28 1000 Containment Structure 2160-3240 0 .080 250 CAH
]' Cooling Fan CAH-FN-10 (P) 900 10-28 1000
- e. so
- m
/
O un1ams l I
.., g p rm
'I jj',hM./.R'1.1.1986 _I.L l SEABROOK - UNIT 1 3/4 8-23 Il ,
-~.
O TABLE 3.8-1 CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT
, PROTECTIVE DEVICES (CONTINUED) ,
Test Verification Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant Overload ,. Overload Overload e III. MOTOR CONTROL CENTERS
- a. Tyoe HE3-Thermal Maonetic Circuit Breaker
!!CC-E512 Containment Structure 450-1400 0 .067 14 CAH Recirculating Filter Fan 150 18-75 1000 CAH-FN-3A MCC-E515 n Thermal Barrier PCCW 450-1400 0 .067 14 CC V Recirculating Pump 150 18-75 1000 CC-P-322A (P)
MCC-E531 l Control Rod Drive ~450-1400 0 .067 ~ ~ 14 CAH " Mechanism Cooling Fan 150 18-75 1000 CAH-FN-2A (P) - Control Rod Drive 450-1400 0 .067 14 CAH Mechanism Cooling Fan 150 18-75 1000 CAH-FN-2C (P) Containment Building Air 450-1400 0 .067 14 SA Compressor 120 18-75 1000 SA-C-4A (P) Lighting Transformer 450-1400 0 .067 200 ED ED-X-16H Feeder (P) 300 18-75 1000 Lighting Transformer 450-1400 0 .067 200 ED ED-X-16F Feeder (S) 300 18-75 1000 MCC-ES22 , O Accem. Tx. en outiet Iso. Valve SI-V3 (P) 450-1400 120 0 .067 18-75 14 1000 st Accum. Tk. 9C outlet Iso. 450-1400 0 .067 14 SI Valve SI-V32 (P) 120 18 '5 ----10 U --, - SEABROOK - UNIT 1 3/4 8-24 j g 1 k;,l3 jjgg ,. pS w
O . TABLE 3.8-1 . CONT,AINMENT PENETRATION CONDUCTOR OVERCURRENT . ! PROTECTIVE DEVICES (CONTINUED) l Test Verification P_enetration System Setpoint Time Withstand Time (Seconds) Powered (amps) (seconds) Instant Instant Instant Overload ' Overload Overload e c III. MOTOR CONTROL CENTERS (Continued) l
- a. Type HE3-Thermal Macnetic Circuit Breaker MCC-E622 0 .067 14 SI Accum. Tk. 9B Outlet Iso. 450-1400 18-75 1000 Valve SI-V17 (P) 120 0 .067 14 SI
-Accum. Tk. 9D Outlet Iso. 450-1400 18-75 1000 Valve SI-V47 (P) 120 .
O MCC-E612 450-1400 0 .067 14 CAH Containment Structure 18-75 1030 Recirc. Filter Fan 150 CAH-FN-3Ej (P) CC-P-3228 (P) ^ - . . . . MCC-E615 " 14 CC 450-1400 0 .067 Thermal Barrier PCCW 18-75 1000 Recirculating Pump 150 CC-P-3223 (P) MCC-631 14 CAH 450-1400 0 .067 Control Rod Drive 18-75 1000 Mechanism Cooling Fan 150 CAH-FN-28 (P) 14 CAH 450-1400 . 0 .067 Control Rod Drive 18-75 1000 Mechanism Cooling Fan 150 : CAH-FN-20 (P) 14 SA Containment Building Air 450-1400 0 .067 18-75 1000 Compressor SA-C-4B (P) 120 ED 450-1400 0 067 200 Lighting Transformer 18-75 1000 300 EO-X-16A Feeder (P) 0 .067 00 R11198dp rg' a Lighting Transformer 450-1400 "\ L\ l
--SR 13 fe86 E5 A 18-75 1 300 ED-X-16A Feeder (5) 3/4 8-25 SEA 8 ROOK - UNIT 1
~T TABLE 3.8-1 (V CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT
, PROTECTIVE DEVICES (CONTINUED) ,
g Test Verification Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant Overload ,. Overload Overload C' III. MOTOR CONTROL CENTERS (Continued)
- a. Type HE3-Thermal Macnetic Circuit Breaker MCC-Ill RC-P-1A Oil Lift 300-980 0 .067 40 RC Pump RC-P-229A 90 5.6-52 1000 RC-P-1B Oil Lift 300-980 0 .067 40 RC
' Pump RC-P-2298 90 5.6-52 -
1000 EC Drain Tank Pump 300-980 0 .067 40 WI.D h WLD-P-33A (P) 90 5.6-52 1000 RC-P-1A Motor 300-980 0-0.67 3.5 RC Space Heater (P) 45 5.6-52 1000 RC-P-lhMotor 300-980 0 .067 3.5 RC Space Heater (S) '90 5.6-52 1000 RC-P-18 Motor 300-980 0 .067 3.5 RC Space Heater (P) 90 5.6-52 1000 RC-P-18 Motor 300-980 0 .067 3.5 RC
, Space Heater (S) 90 5.6-52 - 1000 Lighting Transformer 450-1400 0 .067 3.5 'RC ED-X-16E Feeder (P) 300 18-75 1000 Lighting Transformer 450-1400 0 .067 200 ED ED-X-16E Feeder (S) 300 18-75 1000 Containment Building 300-980 0 .067 0.10 MM Personnel Air-Lock 45 5.6-52 60 MM-MM-30 (P)
ContainmentBuffding 300-980 0 .067 0.10 HM i h' Personnel Air-Lock MM-MM-30 (S) 45 5.6-52 60 I SEABROOK - UNIT 1 3/4 8-26 p~ . ,- {',}s\ MfR ( 'l 3111,986 MM Q g-
O. Te8te 3 8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) . t Test Verification . Penetration Setpoint Time Withstand Time System (amps) (seconds) (Seconds) Powered Instant Instant Instant Overload - Overload Overload f c' III. MOTOR CONTROL CENTERS (Continued)
.a . Type HE3-Thermal Maonetic Circuit Breaker MCC-111 (Continued)
Lighting Transformer 450-1400 0 .067 200 ED' ED-X-16J Feeder (P) 300 18-75 1000
. Lighting Transformer 450-1400 0 .067 .200 . .-. .ED ED-X-16J Feeder (S) . 300 18-75 1000 Incore Detector 300-980 0 .067 3.5 IC
]\ -
Drive A (P) 45 5.6-52 1000 i Incore Detector 300-980 0 .067 3. 5 IC Drive A,(S) 45 5.6-52 1000 0 .067 - 3.5
~~
-Incore Detector -
300-980 - IC Drive B (P) 45 5.6-52 1000 Incore Detector 300-980 0 .067 3.5 IC Drive B (S) 45 5.6-52 1000 Incore Detector 300-980 0 .067 3.5 IC Drive C (P) 41 5.6-52 1000 Incore Detector 300-980 0 .067 3.5 IC
- Drive C (S) 45 5.6-52 1000 MCC-231 RC-P-1C Oil Lift Pump 300-980
- 0 .067 40 RC RC-P-229C (P) 90 5.6-52 1000 RC-P-1C Oil Lift Pump 300-980 0 .067 40 RC RC-P-2290 (P) 90 5.6-52 1000 RC Drain Tank Pump B 300-980 0 .067 40 RC WLD-P-338 90 5.6-52 1000 , _ _
.g AR.lJytL)
SEABROOK - UNIT 1 3/4 8-27 b.l$$$, ,
- ~
O TABLE 3.8-1 - d CONTAINMENT PENETRATION CON 50CTCR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) , Test Verification Penetration Setpoint Time - Withstand Time System (amps) (seconds) (Seconds) Powered Instant Instant Instant Overload ,. . Overload Overload P III. MOTOR CONTROL CENTERS (Continued)
- a. Type HE3-Thermal Maonetic Circuit Breaker MCC-231 (Continued)
RC-P-1D Motor Space 300-980 . 0 .067 3.5 RC Heater (P) 45 5.6-52 1000 RC-P-1D Motor Space 300-980 0 .067 3.5 RC Heater (S) " ~ ~ - - ' 45 5.6-52 1000 RC-P-1C Motor Space 300-980 0 .067 3.5 RC O- Heater (P) 45 5.6-52 1000 RC-P-1C Motor Space 300-980 0 .067 3.5 RC Heater (5) 45 5.6-52 1000
~
Lighting Transformer __ _
~~
450-1400 - - ~ ~ 0 .067 200 ... ..ED..__.. ED-X-16F Feider (P) 300 18-75 1000 Lighting Transformer 450-1400 0 .067 200 ED ED-X-16F Feeder (5) 300 18-75 1000 Lighting Transformer 450-1400 0 .067 200 ED ED-X-16K Feeder (P) 300 18-75 1000 Lighting Transformer 450-1400 0 .067 200 'ED ED-X-16K Feeder (S) 300 18-75 1000 Incore Detector 300-980 0 .067 3.5 IC Drive D (P) 45 5.6-52 1000 Incore Detector 300-980 0 .067 3.5 IC Drive D (S) 45 5.6-52 1000 Incore Detector 300-980 0 .067 3.5 IC Drive E (P) 45 5.6-52 1000
~~ MAIL 311986 ~
- j. p .~ , ,
3/4 8-28 .2 ,/,, !b0 SEABROOK - UNIT i
~ .
O TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT ' 7 PROTECTIVE DEVICES (CONTINUED) - Test Verification Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant Overload - Overload Overload III. MOTOR CONTROL CENTERS (Continued)
- a. Type HE3-Thermal Macnetic Circuit Breaker MCC-231 (Continued)
Incore Detector 300-980 0 .067 3.5 IC Drive E (S) 45 5.6-52 1000 Incore Detector .... _.._ 300-980. 0 .067 3.5 IC - - ' Drive F (P) . 45 . 5.6-52 1000 Incore Detector . 300-980 0 .067 0 Drive F (S) 45 5.6-52 3.5 1000 IC Containment Building 300-980 0 .067 0.10 IC Personnel Air-Lock 45 5.6-52 60 MM-MM-29 (P) _ _ . _ .
~ ~
Containment Building 300-980 0 .067 0.,10 IC Personnel Air-Lock 45 5.6-52 60 MM-MM-29 (S) e O MAR 131986 SEABROOK - UNIT 1 3/4 g.29 9ll)mh..II@T 0 ~ sl A
~
b TABLE 3.8-1 CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) , ! Test Verification _ Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant III. MOTOR CONTROL CENTERS (Continued)
- b. Amcao Motor Circuit Protector MCC-E512 Containment Structure 496-925 0 .067 45 CAH Recir. Filter Fan CAH-FN-3A (S)
Regen. HT Exch Letdown 22-41 0 .067 1000 CS Iso. Valve CS-V149 (P) - - - Regen. HT Exch Letdown 22-41 0 .067 1000 CS m 1 Iso. Valve CS-V149 (S) RC-P-1A to PCCW Iso. Valve 27-30 0 .067 1000 CC CC-V428 (P)
~
RC-P-1A to PCCW Iso. Valve 27-30 0 .067 1000 CC CC-V428 (S) RC-P-10 to PCCW Iso. Valve 27-30 0 .067 100'0 CC CC-V439 (P) RC-P-10 to PCCW Iso. Valve 27-30 0 .067 1000 CC CC-V439 (S) . MCC-E515 Thermal Barrier PCCW 369-689 0 .067 150 CC Recirculating Pump CC-P-322A (S) MC-E531 : Control Rod Drive Mech. 343-640 0 .067 100 CAH Cooling Fan (s) CAH-FN-2A (v') MAR 131986 m ~,
)
7) i SEABROOK - UNIT 1 3/4 8-36 La' AR d.\lsl98B h .% 4. 3. S i.
O Taste 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT ~ PROTECTIVE DEVICES (CONTINUED) . g Test Verification _ Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant III. MOTOR CONTROL CENTERS (Continued)
- b. Amcao Motor Circuit Protector MC-E531 Control Rod Drive Mech. 343-640 0 .067 100 CAH Cooling Fan (C)
CAH-FN-2C (S) Containment Bldg Air 252-470 0 .067 1000 SA - . - Compressor SA-C-4A (S) -- . . .- RC Loop 3 Letdown to " 22-41 0 .067 1000 RC O aesea "x iso vaive RC-V81 (P) RC Loop 3 Letdown to 22-41 0 .067 1000 RC Regen. HX Iso. Valve RC-V81*(S) _ __
' ' ~ '
Letdown Control Valve 5.3-10 0 .067 1000 CS CS-HCV-189 (P) . Letdown Control Valve 5.3-10 0 .067 1000 CS CS-HCV-189 (5) MCC-E521 RC Loop 1 RHR Inlet Iso. 62-115 0 .067 1000 RC Valve RC-V23 (P) RC Loop 1 RHR Inlet Iso. 62-115 0 .067 1000 RC Valve RC-V23 (S) ' RC Loop 4 Pressurizer 39-73 0 .067 1000 RC Press. Relief Iso Valve 3C-V122 (P) , RC ' .. cop 4 Pressurizer 39-73 0 .067 1000 RC I Q Press. Relief Iso Valve _ RC-V122 (S) D& T . :"1 3/4 8-31 a f0 0 . i SEABROOK - UNIT 1 . [
O TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT . PROTECTIVE DEVICES (CONTINUED) ., ! Test Verification. Penetration Setpoint Time Withstand Time System _(amos) (seconds) (Seconds) Powered Instant Instant Instant III. MOTOR CONTROL CENTERS (Continued)
- t. Amcao Motor Circuit Protector MCC-E612 Containment Structure 496-925 0 .067 45 CAH Recirculation Filter Fan CAH-FN-38 (S)
RCP Seal Water Isolation 22-41 0 .067 1000 CS Valve CS-V168 (P) .
