ML19306A015
| ML19306A015 | |
| Person / Time | |
|---|---|
| Site: | Ginna, 05000000 |
| Issue date: | 02/10/1984 |
| From: | Mcculloch W, Mccullouch W SANDIA NATIONAL LABORATORIES |
| To: | Larkins J NRC |
| Shared Package | |
| ML19270C231 | List: |
| References | |
| NUDOCS 8404100065 | |
| Download: ML19306A015 (2) | |
Text
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- t Date: February 10, 1984 -
To: John Larkins, NRC ; From: W. H. McCulloch, 6445 i On February 6-7, 1984 Bill McCulloch, John Aragon, and Don King of SLA visited the NTS Hydrogen Burn Facility to inspect the condition of equipmen: and cable / splice samples after a series of hydrogen burn tests conducted by EPRI. - Also present for the inspection were Jack Haugh and George Sliter of EPRI. - Dick Miller of Westinghouse, Don Randall of Astron, and Jchn Wanless of the NUS l Corporation - - s The calorimeters we had supplied were of special interest to us:, but, at the request of John Larkins, NRC, we also inspected the other equipment and cable / splice samples from their original (new) condition, and we made no attempt to evaluate or interpret the observed conditions of the samples or to assess their operability or reliability. The equipment and samples were tagged and photographed and the condition of each item was noted in an Engineering Record book. The inspection consisted of visually examining the exterior of the equipment and cable / splice samples and the interior of the equipment as their covers were renoved. At times unusual or unexpected " flakes", " crusts" or corrosion were removed from the equipment and placed in sample bags. Tne equipment inspected included absolute and differential pressur'e gages, solenoid valves, limit switches, resistive temperature devices, penetration assemblies, a fan motor, and a motor operated valve. Cable samples were obtained from various manufactures: Kerite, Rockbestos, Samuel Moore, Raychem. Okonite, Boston Insulated Wire & Cable and Anaconda. Most of the cable samples had been spliced into loops using a variety of splicing techniques not all of the type used in nuclear facilities. We are to be informed later by EPRI which were the higher grade splices. In general, the equipment exteriors were scratched, discolored and corroded. - evidence of their handling and exposure to fires and high temperatures. There j were not however, any indications of external damage to the equipment. Upon ' removing the covers e found water condensed on most surfaces inside the pressure gages. The most pn>bable source for this water intrusion is through the feed throughs for the experimental instrumentation (which is not typical of installation in nuciene 16cilities). But there was no way to assure that the
" nuclear grade" gaskets and senis did not allow the intrusion of at least some of the wter. (It should be noted that " nuclear grade" installation procedures i
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1
SchnLarkins 2 1 4 and checks must be meticult asly followed: other techniques judged acceptable ' even by experienced and careful installers, can lead to problems as occured . repeatedly in this test series.) Except for the condensation and small amounts 1 for foreign raterial (scale from evaporated water and bits of construction _. material, e.g., meta,1 filingss) we found no evidence of damage or malfuncti.on. The presence of the water might give some concern if the component were expected to perform over an extended period. There was enough water present to : cause shorting in a printed circuit board if the potting material were absent ,j
.,y.
or compromised. , Damage to the cable / splice samples ranged from virtually none to severe damage to the cable jacket. Except for some splices which may not have been " nuclear grade" our visual inspection revealed no evidence of failure of the conductors or insulators. However, there was in many cases sufficient damage _to the cable outer jacket to seriously doubt its ability to protect the inner cable parts _ . . , from a wet environment. For instances, if the inner insulation shrank back slightly from the splices, reisture could easily cause shorts. In the absence . of defined criteria for survivability, we cannot say that we observed success or failure. Of approximately fifty cable samples, only two were virtually free s of damage. Several showed slight surface melting and a few were charred without significant surface wrinkling or bulging. On many samples the jacket had expanded -
- (either from its own phase change or from pressure generated by high temperture gases inside the jacket) and collapsed giving a heavily wrinkled or knuried '
appearance (like the rough bark on a reture tree). In the more severe cases the jackets had holes and/or splits to reveal the inner layers. All but one of the splices showed. sone damage (usually substantial shrinkage or splitting of the splice jacket). Again, our observations were limited primarily to the cable jacket and we saw no clear indication of damage to internal parts or of a cable's inability to perform its function. . 1 l f
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1 l THE PRODUCTION OF H IN LWR SYSTEMS DURING SEVERE ACCIDENTS i l The early prtduction of H2 in LWR systems during a severe accident is mainly caused This reaction produces
.2by the o'xidation of the Zircaloy cladding by steam.
