ML19242B947
| ML19242B947 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/06/1979 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML19242B944 | List: |
| References | |
| NUDOCS 7908090571 | |
| Download: ML19242B947 (61) | |
Text
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ATTACHMENT "A" Evaluation of necessary changes to Unit 1 Technical Specifications as identified by page number.
Page 1-1 The adjective " principal" shculd be used in the " ACTION" definitier..
Page 1-2 The typo " mechanism" should be corrected.
Page 2-9 The adjective " operating" should be used in the footnote.
Page 3/4 1-9* For overpressure conditions on ECCS pumps specifications should be 1-11* changed to agree with Unit 2.
1-12*
Page 3/41-18* Part length rods, specifications should be changed to agree with 1-19* Unit 2.
1-20*
l- 21
- l-22*
l-28*
Page 3/4 3-20* For degraded grid voltage protection, specifications should be 3-22* changed to cgree with Unit 2 requirements.
3-26a*
3-29*
3-30*
3-33*
Page 3/4 3-26 Item 4.d should identify the T term as Low-Low rather than Low.
avg Page 3/4 3-31 Item 1.f should identify the T avg term as Low-Low rather than Low.
Dage 3/4 3-33 Item 4.d should identify the T,yg term as Low-Low rather than Low.
Page 3/4 3-52 Specification 4.3.3.7.2 should delete reference to " Class A" since circuits do not fit the definition.
Page 3/4 a-26* Update figures for heatup and cooldown curves to agree with figures a-27* previously submitted as part of the overpressure protection review.
-28" Similar to Unit 2, limits for hydrostatic testing are included.
Page 3/4 4-29* Reactor vessel material irradiation surveillance schedule needs revision to be consistent with Unit 2 specifications and WCAP-8771.
Page 3/4 4-30a* Add a new overpressure protection specification and basis for unit 4-30b* which is identical to Unit 2 except for setpoints.
5-3*
5-6*
Page 3/4 6-33 The typo " Volume" should be corrected.
Page 3/4 6-35 The entire surveillance requirement should be changed to agree with 6-36 the proposed Unit 2 requirement, since the hardware is the same equipment for both Units 1 and 2.
D l) b O t]l 7 90800057/ $ .
2_
Page 3/4 7-20 The typo ", and " should be corrected in specification 3.7.6.1.b.l.
Page 3/4 7-21 The specification for the control room habitability systems should be made consistent with proposed Unit 2 requirements, since the control room is a common area.
Page 3/4 7-24 Make surveillance requirements for safeguards area ventilation systems consistent with Unit 2.
Page 3/4 7-68 The entire Specification should be changed to agree with the proposed 7-69 Unit 2 requirements, since the hardware is the same equipment for both Units 1 and 2.
Page 3/4 7-79 The entire Specification should be changed to agree with the proposed Unit 2 requirements, since the hardware is the same equipment for both Units 1 and 2.
Page 3/4 9-10 Remove typographical error.
Page 3/4 10-l* Part length rod specifications should agree with Unit 2.
Page 3/410-2* Remove references to deleted specification.
10-3*
Page B3/4 1-4* Part length rod Basis should be changed to agree with Unit 2.
Page B3/4 4-10* Reactor vessel toughness table has typographical errors which need correction.
Page B3/4 4-11
- Add new Basis fcr overpressure protection to agree with Unit 2.
Page B3/4 7-7 The entire Basis should be changed to agree with the proposed Unit 2 requirements since the hardware is the same equipment for t ith Units 1 and 2.
Page 5-4* Part length rods as a design feature should agree with Unit 2.
Page 6-1 Specification 6.2.2.f should be changed to agree with the proposed Unit 2 specification since it is a requirement for the same people.
Page 6-10 A typo in Specification 6.5.2.7.d.3 should be corrected. Technical Specifications should be capitalized.
Page 6-13 Specification 6.8.1.a should be changed to agree with the proposed Unit 2 specification since Unit I and Unit 2 procedures should be consistent.
Page 6-22 Specification 6.12 should be changed to agree with proposed Unit 2 specification since it is a requirement for common areas.
5.; j 003
ATTACHMENT 3 m.,
533 00i
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i.0 DEFINITIONS
- mme ,,
01 :INED TERMS .,
1.1 The DEFINED TERMS of~ this section acpear in cacitalized type and are acplicable throughout these Technical Saecifications.
THEllAL POWER -
1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2775 MWt.
OPERATIONAL MODE - MODE 1.4 An OPERATIONAL MODE (i.e. MODE) shall correspcnd to any one inclu-sive ccmbination of core reactivity condition, pcwer level and average reactor coolant temperature specified in Table 1.1.
ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each f6c4pe) specificaticn and shall be part of the specifications. pnna:J' OPERABLE - OPERABILITY 1.6 A system, subsystem, train, ccmoonent or device shall be OPERAELE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency
- electrical pcwer sources, cooli :g or seal water, lubricaticn or other auxiliary equipment tha are requ f red for the system, subsystem, train, ccmponent or device to perform its funct# cn(s) are also capable of performing their related support function (s).
~
5.;; Oo5 NORTH ANNA - UNIT 1 1-1
DEFINITIONS l
fEp0RTABLE OCCURRENCE I
si 1.7 A REPORTABLE OCCURRENCE shall be any of these conditicas specified l in Speci fication 6.9.1.8 and 6.9.1.9. -
I
' CONTAINMENT INTEGRITY _ .
I 1.8 CONTAINMENT INTEGRITY shall exist when:
,i h 1. 3 .1 All penetrations required to be closed during accident conditions are either:
]!
- a. Capable of being closed by an OPERABLE containment auto- _.
matic isolaticn valva system, or
- b. Closed by manual valves, blind flanges, or deactivated auto-matic valves secured in their ciosed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1, . .
1.8.2 All equipment hatches are closed and sealed, 1.8.3 Each air lock is OPERABLE pursuant to Specification 3.5.1.3, 1.8.4 The containment leakage rates are within the limits of Specification 3.5.1.2, and y -:s.n 1.8.5 The sealing .mechansim- associated with each penetration (e.g.
welds, bellows or 0-rings) is OPERABLE.
CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter wnich the channel monitors.
The CHANNEL CALIBRATICN shall enccmpass the entire channel including the sensor and alarm and/or trip functions, and shall include the The CHANNEL CALIBRATICN may be performed by CHANNEL FUNCTIONAL TEST.
any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.10 A CdANNEL CHECK shall be the qualitative assessmenc of channel behavior cering operation by observation. This determination shall indication and/cr include, wPere possible, comparisen of the chann, stacus with other indications and/cr status derived frca independent instrument channels measuring the same parameter.
r; LU 1 1-2 b ,; )
NORTH ANNA-JNIT 1
TABLE 2.2-1(Continued),
5 REACTOR TRIP SYSTEM IflSTRUMEllTATI0fl TRIP SETPOINTS y -
NOTATI0ft(Continued)_
jj B Operation with 2 Loops Operation with 2 Loops J- Oseration with 3 Loops (1 loop isolated)*
(no loops isolated)*
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and f S'
- of the power-range nuclear ton chambers; with r;ains to be selected base instrument response during plant startup tests such that:
between - 34 percent and + 10 percent, f (AI) = 0 (i) forq$q-qkndq are percent RATED lilEfutAL ;%iER in Nlhe top and (wher al TH L p0W Q in bottom halvesofthechrerespectivel and qt 4b s
percent of RATED Tile!! MAL POWER (ii) for each percent that tha magnitude of (q - q ) exceeds - 34 porcent, the AT trip setpoint shall be automaticalfy tebuced by 3 percent of its value at RATED TilERMAL POWER.
-q exceeds + 10 percent, (iii) for each percent that the magnitude of (qfy d re, u)ced by 1.25 percent of the Al trip selpoint shall be automatical .
.y its value at RATED TilERMAL POWER.
A < p e, . . \ , ,
Values dependent on NRC approval of ECCS evaluation for these(operation) conditions..
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REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATICN 3.1.2.2 Each of the following baron injection ficw paths shall be OPERASLE:
- a. The flow path from the beric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and
- b. The ficw path from the refueling water sc.crage tank via a charging pump to the Reactor Coolant System.