~
RC'P Seal Water Isolation . 22-41 0 .067 1000 CS Valve CS-V168 (S) RC-P-18 to PCCW Isolation 27-50 0 .067 1000 CC Valve CC-V395 (P) RC-P-1B to PCCW Isolation 27-50 0 .067 1000 CC Valve CC-V39,5 (S) RC-P-1C to PCCW Isolation 27-50 0 .067 1000 CC Valve CC-V438 (P) RC-P-IC to PCCW Isolation 27-50 0 .067 1000 CC Valve CC-V438 (S)
~
Reactor Vent Valve 4.7-9.0 0 .067 1000 RC RC-V323 (P) ' Reactor Vent Valve 4.7-9.0 0 .067 1000 RC RC-V323 (S) MCC-E615 Thermal Barrier PCCW 367-689 0 .067 150 CC Recirculating Pump CC-P-3228 (S) O _ .. .s -
) pI N c It_QMAg 11996 3e
'-~
~' '
T1 M M .1 198'6 - SEABROOK - UNIT 1 3/4 8-33
TABLE 3.8-1 (] CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) .,,' t Test Verification. Penetration Setpoint Time Withstand Time System l (amps) (seconds) (Seconds) Powered Instant Instant Instant III. MOTOR CONTROL CENTERS (Continued)
- b. Amcao Motor Circuit Protector MCC-E631 Control Rod Drive Mech. 343-640 0 .067 200 CAH Cooling Fan CAH-FN-28 (S)
, Control Rod Drive Mech. 343-640 0 .067 200 CAH l Cooling Fan CAH-FN-20 (S) Containment Building Air, 252-470 . 0 .067 1000 'SA Compressor SA-C-4B (S)
- Letdown Control Valve 5.3-10 0 .067 1000 CS CS-HCV-190 (P)
Letdown Control Valve 5.3-10 0 .067 1000 CS CS-HCV#190 (S) 7__.__ i MCC-E621 , RC Loop 1 RHR Inlet 67-125 0 .067 1000 RC Isolation Valve RC-V22 (P)
~
RC Loop 1 RHR Inlet 67-125 0 .067 1000 RC Isolation Valve RC-V22 (S) RC Loop 4 RHR Inlet 67-125 0 .067 1000 RC Isolation Valve RC-V87 (P) RC Loop 4 RHR Inlet 67-125 0 .067 1000 RC Isolation Valve RC-V87 (S) O a 7?EAl[31(986 O
.N . i. \. .C~A Ih .[. . SEABROOK - UNIT 1 3/4 8-34 = .-
O TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT __ PROTECTIVE DEVICES (CONTINUED) ,-, Test Verification Penetration Setpoint Time - Withstand Time System (amps) (seconds) (Seconds) Powered l Instant Instant Instant III. MOTOR CONTROL CENTERS (Continued) c'
- b. Amcao Motor Circuit Protector MCC-E621 (Continued)
RC Loop 1 Pressurizer 67-125 0 .067 1000 RC Pressure Relief Isola-tion Valve RC-V124 (P) RC Loop 1 Pressurizer 67-125 0 .067 1000 RC Pressure Relief Isola-tion Valve RC-V124 (S) Containment Purge 8.6-16 0 .067 1000 O Isolation Valve CGC CGC-V28 (P) Containment Purge 8.6-16 0 .067 1000 CGC Ist 'ation Valve CGI V28 (5) _ , ,_ MCC-111 - RC-P-1A Gil Lift Pump 129-241 0 .067 1000 RC RC-P-229A (S) . RC-P-1A Oil Lift Pump 129-241 0 .067 1000 RC RC-P-229B (S) l Containment Structure 51-95 0 .067 1000 WLD l Sump A Pump WLD-P-5A (P) ' Containment Structure 51-95 0 .067 1000 WLD Sump A Pump WLD-P-5A (S) . Containment Structure 51-95 0 .067 1000 WLD Sump B Pump WLD-P-SC (P) O MARtH986
. n, <\ Tn SEABROOK - UNIT 1 3/4 8-35 /
1 :./ d h I'-) 79 h i
{ (~' TABLE 3.8-1 V CONTAINt:ENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) ,,' ! Test Verification. Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant III. MOTOR CONTROL CENTERS (Continued) p
- b. Amcao Motor Circuit Protector l MCC-111 (Continued)
{
- Containment Structure 51-95 0 .067 1000 WLD
! Sump B Pump WLD-P-5C (S) RC Drain Tank Pump A 185-344 0 .067 1000 WLD WLD-P-33A (S) l Pressure Relief Tank 29-55 0 .067 1000 RC i Cooling Pump RC-P-271 (P) b Pressure Relief Tank Cooling Pump RC-P-271 (S) 29-55 0 .067 1000 RC Excets Letdown Heat 22-41 0 .067 1000 CC l Exchanger Valve CC-434 (P) . _ _ . Excess Letdown Heat 22-41 0 .067 1060 CC Exchanger Valve CC-434 (S) Steam Generator Blowdown 7.7-14 0 .067 ,1000 SB Stop Valve SB-V191 (P) Steam Generator Blowdown 7.7-14 0 .067 1000 SB Stop Valve SB-V191 (S) Steam Generator Blowdown 7.7-14 0 .067 1000 SB Stop Valve SB-V193 (P)
~
Steam Generator Blowdown 7.7-14 0 .067 1000 SB Stop Valve SB-V193 (S) MCC-231 - () RC-P-IC Oil Lift Pump 129-241 0 .067 1000 RC RC-P-229C (S) MAR 131886 SEABROOK - UNIT 1 3/4 8-36
) hk -AR.$l ** 198Q f } M 2 .
O TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT ,. i l PROTECTIVE DEVICES (CONTINUED) ... Test Verification ~ Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered Instant .- Instant Instant c III. MOTOR CONTROL CENTERS (Continued)
- b. Amcap Motor Circuit Protector MCC-231 (Continued)
RC-P-1C Oil Lift Pump 129-241 0 .067 1000 RC RC-P-2290 (5) Containment Structure 51-95 0 .067 1000 WLD Sump A Pump WLD-P-5B (P) Containment Structure 51-95 - 0 .067 1000 WLD Sump A Pump WLD-P-5B (S) Containment Structure 51-95 0 .067 1000 WLD Sump B Pump WLD-P-50 (P) Contairiment Structure 51-95 0 .067 1000 WLD Sump B Pump WLD-P-50 (S) - - - - RC Drain Tank Pump B 185-344 0 .067 1000 SB j WLD-P-338 (5) j Steam Generator Blowdown 7.7-14 0 .067 1000 SB Stop Valve SB-V189 (P) l Steam Generator Blowdown 7.7-14 0 .067 1000 ' SB Stop Valve SB-V189 (S) Steam Generator Blowdown 7.7-14 0 .067 1000 SB Stop Valve SB-V195 (P) Steam Generator Blowdown 7.7-14 0 .067 1000 SB f Stop Valve SB-V195 (S) : 1 L l 1 '
. O ~
._.uan ia.tsas T ., ,r a nx ,
- ~ . . .
G.. ..M/.&31; 986 l .. .- .. SEABROOK - UNIT 1 3/4 8-37 '
- ~ - - - - - -
~_
(' TABLE 3.8-1 CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT
- ~
PROTECTIVE DEVICES (CONTINUED) . Test Verification- Penetration Setpoint Time Withstand Time System (amps) (seconds) (Seconds) Powered 200% 200% 200% 300% - 300% 300% c III. MOTOR CONTROL CENTERS (Continued)
- c. Type G30T Thermal Overload Relays MCC-E512 Regen. Heat Exchanger 2.7 27-98 1000 CS Letdown Iso. Valve 4.0 12-40 1000 CS-V149 (P)
Regen. Heat Exchanger 2.7 27-98 1000 CS Letdown Isa. Valve 4.0 . 12-40 1000 CS-V149 (S) O RC-P-1A Valve to PCCW 3.6 27-98 1000 CC Isolation Valve 5.3 12-40 1000 l CC-V428 (P) RC-P-1A Valve to PCCW 3.6 27-98 1000 CC
~
Isolation Valve 5.3 -- 12-40 1000
~CC-V428 (S)
RC-P-10 Valve to PCCW 3.6 -
-- 27-98 1000 CC Isolation Valve 5.3 12-40 1000 CC-V439 (P)
RC-P-ID Valve to PCCW 3.6 27-98 ~1000 CC Isolation Valve 5.3 12-40 1000 CC-V439 (5) Containment Structure 90 27-98 1000 CAH Recirc. Filter Fan 135 12-40 1000 l CAH-FN-3A (S)
/
O MAR 131988 T^ T nyn R 3.11986; SEABROOK - UNIT 1 3/4 8-38
'h Li S. L.
O
' TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT .
PROTECTIVE DEVICES (CONTINUED) . Test Verification. Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered 200% 200% 200% 300% - 300% 300% c III. MOTOR CONTROL CENTERS (Continued)
- c. Type G30T Thermal Overload Relays MCC-E531 (Continued)
RC Loop 3 Letdown.to 2.7 27-98 1000 RC Regen. HX Isolation 4.0 12-40 1000 Valve RC-V81 P RC Loop 3 Letdowp..to . 2.7 '27-98 1000 RC. - - - Regen. HX Isolation . 4.0 . 12-40 1000 Valve RC-V81 (S) Control Rod Drive 90 27-98 1000 CAH Mech Cooling Fan 135 12-40 1000 CAH-FN-2A (S) Control" Rod Drive 90 27-98 1000 CAH
- . - . Mech Cooling Fan 135 12-40 1000 CAh-FN-26 (S) .
Containment Building Air 68 27-98 1000 SA Compressor SA-C-4A (S) 102 12-40 1000 Letdown Control Valve .8 27-98 .1000 CS CS-HCV-189 (P) 1.3 12-40 1000 Letdown Control Valve .8 27-98 1000 CS CS-HCV-189 (S) 1. 3 12-40 1000 MCC-E521 RC Loop 1 RHR Inlet Iso. 12 : 27-98 1000 RC Valve RC-V23 (P) 17 12-40 1000 RC Loop 1 RHR Inlet Iso. 12 27-98 1000 RC Valve RC-V23 (S)- 17 12-40 1000 C) MAR 131986 i~)..i.1 r; ; n nAR:3.1sl983!n m:.
.- .s. ; \ ;.L .
SEABROOK - UNIT 1 3/4 8-39 d
l TABLE 3.8-1 CONTAINMENT PENETRAIION CONOUCTOR OVERCURRENT
, __ PROTECTIVE DEVICES (CONTINUED) ,
Test Verification Penetration Setpoint Time - Withstand Time System (amos) (seconds) (Seconds) Powered 200% 200% 200% 300% .- 300% 300% e
- P III. MOTOR CONTROL CENTERS (Continued)
- c. Type G30T Thermal Overload Relays MCC-E521 (Continued)
RC Loop 4 Pressurizer 5.3 27-98 1000 RC Press. Relief Iso 8.0 12-40 1000 Valve RC-V122 (P) RC Loop 4 Pressurizer 5.3 27-98 1000 RC
- " Press. Relief Iso 8.0
~
12-40 - 1000 - Valve RC-V122 (S) - b Containment Purge Iso. VLV CGC-V14 (P) 1.1 1.6 27-98 12-40 1000 1000 CGC Containment Purge Iso. 1.1 27-98 1000 CGC VLV CGC-V14 (S) 1.6 12-40 1000 RC Loop 4 RHP. Inlet 12 27-98 1000 RC Iso Valve RC-V88 (P) 17 12-40 1000 RC Loop 4 PH3 Inle't 12 27-98 1000 RC Iso Valve RC-V88 (S) 17 12-40 1000 MCC-E612 - RCP Seal Water Iso. Valve 2.7 27-98 1000 'CS CS-V168 (P) 4.0 12-40 1000 RCP Seal Water Iso. Valve 2.7 27-98 1000 CS CS-V168 (5) 4.0 12-40 1000 RC-P-1B to PCCW Iso. 3.6 - 27-98 1000 CS Valve CS-V395 (P) 5.3 12-40 1000 i RC-P-1B to PCCW Iso. 3.6 27-98 1000 CS j
,, Valve CS-V395 (S) 5.3 12-40 1000 l r 3 )
RC-P-1C to PCCW Iso. 3.6 27-98 1000 CS Valve CS-V438 (P) 5.3 12-40 1000 RC-P-1C to PCCW Iso. 3.6 27 - _1000 C5 Valve CS-V438 (S) 5.3 12-40 m .,_,10170 : 3/4 8-40 ! ' 1 gggMARlkl SEABROOK - UNIT 1 3
. s s .\ m 986ll
Table 3.8-1 Containment Penetration Conductor (V) Overcurrent Protective Devices (continued) . Test verification Pe[!a tra, tion Setpoint Time - Withstand Time System
. (amps) (seconds) (seconds) Powered 200% 200% 200%
300% ,
, 300% 300%
III. MOTOR CONTROL CENTERS (Cont'd.)
- C'
- c. Type c30T Ther al overload Relays -
MCC-E612 (Cont'd.) Contaircent Structure Recire. 90 27-98 1000 CAH Filter Fan CAH-FN-3B 135 12-40 1000 Reactor Vent Valve .8 27-98 1000 RC RC-V323 (P) 1.3 12-40 1000 Reacto~r Vent Valve ' ~ .8 27-98 10 0 0 ' ' '~~" ~ ~ ' RC-RC-Y323 (S) - 1.3 12-40 1000 O b MCC-E631 Control Rod Drive Mech. Cooling 90 27-98 1000 CAM Fan CAH-FN-2B (S) 135 12-40 1000 Control Rod Drive Mech. Cooling 90 27-98 1000 - CAM Tan CAH-FN-2D (S), 135 12-40 1000 Contairment Bldg. Air 68 27-98 1000 SA Co= pressor SA-C-4B (S) 102 12-40 1000 Letdown Control Valve .8 27-98 1000 CS CS-HCV-190 (P) 1.3 12-40 - 1000
~
Letdown Control Valve .8 27-98 1000 CS CS-HCV-190 (S) 1.3 12-40 1000 MCC-621 RC Loop 1 RHR Inlet Iso. 12 27-98 1000 RC Valve RC-V22 (P) 17 I 12-40 1000 RC Loop 1 RHR Inlet Iso. 12 27-98 1000 RC Valve RC-V22 (S) 17 12-40 1000 O hfAR 13 Iggs s u gg..< va.r i w F-wu ) AR[1}9 31
TABLE 3.8-1 () CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) . Test Verification Penetration Setpoint Time . Withstand T'me System (amas) (seconds) (Seconds) Powered 200% 200% 200% 300% 300% 300% III. MOTOR CONTROL CENTERS (Continued)
- c. Type G30T Thermal Overload Relays MCC-E621 (Continued) ,
4: I 1: t; I Cil Dc. 2. > .?7-oo 1000 CS VP;c C" ',4:0 (3) 5.3 22-10 100: RC Loop 1 Pressurizer 5.3 2';-98 1000 RC Relief Iso Valve 8.0 12-40 1000 RC-V124 (P) RC Loop 1 Pressurizer 5.3 27-98 1000 RC rN Relief Iso Valve 8.0 12-40 1000 V RC-V124 (S) RC' Loop RHR Inlet Iso. 12 27-98 1000 RC Valve RC-V87 (P) 17 12-40 1000 RC Loop RHR Inlet Iso. 12 27-98 1000 _RC
- ~
Valve RC-V87 (S) 17 12-40 1Q00
~
Containment Purge Iso. 1.1 27-98 1000 CGC Valve CGC-V28 (P) 1.6 12-40 1000 Containment Purge Iso. 1.1 27-98 1000 CGC Valve CGC-V28 (S) 1.6 12-40 1000 MCC-111 Containment Structure 10 27-98 1000 WLD Sump A Pump WLD-P-5A (P) 15 12-40 1000 Containment Structure 10 27-98 1000 WLD Sump A Pump WLD-P-5A (5) 15 12-40 1000 Containment Structure 10 27-98 1000 WLD Sump B Pump WLD-P-5C (P) 15 12-40 1000 m (,/ Containment Structure 10 27-98 1000 WLD Sump B Pump WLO-P-SC (S) 15 12-40 1000 SEABROOK - UNIT 1 3/4 8-41 D, '
'l N W I : MAR p.11986j,177 MAft la 1886
^d 1
TABLE 3.8-1 d,s CONTAINMENT PENETRATION CON 00CTOR OVERCURRENT
' PROTECTIVE DEVICES (CONTINUED) .