one7 m lecule f r every steam molecule reacted. The geometry of an undamaged 2 (i.e., unmelted) core is very susceptible to oxidation by steam at temperatures above 2200"F (the current Appendix K limitation for LOCA events). The very high surface to volume ratio of the Zircaloy cladding allows sufficient contact area for the oxidation to occur. When an LWR undergoes a severe accident leading to loss of coolant, it , eventually reaches the stage at which the primary water is gradually boiled away leaving the core uncovered. For rapid blowdown events such as a *.arge-break If the low pressure injection LOCA, accumulator injection will refill the core. system then fails, the core water 0111 boil off slowly at ambient pressure. ' 4 Boilaway times from the top of the core to the bottom are of the order of 20-40 _ ininutes depending on the exact sequence if no additional water is injected into ; the core by operator action (Note that at TMI-2 water was injected periodically ,
' t by operator action.)
State of knowledge calculations and direct experimental data (PBF Severe j Fuel Damage Series I, Ex-Pile experiments at KfK in the FRG*) show that ] for an unattenuated boilaway only 15-25% of the Zircaloy cladding will be i $ fully oxidized prior to gross melting of the core (i.e., when temperatures in excess of 5000 F are reached, and failure of the lower core support After l
' plate and lower plenum occur because of the high heating rate incurred.) j gross melting and core collapse into the reactor building lower cavity, the remaining unoxidized Zircaloy may ur.dergo further oxidation cepending upon
)
i
. j
*Kernforschungszentrum-Karlsruhe in the Federal Republic of Germany !
l ENCLOSURE "1"
the availability of oxidant (steam) and the ability of the Zircaloy
~
(present.n$ with, a considerably lower surface to volume ratio within a large mass of molten UO 2 , Ir0 2 , steel, etc.) to come in contact with the steam. Very little infomation is available on the oxidation rate in such a system but experiments are planned to obtain the needed data. For accident sequences where an unattenuated boiloff is prevented by periodic _ water injections (bleed and feed), somwhat more oxidation of the Zircaloy can be achieved prior to or without gmss melting of the core. A situation similar to this occurred at TMI-2 wherew45-50% of the cladding was oxidized. However, there is a limit to the total oxidation possible
+ for this type of sequence also. Maximum or.idation (i.e., maximm H2 production) will occur if the core is bo; led down to certain water levels and held there by feed and bleed operations. However, the water level has to be low enough for considerable oxidation but high enough to provide a sufficient steaming rate to preclude gross melting of the core. Calculations included by the Rogovin Study Group in their report on Three-Mile Island (Volume 2, Part 2 Appendix II.8) illustrate the previous point. Figures 1A,1B, & 1C show core temperatures achieved for uncoveries of 7-feet.
8-feet, and 9-feet respectively for the TMI-2 core. The individual curves on each plot indicate the axial location of the position from the top of the core. That is, 0 = top of core,1 = one foot down from to?, 2 = two feet down from top, etc. All calculaticis assmed a boildown to the indicated level in e
5 20 minutes and holding at that level (presumably by bleed & feed) for an ; additional 60 minutes. To interpret these curves in tems of the production one must derstand that very rapid oxidation of Zircaloy will not occur until the Zircaloy temperature exceeds 2200 F. Using the kinetic equation for 0 oxidation of Zircaloy given in NUREG/CR-0497, TREE-1280, Revision 2, one can calculate the percentage of cladding oxidized in one hour at various 0 temperatures. Table 1 shows the results at three temperatures: 1500 F, , i 0 1830"F and 2200 E. Note that below 1500 only 2% oxidation will occur in 0 , one hour. Since the kinetics are a function of the square root of the time, ; only 2.8% oxidation will occur after 2 hours, etc. Therefore, for those regions ; of the core below say 15000F, it can be safely assumed that very little H2 i will be produced Pless than 3%). p i Figure 1A shows, therefore, that for seven feet of the core uncovered for one-houre4 feet of the cladding is potentially totally' oxidizable. ; That is, a total H2 production of only 4/12 x 100 y 33%. For eight feet ] of the core uncovered (Figure 1B), 5 feet of the core could be totally , oxidized leading to e total possible H2 production from the clad of 5/12 x 100 = 42%. This latter sse is considered to be close to what actually happened at TMI-2. bue that for the above twc situations no l part of the core exceeded 44000 F which. is above the cladding melting of , 0 point but below the fuel (UO2 ) melting point of just over 5000 F. Thus, no gross melting of the core would have taken place except for some dissolution of U0 into the liquid Zircaloy (in the upper two feet only). 2 Figure 1C shows the results for a nine-foot uncovery and hold. In this case 7/12 x 100 = 60% H2 would be produced. However, note that 3.5 feet of the )
-, l
P l core will now reach thd UO2 melting point and core collapse and gross melting l will begin. This large mass may cause failure of the lower plate and lower plenum and lead to a " core on the floor" event. That is, the core would not be recoverable. Note also that even this very severe event oxidized only
^/60% of the cladding prior to core collapse. .