APPLICABILITY : MODES 1, 2, 3 and 4.*
ACTION:
- a. With the ficw path frca the boric acid tanks inoperable, restore the inoperable ficw path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 5Y and barated to a SHUTDCWN MARGIN equivalent to at least 1.77% ik/k at 2CC*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restare the ficw path to CPERA5LE status within the next 7 days er be in COLD SHUTCCWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With the ficw path from the refueling water storage tank inoperable, restore the ficw path to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDCWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE.UIREMENTS 4.1.2.2 Each of the above required ficw paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that the tercerature of the heat traced portion of the ficw path frcm the boric acid tanks is _> 14S"F.
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NCRTH ANNA-UNIT 1 3/4 1-9 L JJJ
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REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUT 00WN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE. 5 APPLICABILITY: MODES 5 and 6.
ACTION:
With no charging pump OPERABLE, suspend all operations involving C ALTERATIONS or positive reactivity changes until one char.;ing pump restored to OPERASLE status.
SURVEILLANCE REOUIREMENTS the above required charging pump shall be demonstrated 4.1.2.3.i At least OPEPJBLE by verifying, that on recirculation ficw, the pump d discharge pressure of >
4.0.5.
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l REACTIVITY CONTROL SYSTEMS CHAPGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATIOJ 3.1.2.4 No more than two charging pumps shall be OPERABLE.
W l /\PPLICABILITY : f10 DES 1, 2, 3 and 4*.
ACTION:
With only one charging pump OPERABLE, restore a second charging pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and barated to a SHUTDOWN tiARGIN equivalent to at least 1.77% ak/k at 200*F within the rext 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; restore a second charging pump to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Thr provisions of Specification 3.0.4 are not applicable for one hour pricr tc babbic farm;tica cr callapse in "CC: 4
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SURVEILLANCE REQUIREMENTS 4.1.2.4 The above required charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a discharge 12 pressure of > 2410 psig when tested pursuant to Specification 4.0.5.
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REACTIVITY CONTROL SYSTEMS 3 /4.1. 3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3 .1. 3 .1 All full length (shutdown and control) rods which are inserted in the core shall be OPERABLE and positioned within i12 steps (indicated position) 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 im.ovable as a result of excessive friction or mechanical interference or known to be untrippable, determine, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> that the 'HUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With mere than one full length rod inoperable or l misaligned from the bank step counter demand position by more than t 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- c. With one full length rod inoperable due to causes l other than those addressed by ACTION "a" above or misaligned frcm its bank step counter demand height by more than + 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The rod is declared inoperable and the SHUTDOM 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 previous analyzed results of these accidents remain valid for the duration of operation under these conditions, and "See Special Test Exceptions 3.10.2 and 3.10.3.
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~3 kk NORTH ANNA-UNIT 1 3/4 1-18
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) b) The SHUT 00WN MARGIN requirement of Specificatic . 3.1.1.1 is determined at least once per 12 hours. c) A power distribution map is obtajned from the movable - incare detectors and F (Z) and F' a within their limits wi9hin 72 hads,re or verified to be d) Either the THERMAL POWER level is reduced to < 75% of RATED THERMAL POWER within one hour and within the next 4 hours the high neutron flux trip setpoint is reduced to < 85% of RATED THERMAL POWER, or e) The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperaole rod within the hour while maintain-ing the rod seraence and insertion limits of Figures 3.1-1 and 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specificaticn 3.1.3.6 during subsequent operation. SURVEILLANCE REGUIREMENTS 4 .1. 3 .1.1 The position of each full length rod shall be deter-mined 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 l positions at least once per 4 hours. 4 .1. 3 .1. 2 Each full length rod not fully inserted shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 i days. [
/a NCRTH ANNA-UNIT 1 3/4 1-19 c7),
JJ ( /4
TABLE 3.1-1 ACCIDENT ANALYSES REOUIRING REE'/ALUATION IN THE EVENT OF AN lh0PERA6LE FULL LENGTH RCD Rod Cluster Control Assembly Insortion Characteri stics Rod Cluster Control Assembly Misalignment loss of Reactor Coolant From Small Ruptured Pipes Or Frem Cracks In large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assemoly Withdrawal At Full Power Ehjor Reactor Coolant System Pipe Rupture (Loss Of Coolant Accident) Major Secondary System Pipe D:pture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) NORTH ANNA-UNIT 1 3/4 1-20 . .. [); b l
REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 All shutdown [$5kret control rod position indi-cator channels and the demand position indication system shall be OPERABLE and capable of determining the control rod positions within + 12 steps. APPLICABILITY: MODES I and 2. ACTICN:
- a. With a maximum of one rod position indicator channel per group inoperable either:
- 1. Determine the position of the non-indicating rod (s) in-directly by the movable incore detectors at least once per 8 hours and imnediately after any motion of the non-indicating rod whict exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER TO < 50% of RATED THERMAL POWER within 8 hcurs.
- b. With a maximum of one demand position indicator per bank inoperable either:
- 1. Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per S hours, or
- 2. Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within 8 hours.
SURVEILLANCE REOUIREMENTS , _ 4.1.3.2 Each red position indicator channel shall be determined to be OPERABLE by verifying the demand position indication system a. nd the rod position indicator channels agree within 12 steps at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indicator channels at least once per 4 hours. i - f' ,) a 3/ 4 1 -21 JJA' ' NORTH ANNA-UNIT 1
REACTI'!ITY CONTROL SYSTEMS
, POSITION INDICATOR CHANNELS-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 At least one rod postion indicator channel (excluding demand position indication) shall be OPERABLE for each shutdown , control rod not fully inserted.
c APPLICA3ILITY: MOCES 3*f, 4*# and 5*#. ACTION: With less than the above required position indicator channel (s) OPERABLE, immediately open the reactor trip system breakers. SUR'/EILLANCE RE0tjIRE? DENTS 4.1.3.3 Each of the above required rod position indicator channel's) shall be determined to be OPERABLE by performance of a CHANNEL FUT.-.'IONAL TEST at least once per 18 months.
'W1:n tne reactor trip system breakers in the closed position.
#See Special Test Exception 3.10.5.
~. NORTH ANNA-UNIT 1 3/4 1-22
;) .; )
REACTIVITY CONTRCL SYSTEMS PART LENGTH RCD INSERTION LIMITS gg _g_ g LIMITING CONDITION FOR OPERATION 3 .1. 3 .7 All part length rods shall be fully withdrawn. APPLICASILITY: M00E5 1* and 2* ACTION: With a maximum of one part length rod not fully withdrawn, within one hour either: [
- a. Fully withdraw the rod, or
- b. Se in HOT STANDBY within the next hours.
SURVEILLANCE RECUIREMENTS 4.1.3.7 Each part length r shall be determine'd to be fully withdrawn by:
- a. Verifying th/ position of the part length rod prior to increasing THERMAL POW _R above 5% of RATED THERMAL POWER, and b.
/
Verifyi,;tg, at least once per 31 days, that the breaker that suppljeselectricpowertothedrivemechanismislockedin the ff position.
- See 9ecial Tesc Exceptions 3.10.2 and 3.10.3.
/
NORTH ANNA-UNIT I 3/4 1-28 c, JJs Dnio
TABLE 3.3-2 RJACTOR TRIP SYSTEli IllSTRUf4EllTATIO!I RESP 0!lSE TiHES RESPONSE TIME FUNCTIONAL UNIT Manual Reactor Trip fl0T APPLICABLE 1.
- 2. Power Range, Neutron flux 1 0.5 seconds *
- 3. Power Range, Neutron flux, liigh Positive Rate NOT APPLICABLE
- 4. Power Range,lieutron Flux, liigh flegative Rate 1 0.5 seconds
- Intermediate Range, Neutron Flux fl0T APPLICABLE 5.
Source Range,lieutron Fl'n NOT APPLICABLE 6.
- 7. Overtemperature AT 1 6.0~ seconds
- Overpower AT NOT APPLICABLE 8.
- g. Pressurizer Pressure--Low 1 2.0 seconds
- 10. Pressurizer Pres:ure--liigh 1 2.0 seconds
- 11. Pressurizer Water Level--High NOT APPLICABLE Neutron detectors are exenpt from response tine testing. Response of the neutron flux signal portion df the channel time shall be measured from detector output cr input of first electronic component in channel.