. .. t Test Verification Penetration Setpoint Time _ Withstand Time System (amos) (seconds) (Seconds) Powered 200% 200% 200%
300% 300% 300% III. MOTOR CONTROL CENTERS (Continued)
- c. Tyoe G30T Thermal Overload Relays Pressure Relief Tank 7.2 27-98 1000 RC Cooling Pump RC-P-271 (P) 11 12-40 1000 Pressure Relief Tank 7.2 27-98 1000 RC Cooling Pump RC-P-271 (S) 11 12-40 1000 RC-P-1A Oil Lift Pump 35 27-98 1000 RC RC-P-229A.(S).__ 52 12-40 1000 ... . .
~
RC-P-1A Oil Lift Pump 35 27-98 1000 RC C RC-P-2298 (S) 52 12-40 1000 Excess Letdown Heat Exch 2.7 27-98 1000 CC Valve CC-V434 (P) 4.0 12-40 1000 Excess-Letdown Heat Exch 2.7 ~ 27-98 1000 CC Valve CC-V434 (S) 4.0 12-40 ~ 1000 - _. Reactor Coolant Drain 48 27-98 1000 WLD Tank Pump WLD-P-33A (1) 71 Steam Gen. Blowdown Stop 1.1 27-98 1000 SB Valve SB-V191 (P) 1.6 12-40 ,1000 , Steam Gen. Blowdown Stop 1.1 27-98 1000 SB Valve SB-V191 (S) 1. 6 12-40 1000 Steam Gen. Blowdown Stop 1.1 27-98 1000 SB Valve SB-V193 (P) 1.6 12-40 1000 Steam Gen. Blowdown Stop 1.1 : 27-98 1000 SB Valve SB-V193 (S) 1.5 12-40 _ 1000 MCC-231 , Containment Structure 10 27-98 1000 WLD f} v . Sump A Pump WLD-P-58 (P) 15 12-40 1000 MAR 131988_ n r1 , :n SEABROOK - UNIT 1 3/4 8-42 4
;i .
'p jggy ;m[
j .14 - .1 1
l (# TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT . PROTECTIVE DEVICES (CONTINUED) - Test Verification. Penetration Setpoint Time Withstand Time System (amos) (seconds) (Seconds) Powered 200% 200% 200% 300% < 300% 300% e III. MOTOR CCNTROL CENTERS (Continued)
- c. Tyoe G30T Thermal Overload Relays MCC-231 (Continued)
Containment Structure 10 27-98 1000 WLD Sump A Pump WLD-P-5B (S) 15 12-40 1000 Containment Structure 10 27-98 1000 WLD Sump B Pump WLD-P_50 (P) 15 12-40 1000 . . . _ . . Containment Structure 10 27-98 1000 WLD Sump B Pump WLD-P-50 (S) 15 12-40 1000 RC-P-1C Oil Lift Pump 35 27-98 1000 RC RC-P-229C (S) 52 12-40 1000 RC-P-1D Oil Lift Pump 35 ~~ 27-98 1000 RC RC-P-2290~(S) - 52 12-40 1000 - Reactor Coolant Drain 48 27-98 1000 WLD Tank Pump WLD-P-338 (S) 71 12-40 1000 Steam Gen. Blowdown Stop 1.1 27-98 1000 SB Valve SB-V189 (P) 1.6 12-40 ,1000
~
Steam Gen. Blowdown Stop 1.1 27-98 1000 SB Valve SB-V189 (S) 1.6 12-40 1000 Steam Gen. Blowdown Stop 1.1 27-98 1000 SB Valve SB-V195 (P) 1.6 12-40 1000 Steam Gen. Blowdown Stop 1.1 : 27-98 1000 SB Valve SB-V195 (S) 1.6 12-40 1000 MCC-E515 ,r} Thermal Barrier PCCW 98 27-98 1000 CC 'v Recirculating Pump 146 12-40 1000 CC-P-322A (S) c MAR 171988 - , SEABROOK - UNIT 1 3/4 8-43 7O h M1/-[ ; i l
- 1/ 5.MARJ1\l98 i
-a - n, t ;
{
q TABLE 3.8-1 V CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) - 1 Test Verification Penetration Setpoint Time - Withstand Time System (amos) (seconds) (Seconds) Powered i 200% 200% 200% 300% , 300% 300% III. MOTOR CONTROL CSNTERS (Continued) e
- c. Type G30T Thermal Overload Relays MCC-E522 Accumulator Tank 9A Outlet 43 27-98 1000 SI Iso. Valve SI-V3 (S) 64 12-40 1000 Accumulator Tank 9C Outlet 43 27-98 1000 SI Iso. Valve SI-V32 (S) 64 12-40 1000 MCC-E615 - --
O Tnermei Berrier eCCw 98 27-98 1000 CC Recirculating Pump 146 12-40 1000 - CC-P-3228 (S) MCC-E622
~ ~ ~
Accumulator Tank 98 43 27-98 - 1000 - SI'-- Outlet Isolation 64 12-40 1000 Valve SI-V17 (S) Accumulator Tank 90 43 27-98 1000 SI Outlet Isolation 64 12-40 . 1000 Valve SI-V47 (S) - MARJ31ggg ,
-;m -
,1 - - - , ,
l SEABROOK - UNIT 1 3/4 8-44 AR3$986fY
TABLE 3.8-1 - I_I CONTAINMENT PENETRAfl0N CONOUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) , Test Verification Penetration Setpoint Time - Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant __ Instant Overload ,
, Overload Overload P
III. MOTOR CONTROL CENTERS (Continued) C d Type JL Thermal Macnetic MCC-E521
~ '
Hydrogen Recombiner 563-1050 0 .067 600 CGC CGC-MM-284A (P) 375 32-160 1000 Hydrogen Recombiner 563-1050 0 .067 600 CGC CGC-MM-284A (S) 375 32-160 1000 MCC-E621 {J Hydrogen Recombiner CGC-MM-2848 (P) 563-1050 375 0 .067 32-160 600 1000 CGC Hydrogen Recombiner 563-1050 0 .067 600 CGC CGC-MM-J848 (S) 375 32-160 1000 e
' - MAR 131986 O
me J n -3m i
~
" aMARJ1\l98lil t
SEABROOK - UNIT 1 3/4 8-45 ,
TABLE 3.8-1 O_ CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT
- PROTECTIVE DEVICES (CONTINUED) 7 ..
Test Verification Penetration Setpoint Time - Withstand Time System (amos) .(seccnds) (Seconds) Powered Instant Instant Instant Overload , Overload Overload IV. PRESSURIZER HEATERS (Continued) e Backuo Groun A - RC-PP-6A
' Type FJ-Thermal Macnetic Heaters 1, 2 and 22 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 He es 1, 2 and 22 563-1050 0 .067 600 RC acondary Breaker, (S) 375 28-170 1000 Heaters 5, 6 and 27 --563-1050 ' 0 .067 600 RC Primary Breaker (P) .- 375 28-170 1000
/_T Heaters 5, 6 and 27 563-1050 . 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 9, 10 and 32 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 - Heaters 9, 10 and 32 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 13, 14 and 37 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 RC Heaters 13, 14 and 37 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 17, 18 and 42 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 Heaters 17, 18 and 42 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Backuo Group C - RC-PP-6C Heaters 51,52and24 563-1050 0 .067 600 RC t' Primary Breaker (P) 375 28-170 1000 '] MAR 131886
--~3 n 9 T i
/\ l } P~i.~Q :
SEABROOK - UNIT 1 3/4 8-46 D) ]'i,:) 1
;jg j,
.r. -. - - -
,,y __ - , , - , - -, - - . , - - - - - - - , ,,w-
Q ~ TABLE 3.8-1 CONTAINMENT PENETRATION CON 00CTOR OVERCURRENT . PROTECTIVE DEVICES (CONTINUED) . Test Verification Penetration Setpoint Time Withstand Time System (amps) (seconds) (Secor.ds) Powered Instant Instant Instant Overioad - Overload Overload
, e IV. PRESSURIZER HEATERS (Continued)
, Backuo Grouc C - RC-PP-6C (Continued)
Heaters 51, 52 and 24 563-1050 0 .067 6CO RC Secondary Breaker (S) 375 28-170 1000 Heaters 57, 58 and 29 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 ,. , Heaters 57, 58 and 29 563-1050 0 .067 600 RC Secondary Breaker (S) - 375 - 28-170 1000 Heaters 63, 64 and 34 563-1050 0 .067 600 RC O- . Primary Breaker (P) 375 28-170 1000 Heaters 63, 64 and 34 563-1050 0 .067 600 RC Secondary Breakers (S) 375 28-170 1000 Heaters 69, 70 and 39 563-1050 - - - - 0 .067~ 600 RC~ ~ Primary Breaker (P) 375 28-170 1000 Heaters 69, 70 and 39 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 75, 76 and 44 563-1050 0 .067 600 RC Primary Breaker (P) - 375 28-170 1000 Heaters 75, 76 and 44 563-1050 0 .067 600 RC Secondary Breater (S) .375 28-170 1000 Control Grouc - RC-PP-6E Heaters 47, 48 and 21 563-1050. 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 Heaters 47, 48 and 21 563-1050 0 .067 600 RC Secondary Breaker'(S) 375 28-170 1000 ! O Heaters 53, 54 and 26 Primary Breaker (P) 563-1050 375 0 .067 28-170 600 1000 RC MAR 131986 , 77 5 m r.- SEABROOK - UNIT 1 3/4 8-47 j s RA3gS86!.f
..h L, N ',7 i
; TABLE 3.8-1 v' CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT
__ PROTECTIVE DEVICES (CONTINUED) ,,' Test Verification Penetration Setpoint Time - Withstana Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant Overload .- Overload Overload IV. PRESSURIZERHEATERS(Continued) Control Grouc - RC-PP-6E (Continued) Heaters 53, 54 and 26 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 59, 60 and 31 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 Heaters 59, 60 and 31 563-1050 0 .067 600 RC
~ Ssc6ndary Breakers (S) . 375 28-170~'~~~ 1000 ~
r Heaters 65, 66 and 36 563-1050 0 .067 600 RC V Primary Breakers (P) 375 28-170 1000 Heaters 65, 66 and 36 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 71, 72 and_41 _ ;_. 563-1050 ~ 0 . 067 - - 600.. RC Primary Breakers (P) 375 28-170 1000 Heaters 71, 72 and 41 563-1050 0 .067 600 -- RC Secondary Breaker (S) 375 28-170 1000 Heaters 77, 78 and 46 565-1050 0 .067 600 RC Primary Breakers (P) 375 28-170 - 1000 Heaters 77, 78 and 46 563-1050 0 .067 600 'RC Secondary Breaker (S) 375 28-170 1000 Backuo Group D - RC-PP-6D Heaters 3, 4 and 25 563-1050 0 .067 600 RC Primary Breakers (P) 375 28-170 1000 Heaters 3, 4 and 25 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 (y V Heaters 7, 8 and 30 563-1050 0 .067 600 RC Primary Breakers (P) 375 28-170 1000 1 Th 731ARjlIBT SEABROOK - UNIT 1 3/4 8-48 h/ ,ing ,4, g ;; d I
- -ME]R86 -
p TABLE 3.8-1 V CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT . PROTECTIVE DEVICES (CONTINUED) . Test Verification - Penetration Setpoint Time Withstand Time System , (amps) (seconds) Fowered (Seconds) Instant Instant Instant Overload .- Overload Overload e IV. PRESSURIZER HEATER 3 (Continued) Backun Grouc D - RC-PP-60 (Continued) Heaters 7, 8 and 30 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 11, 12, and 35 563-1050 0 .067 600 RC Primary Breaker (P) 357 28-170 1000 Heaters 11, 12 and 35 563-1050 0 .067 600 . RC Secondary- Breaker TS) . 375 - 28-170 1000 fs Heaters 15, 16 and 40 563-1050 0-067 600 RC U Primary Breaker (P) 375 28-170 1000 Heaters 15, 16 and 40 563-1050 0 .067 600 RC Secondary Breakers (S) 375 28-170 1000 Heaters 19,_29 and_45 563-1050 0 .067 600 - - - - - RC Primary Breaker (P) 375 28-170 1000 Heaters 19, 20 and 45 563-1050 0 .067 ~ 600 RC Secondary Breaker (S) 375 28-170 1000 Backuo Grouc 8 - RC-PP-68 Heaters 49, 50 and 23 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 Heaters 49, 50 and 23 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Heaters 55, 56 and 28 563-1050 0 .067 600 RC Primary Breaker (P) 375
~
28-170 1000 Heaters 55, 56 and 28 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 ! O weeters 81. 82 ema 23 ss3-1oso o .os7 soo RC Primary Breaker (P) 375 28-170 1000
- - - MAR 131986 -
,.? ... - , .\ ,-
-{
SEABROOK - UNIT 1 3/4 8-49 -
$ -} '
TABLE 3.8-1 V CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) Test Verification Penetration Setpoint Time - Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant Overload , Overload Overload IV. PRESSURI[ERHEATERS(Continued) Backuo Groua B - RC-PP-6B (Continued) Heaters 61, 62 and 33 563-1050 0 .067 600 RC Secondary Breaker (S) 375 28-170 1000 Control Grouc 0 - RC-PP-6D Heaters 67, 68 and 38 563-1050 0 .067 600 RC l l Primary Breaker (P) 375 28-170 1000
- e. _ _.
Heaters 67, 68 and 38 " 563-1050 ' 0 .067 600 RC
,e Secondary Breaker (S) 375 28-170 1000 Heaters 73, 74 and 43 563-1050 0 .067 600 RC Primary Breaker (P) 375 28-170 1000 .