Therefore, ~it seems highly unlikely that an event can be constructed to
~
result in oxidation of more than about 50-60% of the cladding and still maintain the core in a recoverable state (i.e., no gross melting and relocation). , Experimental data to confirm such analyses will be performed on full-length fuel bundles by NRC in the NRU reactor in Canada over the next year. Follow-on tests will be needed to confirm the calculations for various uncovery levels. ; Table 1. Percent Cladding Oxidized at Va,rious Temperatures in One Hour Temperature Present Clad ; (OF ) Oxidized . - 2200 (1200C) 20.5 o , 1830 (1000C) 7.8 0 2.1 1500 (8000) I
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Accident Likelihoed and Potential for " Arrested" Secuences in PWRs with Larce, Dry Containments Since the tire of the original staff proposals on interim hydrogen rulemaking (SECY 81-245A), a considerable amount of analysis has been performed with re-spect to the likelihood of severe accident sequences in PWRs with large dry containments. A number of PRAs have been perfomed under the sponsorship of both the NRC and utilities, and have identified those sequences which contribute most importantly to the respective estimates of the frecuency of core melting. Such information has been. catalogued and studied as one task of the RES-sponsored Accident Sequence Evaluation Program (ASEP). Based on this catalogue, it has.been determined that transient-initiated accidents and small 4
- LOCAs are, in general, the most important types of sequences in most PWRs.
More spet.!fically, ASEP identified five accident sequences which are generally important to the characterization of the core melt frequency in PWRs. These sequences are:
- Transient without early emergency core cooling ! - Transient-induced LOCA without early emergency core cooling i
2
- Small pipe treak LOCA without early emergency core cooling
- Transient without reactor shutdown (ATWS)
' - Transient without early emergency core cooling and without contain-ment heat removal. ENCLOSURE "J" l
. T:
2 An important characteristic of these sequences (except KTh'S) is that the loss of water fr6m ihe reactor coolant systen is rather gradual, and thus the pro-
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cesses of core uncovery, core damage, and gross core nelting can be protracted. For this reason, there can be a significant potential for operatcr intervention to restore cooling of the fuel. In general, this recovery potential has not ; been included in published PRAs. One exception to this is the incorporation of I the recovery of offsite electric power during a station blackout accident r,equence. In this sequence, a recovery probability (as a function of time) based on historical data is typically included. The potential benefit of accounting for operator recovery actions has been evalnated as part of the NRC-sponsored IREP PRA of the Arkansas Huclear One (ANO) Unit 1 plant. In this study, faults which by themselves or in com-
- bination with cther faults cou;d lead to failures of important systems were categorized as recoverable or non-recoverable (e.g., an inadvertently closed :
valve was recoverable, a plugged valve was not). For each recoverable fault, a l recovery location was identified and a critical time developed, based on how , quickly the specific associated function had to be restored. The latter time was based on the onset of an " unacceptable" core condition. For the IREP j study, the unacceptable condition was specified as core uncovery, which for transient events and sum 11 LOCAs occurred in roughly forty minutes. Using this , data and general data on the probabilities cf human restoration actions as , { functions of time, a recovery probability was developed for each recoverable 4 ! I fault. f i
to - l 3 i ! Since the publication of the AND IREP, this analysis has been extended to con-4 sider restoration times correlated to an unacceptable core condition defined as onset of core melting, rather than core uncovery. in effect, this extended the recovery time for small LOCAs and transients (includirg loss and subsequent l recovery of offsite power) from roughly forty minutes to a time on the order of seventy minutes. Table 1 displays the results of these analyses. l l In addition to these~ assessments of recovery potential, the AND IREP also iden-l tified two accident sequences in which it was possible that the operable core