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TABLE 3.3-3 (Continued) ACTION 17 - With the number of CPERACLE Channels one less than the. Total Number of Channels operation may proceed provided the inoperable channel is placed in the tripped condition - within i hour and the Minicum Channels OPERABLE requirement is met, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.1.
'ACTICN 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANCSY within the next 6 hours and in COLD SHUTCCWN within the following 30 hours.
Wn :ne n-c e- er cpEmt.E Cnanow s cae 'ess nm l-2TICN 19 - be Tcza E en' e e. C.n " c.n cs s.s , E>TmTuo ord /o,- OOWE:t n--K , a iCN cocq orece:c cre v, cec we - % ,.s.a J v: e i
-;sc.:-ns c~ snsDO ct. ine in co e,*, e c.,2c c e . is a scen .c _,
. s
+ \..
tn a zec\ ccnc : son vv cmtn L n e ar. t i 1 m
- c. \ne Minimam On o nne. s OPERre _r_ equ. cements LS met ,. . n m eve.e , cne cec m:sc o c,'s c mc nne,
'e
, n encw ce, a3 oc ssec ce un -
- 2. n e c es ce t i due v ec. ,coc.e 2sc u.ma ner sc u, s coc .- -- 3, -. ,_ . 1. _ . ;
-) i NORTH ANNA-UNIT 1 3/4 3-22 -
\ \ j, r l
' v L s J .> -)
TABLE 3.3-4 (Continued)_ h -EtlGillEERED SAFETY FEATURE ACTUATION SYSTEM IflSTRUMEllTATI0ff TRIP SE s TRIP SETPolflT ALLOWABLE VALUES - b , FullCTI0tlAL UtilT E 6. AUXILIARY FEEDWATER PUMP START G > 4% of narrow range
- a. Steam Generator Water 1 5% of narrow range instrument span each a instrument span each Level Low-Low steam generator steam generator See 1 above ( All S.I. Setpoints)
- b. S.I.
1 57.5% Transfer Das Voltage > S2.5% Transfer Dus Voltage Station Blackout c. fl . A . ft. A.
- d. Trip of Hain Feed Pump w
viu 7 LO'E OF POWER VU ,b kknd.r'io\ktu N YI IN V'J\l5 VNk'n CA 2'iQ 1 (p yg\W m\\g q ca G howc t, < J'i t. ay cf W\\. g.) 9.'? i O.03 Second ki<ne Si dAO.03
/ second Ilow dela/
( Lo c+. c o . ( .I) \0j \ k $. k 1 , e --
'IE t 3 cru nd L ~ det y
( De3< nded %D < yl U>i3 wonA Uom dA j
TABLE 3.3-5 (Continued) ENGIf!EERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME It! SECONDS
- 6. Steam Flow in Two Steam Lines-Hich Cointicent witn Steam Line Pressure-Low
- a. Safety Injection (ECCS) 5,13.0#/23.0##
< 3.0
- b. Reacter Trip (fecm SI)
- c. Feedwater Isolation 5, 8. 0
- d. Containment Isolation-Phase "A" y,18.0#/23.0##
- e. Auxiliary Feedwater Pumps 3,60.0
- f. Essential Service Water System Not Acplicable 9 Steam Line Isolation 5, 8. 0
- 7. Containment Pressure--High-High
- a. Containment Quench Sprav < 60.0
- b. Containment Isolation-Phase "B" 3,60.0
- 8. Containment Pressure-Intermediate Hign-Hign
- a. Steam Line Isolation 3, 7.0
- 9. Steam Generator Water Level Low-Low
- a. Auxiliary Feedwater Pumps 3, 60.0
- 10. Station Blacke't.t.
- a. Auxiliary Feedwater Pumcs Not Applicable
- 11. Main Feedwater Puma Trio
- a. Auxiliary Feedwater Pumps Not *pplicable
- 12. Steam Generator Water Level--High High
- a. Turbine Trip - Reacter Trip 5, 2. 5
- b. Feedwater Isolation 3. 11 .0 5tt -ru c.n tc. b Cen NORTH ANNA - UNIT 1 3/4 3-29
-01 r '7 IEO.L "I ) )
9 e kO
;. _. ; - 5 C- " O n +_c o s* ~
tl 4 a^ " _- c := O. . .' 4i '# cCO TOC', O db At'.C?_. C COi ~ O O'
- ~
, l ,
f, s " == m,' \ 9
- m. ,mm NWw'wD
%=.sr' Y ame t =%&M e r . = m -
W
- 3
.~. da b kObTOCN ' b , w 'd -
4 - * = = ** * *
- l w sa: '
v _ c ,. _ma m.. ,. )'.
. u _ ~, ,,,
b g .'s ta. . ir\ G C, . b T. e. ~ s u 3 - .,'h u ;;t >i: .'a. ;);jr e - t 3 ;
- - i 4.2;a q fu ,t t- g ,; . I ,, r ; '3 b Yd/ b .p;j y 'N U {(N U U d bj d
[j; dD[,1731 r, c-, _ 5. , s
-al 6
s L
TABLE 3.3-5 (Continued)
-:ABLE NOTATION
- Diesel generator starting and sequence loading delays included.
Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, and Low Head Safety Injection pumos. Diesel generator starting and sequence loading delays not_ included. Offsite pcwer available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps
#1 Diesel generator starting and sequence leading deiays included.
Resacnse time limit includes opening of valves to establish SI path and attainment o" discharge pressure for centrifugal charging pumps. x T v e_
' ne reseccia dmes Sn mn co e caed
._ \
on
- s. e .: m.a hm when -ne stam a v eames t _ re:c u=- E u.p pW _n c; S e. e me n em,j
'.195 fe c ese. Omec:cw i/ J - d <
4
.\
C ut4
,1 n1 0 0 ',
)
NORTH ANNA - UNIT I 3/4 3-30
TABLE 4.3-2 (Continued) ENGillEERED SAFETY FEATURE ACTUATI0il SYSTEf1 IflSTRUMEtlTATI0fl G SURVEILLAfiCE REQUIREllErlTS f? m CllAfitlEL LODES Ill WilIClf
;r: CllAtlNEL CilAtlNEL FUtlCT10flAL SURVEILLANCE g FUNCTIONAL Uti!T CllECK CAL IllRATION TEST REQU_ IRED C
- 4. STEAM LINE ISOLATION
] a. Manual fl. A. fl . A . R 1, 2, 3
- b. Automatic Actuation Logic fl. A. N.A. M(2) 1, 2, 3 ,
- c. Containment Pressure-- S R M 1, 2, 3 InterrTdiate liigh-liigh
- d. Steaat Flow in.Two Steam S R !! 1, 2, 3 Lines--liigh Coincident with oT2 .c-L.n-L<.., --40 or Steam Line D I (TPlYNsure-@-Low I' S. TURBIflE TRIP AND FEEDWATER d ISOLATI0tl
- a. S team Ger.erator Water S R M 1, 2, 3 L eve 1 --Iligh-lli r')
- 6. AUXILI ARY FEEDWATER PUMPS
- a. Steam Generator Water S R H 1, 2, 3, 4 Level--Low-Low 3 b. S.I. See 1 above (all S.I. Surveillance Requirements)
$j c. Station Blackout N.A. R ti.A. 1,2,3,4
- d. Main Feedwater Pump Trip d.A. N.A. R 1, 2 '
a (.; . () g M F_ R o 4 !I kV Erno y m/ M t N A- R iA . l.2 ,3
) b \,,, s.ou , 37b e -<!%hd 3 o,, R 1.2,3 h 4 K LV E m.ic.,.4,< 7 1i A. ta Un.,m %,./(D~,^^4 C e) a e
X O - C @L E
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t C 6 LJ C CJ +J E O >= ma
>= C % m ed m < rc H: C 3 -M = =
Wl u, - m o3a Q eJ m 3 GJ %
<. Z cG G & *J C,. LJ O D'
C 3 b = G. +.a
- ** C C CJ U r3 O U w= % C O L w & c)
< < J3 +c== =
C4 C W J C c4 3UJ oC
< CJ O O U LJ M - EO -E C. C --
Z
- - *J Cw= W C ec - =
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- E m = w.a
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=
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= w to O >- . . . .
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u., v tiORTH AMA-U?ili 1 3/4 3-26 .-
. i. (\vu Q .s g3 ~Y JJ
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=== w .e-
=<w . . . .