1 Heaters 73, 74 and 43 563-1050 0 .067 600 RC __ Secondary Breaker (S) 375 28-170 1000 __ . (125 V dc & 120 V ac)
- Ty'ce E2 Thermal Macnetic 125 V dc Distr. Panel 111A Circuit #4 (P) 300-980 0 .067 0.6 ,CAH 45 4.5-38 1000 Circuit #6 (P) 300-080 0 .067 0.6 NG 60 4.5-38 1000 125 V dc Distr. Panel 112A Circuit #2 (P) 300-980 0 .067 0.6 RH 60 4.5-38 1000 Circuit #7 (P) 300-980 0 .067 0.6 SI
( ) 60 4.5-38 1000 Circuit #19 (P) 300-980 0 .067 0.6 RC 60 4.5-38 - ---100 0 g SEABROOK - UNIT 1 3/4 8-50 I'
\ -MA,R la,1886 m 77 il)]NhRN1k0$] .
I O TABLE 3.8-1
\ CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) -
Test Verification Penetration Setpoint Time - Withstand Time System , (amos) (seconds) (Seconds) Powered Instant Instant Instant Overload , Overload Overload V. LOW VOLTAGE CIRCUIT BREAKERS s' (125 V dc & 120 V ac) 125 V de Distr. Panel 113A Circuit #4 (P) 300-980 0 .067 0.6 CC - Circuit #5 (P) 300-980 0 .067 0.6 RC 60 4.5-38 1000 Circuit #7 (P) . 300-980 . 0 .067 0.6 SI 60 4.5-38 1000 n
'G Type E2 Thermal Maanetic 125 V de Distr. Panel 1118 Circuit #4 (P) 300-980 0-0.67 0.6 CAH 45 . . . .
4.5-38 1000 Circuit #6 (P) 300-980 0 .067 0.6 NG 60 4.5-38 1000 Circuit #15 (P) 300-980 0 .067 0.6 CAH 45 4.5-38 1000 125 V dc Lichtina Distr. Panel XL10 Circuit #20 (P) 450-1260 0 .067 300 ED 150 4.5-38 1000 Circuit #20 (S) 450-1260 ' 0 .067 300 ED 150 4.5-38 1000 tO MAirqiggg_ _
! g ' MAR 3.11986 '
' ' ~ ~ ~ ' " "
SEABROOK - UNIT 1 3/4 8-51 i- k
/ \,
s' TABLE 3.8-1 _ CONTAINMENT PENETRAT10ti C0fiDUCTOR OVERCURRENT , 3
- ~
. ~
PROTECTIVE DEVICES (CONTINUED) Test Verification Penetration Setpoint Time Withstand Time System
~(amos) (seconds) (Seconds) Powered Instant . Instant Instant Overload Overload Overload c
V. LOW VOLTAGE CIRCUIT BREAKERS (Continued) (125 V dc & 120 V ac) Tyoe E2 Thermal Macnetic 125 V dc Distr. Panel 1128 Circuit #1 (P) 300-980 0 .067 0.6 RC 60 4.5-38 1000 Circuit #2 (P) 300-980 . 0 .067 0.6 RH
~ ~ 60 4.5-38 1000 0 Circuit #3 (P) 300-980 0 .057 0.6 C0P 60 4.5-38 1000 Circuit #5.(P) 300-980 0 .067 0.6 WLD 60 4.5-38 1000 _
Circuit #7 (P) . 300-980 0 .067 0.6 SI 60 4.5-38 1000 Circuit #15 (P) 300-980 0 .067 0.6 VG 60 4.5-38 1000 Circuit #16 (P) 300-980 0 .067 0.6 .RC 60 4.5-38 1000 Circuit #19 (P) 300-980 0 .067 0.6 RC 60 4.5-39 1000 125 V dc Distr. Panel 1138 Circuit #4 (P) 300-980 0 .067 0.6 CC 60 4.5-38 1000 m Circuit #5 (P) . 300-980 0 .067 0.6 RC ( _) 60 4.5-38 1000 Circuit #7 (P) 300-980 0 0 SI
., . s. :- 7-7 '
SEABROOK - UNIT 1 3/4 8-52 31p6 y-j ' ' !
/~N TABLE 3.8-1 O CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT
__ PROTECTIVE DEVICES (CONTINUED) , g Test Verification Penetration Setpoint Time Withstand Time System
, (amps) (seconds) (Seconds) Powered Instant Instant Instant Overload .- Overload Overload ?
e V. LOW VOLTAGE CIRCUIT BREAKERS (Continued) (125 V dc & 120 V ac) Type E2 Thermal Maanetic 120 V ac Vital Instrument
> Distr. Panel 1E Circuit #3 (S) 300-980 0 .067 0.6 ML 45 4.5-38._ ._ . _.1000 Circuit #9 (P) - 300-980 ' 0 .067 0.6 SI 45 4.5-38 1000 (o'_) '
Circuit #16 (P) 300-980 0 .067 0.6 RC 45 4.5-38 1000 120 V ac Vital Instrument Distr. Panel lf - - - - - Circuit #3 (S) 300-980 0 .067 0.6 ML 45 4.5-38 1000 Circuit #9 (P) 300-980 0 .067 0.6 SI 45 4.5-38 , 1000 Circuit #16 (P) 300-980 0 .067 O. 6 .RC , 45 4.5-38 1000 Type 80 Thermal Macnetic 120 V ac Vital Instrument , Distr. Panel llE . Circuit #3 (P) 263-700 0-067 1.2 RM : 45 3.5-38 1000 G V MAR.131986
)
\
D f T1 'm ._l ' D Ph0S SEABROOK - UNIT 1 3/4 8-53 L II)M - L
TABLE 3.8-1 (~) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT
._ PROTECTIVE DEVICES (CONTINUED) , ,' ;
Test Verification Penetration Setpoi.nt Time - Withstand Time System (amos) (seconds) (Seconds) Powered Instant Instant Instant Overload . Overload Overload c V. LOW VOLTAGE CIRCUIT BREAKERS (Continued) (125 V dc & 120 V ac) 120 V ac Vital Instrument . Distr. Panel 11F Circuit #3 (P) 263-700 0 .067 1.2 RM 45 3.5-38 1000 Tyoe BQ Thermal Macnetic MCC-111, 120 V ac Distr. Panel Circuit #1 (S) 135-308 0 .067 7. 0 SB/SF/ 45 6.5-45 1000 CC Circuit #2 (S) 135-308 0 .067 7.0 WLD
. 45 6.5-45 1000 Circuit #12 (S) 263-700 0 .067 1.2 CAH 60 3.5-38 1000 Circuit #21 (5) 135-308 0 .067 7. 0 RC 45 6.5-45 1000 Circuit #28 (P) 263-700 0 .067 ~ 1. 2 IC 105 6.4-45 1000 Circuit #31 (P) 135-308 0 .067 7.0 RMW 45 6.5-45 1000 MCC-231, 120 V ac Distr. Panel Circuit #4 (P) 135-308 0 .067 7.0 SB/WLD 45 6.5-45 1000 Circuit #5 (P) , 263-700 0 .067 1.2 IC q 105 6.5-45 1000 k~/
Circuit #14 (P) 263-700 0 .067 1.2 MARL 3\$80IC 105 6.5-45 1000 p, ; q ;3. 'r, ,' i :) SEABRuGK - UNIT 1 3/4 8-54 1 M R s3'U986 .d
/ _Ls 1\. 3.
fs TABLE 3.8-1 t
, ) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (CONTINUED) r -
Test Verification Penetration Setpoint Time Withstand Time System (amps) (seconds) (Seconds) Powered Instant Instant Instant Overload Overload Overload V. LOW VOLTAGE CIRCUIT BREAKER 3 (Continued) e (125 V dc & 120 V ac)
~ Type BQ Thermal Magnetic MCC-E512, 120 V ac Distr. Panel Circuit #14 (S) 135-308 0 .067 7.0 CC/CAH 45 6.5-45 1000 Cirguit #15 (S) 135-308 0 .067 7.0 CC
' 45 6.5-45 1000 MCC-E521, 120 V ac Distr. Panel Circuit #7 (P) 135-308 0 .067 .7. 0 SI 45 6.5-45 1000
-. Circuit #10 (P) 135-308 0 .067 7.0 CC
.J.J: . _"~ 45 -
- .6.5-45 1000 Circuit #13 (S) 135-308 0 .067 7.0 CGC 45 6.5-45 1000 MCC-E531, 120 V ac Distr. Panel Circuit #2 (S) 135-308 0 .067 - 7.0 RC/CS 45 6.5-45 1000 SA/CAH Circuit #11 (P) 135-308 0 .067 7.0 CC 45 6.5-45 1000 Type BQ Thermal Magnetic MCC-E612, 120 V ac Distr. Panel Circuit #13 (5) , 135-308 0 .067 7.0 CAH 45 6.5-45 1000 Circuit #14 (S) 135-308 0 .067 7.0 RC 45 6.5-45 1000
-, MAR 1.33986 t
.h /\ I SEABROOK - UNIT 1 3/4 8-55 :
I {'i _LMeB10.98h. 1.
TABLE 3.8-1 hs CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT
. PROTECTIVE DEVICES (CONTINUED) -
r - Test Verification Penetration Setpoint Time . Withstand Time System (amps) (seconds) (Seconds) Powered Instant Instant Instant Overload Overload Overload
- V. LOW VOLTAGE CIRCUIT BREAKERS (Continugd)
(125 V dc & 120 V ac) Tyoe BQ Thermal Magnetic MCC-E612, 120 V ac Distr. Panel (Cont'd) Circuit #15 (S) 135-308 .0 .067 7.0 CC 45 -6.5-45 1000 MCC-E621, 120 V ac Distr. Panel , Circuit #4 (P) 1 5-308 06 SI Circuit #6 (P) 135-308 0 .067 7.0 CC 45 6.5-45 1000
' - ~~ Circuit #13 (5) 235-308 0 .067 7.0 CGC 45 --- 6.5-45 1000 MCC-E631, 120 V ac Distr. Panel Circuit #1 (S) 135-308 0 .067 7.0 CS/CAH 45 6.5-45 1000 Circuit #2 (S) '135-308 0 .067 " 7.0 SA 45 6.45-45 1000 -
Circuit #10 (P) 135-308 0 .067 7.0 CC 45 6.5-45 1000 Type BQ Thermal Magnetic 120/240 V ac Distr. Panel PP-8C Circuit #5 (S) ,. 263-700 0 .067 1.2 CAH 45 3.5-38 1000
-. 131986 m l SEABROOK - UNIT 1 3/4 8-56 (1MAR l
i'M
.. a kR(1\980i g
;L 1 21 n l l
l
j TABLE 3.8-1 O- - CONTAINsEnT eEneraA110N CON 0uC10a OveaCuaaEnr . PROTECTIVE DEVICES (CONTINUED) .. 7 - Test Verification Penetration
. Setpoint Time - Withstand Time System (amps) (seconds) (Seconds) Powered Instant Instant Instant Overload Overload Overload 4
V. LOW VOLTAGE CIRCUIT BREAKERS (Continued) MCC-E515, 120 V ac Distr. Panel Circuit #13 (S) 135-308 0 .067 7.0 CC 45 65-45 1000 4 MCC-E615, 120 V ac Distr. Panel Circuit #13 (S) 135-308 0 .067 7.0 45 6.5-45 1000 1 O. 1 1 s
/
O . MAR 131986 ov "d" n ra ii SEABROOK - UNIT 1 3/4 8-57 1V .fi'MARi~
.k .i 1986
- 21. Ja
4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES G MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION . LIMITING CONDITION FOR OPERATION 3.8.4.3 The thermal overload protection of each valve given in Table 3.8-2 shall be OPERABLE. APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE. ACTION: With the thermal overload protection for one or more of the above required valves inoperableT bypass the inoperable thermal overload within 8 hours"; restore the inoperable thermal overload to OPERABLE status within 30 days or declare the affected valve (s) inoperable and apply the appropriate ACTION Statement (s) for the affected system (s). SURVEILLANCE REOUIREMENTS
~ " --"4:8.74'.3'"The thermal overload protection for the above required valves shall be demonstrated OPERABLE at least once per 18 months and following maintenance _ @-
on the motor starter by the-performance-of-a-CHANNEL--CAL-IBRAT-IONof a represen-p V tative sample of at least 25% of all thermal ov.erloads for the above required valvesad repktctg nem St% P'e cAhloaded devices.
~
~
- A hm.d ovesload p olecf[o,t deuke ts consedered inoperable tT -the vdky opeta.Hy %e ts nof b1 accordane. cat % the c fert% esinblished th Ae applica.bW relay curve. '
g.g cuedad deutdes u>ktc4 pmvih ,penchaf[on profe.fdy per gectg'ca/rd f 3647. sha.it mot he h Passed O -MAR 131986 -- n, i , 4 1 <
.n) J7 u,
SEABROOK - UNIT 1 3/4 8- a l
TABLE 3.8-2 . ." i MOTOR OPERATED VALVES WITH TERMAL OVERLOAD PROTECTION DEVICES OPERABLE AT ALL TIMES Valve Number Function / System MS-V204 Main Steam Isolation Bypass c MS-V205 " Main Steam Isolation Bypass MS-V206 Main Steam Isolation Bypass - MS-V207 Main Steam Isolation Bypass MS-V22 Reactor Coolant Loop 1 (RHR) . RC-V23 Reactor Coolant Loop 1 (RHR) RC-V87 Reactor Coolant Loop 4 (RHR) RC-V88 Reactor Coolant Loop 4 (RHR) O RC-v122 Reector Cooient eressurizer RC-V124 Reactor Coolant Pressurizer RH-V35 . Residual Heat Removal
~ ~
RH-V22 Residual Heat Removal . RH-V14 Residual Heat Removal RH-V70 Residual Heat Removal
~
RH-V36 Residual Heat Removal RH-V21 Residual Her.t hemoval ~ RC-V323 Reactor Vessel Head Vent RC-V26 Residual Heat Removal : , RC-V32 Residual Heat Removal R-PCV-611 Residual Heat Removal R-PCV-610 Residual Heat Removal O SI-V3 Safety Injection Accumulators MAR 131986 7 ys *^,1 i
. . A.B ? ) .B.H_. ,js SEABROOK - UNIT 1 3/4 8-58 i . . _ . . _ . ,
i G V TABLE 3.8-2 (Continued) , ,, , Valve Number Function / System - SI-V17 Safety Injection Accumulators - SI-V32 Safety Injectica Accumulators SI-V47 Safety Injection Acc'umulators c , CHS-V47 Safety Injection Cold Leg ' CHS-V49 Safety Injection Cold Leg CS-V460 Safety Injection Cold Leg CS-V461 Safety Injection Cold Leg CS-V475 Safety Injection Cold Leg
- CS-V90 Safety Injection Cold Leg
. . ~ , . .. . .