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I cooling equipment might be adequate to prevent gross core melting, even though l it would apparently not prevent core xncovery. In Tabit 1, the last column (" alternative success criterion") includes this possible effect for the two j specific secuences. l The table I results suggest several conclusions. With respect to operator f recovery actions, it appears that, for this AND case study: i i J l
- operator recovery acticas before core uncovery can have a significant t i >
effect on estimates of core melt frequency; and ; I I i i
- if recovery actions have not occurred by the time of core uncovery, it is unlikely that such actions will be taken before the onset of gross core ~
melting. I i 1
a ,7 7 4 The second conclusion -esults from two factors. First, many recoverable faults can be rectified very quickly, and, as such, are accounted for in the pre-core uncoveiy phase (thus the first conclusion). As tima proceeds, however, the incremental changes in perhability due to recovery actions decrease rapidly. 1 That is, the recovery model corresponds to the intuitive concept that if the operators have been unable to correctly diagnose and account for initial faults : by the time the core is firct uncovered, then the likelihood of such actions:by , i
-the time of gross core melting is relatively low. In addition, as such credit for recovery actiers decreases the effect of recoverable faults, non-recoverable 4
faults become increasingly more important and finally prevent any further gain. , ! Thus, as Tabie 1 shows, the only ' incremental reduction in probability occurs in
- l the station blackout sequence (T(LOP) ATilTCC 3FEX1T EC), due to the sarawhat s'
greater likelihood of offsite power restoration prior to core melting. The inclusion of the probability reduction due to an alternative success criterion t l j for two sequences has a more pronounced effect, but the overall reduction still remains rather small. , I In summary, for this case study, it cdn he concluded that recovery actions are l a significant fact:r in preventing severe core damage. However, such actions are most likely to occur before any damage has occurred (i.e., before core i uncovery). If the accident has progressed to the point of initial core j uncovery, then, under the assumptions used for this case study, it has rela-tively little chance of be'ng arrested bt. fore gross core melting. i i 1 ! l 1 j i .
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v; TABLE 1(continued) ; TERMS . i l 1 SLOCAI;1.2) Small LOCA 4" - 1.2")
- 51,0CA I 1.7) Small LOCA 1.2"-1.7") i St.0CA d4) Small LOCA 1.7"-4")
Tl; LOP) Transient initiated by loss of offsite power ' Transient initiated by failure of 4160 VAC bus A3 - T(A3) Transient initiated by failure of 125 VDC bus D01 il D01 T[D02 Transient initiated by failure of 125 VOC bus 002 ' Tl FIA Transient with all front-line systems (ECC, AFW, etc.) initially available ,. j EC Failure of emergency core cooling system in injection mode ECW Failurs of emergency core cooling system in recirculation mode SPRAYS Failure of containment spray system , 57RV stuck-open safety / relief valve NW Failure of auxiliary feedwater system RBC Failure of containment fan cooling system s I l m ,, _.... ,, , _ _ _ _r - ,. .,. _ . - . . . _ _ , , , . . . . . .,_....___..-._,_,....m , _m. . _ . _ , _ _ . . - ., , .,-,...._m.,._, . _ . . . , . - . .. ,,,. . . - ,
, , Y. ~r . - CLOSV$6 h 3f'f 8 '
M6075 h . For: The Conrnissioners zle BY From: - - k'illiam J. Dircks Executive Director for Operations
Subject:
STATUS OF HYDROGEN CONTROL 3 SLUE AND RULEMAKING RECOMMENDAT30h5 IN SECY-53-357-A
Purpose:
To inform the Comission of the background and basis for the 1roposal that rulemLking with regard to hydrogen control for
, JRs with large, dry contairments can be safely deferred.