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==
w m m m m =. Lae LJC:C N N N N N N N N N
== . . . . .
em - - - - - - - -
=
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c ;< - -
- w= - -
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>-- =cw .N. - -
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= = = = = = x r w <-w r
= =or w u=
=
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=
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c x w m _a - ws . . . .
~ y= =< < < < <
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mly < =.
= m-
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cc,a =w . .
=
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<= = = m m m m t- w =u
=}>- v
<l=
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- u. = c m =
>- . = i w .c .a .=
w c=. U .= 3N C - U Q
< c: - :n o- 0 3 - -
. m r e - ;u - a e e =, a = s = = ea c- c O uJ J G i W "> l
. , CJ Ce a i w = c am a eJ a cJ c: - c - ua =w m= - u w e o = =u me m =
-s c e =
o w c c - e w o c_ w oU c o a m
= a a 3 >- 3=v
- =.-=
P - m c; c .= o uJ H-a e
=
c; u C +J = u u . ' .a c_ Om 4.J
= .O - m a c. c. - e e eu LC 3 >-
m a U c. L.J =>-- e fJ 3 i e ec - - a e a L - e cJ .= aa c: a < a o. e- <J - C CJ u O - a 3OmI C =r= C HO C U C N C > am Qel31I m C U cJ um - - sc -ea = - s= u =,.cio n u w- .a s. = a
=-
oe or, a. u .= = n
- w. , - n - u :,
sm e
- = ~__ m e a e eO s ce . m
> =M u
=
m
= C .cJ J:".
= - L.J = 0 M C CJ u 3 c0
= w e a = c-a ua e> = >v = - a = =n
>- < n = c uom - c3 a -o e - n = c-
_a r2 x < u c. u = c c= m Jw .c < r < u=
< wc -
=
= u. w a
a <w . . . . . .
'. u
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- La L C = U "O C % m U v ~,
n .
= .
- N
- u. . -
a l NORTH ANNA - UNIT 1 3/4 3-31 'T
- 7 o
',]' y_ j I CU
INSTRUMENTATION FIRE OETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum, the fire detection instrumentation for each fire l detection zone shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: Whenever equipment in that fire detection zone is required to be OPERABLE. ACTION: With one or more of the fire detection instrument (s) shown in Table 3.3-11 inoperable:
- a. Within I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, and 2
- b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or, in lieu of any other report required by Specifica-tion 6.9.1, prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 within the next 30 days outlining tne action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not a ppl icabl e.
SURVEILLANCE RECU! G ENTS ,
- 4. 3. 3. 7.1 Each of the above required fire detection instruments shall be deconstrated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST. i 4.3.3.7.2 The NFPA Code 72D supervised circuits supervision j 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 non-supervised circuits betwaen the local panels in Specification 4.3.3.7.2 and the control room shall be f amonstrated OPERABLE at least once per 31 days. , NORTH ANNA - UNIT 1 3/4 3-52 -
REACTOR CCOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR CCOLANT SYSTEM , LIMITING CONDITION FOR OPERATION 3 . 4 . 9.1 The Reacter Coolant System (except the pressuri:er) temperature and pressure shall be limited in accordance with the limit lines showr. on Figures 3.?-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 one hour period.
mb. A maximum cooldown of ICC*F in any one hour period. A PPL ICABILITY : 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 cut-of-limit condition en the fracture toughness properties of the Reactor Coolant System; 4termine that the Reactor Coolant System remains acceptable for cont .ed operations or be in at least HOT STAND 3Y within the next 6 hou:' . and reduce the RCS T and pressure to less than 200*F and 500 psig, respectively, within t*39 following 30 hours. SURVEILLANCE REQUIREMENTS _ _ _ _ , 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be deter: lined to be within the limits at least once per 30 minutes during system heatup, cooldewn, and inservice leak and hydrostatic testing operations. 4.4.9.1.7 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material prope. ties, at the intervals shewn in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. NORTH ANNA - UNIT 1 3/4 4-25 533 080
e e r% f g%
.,% f &I [ I ( e e a
d '^ tQJW TJ >
]- r t ,~, Cn; O r, .- ncy z _. r:a n
\
i
+ ,
- *- N 0 h' (&s V
' \ (P {
- Q gf% yQ
~- l D E J:' 4 C C6 t' C.M A L G.QC,n -j . t ntn- y - 3
.d t- ., ,.
Cccacun , ; n w_ i toes
, g ,,,. , -
,gp r -. p L .
~Js w
; a, '1,'
G;L ~! i l i . ; 7l /js>, (
" '41L.)U 'i lj-p lt 3000 -.
NATERI AL PROPERTY 2 ASIS CONTROLL NG MATER I AL: FORGED METAL COPPER C*NTENT: 0.16 WT'. PHOSPHORUS CONTENT: 0.019 WT! RT INITIAL: 38 F -' NOT RT MDT A 5Em: I/4T. 134 F 3/4T. 104 F E 2000 - LEAK TEST LIMIT en i u 8 E cs U
-t UNACCEPTABLE ACCE PT A BLE '
c.) OPERATICN g CPERATION v:: u 12,3. . ~ - 1000 - HEt.fuP RITE UP TO 100 ( F/MR) c C R I T I C A L I TY LIMIT 100 m 1 m I c e.e .w, C = e .-a - Wee % .. y =- l i i 0 0 100 200 300 400 500 ~ INDICATED TEMPERATURE ( F)
~
= . . - ..,n. ,.- e ~ .: - ,: 5 e z. s - c .. = ~. - -
Figure 3.4-2 -North-Anna-Power Station,No.-1-Reactor Coolant Systerrmeatup
-bimitations- Applicable-to+E f f ective-Full-Power-Years end{cniains WrewcHOO -end-6&PSfG4er-Pessibledestrument F Errory
-1 =*
\, , [* , ,,.- ., y -
d ~ ~~ d _,.
3000 NATERI AL PROPERTY BASIS CONTROLLING MATERI AL: FORGED HETAL COPPER CONTENT: 0.16 *T% PHOSPMORUS CONTENT: 0.019 WT% RT IMITIAL: 35 F NDT RT AFTER 5 EFPY: NDT I/47 134 F 3/4T. 104 F 2000 - 2 E'
~
a UNACCEPTABLE OPERATION S
.=ci 0; 1 5
l f2
'93 1000 -
3 3 COOLDOWN RATES ( F/HR) .[
) ACCEPTABLE OrERATION 3 25 3
- 50
^
100
, , ~)
d I I l \ m 0 400 500 300 c]
. . ~ . .
O 100 200 DI 3 INDICATED TEMPERATURE (OF) Q TM Figure 3.4-3 North Anna Power Station No.1 Reactor Coolant System Cooldown Limitations App +teet.m :o 5 Ef fectivefttWPowerWeerstnd C:.mtems-Merrnsef4GOMEG-PS!C for P xtNe4nst+ttenent-Erret-
,_:_ 3 ,
3 /4 ' - 15 . - 1 2 \ \ ;3 \ gjJ
TABLE 4.4-5 5 REACTOR VESSEL FIATERIAL IRRADIATION SURVEILLANCE SCllEDULE U g CAPSULE REF10 VAL INTERVAL [ V 1st Refuell:10 z , U X 10 years W 10 years (Reinsert in Location V) Y 10 years (Reinsert in tocatloa X) W 20 years Z 20 years (Reinsert in Location V) Y 30 years U 30 years (Reinsert in Location X) 4,- m T (Extra) S (Extra) c1
'-~" I
- c -
- a1 g lW [ ;
' , ,.;, h 'auN e q.,
hi/ sq.,, ., .J, ' > .s--\
@I, -
y g k
REACTOR C00LAtiT SYSTEi1 OVERPRESSURE PROTECTIO.'! SYSTEMS LIllITIflG C0lDITIO1 FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE: Two power operated relief valves (PCRVs) with a li ft setting of a. lest than or equal to,375 psig*, or Sof
- b. A reactor coolant system vent of greater than or equal to 2.07 square inches, or
- c. A maximum pressurizer water volume of 457 cu. ft.'
APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 2400 F, except when the reactor vessel head is unbolted. 3:o*# ACTIO:1:
- a. With one PORV inoperable, either restore the inoperable PORY to OPERABLE status within 7 days or depressurize and vent the RCS through a 2.07 square inch vent (s) within the next 8 hnurs; main-tain the RCS in a vented condition until both PORVs have been restored to OPERABLE status.