. .--.-s..~~-
SI-V112 Safety' Infection C'old Leg O SI-v114 Sefety infection Coid teo CRS-V51 Safety Injection Cold Leg CRS-V53 . Safety Injection Cold Leg
~
SI-V89 Safety Injection Cold Leg ,
"~
FW-FV-4214A Feedwater FW-FV-42148 Feedwater FW-FY-4224A Feedwater - FW-FY-42248 Feedwater FW-FV-4234A Feedwater FW-FV-4234B Feedwater FW-FV-4244A Feedwater FW-FY-42448 Feedwater SI-V93 Safety Injection O SI-V111 Safety Injection _-g gg ' 1 Q }il f p f* ' ij 3/4 8-59 jj_y t- " A SEABROOK - UNIT 1 b e.h 31486 P a
- ~
TABLE 3.8-2 (Cont.inued) .. i Valve Number Function / System CS-V65 Boren Injection - CS-V66 Boron Injection t SI-V138 Baron Injection l c SI-V139 Baron Injection " SI-V102 Baron Injection SI-V77 Boron Injection CS-V142 Chemical and Volume Control CS-V143 Chemical and Volume Control CS-V149 Chemical an_d Volume Control
~
CS-V154 Chemica1 and Volume Control O CS-v158 Chemice, end voiume Contrei CS-V162 Chemical and Volume Control CS-V166 - Chemical and Volume Control
~ ~ ~ ~ ~ ~
CS-V167 Chemical and Volume Control . CS-V168 Chemical and Volume Control CS-LCV-1173 Chemical and Volume Control CS-LCV-112C Chemical and Volume Control - CS-V844 Baron Injection CS-V845 Baron Injection CS-V846 Boron Injection . CS-V847 Baron Injection ' CS-V196 Chemical and Volume Control CS-V197 ' Chemical and Volume Control CS-V426 Chemical and Volume Control - Eemergency Boration . MAR 31198.6._._ ... . . i. SEABROOK - UNIT 1 3/4 8-60 g . MAR 131986 j i
L TABLE 3.8-2 (Continued)
' .- i Valve Number Function / System
- SW-V54 Service Water _
SW-V25 Service Water SW-V4 Service Water - SW-V5 Service Water SW-V74 Service Water SW-V76 Service Water SW-V19 Service Water SW-V2 Service Water SW-V22 Service Water SW-V29 Service' Water SW-V31 Service Water []) SW-V20 Service Water SW-V27 Service Water SW-V56 Service Water . SW-V23 Service Water . SW-V34 Service Water SW-V15 Service Water . SW-V17 Service Water SW-V26 Service Water SW-V55 Service Water CGC-V14 Combustible Gas Control CGC-V28 Combustible Gas Control CBS-V38 ECCS/CS Fluid Supplies O CBS-vu cCCS/CS Fluid Supplies _ _ _ . . MAJ{.131886 ---- Th. n ,'\, j 7 ~1 7]^l i i , SEABROOK - UNIT 1 3/4 8-61 , ./ -MAR 3fl986'.1 i-
G O . TABLE 3.8-2 (Continued)
. t*
Valve Number , Function / System CS-LCV-1120 ECCS/CS Fluid Supplies , CS-LCV-112E ECCS/CS Fluid Supplies . CBS-V2 ECCS/CS Fluid Supplies CBS-V5 ECCS/CS Fluid Supplies CBS-V8 ECCS/CS Fluid Supplies C85-V14 ECCS/CS Fluid Supplies CBS-V11 ECCS/CS Fluid Supplies CBS-V17 ECCS/CS Fluid Supplies CC-V266 Primary Component Cooling Loop B CC-V272 Primary' Component Cooling Loop B CC-V145 Primary Component Cooling CC-V137 Primary Component Cooling CC-V395 Primary Component Cooling Thermal Barrier - CC-V428 Primary Component Cooling Thermal Barrier - CC-V438 Primary Component Cooling Th.ermal Barrier CC-V439 Primary Component Cooling Thermal Barrier CC-V1092 Primary Component Cooling Thermal Barri,er CC-V1095 . Primary Component Cooling Thermal Barrier - CC-V1101 Primary Component Cooling Thermal Barrier CC-V1109 Primary component Cooling Thermal Barrier I 1
. JAR 13.1986 _. ;
g .7 - c. - SEABROOK - UNIT 1 3/4 8-62
!.(AL3.]._8.
h
3/4.9 REFUEL.ING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITI0Id FOR OPERATION 5 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met; I either: '
- a. A K,7f of 0.95 or less, or
- b. A boron concentration of greater than or equal to C000 ppm.
APPLICABILITY: MODE 6.* ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its is reduced to less than or equal to 0.95 or the boron equivalentuntilK[(oredtogreaterthanorequalto2000 ppm concentrationisrI
~
, whichever is Q the more restrictive.
SURVEILLANCE REOUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity condi'tions shall be determined prior to:
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full- %ngth control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours. , , ._ ,
-4 : 971--3 -Val ve's-I s o l a ti o n - o f-unb o ra te d -wa te r-s o u rc e s -s ha l 1-be -ve ri fi ed - c l o s ed --
and_ secured-in position ~by mechanical stops-or-by-removal-of-air-or-electrical-
-power atrieast-once pfr=31rdaysn
*The reactor- shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
.- - m u aissa 3/4 9-1 Q T--
SEABROOK - UNIT 1 l } l.
- l. L J . L :. ,. .
REFUELING OPERATIONS E4.9.2 INSTRUMENTATION LIMITING CONDITICId FOR OPERATION l 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERA 8LE, each with continuous visual indication in the control room and one with audible indication in the containment and control room. , APPLICABILITY: MODE 6. c. ACTION: a. With one of the above required monitors inoperable or=not-operating,- l
-immediately-suspend-all-operations-involving-v ce CORE'5ALTERATIONSg-0 gg lg nts: men r se contra7l roam posit-ive-reactivity ,s e +e changes lpor,nke. ;Fa. bnfom? 7 t.T& $, A T/D MS o r POSITIVE .RfAC.To ysTY,ssaewsy kver ymmesida1 <uspend coM b.
ith both of the above required monitors inoperable or not operating, determine the baron concentration of the Reactor Coolant System at least once per 12 hours. O SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:
- a. A CHANNEL CHECK at least once per 12 hours,
- b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
- c. Ar. ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
O MAR 131986 iD A SEABROOK - UNIT 1 3/4 9-2 D 'j.a y.... ,. 1 QIrg P]j,
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. i. _.
REFUELING OPERATIONS DECAY TIME O . 3/4.9.3 , LIMITINGCONDITIO5FOROPERATION 3.9.3 The reactor shall be subcritical for at least 100 hours. APPLICABILITY: During movement of irradiated fuel in the reactor vessel. ACTION: g, With the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel. O suaveittANcs REauraEMENTs 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours by verification of the date and_ time of subcriticality prior to movement of irradiated fuel in the reactor vessel. . O u n m s8a SEABROOK - UNIT 1 3/4 3-3 k_ _ _ j A.
REFUELING OPERATIONS
/m 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS L) .
- 5 LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment door closed and held in place by a minimum of four bolts, -
- b. A minimum of one doo9 in each ai: lock is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
gor camble e>C bety clos <h
- 1) ~ Closed"by an isolation valve, blind flange, or manual valve, or
- 2) Be capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. - ACTION:
~N (O With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building.
SURVEILLANCE REOUIREf*ENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERA 8LE automatic containment purge and exhaust isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building moeste nG be% closel oc]
- a. Verifying the penetrationsVare in their closed / isolated condition, )
or '
- b. Testing the containment purge and exhaust isolation valves per the applicable portions of Specification 4.6.3.2.
O ___ MAR 131986 rV SEABROOK - UNIT 1 3/4 9-4 ,
*)
1
, u 1.1 -.
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~ REFUELING OPERATIONS , 3/4.9.5 COMMUNICATIONS -
LIMITING CONDITIO3 FOR OPERATION
- 3.9.5 Direct communications shall be maintained between t'he control room and personnel at the refueling station.
APPLICABILITY: During CORE ALTERATIONS. - ACTION: When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. Q SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS. O MAR 13 Is8s n1 r SEABROOK - UNIT 1 3/4 9-5 1 }, 8I # A 6. 4, . L
\
REFUELING OPERATIONS O. 3'4.9.e ReguEL1No nACa1NE
. 1
~
LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine and auxiliary hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERA 8LE with:
- a. TherefuelingmachineusedforYovementoffuelassemblieshaving: ^
- 1) A minimum capacity of 4000 pounds, and 32w
- 2) An overload cutoff limit less than or equal to-3900' pounds.
- b. The auxiliary hoist used for latching and unlatching drive rods having:
- 1) A minimum capacity of 3000 pounds, and
- 2) A load indicator which shall be used to prevent lifting loads in excess of 1000 pounds APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor vessel.
O'~ ACTION: With the requirements fe" refueling machinc and/or hoist OPERABILITY not satis-fied, suspend use of any inoperable refueling machine and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor vessel.
- SURv'EILLANCE REQUIREMENTS The 4.9.6.1 -Each refueling machine used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 4000 pounds and demonstrating an automatic load cutoff when the refueling machine load exceeds -3900-pounds.
3 00 l e. 4.9.6.2 E -auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 3000 pounds. O mAa-131988 .. 7 , I SEA'8R00K - UNIT 1 3/4 9-6 a .
REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS V - LIMITING CONDITION FOR OPERATION - 3.9.7 Leads in excess of 2100 pounds shall be prohibited from travel over fuel assemblies in the storage pool. APPLICABILITY: With fuel assemblies in t,he storage pool. ACTION: g,
- a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.
- b. The provisions of Specifications 3.u.3 and 3.0.4 are not applicable.
i O Suave 1tLxNcE Reau1ReMENTS 4.9.7 Crane interlocks which prevent crane travel with loads in excess of 2100 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter durihg crane operation. 80 SEABROOK - UNIT 1 3/4 9-7 k 'A .bi E
.-_ ._ _ _ . . . , . . _ _ _ _ _ _ _ _ _ . _ . _ _ , , . . , _ . _ _ _ _ . _ _ _ _ . m, _ , . _ , _ . _ . _ _ _ - _ _ , _ _ - . . . . . . _ . _ _ _ _ _ . , . . . _ , . _ . , _ . . _
REFUELING OPERATIONS
/] , 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL- .
t, LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat remov,al (RHR) loop shall be OPERABLE and in operation.* AP3LICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.
- ACTION:
With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor fecay heat load or a reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containstnt penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. m U SURVEILLANCE REQUIREMENTS
- 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at least once per 12 hours. , 7754
*The RHR loop may be removed from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of *he reactor vessel O h t 1*95- MAR 131986 ,
l SEABROOK - UNIT 1 3/4 9-8
]TAK. -
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--- --- w - e,- - - -
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w -nm,- ,,n -- w ~ -- ---- -
REFUELING OPERATIONS , i LOW WATER LEVEL LIMITING CONDITION FOR OPERATION . [ 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.* APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet. '
- ACTION
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.
O V SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coola t at a flow rate of greater than or equal to 2800-gpm at least once per 12 hours. 4 75o
- Prior to initial criticality, the RHR loop may be removed from operation ':r up to 1 hour per 8-hour period during the performance of CORE ALTERAT!0NS in the vicinity of the reactor' vessel hot legs.
MAR 131986 O - T' SEABROOK - UNIT 1 3/4 9-9 ]1-b l* . A. . n s-,w. w 4 m o ,,,7-- , w, -------wm--,,-- 7 , - - - gn g , g q -----gn % e~ ~ - ~ '
REFUELING OPERATIONS c 3. .9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ' LIMITING CONDITION FOR OPERATION - 3.9.9 The Containment Purge and Exhaust Isolation System shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel withir the containment. . ACTION:
- a. With the Containment Purge and Exhaust Isolation System inoperable, close each of the purge and exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere. '
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.9 The Containment Purge and Exhaust Isolation System shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATICNS by verifying that containment purge and exhaust isolation occurs on manual initiation and on a High Radiation test signal from each of the containment- radiation monitoring instrumentation channels. (Minif $1A lo r c ra.n e. ojea. MAR 131986 O - gn gy 1 SEABROOK - UNIT 1 '4/4 9-10 N f
REFUELING OPERATIONS O. 3/4.9.10 WATER LE',2L - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flange. APPLICABILITY: During movement of fuel a'ssemblies or control rods within the containment when either the fuel assemblies being moved or thg fuel assemblies seated within the reactor vessel are irradiated while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel. O' SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies or control rods. , 7 50l$8$ m . ,
, f. -
SEABROOK - UNIT 1 3/4 9-11 s iJ r" :s i _L l1d 1M
REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL (-)> LIMITING CONDITION FOR OPERATION - 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool. ACTION: e
- a. With the requirements of the above specification not satisfied, suspend all movement cf fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
I)- s SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool. , l l 1 i ht4R 131986 PD%GT J M 11 a SEABROOK - UNIT 1 3/4 9-12 W
REFUELING OPERATIONS 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING LIMITING CONDITI0r FOR OPERATION
- 5 3.9.12 Two independent trains of the Fuel Storage Building Emergency Air Cleaning System shall be OPERABLE whenever irradiated fuel is in the storage pool and shall be OPERABLE and-operating during fuel movement.
APPLICABILITY: ulbh % bran - e ACTION:
- a. With one train of the Fuel Storage Building Emergency Air Cleaning System inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided the OPERABLE train of the Fuel Storage Building Emergency Air System is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers,
- b. With no trains of the Fuel Storage Building Emergency Air Cleaning System OPERABLE, suspend all operations involving movement of fuel within the storage pcol or crane operation with loads over the O~ storage pool until at least one train of the Fuel Storage Building Emergency Air Cleaning System is restored to OPERABLE status and is in operation.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.12 The above required trains of the Fuel Storage Building Emergency Air Cleaning System shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:
O - m isas SEABROOK - UNIT 1 3/4 9-13 j . % f* Q %"
REFUELING OPERATIONS
^
3 (G . SURVEILLANCE REQUIREMENTS (Continued) .
- 1) Verifying that the cleanup system satisfies _the in place penetrationangbypassleakagetestingacceptancecriteria J.o 4f less than'0iO5% and uses the test procedure guidance in Regulatory Positions C.S.a, C.5.c, .and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978*, and the system flow rate is 17,700 cfm i 10%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory . Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,& meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,* for a methyl iodide penetration of less than 1.0%;
and o CMS / y g ayo
- 3) Verifying a system flow ratekof 15,2009cfm i 10% during system operation when tested in accordance with ANSI N510-1980.
- c. After every 720 hours of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a O representative carbon sample obtained in accordance with Regulatory U Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,*
meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,* for a methyl iodide penetration of less than 1.0%.