t Discussion: In sty memorandum of March 15,1984, (see Enclosure "A"). I i infonred you of recent test infomation developed at the Nevada Test Station (ETS) and indicated that it warranted further con- > sideration prior to your decision on the -Hydrogen Control Rule - - 1 (SECY-85-357). The two major issues concern ections to be taken on the Pzrk III BWRs and ice condenser PWRs included in the ; Hydrogen Control Rule, and on the LWRs with large, dry contain- . ments. With this Inforr.ation Paper, the staff provides the
. background and basis for the proposal that rulerraking with re-gard to hydrogen control for LWRs with large, dry containments can be safely deferred pending completion of the ongoing NRC-and industry-sponsored research programs. .
Since my March 15, 1984 memorandum was written, the staff, in ' conjunction with industrial participants, has completed a pre-liminary review of the NTS test data. A summary and description of preliminary results from the NTS program is given in Enclo- < sure "B". The program involved a series of hydrogen combustion ! tests at premixed hydrogen concentrations varying from 5 to 13 l volume percent hydrogen, with different amounts of steau. pre-sent, as well as tests in which hydrogen was continuously in-j jetted at different rates. As indicated in my March 15, 1984
- remorandum, the electrical cables appeared to suffer damage dur-ing the tests at the higher hydrogen concentrations. Post test 1 examination tests have since been performed on the electric 1
cables to deterrine the extent of this, danage. The stait review of the results of the post test examination of the cables is given in Enclosure *C". because the cables "were not erergized during the hydrogen burns, it is not possible after the~ ihet to , i
- CONTACT: !
M. F!sishnan, RES * .. i 443-7616 4
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16 Coccissitners I d h, 4, %dCM g4 l
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! demonstrate conclusively that the c les would have functioned during the burn. ,However, while the e was clearly external dam-age to cables that burned at both 1 - and 13 volume percent hy-drogen, the post test examination revealed that the class IE ' . cables were still able to function. As indicated in Enclosure
'C', the cable test procedures included inaersion of the cables i . ...in water while AC and DC voltage was applied. Some of the
- . cables in the HTS tests had jackets that were damaged in a sini-l lar sanner to those at TMI-2 which also passed subsequent elec- l j trical tests.
In addition to the post test examinations discussed above, the ' staff has performed its own preliminary review and analysis of the HTS test data and has had zany discussions with EPRI regard- /g e ing the interpretation of the data. The staff has concluded - 4 that the NTS experiments were not a dW+ in .;t. eum. 4 M j! ' equipment survivability in the ondenser PWRs and Mark III ( BWRs because of differenc etween the ' test conditions and : r'equired by the HydrogenJfIe hydrogen corgi actual plant conditions. t tontrol Rule and by _.gystemac-
._. be ng -;
tice will effectively concentrations of roughly limit 5 to& 9Wpercent. combustionfo There washydrogen ou .iwiiii- (ef , k hb3 EiT/ - cent cable tests. Equis.entburuing failures at lower below hydrogen10 concentrations volune percent in h k 5 =i 1 m us tests are believed +to be the result of
- system" fail- ;
/ ures, such as improper installetion, and were not related to hydrogen combustion. The NTS data should, therefore, prove to '
be most useful not for a direct indication. of equipment survivability, but as a validation of the analytical models that i are being used to predict hydrogen burn phenomena and equipmeni;- ' response. No information has been uncovered to date which would lead the staff to conclude that the requirements of the Hydrogen
. Control Rule could not be reasonably met by the ice condenser
- PWRs and Mark III BWRs. The HTS data is still undergoing inten- !
l sive review and additional testing and analysis is planned. The i results are bekg closely monitored regarding their implications ; relative to the Hydrogen Control Rule. l \ i To put the NTS tests into perspective with regard to the Hydrogen i Control Rule now before you, the hydrogen concentration result-ing from the oxidation of 75% of the cladding in a PWR with a i
+
large, dry or subatmospheric containment would vary from approx-2 imately 9.7% for r plant like lion to 14.8% for a plant like !