- b. With both PORVs inoperable, depressurize and vent the RCS through a 2.07 square inch vent (s) within 8 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status.
. c. In the event either the PORVs or the RCS vent (s) are used to miti-gate a RCS pressure transient, a Special Report shall be prepared and s _t i tted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable.
43 0 '
*Less than or equal to f)?f psig with an RCS cold leg temperature less than 1400F.
Wif _. _ (J , sqq
!Caly applicable for RCS cold leg temperatures > 1200F. 5j3
; _ ___ 3 _,_
jE/.CTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: ,
- a. Performance of a CHANNEL FUNCTIONAL 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 thereaf ter when the PORV is required OPERA 3LE.
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel, at least once per 18 conths.
- c. Verifying the PCRV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.
- d. Testing pur uant to Specification 4.0.5.
4.4.9.3.2 Tha RCS vent (s) shall be verified to be open at least once per 12 hours
- when the vent (s) is being used for overprccsure protection.
"Except when the vent path./cy is provided with a valve which is locked,
. sealed, or otherwise secured in the open position, then verify-these-
"ic-1 'l c 3 c p : n a t-ksst-owce -peM>4--days,- c _; y2, a s , 2 .; .- cs- ec t:
C n C E. Orc 3 L Co .t, Wn,. ce,-- f_ e c , k, cge;ce <A n - C ' '. . T **.C? C. f t . ~- NORTH ANNA - UNIT i 3/4 4- lob r"7 - JJ) I) ' I ?.
EMERGE! ICY CORE C00LIfiG SYSTEMS avg > 350 F ECCS SUBSYSTEMS - T LIMITIriG C0fiDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPEPABLE centrifugal charging pumo,
- b. One OPERABLE low head safety injection pump,
- c. An OPERABLE flow path capable of transferring fluid to the Reactor Coolant System when taking suction from the refueling water storage tank on i safety injection signal or frca the containment sump when suction is transferred during the recirculation phase of operation or from the disch=.rge of the 2 outside recirculation spray pump.
APPLICABILITY: MODES 1, 2 and 3. ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable sub-system to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation
, and the total accumlated actuation cycles to c* ate. C. % 2-wsw ns J S2ecicc.:2c 3. D. 2 a s n c-- aaon cao'e. ic, 3. 5. 2. ct c~nc 3.E.2.'o Oc c c c e. ntue SCmc una hecc.ac
, ,J o'ce s e. 3 2.C ' C- cr octc c he C_C Cj G O A n O t . C '.N 3 o :.
NORTH ANNA-UNIT 1 3/4 5-3 J3b
, EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T ava < 350 F v, ,
i.IMITING CONDITION FOR OPEPATION 3.5.3 As a minimum, one ECCS subsysten comprised of the fallowing shall I be OPERABLE:
- a. OneOPERABLEcentrifugalchargingpumpf
- b. One OPERABLE low hea/. safety injection pump, and
- c. An OPERABLE flow pa:h capable of transferring fluid to the i
reactor coolant system when taking suction from the refueling water storage tank upon being manually realigned or from the l containment sump when the suction is transferred during the recirculation phase of operation or frcm the discharge of the 2 outside recirculation spray pump. APPLICABILITY: MODE 4 [ , I_I__.T.I0ti:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE states within 1 hour or be in COLD SHUTDOWN within the next 20 ho" s.
- b. With no ECCS subsystem OPERABLE s'use of the inoperability of the low head safety injection pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant Systam T l heat removal met $33s.ess than 350*F by use of aiternate
- c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Ccmission pursuant to Specificatica 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.
See b :ovec b e. u-SURVEILLANCE REOUIREMENTS 4.5.3.iThe ECCS subsystem shall be demonstrated OPEPABLE per the applicable,z l g} Surveillance Requirements of 4.5.2. J .; s , h A _- acrw E, , - . . f NORTH ANNA-UNIT 1 3/4 5-6 ;
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CCNTAINMENT SYSTEMS 3/4.6.2 CCMBUSTISLE GAS CONTROL _ , HYOROGEN ANALYZERS .. L IMITING CONDITION FOR OPERATION Two independent ccntainment hydrogen analyzers (shared with . _ 3 . 6 . 4 .1 Unit 2) shall be OPERABLE. A?FLICA3ILITYL: M00E5 ' and 2. ACTION _: _ With one hydrogen analy:er incperable, restore the inoperable STANO3Y witnin analyze to operable status witnin 30 days or be in at least HOT the next 6 hours. . m....wemes m. SURVEILLLANCE REQUItEMENTS 4 . 6 . 4 .1 Each hydregen analyzer shall be demonstrated EOPERABLE at lea once per 92 days on a STAGGERED TEST 3 ASIS by performing a C CALISPATICN using sample gases containing: a. One volume percent (t .25%) hydrogen, balance nitregen, and Fourvsam e. b.
.'lo4erar percent (+ .25%) hydrogen, balance nitroger. .
3/a 6-33 NORTH ANNA - UNIT 1 J5 bYU
CONTAIMME?iT SYSTE'45_ W ASTE GAS CHARCCAL FILTER SYSTEM L IMITING CONDITION FOR OPERATION 3.6.t.3 A waste gas charcoal filter system (shared with Unit 2) shall be OPERABLE. . APPLICABILITY: MCDES 1 and 2. - ACTICN_:
~
With the waste gas charcoal filter system incoerab - HO' STAL.' within the next 6 hours. SURVEILLANCE REOUIREMENTS 4.6.4.3 The waste gas charcoal filter system shall be demonstrated OPEPJ, ELE: C a. At least once per 31 days by: (
- 1. Initiating ficw threugh the HEPA filter and charcoalblewers and ve adsorber train usin] the process vent fying that the purge system ocerates for at least 15 minutes, b.
At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charecal adsarber housings, or (2) follcwing painting, fire or chemical release in any i ventilation ene ccmnunicating w th the system by:
- 1. Verifying that the cleenuo system satisfies the in-place testing acceptance criteria and uses the test oracedu es of Reculatory Positions C.5.a. , C.5.c and C.S.d of Pegula-
/~ tory Guide 1.52,\ie@jen (._{ubjd93+
cu s n c.s , and n tnthe sys :em ficw rate is 300 cfm + 101, (e,.mt
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n ,:: . Navsen ~_ , N.mecn 3/4 6-35 NCRTH ANNA - UNIT 1 {;a/ G' 1, f; ); v
CCNTAINMENT SYSTEMS SURVEILLAt;CE RECUIREMENTS (Continued)
- 2. Verifying within 31iays af ter removal that a laboratory analysis of a representative caroon sample obtained in accordance with Reculatory Position C.6.b. of Regulatory Guide 1.52, .'Rs,isicr 1, L1,10M meets the laboratory testing criteria of Regulatory Position C.6.a. of Regula-]/
tory Guide 1.52,IRev4sion ., Ely .C~is %w n _ ,7 nmc& _n ;.
- 3. Verifying a system flow rate of 300 cfm 210% during -
system coeration when tested in accordance with ANSI N 510-1975.
- c. After every 720 hours of charcoal adsorber operation by verifying within 31 days af ter removal that a laboratory analysis of representative carbon sample obtained in accordance with ~ .. o ReJulatory Position C.6.b of Regulatory Guide 1.52,ihtamMJ,"'""
IJulFT_6, meets the la'coratory testing cr iteria of Regulatory h SE 'To_sition C.6.a of Regulatory Guide 1.52,(Re,isicr.1,f Li 1:7?.
- d. At least once ;,er 18 months by: "1' "
- 1. Verifying that the pressure droa across the HE?A filter and charcoal adsorber a;sembly is C 8 while operating the filter tram a.5 inches t a flowWater rate ofGauge 300 cfm I
-+ 10"..
l ew m, , c.n
- e. After each complete or partial replacement of a HEPA filter bank by verifying that the bEPA filter banks remove > 99'. of the DOP when they are usted in-place in accordance with ANSI N510-1975 while operati ng the system at a flow rate of- 300 cfm i 10P..
- f. After each complete or partial replacement of a charcoal adsorber bank by verifying that tne charcoal adsorbers remove > 99", of a halogenated hydrocarbon refrigerant test gas when tney are tested in-place in accordance with ANSI N510-1975 while aperating the system at a ficw rate of 300 cfm i 10%.