- d. At least once per 18 months by: ,
- 1) Veritying that the pressure drop across the combined HEPA g*3 filtersandcharcoaladsorberbanksislessthang6nches Water Gauge while operating the system at a flow ratejof 16,000,cfm 10%, re-t *-
os1,1oo
- 2) Verifying that the system maintains the spent fuel storage pool g,;zS area at a negative pressure of greater than or equal to .1/49 nch Water Gauge relative to the outside atmosphere during system operation, c.ross connect
- 3) Verifying that the filter cooling bypass valvey can be manually ,
opened, and
- 4) Verifying that the heaters dissipate 95 i 11 kW when tested in accordance with ANSI N510-1980.
O
- ANSI f $10-1980 shall be used in place of ANSI'510-1975 as referenced in Regulatory Guide 1.52, Rev. 2, March 1978. g-$ggg D Q D ,M SEABRO3K - UNIT 1 3/4 9-14 [ --
[g { J
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)
- e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the-in place penetration and bypass leakage testing acceptance criteria of less than 0.'05% in,I O accordance with ANSI N510-1980 for a 00P test aerosol while operating the system at a flow rate of 17,700 cfm i 10%.
- f. After each complete or partial replacement of a charcoal adsorber t.!
bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than -Or,05% in accordance with ANSI N510-1980 for a halogenated I.0 7 ydrocarbon refrigerant test gas while cperating the system at a flow rate of 17,700 cfm i 10%. i ! L i
. l O
i
/
O -- MAIL 13JS86_ D I"') ?. p r g SEABRC0K - UNIT 1 3/4 9-15 .l A .'s ( l
3/4.10 SPECIAL TEST EXCEPTIONS 4 3/4.10.1 SHUT 00WN MARGIN LIMITING CONDITION'FOR OPERATION - 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3 1.1.1 may be suspended for measurement of control rod worth and SHUTOOWN MARGIN provided reactivity equivalent to at least the highest estimat?d control rod worth is available for trip insertion from OPERABLE control rod (s). APPLICABILITY: MODE 2. e ACTION:
- a. With any full-length control rod not fully inserted and with less than the above reactivity equivalent available % trip insertion, immediately initiate and continue boration n greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
- b. With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immedi-ately initiate and continue boration at greater than or eytal to n 30 gpm of a solution containing greater than or equal to 7000 ppm O boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. _ . _ l$h0 t , r, 3, m >< 1 SEABROOK - UNIT 1 3/4 10-1 J k I 'k. .[L
SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBI, TION LIMITS
~
LIMITING CONDITION-FOR OPERATION . 3.10.2 The group height, insertion, and power dist ributfort limits of Specifications 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, 3. 2.1, ar.d 3. 2. 4 may be s uspended during the performance of PHYSICS TESTS provided: 4
- a. The THERMAL POWER is maintained 'less than or equal to 85% of RATED THERMAL POWER, and- c
-b. The-1-imits of Specifications-3.-2r2-and-3.-2.-3-are-maintained and determined-at-the-frequencies _specif.ied-in Specification-4:-10.-2. 2 - be1 ow.~
APPLICARILITY: MODE 1. ACTION-reder f }an $5 % RTP elt her: Withb)yggpfjJ.Sm&E of-the-4-imits-of- pecification-3.-2r2-or-3.-2.-3-being-exceeded.while_. the-requi rements-o f-Speci f i cati ons -3. l. 3.1,11.3. 5,_3 1 3. 6,_3. 2.1,_and _3. 2. 4-are-suspended 7either-~ g o f, g g J,a se e g a.zl t o 9 5 90 S TA U bhlA i h * " 'r r p a. Reduce THERMAL POWER suf f.ickent_toJatisfy the-ACTION-requirements-
\
of-SpecTfications 3-2-2 and-3 2.-3;-or-
- b. Be in HOT STANDBY within 6 hours.
SURVEILLANCE REOUIREMENTS 4.10.2.1 The THEFMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 10:2r2Mhe-requirements of-the-below listed specifications shall-be performed-at-least-once per-12-hours-duririg PHYSICS TESTSP
.a - --Speci fications -4.-2:2;-2 and-4.-2:2r3;-and b.-Specification 4.2.3. 2. -
1 MAR 131986
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SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4., 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and
- c. The Reactor Coolant System lowest operating loop temperature (T**9) is greater than or equal to 541*F.
APPLICABILITY: MODE 2. ACTION: .
- a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakerr.
- b. With a Reactor Coolant System operating loop temperature (T""9)
O 1ess taea 541 r restore T to ~5taia $ts 14 $t witata 15minutesorbeinatleafl9 HOT STANDBY within the next 15 minutes. SURVEILLANCE REOUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equal to 541 F at least once per 30 minutes during PHYSICS TESTS. l 0 MAR 1338H 7pm SEABROOK - UNIT 1 3/4 10-3
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SPECIAL TEST EXCEPTIONS
- g. 3/4.10.4 REACTOR COOLANT LOOPS V. .
LIMITING CONDITI0tt FOR OPERATION - 3.10.4 The limitations of the-following
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-requirements-may-be-suspended:-
.yy ,i.1. i 3.y.t.3,aAd 3 4.3.'l *of k w r u~b/
'ai Specification 3.4.1.1; - During performance of startup and PHYSICS TESTS in MODE-1-or-2-provided:-
Provt Aa ' CC1) The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and c b'2) The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
- b. Specification-374M'-2 -During-the performance-of-hot-rod-drop-time -
-measurements-in-MODE-3-prov.ided_at leas _t_.two-reactor-coolant-loops----
as-listed-in Spdi-TficatWn~J.-4:-172-are-OPERABLE --- APPLICA8ILITY: During operation below the P-7 Interlock Setpoint,or performance -- o f-ho t--rod -d rop -ti me -me a s u rements -- ACTION:
- a. With the THERMAL POWER greater than the P-7 Interlock Setpoint during the performance of startup and PHYSICS TESTS, immediately open the Reactor trip breakers.
-b . With-less-than-the above requWed reactor-coolant-loops-OPERABLE-
-during-the performance-oflot rod drop-time measurements, immed i a te ly-op en -the -re a c to r-tri p - b re a ke rs -and -comp l y -wi th -the -
ACT ION -s tateme nts -o f- Spe c i fi ca ti o n -3.-4.-172. SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL PG',lER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during STARTUP and PHYSICS TESTS. 4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating startup and PHYSICS TESTS. 4r10.-4.-3-At-least-the-above-required -reactor-coolant-loops shal1-be - determined _0PERABLE-within-4-hours-prior-to-initiation of-the -hot rod ' drop - time-measurements-and-at-least-once-per-4 -hours-during -the-hot -rod. drop _ time - measurements-by-verifying-correct-breaker-alignments and -indicated power-x avallabil.ity.and -by-verifying-secondary side-narrow range wateggtg - greater than _or - equal-to --- %.
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SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN LIMITING CONDITION 70R OPERATION . g 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual full-length shutdown and control rod drop time measurements provided;
- a. Only one shutdown or control bank is withdrawn from the fully inserted
position at a time, and
- b. The rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements. ACTION: With the Position Indication Systems inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers. O SURVEILLANCE REQUIREMENTS -. 4.10.5 The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Po'sition Indication System agree within 12 steps when the rods are stationary.
*This requirement is not applicable during the initial calibration of the Digital Rod Position Indication System provided: (1) K is maintained lessthanorequalto0.95,and(2)onlyoneshutdownoNontrolrodbank O is withdrawn from the fully inserted position at one time.
MAR-131886-SEABROOK - UNIT 1 3/4 10-5 3h < l i. A. 1.
RADIOACTIVE EFFLUENTS 3/4.11. SOLID RADI0 ACTIVE WASTES
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LIMITING CONDITION FOR OPERATION 3.11.3l Radioactive wastes shall be solidified or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when receiveu at the disposal site. APPLICABILITY: At all times. , ACTION:
- a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures, and/or the Solid Waste System as necessary to prevent recurrence.
- b. With SOLIDIFICATION or dewatering not performed in accordance with i the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to prevent recurrence. '
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS O E::aF l 4 111 A PRCCEss CON NA PRO &eM dah.L w g gg , V W { W W O *It A & \ of 10 CPR Pa&El ad W ' 't) l 15 % W idZ. d 4
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l I O-F - G SECTION 5.0 DESIGN FEATURES O MAR 131986 O 4. Tm rn n!: v n
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EXCLUSION AREA -- 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1. 2 The Low Population Zone shall be as shown in Figure 5.1-2. c. SITE BOUNDARY FOR CASEQUS EFFLUENTS 5.1.3 The site boundary for gaseous effluents shall be as shown in Figure 5.1-1. DISCHARGE BOUNDARY FOR LIQUID EFFLUENTS l 5.1.4 The discharge boundary for liquid effluents shall be as showp in Figure 5.1-3. O 5.2 CONTAINMENT 1 CONFIGURATION 5.2.1 The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and havir.g the following. design features:
- a. Nominal inside diameter = 140 feet.
.2. I9
- b. Nominal inside height = -205- feet,
- c. Minimum thickness of concrete walls = 4 feet 6 inches.
- d. Minimum thickness of concrete dome = 3 feet 6 inches.j
- e. Minimum thickness of concrete floor pad = 10 feet,tUll Y fas t R il m t.
- f. Nominal thickness of steel liner = 1/4, 3/8, and 1/2 inch for the floor, wall, and dome, respectively.
- g. Net free volume = 2.704 X 108 cubic feety DESIGN PRESSURE AND' TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a x maximum internal pressure of 52.0 psig and a temperature of 296*F.
SEABROOK - UNIT 1 5-1 Trhn MAR-13d 5
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. .. MAIL 13.1986
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c This figure shall consist of a map of the site area showing the perimeter of the site and locating the poini where gaseous effluents are released. If O- onsite iend erees ses3ect to redioective meteriais in gaseous waste are utilized by the public for - recreational or other purposes, then these areas - shall be identified by occupancy factors and the licensee's method of occupancy control. The figure shall be sufficiently detailed to allow identification of structures and release point locations, and areas within the SITE BOUNDARY that are accessible by members of the general public. See NUREG-0133 for additional ' guidance.
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SEABROOK - UNIT 1 5-4 . l-OM M ,N I }G
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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEM8 LIES _ 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly 4 containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have I a nominal active fuel length of 144 inches and contain a maximum total weight
/7/;fT1'748-grams uranium. The initial core loading shall have a maximum enrichment of 3.1 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 5.O weight percent U-235. * -'
e l CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80% silver, 15% indium, and 5% cadmium. All control rods shall be clad with stainless steel tubing. 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and I c. For a temperature of 650 F, except for the pressurizer which is 680 F.
I VOLUME l 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,350 j cubic feet at a nominal T,yg of 588 F. 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1. i MAR 131986 O .... _ SEABROOK - UNIT 1 5-5 i Th '.'; 3 k D 77 i , , , - s ., d,.1 l dA. 1L u
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DESIGN FEATURES 5.6 FUEL STORAGE O CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. A k,77 equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 1.5% ak/k for uncertainties as described in Section 4.3 of the FSAR, and
- b. A nominal 10.35 inch center-to-center distance between fuel assemblies placed in the storage racks.
- 5. 6.1. 2 The k,7f for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 14 feet 6 inches. CAPACITY O 5.6.3 The spent fuel stor ge pool is designed and shall be maintained with a storage capacity limited to no more than 1236 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMI_T 5.7.1 The componen(ts identified in Table 5.7-1 are designed an.d shall be maintained within the cyclic or transient limits of Table 5.7-1. MAR 13Isss p O I
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O O O TABLE 5.7-1 S COMPONENT CYCLIC OR TRANSIENT LIMITS
- o O CYCLIC OR DESIGN CYCLE 7 COMPONENT TRANSIENT LIMIT OR TRANSIENT -
E Reactor Coolant System 200 heatup cycles at < 100*F/h lleatup cycle - T *U
- from < 200*F
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G and 200 cooldown cycles at to 1 550*F. w
-< 100*F/h. Cooldown cycle - T from 3550*Fto<200*FNU z
200 pressurizer cooldown cycles Pressurizer cooldown cycle at < 200*F/h. temperatures from 3 650"F to
< 200 F.
80 loss of load cycles, without 3 15% of RATED TilERMAL POWER to immediate Turbine or Reactor trip. 0% of RATED TilERMAL POWER. T 40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical A.C. electrical power. ESF Electrical System. 80 cycles of loss of flow in one loss of only one reactor reactor coolant loop. coolant pump. I _ _ _ _ h M 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. kg g 10 auxiliary spray Spray water temperature differential actuation cycles. > 320*F.
- un p , E 200 leak tests. Pressurized to 1 2250 psig.
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P 10 hydrostatic pressure tests. Pressurized to 3,3106 psig.
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Secondary Coolant System I steam line break. Break in a > 6-inch steam line. 10 hydrostatic pressure tests. Pressurized to 1 1481 psig. N k [f a -
O-F . t. O SECTION 6.0 ADMINISTRATIVE CONTROLS O MAR 13 less
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ADMINISTRATIVE CONTROLS
. 6.1 RESPONSIBILITY ,
6.1.1 The Statiorrtianager shall be responsible for overall station opera- , ,
' tion and shall delegate in writing the succession to this responsibility during his aosence.
~
6.1. 2 The Shift Superintendent (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Vice President, Nuclear Production shall be reissued to al-1 station personnel on an annual basis. 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for station management and technical support shall be as shown in Figure 6.2-1. STATION STAFF 6.2.2 The station organization shall be as shown in Figure 6.2-2 and:
- a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1; O b. At least one licensed Operator shal' be in the control room when fuel is in the reactor. In additio1, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;
- c. A Health Physics Technician
- shall be on site when fuel is in the reactor;
- d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
- e. A site Fire Brigade of at least five members
- shall be maintained on site at all times. The Fire Brigade shall not include the Shift Superintendent and the three other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and
*The Health Physics Technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours, in order O te accemmedate une nected eeseece. previded immediete action is texen to fiii the required positions.
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MAR 131986 p> c a T) x , 6-1 a AN h A SEABROOK - UNIT 1 l Immaamm-
ADMINISTRATIVE CONTROLS STATION STAFF (Continued)
- f. Administrative procedures shall be developed and implemented to limit .
the wotking hours of station staff who perform safety-related func,- , J tions,*e.g., licensed Senior Operators, licensed Operators, health ~ physicists, auxiliary operators, and key maintenance personnel. The amount of overtime worked by -unj't-staff members performing safety-related functions shall be limite'd in accordance.with the NRC Policy Statementonworkinghours(Gen 7'ricLetterNo.82-12).
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i - MAR 131986 O. , rw 7n $ i SEABROO/,- UNIT 1 6-2
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TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION . r - POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MOD'E 5 or 6 SS 1 1 SRO 1 None RO 2 1 A0 2 , 1 STA 1* None SS - Shift Superintendent with a Senior Operator license on Unit 1 SRO - Individual with a Senior Operator license on Unit 1 RO Individual with an Operator license on Unit 1 AO - Auxiliary Operator STA - Shift Technical Advisor The shift crew composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shif t change due to an oncoming shif t crewman being late or V absent. During any absence of the Shift Superintendent from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with a valid Senior Operator license shall be designated to assume the control room comand function. During any absence of the Shift Superintendent from the control, room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command function.