. Surrey-3. These figures for volume percent hydrogen do not in-
- clude the effects of any steam which would' be released during ;
the accident raising the containment prissure and reducing the hydrogen concentration. Hydrogen is assumed to be homogeneously mixed when calculating the volume percent, however, the issue of
, hydrogen stratification is receiving study in the research pro-Dram. Recen:. r.nalyses of the accident a TMI-2 estimate the j - ... . . -
l :
~
j D e Cont..issioners ' total hydrogen released at approximately 900-1,000 lbs or the equivalent of 45-50% clad-oxidation. This number probably ap-proaches an upper limit for a degraded-core accident, that is, higher clad oxidation percentages are only to be found with core-melt accidents. Enclosure "D" provides the current staff understanding relative to hydrogen produr ' son in LWR systems during severe accidents. It is seen that A would be unlikely to have more than about 60% clad-oxidation prior to gross nielting of the core. Furthennore, for this to occur, it re-quires a feed and bleed type of situation to maximize oxidation while preventing full core melting. The water level is thus maintained at some optimum position for saximum hydrogen produc-tion, as occurred during the TMI-2 accident. If no water is irdected and the core proceeds directly to meltdown, it is ex-pected that significantly less than 6Di clad-oxidation would take place (see Enclosure "D"). Recently completed PRA studies on PWRs with large, dry containments (Enclosure "E") indicate that, notwithstanding what occurred at TMI-2, it is relatively unlikely that recovery will take place once initial core uncovery has commenced. .Thus, while in the hydrogen rule the . - staff has conservatively assumed a 75% metal-water reaction for the purptses of design of the hydrogen control system and equip-ment qualification, the staff believes that it is unlikely thct a 7% clad-oxidation can occur. Therefore we still believe
~
that a 751 clad-oxidation provides a conserva,tive basis for the rule. i Our proposals with regard to the two issues or classes of reac-tors are as follows:
- 1. Regarding the Mark III BWRs and ice condenser PWRs, !
SECY-83-357A recomends that the Comission approve the l Hydrogen Control Pale. . , t
- 2. On page 4 of Enclosure *F", SECY-83-357A, the third sen-tence in the middle paragraph has been modified so as to
< read (new words are underlined), " Dry containment designs have a greater inherent capability to accomodate large quantities of hydrogen because of their high design pres-sure and large volume; therefore, for these designs the Comission believes that rulemaking with regard to hydrogen control can be safely deferred pending completion of NRC-and industry-sponsored research which includes studying the g effects othydrocen burning :t M ;t:- = =t = &: to i D_ - __' Zef fects on ecuipment survivability." ! gyh We will continue to monitor the evaluation of the NTS data concerning any implications regarding equipment i survivability.
@5ed I o !:7. hT$ ddais ifith hydrogen combustion at high concentrations.Regarding
) Because of i
l
~ -
The Comissioners 4 . i the preliminary review indicating that/mecJ, t equipment could sur-vive a burn at high concentration and 5ecause of the small like-
. lihood that a high concentration of hydrogen could be achieved, we recommend that the Commission defer action on the large, dry.
containment LWRs until completion of the NRC.- ' ' ' 1., exper ;
~
imental and analytical programs currently underwey. The NRC _ . ...- prograr is intended to supplement the NTS experiments and pro ;j vide Turther experimental data for deteming equipment.
. survivatility in event of a hydrogen burn. Heat transfer to equipmem'. and the resulting temperatures measured at locations, in the dwar of the NTS experiments are to be evaluated this, sumer as EPRI completes their data recuttion and makes ass-surd Cd ph4s available to the NRC staff. The results of M pot 6:sf analysis wiu be used to validate or revise the themal models
- / in the HECTR and HYBER Equipment Response Codes used for pre -
d dicting equipment temperatures in a hydrogen burn. Separate. effects experiments are to be run by the NRC at the Sandia solar: radiation facilities exposing cables and equipnent to time de-pen' dent heat fluxes, environments, and initial temperatures sim-
- ulating those measured in the NTS experiments. The solar facil ,
ity will then be used to expose energized cables and act',ve' equipment in prototypic configurations to a heat flux predicted > for a hydrogen burn in a full-scale containment compartmant. Additional tests will be made with cables and equipment to as-sess the protective factor provided by enclosing cables in can duits, trays or using heat shields; the cooling provided by con- Mb. .: tainment sprays; and the safety margin between the heat 't which equipment fails or malfunctions and that predi d for a, or hydrogen burn in a full-scale containment. The cod evelopment I andmarate effects experimental results will provide input to( MOrm anaIybl *
#- N y g f, , . . . .. : :. ::-i..... . _ ,_ .' . 2"' 2 " M i emessWen The separate *ffects tests are scheduled to be com-Alvafe.$e. Mer*x I pleted in February 19bi, and the analytical evaluations for Ype. of egvef%mh ;g equipment survivability in full-scale containments, usin e idated codes, by the end of May 1985. ,
k 1 will continue to keep you advised of any significant develop-l ments related to the hyd;., gen control program and will provide I recomendations far rulemaking related to LWRs with large, dry 2 contairaients _ at P,e conclusion of the ongoing research in early g 1985. William J. Dircks Executive Director for Operations y e m + m ee e 4 e e + e a e e #
< e._
. 'The Comissioners ,
Enclosures:
*A* - Memo, dtd 3/15/84, W. J. Dircks to Comissioners (w/o attachments)
*B" - Sumary of Preliminary Results from the EPRI Nevada Test Station Program ,
"C" - Memo, dtd 5/16/S4, k'. S. Famer to Distribution, !