3/4 6-26 7 NORTH ANNA - UNIT 1 c.J JJ ' 1:~ (V ')
PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION , LIMITING CONDITION FOR OPERATION 3.7.6.1 Flood protection shall be provided for all safety related systems, components and structures when the water level of the North Anna Reservoir exceeds 256 feet Mean Sea Level USGS datum, at the main reservoir spillway. APPLICABILITY: At all times. ACTION: With the water level at the main reservoir spillway above elevation 256 feet Mean Sea Level USGS datum:
- a. Be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours, and
- b. Initiate and complete within 36 hours, the following flood protection measures:
- 1. Stop the circulating water pumps, and i
- 2. Close the condenser isolation valves ,
SURVEILLANCE REQUIREMENTS ; I 4.7.6.1 The water level at the main reservoir. spillway shall be deter- i mined to be within the limits by;
- a. Measurement at least once per 24 hours when the water level is ,
below elevation 255 feet Mean Sea Level USGS datum.
- b. Measurement at least once per 2 hours when the water level is equal to or above 255 feet Mean Sea level USGS datum.
7
-J I) J
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NORTE ANNA - UNIT 1 3/4 7-20
- PLANT SYSTEMS _
_3/4.7.11_ SEALED SCURCE CONTAltINATION~ LIMITING CONDITION FOR OPERATICN.- I i l either in , Each sealed source containing i radioactive mater al l' ;t+in l 10 Cfe-3O 3.7.11.1 ; j
+-G4-*kr' c*ief-any-otherweMa4,-mcexcess de contaminaskn, of.Wase-q 2 te-fme-ef -u'M}O5-eie,reewies-of - - x. -ro+a , -
g T I APPLICABILITY: At all times. ACTICN: cess of the e and:
- a. Eacn above limits shall be irmediately withdrawn frca contamination sealed source with removable us 1.
Either decontaminated and repaired, or l i s 2. Disposed of in accordance with Ccemission Regu at on . 4 are not b. The provisions of Specification 3.0.3 and 3.0. i applicabl e. l _ SURVEILLANCE REOUIREMENTS shall be tested for Each sealed source 4.7.11.1.1 Test Recuirements leakage and73r contamination by:
- a. The licensee, or i or an b.
Other persons specifically authorized by the Commiss on Agreement State. f at least 0.005 The test method shall have a detecticn sensitivity o microcuries per test sample, (excluding Each categoryl of sealed sourcessubjected to core flux 4.7.11.1.2 Test Frecuencies l startup sources anc fission detectorsll previous sealed yshall b At least once per six months for a J
- a. Sources in use l sources containing radicactive materials.
l l Tl 3/4 7-65 e ~ ., UNIT 1 *OR
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- SURVEILLANCE REOUIRE?iEfiTS (Continued) ding Hydrogen 1.
With a half-life greater than 30 days (exclu 3),and 2. In any form other than gas. and fission
~ Each sealed sourcetransfer to another ths. Sealed
- b. Stored ceteccorscurces snail not ce in use prior to use or testec licensee unless tested witnin the previous six moni hout a sources and fission detectors transferred w ttested prior to be
^g_ 3 ,3 indicating the last test date shall be vgnj placed into use. [
Each sea led startup
- c. Startuo sources and fission det.d'actors prior to beingand following l source anc fission catector snali ce testesub I repair or maintenance to tne source.
d and submitted to A Special Report shall be prepare 4.7.11.1.3 Recorts f sealed source or fission detector 5 microcuries of removable the Cor: mss 1cn en an annual basis ileakage tests re contamination. 3/4 7-69 r I; NORTH ANNA - UNIT 1
PLANT SYSTEMS .
-LOW PRESSURE CO,- SYSTEMS _
LIMITING CONDITICM FOR OPERATION __ 3.7.14.2 The following low pressure CO2 systems shall be OPERABLE with a minimum of FG-and-4 tons in the storage tanks at a minimum pressure of 275 psig. 5.5
- a. Cable tunnels and vaults
- b. Charcoal filters
-c. Emergency diesel generator rooms APPLICABILITY _: Whenever equipment in the low cressure CO2 pr tectedsareas is required to be OPERABLE.
2 ACTION: a' . With one or more of the above required low pressure C0 7systems inoperable, establish a continuous fire watch with bacRup fire suppression equipment for the unprotected area (s) within i hour; restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuan to Speci-fication 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule
'for restoring the system to OPERABLE status.
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.14.2 Each of the above required low pressure CO2 systems shall be demonstrated OPERABLE:
- a. At l' east once per 7 days by verifying CO2 storage tank level and pressure, and .
- b. At least once per 18. months by verifying:
- 1. The system valves and asscciated ventilation dampers actuate manually and autcmatically, upon receipt of a simulated actuation signal, and
- 2. Flow from each nozzle during a " Puff Test."
l NORTH ANNA - UNIT 1 3/4 7-79
REACTIVITY CONTROL SYSTEMS BASES I 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specificatic- of this section ensure that (1) acceptable power distribution limits u e iaintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion 1Laits. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peakin.g factors or a restriction in THERMAL POWER; either of these restrictions provides assurance of fuel rod integrity during continued operation. In addition those accident analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's.are satisfied.
-The-res t ric t i o n-p ro hiMt4ng-pa r t-l eng th-rod-i n s e rt i o n-e n s u r e s-t ha4-
-adversc-powe r-cha pes-o nd-ra pid-loc A-powes ha nges-whic h-may-ef-fec t-DN 3-
-co nsMerc t i o n s - do- no t-occ u r- a s-a-res et-o f-pa r t-l e n g t h-rod-ins e r t4 sn-
-dwirrg-cporation.
The maximum rod drop time restriction is consistent with the as-sumed rod drop time used in the accident analyses. Measurement with T > 500 F and with all reactor coolant pumps operating ensures that t$39 measured drop times will be representative of insertion times ex-cerienced during a reactor trip at operating conditions. NORTH ANNA - UNIT 1 B 3/4 1-4 Ebb \Of
IAutt n J/4 4-1 b,5, lil' ACID:t Vl55[i 10._0Qi_nf 55 I Act f {tal..li.11 Htnisons 50 f t-lb/ Average N 35 mil Teap. (*f) upper shell (ft-lb) p Parallel to IIJJor Normal to Hajor ill Parallet tu Hajor 10:341 to Hajor Heterial Cu P NDIT 2: y Component lleat No. Typ? (1) (1) LI] Working Direction Wuskt;g Disection (*b' Wor kindircGlon Wor k in 3 Direction 8 C l . Ild. Dome 53565-1 A513,0,Cl.1 -31 14 34* -26 stib' A500,Cl.2 -40 <-16 -%* 40 l(I c- Cl. Ild. flan 9e 4904 -22 161
;: Ves, flange 4902 A500,Cl.2 -22 <-10 -50' ' '
100' y inlet Nozzle 4964 A500,Cl.2 -31 14 34* 30*
-2C
-22 UnL' Inlet Norste 4966' A500,C l .2 -22 10 l ' '
" A500,Cl.2 -22 41 61* 1 tia Inlet Norale 4968 t,1 * -) 100' Outlet Horsle 496) A500 CI .2 -11 41 A500,Cl.2 -22 34* -22 90L*I Outlet Harate 4965 14 90' Gallet Hustle 4961 A500,C l .2 -4 34 $P -4 '
Upper Shell 4952 A500,Cl.2 2 46 66 5 6 60f 9f
*'l inter. Shril 4958 A500,Cl.2 0.12 0.009 -11 20 71 9a 5
17 3a
- 14hd?'
92f 05 iower Shell 4979 A500,C l . 2 0.16 0.019 -11 40
- Dot.IlJ. Seg. 53647-3 A533,0,C1.1 -31 21 41* -1J '
.. Dot. HJ. Seg. 53640-4 A513,0.Cl.1 -11 21 47*
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0-06 0.020 -I] 7 "> 19 142 lia r -22 39 -21 L o lia r ( ,I t oF L IM Me,o.l <e E- 1 7._.. pe EstimatcJ teniperature ba6ed-on ilo9ulatury Review Plan #canch-lechnital-Pusition Hiiu-5-2
%ene (4) liintamaa energy at highest test teinperature (* f.0*f) - 1 sheer not reported (s) Average tranverse data obtaisied frans surveilleico prograin.