*The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift Superintendent or the individual with a Senior Operator license meets _the qualifications for the STA as required by the NRC.
SEABROOK - UNIT 1 6-5 h
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l . ADMINISTRATIVE CONTROLS 6.4.1 STATION OPERATION REVIEW COMMITTEE (50RC) O FUNCTION l 6.4.1.1 The 50RC shall function to advise the Station Manager on all
- I l matters related to nuclear safety.
COMPOSITION - l 6.4.1.2 The 50RC shall be composed of the: Chairman: Station Manager Member: Assistant Station Manager Member: c Operations Manager Member: Technical Services Manager Member: Maintenance Department Supervisor Member: Instrumentation and Control Department Supervisor Member: Reactor Engineering Department Supervisor Member: Health Physics Department Supervisor , Member: Tec.4nica. I) --->Engfrieering-Services Department Supervisor l Member: h ff"U Chemistry Department Supervisor ALTERNATES 6.4.1.3 All alternate members shall be appointed in writing by the 50RC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in 50RC activities at any one time. O acertso eaccuescv . 6.4.1.4 The 50RC shall meet at least once per calendar month and as convened by the 50RC Chairman or his designated alternate. QUORUM 6.4.1.5 The quorum of the 50RC necessary for the performance o'f the 50RC l responsibility and authority provisions of these Technical Specifications I shall consist of the Chairman or his designated alternate and four members including alternates. RESPONSIBILITIES 6.4.1.6 The 50RC shall be responsible for:
- a. Review of: (1) all proposed procedures required by Specification 6.7 and changes thereto, (2) all proposed programs required by Specification 6.7 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Station Manager to affect nuclear safety;
- b. Review of all proposed tests and experiments that affect nuclear safety; -
O !
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4 SEABROOK - UNIT 1 6-7 dc taw.1.3.1886 4 MAR A --
ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) O. c. Review of all proposed changes to Appendix "A" Technical Specifications; 4 Y , gf.dlan
~
- d. Review of all proposed changes or modifications to unit-systems or equipment that affect nuclear safety;
- e. Investigation of ell violations of the Technical Specifications, including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence, to the Vice President-Nuclear Production and to the N'SARC;
- f. Review of all REPORTABLE EVENTS;
- g. Review of station operations to detect potential hazards to nuclear safety;
- h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Station Manager or the NSARC;
- i. Review of the Security Plan and implementing procedures and submittal of recommended changes to the NSARC;
- j. Review of the Emerg'ency Plan and implementing procedures and submittal of recommended changes to the NSARC; O k. Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President-
, Nuclear Production and to the NSARC; and ,
- 1. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.
6.4.1.7 The 50RC shall:
- a. Recommend in writing to the Station Manager approval or dis-approval of items considered under Specification 6.4.1.6a. through d;
- b. Render determinations in writing with regard to whether or not each item considered under Specification 6.4.1.6a. through e, constitutes an unreviewed safety question; and
- c. Provide written notification within 24 hours to the Vice President-Nuclear Production and the NSARC of disagreement between the 50RC and the Station Manager however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.
SEABROOK - UNIT 1 6-8 nnA 2 = pm '
4 i ADMINISTRATIVE CONTROLS RECORDS 6.4.1.8 The 50RC shall maintain written minutes of each 50RC meeting that, ' at a minimum, docLment the results of all SORC activities performed under tMe , responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President-Nuclear Production and the NSARC. i 6.4.2 NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC) FUNCTION 6.4.2.1 The NSARC shall function to provide independent review and audit of j designated activities in the areas of:
! a. Nuclear power plant operations, i b. Nuclear engineering, i' c. Chemistry and radiochemistry,
- d. Metallurgy, 7
- e. Instrumentation and control, j f. Radiological safety, i g. Mechanical and electrical engineering, and 3 h. Quality assurance practices.
ssoIP L
- The NSARC shall report to and advise thelVice President-Nuclear-Productiort
- on those areas of responsibility specified in Specifications 6.4.2.7 and 6.4.2.8.
O COMPOSITION 6.4.2.2 The NSARC shall be coreposed of the -Af Jea.st Fw didend D ' i i
-Director: Cha.)r tw1, Nuclear-QuaTity l/lcx CMeru n An d m em he's,Ch( fl bencludlnen MTdi'ger-(PSNH) des 1 na.
p,,3nfedin kerl6in fed allernd hy Member. -Manager-Electrical-Engineering -(YNS0) #'c 6 G i" Vic e Pr eside f" i J! ember: Director-Environmenta'-Enginetr'irig-(YNSD) Colice five //, 6 4 = gember:
;4 Member; Membe r-:
Principat UperatTorTEngineer-(PSNH) Engineer, NUcTeir Engineering-Manager-(PSNH)- Engineering-(YNS0)ln Q4g aj5M She
- Member. ,g gO en4 c/ rue,et Member:
S e a b ro o k-S t a t i o n-Ma na ge r-( PS NH )- Ma n a g e r;-N uc l e a r-P roj e c t s-( PS NH )~
' ggffIt', ed
.de in 4,y 2 /, geA l, 'M ber shall m e e t iIs e, ,
^I ALTERNATES h_ psy 6.4.2.3 All alternate members shall be appointed in writing by the Senior Vice President Director to serve on a temporary basis; however, no more than
- -two-alternates shall participate as voting members in NSARC activities at any T,
one time. A nisigge,g
- y l CONSULTANTS l
6.4.2.4 Consultants shall be utilized as determined by the NSARC Of rector-O to Previce e Pert ovice to the "saac. ___ _ l SEABROOK - UNIT 1 6-9 l p,{dVmR j{ ( b4Iig Jt m k
-~'v-w- -
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I . ADMINISTRATIVE CONTROLS i O . "eert"o raecue"cv - I 6.4.2.5 The NSA$ shall meet at least once per calendar quarter during the . t initial year of unit operation following fuel loading and thereafter at least 4 once per 6 months -thereaf ter, r 2 SQ, i QUORUM f 6.4.2.6 The quorum of the NSARC necessary for the performance of the NSARC ] review and audit functions of these Technical Specifications shall consist of
- the Director Chairman or Vice-Chairman and at least four NSARC members includ-ing alternates. No more than a minority of the quorum shall have line re,sponsi
i bility for operation WseM ("^ n ut rJMRofCthe .ukunit. & jaw edu LcLknsiti lla /q;dwt. VGulpdr 4~1L REVIEW
- 6.4.2.7 The NSARC shall be responsible for the review of
i l a. The safety evaluations for: (1) changes to procedures, equipment, i or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an j unreviewed safety question; l
- b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in 10 CFR 50.59;
- c. Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59;
- d. Proposed changes to Technical Specifications or this Operating License;
- e. Violations of Codes, regulations, orders Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;
- f. Significant operating abnormalities or deviations from normal and expected performance of station quipment that af fect nuclear safety;
- g. All REPORTABLE EVENT 5;
- h. All recogni:cd indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and j
- i. Reports and meeting minutes of the 50RC. -
l AUDITS / 6.4.2.8 Audits of station activities shall be performed under the cognizance of the NSARC These audits shall encompass:
<trL fu k & N N Q VL tyw ylA Y N ** A + Mfg j} lggg Q M _ nj myn awk h~n om me. <' -
lm .z~jil m M M 1 H s ' 3 G XL 4"'Y D '\ 7~g , SEABROOK - UNIT 1 6-10
", . u ,...
p$ t i l l
. .r. .f , A L
ADMINISTRATIVE CONTROLS AUDITS (Continued)
- a. The cortformance of station operation to provisions contained within !
the Technical Specifications and applicable license conditions at least once per 12 months;
- b. The performance, training, and qualifications of the entire station staff at least once per 12 months;
- c. The results of actions taken to correct deficiencies occurring in station equipment, structures, systems, or method of operation that affect nuclear safety, at least once per'6 months;
- d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix 8, 10 CFR Part 50, at least once per 24 months;
- e. The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA personnel;
- f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified of fsite licensee fire protection engineer or an outside ir. dependent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year;
- g. The Radiciogical Environmental Monitoring Program and the results thereof at least once per 12 months;
- h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months; -
- 1. The PROCESS CCNTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months;
- j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months; and
,k,.
ffa $6'1 Any other area of-unit operation considered appropriate by the 1" NSARC or the Senior Vice President.
, he $ooter ?IA A An d infleh tdlq f to te0VCS Y I" f * # # *
"*'j.7b$'jenefbio a s ;~ pieme dl,, eaceda rec oJ^ le#' /*"'" l'"
6.4.2. Rec $rTs$ ARC activities shall be prepared, approved, and dis-tributed as indicate,d below:
- a. Minutes of each NSARC meeting shall be prepared, approved, and forwarded to the Senior Vice President within 14 days following O each meeting; _
Q , h e, a S' t. i 31 SEABROOK - UNIT 1 6-11 .. / . L g-({$ l
_ _ . . - . _ _ ~ . -. - 1 ADMINISTRATIVE CONTROLS O RECORDS (Continued) w MNb ho*wILJ wbt NO M [N14 _;4 da Acpan% hy,
- b. Reportsofreviewsencompassed~bySpecifica*. ion 6.4.2.7shallbh, ,
prepared. approved,-and-forwarded-to the Senior Vice President " within 14 days following completion of the review; and
- c. Audit reports encompassed by Specificction 6.4.2'8 shall be forwarded to the Senior Vice President and to the managetaent positions respons-ible for the areas audited within 30 days after completion of the audit by the auditing organization.
e 6.5 REPORTABLE EVENT ACTION > 6.5.1 The following actions shall be taken for REPORTABLE EVENTS:
- a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part f.0, and
- b. Each REPORTABLE EVENT shall be reviewed by the 50RC and the results of this review shall be submitted to the NSARC and the Vice President-Nuclear Production.
6.6 FAFETY LIMIT VIOLATION 6.6.1 The following actions shall be taken in the event a Safety Lin.it is O violated:
- a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour. The Vice President-Nuclear Production and the NSARC shall be notified within 24 hours;
- b. A Safety Limit Violation Report shall be prepared. Th'e report shall be reviewed by the 50RC. This report shall describe: (1) applicable circumstances preceding the violation, (2) ef fects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;
- c. The Safety Limit Violation Report shall be submitted to the Conunission, the NSARC, and the Vice President-Nuclear Production within 14 days of the violation; and
- d. Operation of the station shall not be resumed until authorized by the Commission.
6.7 PROCEDURES AND PROGRAMS 6.7.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
- a. The .pplicable procedures recommended in Appendix A of Regulatory O c#i'- 123.a <ie 2.r8r"rx2978; pp Mq 1 q
6m SEABRC0K - UNIT 1 6-12 p, d, - M
ADMINISTRATIVE CONTROLS O. PROCEDURES AND PROGRAMS (Continued) b. The energency operating procedures required to implement the require-' , ments 67 NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Ge'neric - Letter No. 82-33; .
- c. Security Plan implementation; -
d. Emergency Plan implementation;
- e. PROCESS CONTROL PROGRAM impleme'ntation;
- f. OFFSITE DOSE CALCULATION MANUAL implementation; and
- g. Quality Assurance Progras for effluent and environmental monitoring.
6.7.2 Each procedure of Specification 6.7.1, and changes thereto, shall be reviewed by the 50RC and shall be approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative procedures. 6.7.3 Temporary changes to procedures of Specification 6.7.1 may be made pro-vided:
- a. The intent of the original procedure is not altered; O a. Tae che#se is aPProvee er two members or the ai at a semeat sterr.
at least one of whom holds a Senior Operator license on the unit affected; and
- c. The change is documented, reviewed by the 50RC, and approved by the
, Station Manager within 14 days of implementation. ,
6.7.4 The following programs shall be established, implemented, and maintained:
- a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the RHR and containment spray, Safety Injection, chemical and volune control, and-hydrogen-recombiners.- The program shall include the following:
- 1) Preventive maintenance and periodic visual inspection requirements, and
- 2) Integrated leak test requirements for each system at refueling cycle intervals or less.
O :-MAR 4a isas . SEABROOK - UNIT 1 6-13 -
? l.
- 2. . b. .Il
&T1 d
ADMINISTRATIVE CONTROLS (a q ,: . PROCEDURES AND PROGRAMS (Continued)
- b. In-Plant Radiation Monitoring , ;
A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
- 1) Training of personnel,
- 2) Procedures for monitoring, and
- 3) Provisions for maintenance of sampling and analysis equipment.
- c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:
l 1) Identification of a sampling schedule for the critical variables and control points for these variables,
- 2) Identification of the procedures used to measure the values of the critical variables,
,m , 3) Identification of process sampling points, which shall include
() monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,
- 4) Procedures for the recording and management of data,
- 5) Procedures defining corrective actions for all off-control point chemistry conditions, and
- 6) A procedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action. -
- d. Backuo Method for Determining Subcooling Margin A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:
- 1) Training of personnel, and
- 2) Procedures for monitoring.
- e. Post-Accident Sampling A program which will ensure the capability to obtain and analyze
(,)
" reactor coolant, radioactive iodines and particulates in plant gaseous ef fluents, and containment atmosphere samples under accident conditions. The program shall include the following:
SEABROOK - UNIT 1 6-14 7 7 MAR }@19g rp l
, ./ 4. L N
ADMINISTRATIVE CONTROLS
, PROCEDURES AND PROGRAMS (Continued)
, 1) Tcaining of personnel, . 3
- 2) Procedures for sampling and analysis, and
, 3) Provisions for maintenance of sampling and analysis c.quipment.
6.8 REPORTING REQUIREMENTS ROUTINE REPORTS e 6.8.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted. STARTUP REPORT
- 6. 8.1.1 A summary report of station startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the station.
The Startup Report shall address each of the tests identified in the Final O- Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license condi-tions based en other comitments shall be included in this report. Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all tnree events have been ccmpleted. W $ 0 ' SEABROOK - UNIT 1 6-15
- D R AJN, -
l ADMINISTRATIVE CONTR0tS ANNUAL REPORTS * .
. i~
6.8.1.2 Annual Reports covering the activities of the station as described below for the previcus calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March ~ 1 of the year following initial criticality. Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions ** (e.g. , reactor operations and surveillance, inservice inspection, routine maintenance, special 4
maintenance [ describe maintenance], waste processing, and refueling). The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions;
- b. The results of specific activity analyses in which the primary 4
coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting
. O 48 hours prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, i results of analysis while limit was exceeded and results of one I analysis after the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiotodine concentrations; (3) Clean-up flow history starting l 48 hours prior to the first sample in which the limit was exceeded;
}' (4) Graph of the I-131 concentration (pCi/gm) and one other radio-iodine isotope concentration (pCi/gm) as a function of time for the
- duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
- c. Documentation of all challenges to the pressurizer power-operated relief valves (PORVs) and safety valves.