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/326
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The OPERABILITY of two PORVs or an RCS ven.t opening of greater than 2.07 square inches ensures that the RCS will b,e protected from pressure tran-sients which could exceed tne limits offppendix G to 10 CFR Part 50 when one or more of the RCS cold legs is < 'WC0F, and the Reactor Vessei Head is bolted. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator < 500F above the RCS cold leg temperature, or (2) the start of a charging pump and its injection into a water solid RCS. When the Reactor Vessel Head is unbolted, a RCS pressure of less than 100 psig will lift the head, thereby creating a relieving capability equivalent to at leas t one PORV. 3o0 3 to Uhen the temperature of the RCS cold legs is between 3200F and 340 0 F, over-pressurs protection can also be provided by a bubble in the pressurizer. In such a case, a maximum pressurizer water volume of 457 cu. ft. has been selected to provide at least 10 minutes for operator response in the event of a mal function resulting in .naximum flow from one charging pump. l G b 8 e bf rJJ: J, f (,) , ,y
I PLANT SYSTEMS SASES 3/4.7.11 SEALED SOURCE CONTAMINATION The limitatiens en sealed source removable contamination ensure that tie total bcdy or individual crgan irradiation dces not exceed allowable timits in the event of incestien er innalation of the scurce material. The limitations en removable ccntamination for sources requiring leak testing, including alpha emitters, is based en 10 CFR 70.39(c) limits l for plutenium. Leakage of scurces excluded frcm the retuirements of this specificat'an represent less than one maximum permissible bcdy burden for total bcdy irradiation if the scurce material is inhaled cr ingested. % ? _w z .. % e_ 3/a.7.12 SETTLEMEN7 0F CLASS 1 STRUCTURES In crder tc assure .nat settiemen: does not exceed predicted and allcwable settlement values, a progcam has been established tc conduct a survey of a specified number of points at the site en a semi-annual basis. The first survey was ccnducted in May 1975 to establish base-line elevaticns for mest of the points. Where applicable, the base-line elevaticas of points established subsequent to the May 1976 survey have been adjusted to tne May 1976 survey by an evaluaticn of the settlement records of settlement points on the same structure or on nearby struc-tures. Baseline elevatiens for some points were established en dates other than May 1975 as indicated in Table 3.7-5. Additicnal surveys will be performed semiannually.2 The determinaticn of the elevation of the points shall be by precise leveling with seccnd order Class II accu-racy as defined by the U. S. Oecartrent of Ccamerce, Nationci Oceanic anc Atmescheric Acministra:1cn. Nac1cnal Ocean Survey, 1974. Wnen any g se :;ement point listec in Tacle 3.7-51s inaccessiale during a survey, ccmparisen to allcwable settlement shall be based on settlement of other points on the same structure or on nearby structures having similar founcation conditions. When any settlement point listed in Table 3.7-5 is dislocated or replaced, a documented review of the settlement reccrds of points en the same structure and additionally points en nearby structures having similar foundaticn conditicns snail prcvide a new reference elevaticn for the point that reflects the estimated settlement since the base-line survey. If the total settle-ment or differential settlement exceeds 75 percent cf the allcwable value, the frecuency of surveillance shall be increasec as dictated by the engineering review. Allcwable differential movement is ccntrclied by pipe deflecticns permitted by fixaticn points in buildings. The items limiting differen-tial settlement are as follcws: l NORTH ANNA - UNIT 1 3 3/4 7-7 < .( .+., O .) J I
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CESIGil FEAT 1JRES DESIGil PRESSURE A 10 TEMPERATURE 5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 45 psig and a temperature of 230'F. 5?3 REACTOR CORE FUEL ASSEM8LTEl 5.3.1 The reactor core shall contain 157 fuel assam 511c. with each fuel assembly contair;ing 264 fuci rods clad with Zircaloy -4. Each fuel rod shall have a ncminal active fuel length of 144 inches and contain a maximum total weight of 1780 grams uranium. The initial core loading shall have a maximum enrichment of 3.2 weight percent U-235. Reload fuel shall be similar in pnysical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235.
, .C0ftTROL R00 ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length control rod assemblies. The full length control rod assemblies shall contain a ncminal 142 inches of absorber material.
The ncminal values of absorber mc erial i shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. The <1ance ;
- of the void length in the part length rods shall ccntain aluminuc. vxide. l 5.4 REACTOR CCOLAf4T SYSTEM DESIGti PRESSURE AMO TEMPERATURE 5.4.1 The reactor coolant syst3m is designed and shall be maintained:
NORTH A?i1A - U? LIT 1 5-4 D .~2 b
6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence. 6.2 ORGANIZATION OFFSITE . 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1. FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:
- a. Each on ducy shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
- b. At least one licensed Operator shall be in the control room when fuel is in the reactor.
- c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery frcm reactor trips.
- d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
- e. All CORE ALTERATIONS shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
~
- f. A Fire Brigade of a . least 5 mer.bers shall be maintained onsite at all timesf The Fire Brigade shall not include the minimum shift crew shown in Table 6.2-1 or any personnai required for other essential functions during a fire emergency.
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u - ACMINISTRATIVE CONTROLS REVIEW (Cc,nt'd). L t. , a dem - m
- 1. Violations of apolicable codes, regulations, orders , tesh44tal speci'4:ations, license requirements or internal procedures or .
instructions having safety significance;
- 2. Significant operating abnormalities or ceviaticns form normai or expected performance of station safety-related structures, systems, or ccmponents; and 7, _ . . .m
- 3. Reportable occurrences as defined in the station tcchnic-a4
#s90C i # i ca ti }RG.
5:xu:u. , Review of events covered under tnis paragraph shall include the results of any investigations made and reccamendaciens resulting frem such inves-tigations to prevent or reduce the p. obability of recurrence of the event.
- e. Any other matter involving tafe operation of the nuclear power staticr.s which a dul~v apcointed subcommittee or committee member c,eems approcriate for consideraticn, or which is referred to the SyNSOC by the Station Nuclear S'afety and Operating Ccmmittee.
AUDITS 6.5.2.8 Audits cf station activitias ; hall be performed under the cognizance of the SyNSOC. These audits shall enccmpass:
- a. The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 mont:.is.
- b. The performance, training and qualificatiens of the entire facility staff at,least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems cr method of operation that :rfect 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 "B", 10 CFR 50, at least once per 24 months.
- e. The Station Emergency Plan and implementing procedures at least once per 24 months.
- f. The Station Security Plan and implementing procedures at least once per 24 months.
NORTH ANNA - UNIT 1 6-10 rr7 1 l;4 .5 JJJ i
AMINISTRATIVE CONTROLS 6 .8 PROCEDURES 6 .8.1 Written procedures shall be established, implemented and main-t ained covering the activities referenced below:
"A" of
- a. The applicable procedures reccomended in Appendix I Regulatory Guide 1.33,4buen5e".1972. '
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan implementation.
- e. Emergency Plan implementation.
- f. Fire Protection Program Implenentation.
6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the SNSOC and approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative procedures. 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
- a. The intent of the original procedure is not altered.
- b. The change is approved by tuo members af the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
- c. The change is documented, reviewed by the SNSOC a-d approved by the Station Manager within la days of implementation.
6.9 REPORTING REOUIREMENTS
- ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.
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NORTH ANNA - UNIT 1 6-13 533 115
ACMINISTPATIVE CONTRCLS
- n. - == m
, o c u.- c ma'CS w 6.12 HIGH RADIATION AREA / -[cy woc mcem/ ~
I - 6.12.1 In lieu of the " cont'r ol device" or "alam signal" required by paragraph 20.203(c)(2) of 10fFR 20, eacn high radiation area in which the intensity of radiation i stlCCO ..r:m/hr or im]shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Pemit.' Any individual or group of individuals pemitted to enter such areas . shall be provided with or accompanied by one or more of the following:
- a. A radiation montoring device whicn continuously indicates the radiation dose rate in the area.
- 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 device may be made after the dose rate level in the area has been established and personnel have teen made kncwledgable of them.
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Femit. 6.12.2 The requirements of 6.12.1, above, shall also apply to each high radiaticn area in wnich the intensity of radiaticn is greater than 1000 mrem /hr. In addition, locked doors %11 be proviced to prevent urauthorized entry into such areas and th, .eys shall be maintained und- the administrative control of the Shif t Supervisor on duty and/or the 91 ant Health Physicist, t .
Healtn Pnysics personnel shall be exempt frcm the RWP issuance require-ment during the perfomance of their assigned radiation protectic..
duties, provided they comply with approved radiation protection pro-cedures for entry into high radiation areas.
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. NORTH ANNA - UNIT 1 6-22 e . s ;
PLANT SYSTEMS 3/4.7.7 CONTRCL ROOM EMERGENCY HABITABILITY SYSTEMS LIMITING CONDITICN FOR CPERATICN 3.7.7.1 The following control recm emergency habitability systems shall be OPERASLE:
- a. The emergency ventilation system. .
- b. The bottled air pressurization system.
- c. Two air conditioning systems.
APPLICABILITY: MCDES 1, 2, 3 and 4 ACTICN: % 4-e gg WMn cither-th mergency-vent-iiat4cn-systemar-tne -boteed-a+r-;>res-suri::ti:n sjstes ,v. d-en: ni r :end4+i-onkg-systes-inoperab1:, rc: tere the :n;;;ec;ble system: t: Cr:1?dlL: rat.Mbin 7 d ys er be i ; at least-HOT : AN :Y ei tnin the-next - h ars-end in CCL: ::=:,,N .i th e r th: foi4ew4WA-hoursv SURVEILLANCE RECUIREMENTS -- 4.7.7.1 Eacn control rocm emergency ventilation system shall be demon-strated OPEFABLE:
- a. At least once per 31 days on a STAGGERED TEST SASIS by initi-ating, from the centrol recm, ficw through the HEPA filters and charcoal adscrbers and verifying that the system Operates for at least 10 hour; with the heaters on.
- b. At least once per 18 months or (1) af ter any structural main-tenance on the HEPA filter or charcoal adsorter housings, or (2) following painting, fire or chemical release in any venti-lation zone communicating with the system by:
- 1. Verifying that the cleanuo system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a, C.S.c and C.5.d of Regula-tary Guide 1.52, Revision 1, July 1976, and the system-
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l PLANT SYSTEMS 3/4.7.8 SAFEGUARDS AREA VENTILATION SYSTEM _ , L IMITING COND! TION FOR OPERATION 3 .7.3.1 Two safeguards area ventilation systems (SAVS) shall be OPER with: .
- a. one SAVS exhaust fan b.
one auxiliary building HEP;. filter and charcoal adsorber assemoly (shared with Unit 2) APPLICABILITY: MODES 1, 2, 3 and 4 ACTICil-With one SAVS inoperable, restore the inoperabl hours and in COLD SHUTCOWN within the folicwing 30 bcurs. SURVEILLAtCE REOUIRE'4ENTS Each SAVS system shall be demonstrated OPERABLE: 4 .7 . 8 .1
- a. At least once per 31 days en a STAGGERED TEST BASIS by:
- 1. Initiating, from the control recm, ficw through the auxiliary building HEPA filter and charcoal adsorber assembly and verifyir.g that the SAVS operates for at least 10 hours with the heater on.
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) folicwing painting, fire or chemical release in any ventilation :ane ccmmunicating with the system, by:
- 1. Ve. ifying that the cleanup system satisfies the in-dia e testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.S.d of Regula-tory Guide 1.52, Revisicn 1, July 19'16, a,..
and the system flow rate is 6,300 cfm - 10 ' . ! . . ;; - 4 . r . . . m.y Ew - m u. .- NORTH At;NA - UNIT 1 3/4 7-2a
- x. ( .\>'9 l
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REFUELING OPERATIONS l l WATER LEVEL - REACTOR VESSEL I LIMITING CONDITION FOR OPERATION . 3.9.10 At least-23 feet of water shall be maintained over the top of irradiated fuel ' assemblies seated within the reactor pressure vessel - APPLICABILITY: Curing CORE ALTERATIONS while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all CORE ALTERATIONS. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determin ed to be at least its minimum recuired depth within 2 hours prior to the start of and at least once per 24 hours thereafter during CORE ALTERATICNS. NORTH ANNA - UNIT 1 3/4 9-10 r.x J J .) 17-l i.
3/4.10 SPECIAL TEST EXCEPTIONS SHUTCOWN MARGIN LIMITING CONDITION FOR OPERATION . 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended #ar measurement of control rod worth and SHUTOOWN MARGIN provided,
- a. Reactivity equivalent t.o at least the highest estimated con-trol rod worth is available for trip insertion from 092PABLE control rod (s), and-
-b. A11-part4en gt h-rod s-a re-wi-th d rawn-to-a t-least-t he-180-s tep-
-pos4tton-and-GFERABL'.
APPLICABILITY: MODE 2. ACTION:
- a. With any full length control rod not fully inserted and with less than the above eactivity equivalent available for trip insertion or the part length rods not within their withdrawal limits, in-itiate and conti ue baration at > 10 gpm of at least 20,000 ppm boric acid solutit": or its equivalent until the SHUTCOWN MARGIN required by Specification 3.1.1.1 is restored.
- b. With all full length control rods inserted and the reactor sub-imediately critical by less than the above reactivity equivalent, initiate and continue baration at > 10 gpm of at least 20,000 ppm boric acid solution or its equivalent until the SnUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE RECUIREMENTS 4.10.1.1 The position of each full length and-part4cngth rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full length rod that is not full / inserted shall be demonstrated capable of full insertion when tripped from at least 50". withdrawn position ,d thin 24 hours prior to reducing the SH';TDOAN MARGIN to less than the limits of Specification 3.1.1.1. 4d 0-1.3 The
;;act 1sagth r;ds-that-are-part4cily14mW,-of-demonstratsd OE.3st 4thdrawvsha4
-hours-pr4or-to-reduc 4ng-the-SHUTGCWM-MARGUMo less-than-the- i 5;;eci fic;tica 2.1.1.h 3/4 10-1
,. NORTH ANNA - UNIT 1
_ 3-D
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SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION - 3.10.2 The group height, insertion and power distribution limits of Speci fications 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, 3.1. 3. 7, 3. 2.1 and 3. 2. 4 may b e suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER is maintained < 85% of RATED THERMAL POWER, and
- b. The limits of Spec 'ications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Speci fication 4.10.2. 2 below.
APPLICASILITY: MODE 1. ACTION: With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.S, 3.1.3.6, , 3Fr377, 3.2.1 and 3.2.4 are suspended, either-
- a. Reduce THERMAL tPOW' : sufficient to satisfy the ACTION require-ments o f Speci fications 3.2.2 and 3.2.3, or
- b. Be in HOT STAND 3Y within 6 hours.
SURVEILLANCE RECUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85% of RATED THERMA' PCWER at least once per hour during PHYSICS TESTS. 4.10.2.2 ine Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS:
- a. Specification 4.2.2 - At least once per 12 hours,
- b. Specification 4.2.3 - At least once per 12 hours.
NORTH ANNA - UNIT 1 3/4 10-2 L , s; 1DT
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SPECIAL TEST EXCEPTIONS _ PHYSICS TESTS _ LIMITING CONDITION FOR OPERATION 3.10.3 The limi tations o f Speci fications 3.1.1.t, 3.1. 3.1, 3.1.3'.5,3 ^; 3.1. 3. E x ed 3.'. 3.7 may be suspended during the performance of PHYSICS ,, s TESTS provided: '
- a. The THERMAL POWET does not exceed 5% of FATED THERMAL PCWER, ,
and l b. The reactor trip setpoints en the CPERABLE Intermediate and i Power Range Nuclear Channels are set at < 25% of RATED THERMAL I POWER. i APPLICASILITY: MODE 2. i l ACTION: With the THERMAL POWER > 5% of PATED THERMAL POWER, imediately open th f, reactor trip breakers. 1 SURVEILLANCE RECUIREMENTS 4.10.3.1 The THERMAL POWER shal' be determined to be < 5% of RAT TnE.WAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Inter ediate and Power Range Channel shall be subjected PJNCTICNAL TEST witnin 12 hours prior to initiating ?HYSICS to a CHANNEL I TESTS. ,
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3/4 ' 0-3
' NORTH ANNA - UNIT I
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