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT *** 6.8.1.3 Routine Annual Radiological Environmental Operating Reports covering the operation of the station during the previous calendar year shall be submitted
*A single submittal may be made for a multiple unit station. The semittal should combine those sections that are common to all units at the station.
- **This tabulation supplements the requirements of $2fL407_nL10 CFR P2rL20 1
1 ***A single submittal may be made for a multiple uni t, ong p73 j SEABROOK - lJNIT 1 6-16 ,,
l ADMINISTRATIVE CONTROLS
~
I . ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) prior to May 1 of each year. The initial report shall be submitted prior to' May 1 of the year following initial criticality.
! The Annual Radiological Environmental Operating Reports shall include M N#W qummaries;-interpretationsrand an analysis of trehds of the restuits bf the' g g g ' ,
i radiologicalenvironmentalsurveillanceactivitiesforthereportperJod', 8 9 U lle OIY/ in'c(udingacomparisonwithpreoperationalstudies,withoperationalcontrols.geef,,,yj as app,ropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the Land Use Ce,nsus required by Specificatio 3.12.2. l The Annual Radiological Environmental Operating Reports shall include the d results of analysilbof all radiological environme6tal samples and of all environmental radiati'ogmeasurements taken during the period pursuant to the 1ccations specified in the table and figures' in the Offsite Oose Calculation l
~
Manual, as well as summari' zed and tabulated results of these analyses and l measurements. In the event that some ' individual results are not available for
! inclusion with the report, the r'epor't shall be submitted noting and explaining the reasons for the missing results N he missing data shall be submitted as soon as possible in a supplementary rep' ort.
l Thereportsshallalso/include the following: \ a summary description of l O the Radioiosicai Envirorrmentai soaitoriaseroorem;'at 4 east two iesibie aas* covering all samp11,ng locations keyed to a table giving' distances and directions frcm the centerline of one reactor; the results of licensed participation in I the Interlaborjtory Comparison Program and the corrective action taken if the specified program is not being performed as required by Specification 3.12.3; reason fo j'not conducting the Radiological Environmental Monitoring Program as requiryd by specification 3.12.1, and discussion of all deviations from the j sampling schedule of Table 3.12-1; discussion of environmental sample measure-i menfs that exceed the reporting levels of Table 3.12-2 but are not the result y i /f plant effluents, pursuant to ACTION b. of Specification 3.12.1; and discussion t t
"of ati analyser-in which-the-tLD required-by-Table-4r12-1 was-not-achievable.----N SEMIANNUALRADIOACTIVEEFFLUENTRELEASEREPORT*[ ;
6.8.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the station during the previous 6 months of operation shall be i submitted within 60 days after January 1 and July 1 of each year. The period ! l of the first report shall begin with the date of initial criticality. ' ) $~ee JMCerf 1 -*0ne map-sha1Lcover locations-near-the-SITE-BOUNDARY;-a-second.shall-include . the-more-distant locations..
- [*Asinglesubmittalmaybemadeforamultipleunitstation. The submittal !
l should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall O specify tae reieeses or reotoective meteriei from eech #ait. n,---..-,---_.-_,.,,,_-_,v,__,---n- . - - ,_ , ,_ - __ - --, , , , . , , , _ _ _ _ , , - , , . - _ , _ , . -
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_ --'i-h-e --un ege mwm e - e e ==mpamami-- w e e ah+-* h ---==4--m-w--emm e e-==% e-mmm,--,-- a .meAe + = = A- e~, ,--=-+-w., --g.-m
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- em .<e - -- ---
_.- - - - - -w-g+ -+-,me.e-- m- w gummy .e4.- ..-,- e- w.-, .m ,- ga es-- -.- m.wo--.eme-- a--e 4 -
-w-i- - -~ e- --a e- = + .- wm .
&+-*-w-e - m+-m..+,.w -e.mwe-.ga---*emp-wN #- -- =--- ==' -*N-6 +-9 + * -** -* * --- -
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96M-- 6m. w - e ww w a S uun m
ADMINISTRATIVE CONTROLS 4
. \EMIANNUAL-RADIOACTIVEEFFLUENTRELEASEREPORT(Continued)-
4 The Semiannu21 Radioactive Effluent Release Re M 'shall include a - summary of the quantities of radioactive liquid and gaseous effluents irid ' ' solid wliste released from the station as outlined in Regulatory Guide 1;21, j "Measurin,g, Evaluating, and Reporting Radioactivity in Sol.id Wastes and Radioactive Materials in Liquid and Gaseous Ef fluents fro'm Releaseso(CooledNuclearPowerPlants," Li ght-Wa ter- Revision 1, June 1974,wi'thdata i summarized on\a quarterly basis following the format of Appendix'B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three addit'ional categories: class of solid wastes (as' defined by 10 CFR Part 61), type of' container (e.g., LSA, Type A, Type B, La'rge Quantity) and SOLIDIFICATION ager$t or absorbent (e.g. , cement, urea fo'rmaldehyde).
\
The Semiannual Ra,dioactive Effluent Release Repo/ rt to be submitted within
, 60 days after January 1,of each year shall include'an annual summary of hourly meteorological data collected over the previous,y' ear. This annual summary may be either in the form of ~an hour-by-hour listing on magnetic tape of wind
} speed, wind direction, atm'ospheric stability,' and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction,
- and atmospheric stability.* This same ye' port shall include an assessment of theradiationdosesduetothe\adioactiveliquidandgaseouseffluentsreleased from the unit or station during the, previous calendar year. This same report a shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figur'e 5.1s 3 ) during the report period. All assumptions used in making these assessments, i.e. , specific activity, exposure time, and location, shall,be included try these reports. The meteorological conditions concurrent wjth the time of release of radioactive materials in l gaseous effluents, as, determined by sampling frequency and measurement, shall be used for determining the gaseous pathway , doses. The assessment of radiation a doses shall be performed in accordance with the methodology and. parameters in j the OFFSITE DOSE CALCULATION MANUAL (00CM).
l TheSemia/ nnual Radioactive Effluent Release (teport to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources \,, including doses from j primary [ef fluent pathways and direct radiation, for the, previous calendar
- to show conformance with 40 CFR Part 190, " Environmentals Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the' dose contribution from liquid and gaseous effluents ar'e given in Regulatory Guide 1.109, Rev. 1, Octobe
- 1977.
The Semiannual Radioactive Effluent Release Reports shallsinclude a list anddescriptionofunplannedreleasesfromthesitetoUNRESTRIC{EDAREASof radioactive materials in gaseous and liquid ef fluents made during the reporting period. ~-
/ "In lieu of submission with the Semiannual Radioactive Effluent Release I Report, the licensee has the option of retaining this summary of requ} red i
metiorological_ data on site'iri a^ file ~t6at'shall be.pr,ovided o to the NRC s
; upon request. -N - ! SEABRO UNIT 1 6-18 7"% y, .
P s j Ul .! 3. 81.j A l u_ . _ _
ADMINISTRATIVE CONrROLS O i Trata""ua' a^oto^CrIvE EggtueN1-RELEASE-REPORT _(Continued) - T emiannual Radioactive Effluent Release Reports shall an - changes made'durfng the reporting period to the PROCESS CONTROL PROGRAM and i the 00CM, pursuantsto Specifications 6.12 and 6.134espectively, as well as 1 d Radwaste Treatment Systems any major pursuant change to Yiquid,It Gaseous, to Specificatio'n-6.14. shall or 50)also include a listing of new l tions for dose calculations %nd/or,. environmental monitoring identified by the Land Use Census pursuant ecif on 3.12.2. 1 TheSemiannualJadioactiveEffluentReleaseReportsshallalsoinclude the following: an explanation a's to why thbinoperability of liquid or gaseous effluent monitofing instrumentation was not corrected within the time specified in Specificatif on 3.3.3.10 or 3.3.3.11, respectively;'and description of the events)eadingtoliquidholduptanksorgasstoragetankYexceedingthe. l
- ecificati_o.0_3.11_1.4 or 3.11.2.6, respectively.
MONTHLY OPERATING REPORTS i '~ 6.8.1.5 Routine reports of operating statistics and shutdown experience, i 4ncluding documentat4cn-of-al4-challenges ~to-the-PO~RVs-or-safety-valves > ! shall be submitted on a monthly basis to the Director, Office of Resource j Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a j copy to the Regional Administrator of the Regional Office of the NRC, no later
- than the 15th of each month foll: wing the calendar month covered by the report. <
j RADIAL PEAKING FACTOR LIMIT REPORT 6.8.1.6 The F,y limits for RATED THERMAL POWER (F,RTP) shall be provided to I the NRC Regional Administrator with a copy to Director of Nuclear Reactor
) Regulation, Attention: Chief, Core Performance Branch, U.S. Nuclear Regulatory i Commission, Washington, D. C. 20555, for all core planes containing Bank "0" control rods and all unrodded core planes and the plot of predicted (F Pg ,j) vs Axial Core Height with the limit envelope at least 60 days prior to each cycle initial criticality unless otherwise approved by the Commission by letter.
In addition, in the event that the limit should ct,ange requiring a new substan-tial or an amended submittal to the Radial Peaking Factor Limit Report, it will i be submitted 60 days prior to the date the limit would become ef fective unless ; otherwise approved by the Commission by letter. Any information needed to support F RTP will be by request from the NRC and need not be included in this
*Y j ! report. : }
1 l 1 i I
] ___ ! , y. A n r9 ; ;
SEABROOK - UNIT 1 6-19 M. ; i
ADMINISTRATIVE CONTROLS O. SPECIAL REPORTS - 6.8.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report. 6.9 RECORD RETENTION 6.9.1 In addition to the applicable reco'rd retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.9.2 The following records shall be retained for at least 5 years:
- a. Records and logs of station operation covering time interval at each power level;
- b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety;
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; O e. Records of changes made to the procedures required by Specification 6.7.1;
- f. Records of radioactive shipments;
- g. Records of sealed source and fission detector leak tests and results; and
- h. Records of annual physical inventory of all sealed source material of record.
yAR 131986 0 ' SEABROOK - UNIT 1 6-20 n n ,(h,.fA L1.. fi p'ip A m
ADMINISTRATIVE CONTROLS Q , RECORD RETENTION (Continued) . 6.9.3 The following records shall be retained for the duration of the stati.on i' Operating Licens4:
- a. Records and drawing changes reflecting station design modifications made to systems and equipment described in the Final Safety Analysis Report;
- b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; .
- c. Records of radiation exposure for all individuals entering radiation control areas;
- d. Records of gaseous and liquid radioactive material released to the environs;
- e. Records of transient or operational cycles for those station components identified in Table 5.7-1;
- f. Records of reactor tests and experiments;
- g. Records of training and qualification for current members of the station staff;
- h. Records of inservice inspections performed pursuant to these Technical Specifications;
- i. Records of quality assurance activities required by the Operational O ouaiity Assuraace aaa#ai:
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59;
- k. Records of meetings of the 50RC and the NSARC;
- 1. Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.9 including the date at which the service life commences and associated installation and maintenance records;
- m. Records of secondary water sampling and water quality; and
- n. Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.
6.10 RADIATION PROTECTION PROGRAM 6.10.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. O e
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ADMINISTRATIVE CONTROLS O - 6.11 HIGH RADIATION AREA 6.11.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the
" control device" or " alarm signal" required by paragraph 20.203(c),each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radia-tion is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g. , Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radia-tion areas with exposure rates equal to or less than 1000 mR/h, provided they are I
otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area; or
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring l
device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or O c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing l positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.
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6.11.2 In addition to the requirements of Specification 6.11.1, areas accessible to personnel with radiation levels greater than 100C mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys
- dec./ths,C/4 hall e maintained under the dministrative control of the Shift Superintendent uty and/or health physics per.v.is ion. Doors shall remain locked except en dud / during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and -the-max t-i me -f o r-i nd i v i d u a l s-i n -t h a t-a r e ar Inlieuofthestaytim)/ e um-allowable-stay-the RWP, direct or remote (such as closed circuit TV cameral)continuous pecification of surveillance may be made by personnel qualified in radiat on protection procedures to provide positive exposure control over the activities being performed within the area.
For individual high radiation areas accessiblejto personnel with radiation within large areas, such as levelscontainment, PWR of greaterwhere than no1000 mR/hexists enclosure that are for t located [rposes, and of locking where no enclosure can be reasonably constructed ago d the individual area that O iadivieuei eree sheii ee b,rriceeed. conspicuous Posted light shall be activiated as a warnin , device y e m 4. ead a riesaia9(/ k udee d; red e,,dre / e G herH h physics o o rcs' e l-
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ADMINISTRATIVE CONTROLS 6.12 PROCESS CONTROL PROGRAM (PCP)
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. t 6.12.1 The PCP si1all be approved by the Commission prior to implementation.
6.12.2 Licensee-initiated changes to the PCP: _
- a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
- 1) Sufficiently detailed information to totally' support the rationale for the change without benefit of additional or supplementai information;
- 2) A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- 3) Documentation of the fact that the change has been reviewed and found acceptable by the SORC.
- b. Shall become effectivt upon review and acceptance by the SORC.
6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.13.1 The ODCM shall be approved by the Commission prior to implementation. 6.13.2 Licensee-initiated changes to the ODCM: dm t sb,uaJa.U K.b Sha m be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain: I 1) 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 changed with each page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying the change (s); l
- 2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and
- 3) Documentation of the fact that the change has been reviewed and found acceptable by the 50RC.
'bC. ca wmar Jan ShaW become effective upon review and acceptance by the SORC.
l O a.chu-Ase roJa>6 M hd Q) 9-daMc &$ parto indw& una 131986
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ADMINISTRATIVE CONTROLS 6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS
- 6.14.1 Licensee-initiated major changes to the Radwaste Treatment Systems -
(liquid, gaseous, and solid):
- a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the 50RC. The discussion of each change shall contain:
- 1) A summary of the evaluatio'n that led to the determination that the change coul,d be made in accordance with 10 CFR 50.59;
- 2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- 3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;
- 4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto; p 5) An evaluation of the change, which shows the expected maximum V exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;
- 6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid was,te, to the actual releases for the period prior to when the change is to be made;
- 7) An estimate of the exposure to plant operating personnel as a result of the change; and
- 8) Documentation of the fact that the change was reviewed and found acceptable by the 50RC.
- b. Shall become effective upon review and acceptance by the 50RC.
- Licensees may choose to submit the information called for in this Specification as part of the annual FSAR update.
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