or.t of May 10 Meeting with EPRI and Electro-Test i _ Rep' to. Review Hydrogen Burn Cable Post-Test Examination , Results !
"D" - The Production of H in 2
LWR Systems During Severe Accidents
*E" - Accident Likelihood and Potential for " Arrested * .
Sequences in PWEs with large, Dry Contairments , I
+
e -. ee e e h
*M h
.o '
l O O s 1 l
= ' -
The Commissioners 5 t Distribution * . RAVAB r/f '- Circ /Chron . - - < RES Central Files ' Subj Files MFleishman - JMalaro . MErnst PSernero MCunningham GMarino RCurtis ! _ 0Bassett i WFarmer , WMarris - GArlotto FGillespie ~ Dross-
- RMinogue :
MErnst ECase, NRR RMattson, ER WButler, NRR , IRoszteczy, NRR RCleveland, NRR CTinkler, NRR HGarg, NRR VNoonan, NRR RCarter, NRR ' RLaGrange, NRR VBenaroya, NRR KParcieski, NRR TSu, NRR ~ KKniel, NRR MTaylor, EDO , WDircLs, EDO JRoe, EDO TRehm. EDO VStello EDO 1
j.. -
~g UN!TED STATES Endlosure -
. ! , , p, NUCLEAR REGULATORY COMMISSION c - nemmeroes.a.c.ammes y%.*#@. ! EAR I 5 564 - MEMORANDtM FOR: Chaiman Palladino , Commissioner Gilinsky Commissioner Roberts . Commit.ioner Asselstine
- Commissioner Bernthal
- FROM: William J. Dircks Executive Dimeter for Operations , ,
SUBJECT:
REcomENDATION TD DEFER DECISIM ON FINAL HYDRDGEN CONTROL '
~.
1 RULE (SECY-83-357) , ) 1 Tie Consission has scheduled an affimation/ discussion and vote on the final Hydrogen Ccmtrol Rule (SECY-83-357) for March 16, 1984 This memorandum is to infom you of recent test infomation which I believe warrants further consideration prior to your decision on SECY-83-357. As part of a cooperative large-scale eg. 6..tal program at the Nevada Test Station, EPRI has recently completed a series of hydrogen contiustion tests in i a 52-foot diameter dewar. The objectives of this research program were to
- confim test data on hyd ogen combustion taken at smaller scale, assess the effects of igniter location on burning and obtain some generic information on the perfomance and thermal response of selected nuclear plant equipment under
- a range of hydrogen burn conditions. Preliminary abservations (Enclosures 1, 2 and 3) indicate that there may be a potential problem regarding the behavior of electrical cables during a hydrogen burn at a hydrogen concentration of 13 1
percent. It should also be noted that after entry into TMI-2 following the accident, damage was observed to the electrical rables.
~
The NRC staff and the contractors are currently in the process"of analyr'ing l the data in detail and evaluating the uncertainties and implications of the l EPRI tests. This review will include all the equipment identified in l Attachment 1 of Enclosure 1. I will continue to keep you advised on this matter as the post-test analyses are conducted and will provide reconnenda-tions to you on how and when to proceed with SECY-83-357.
~
gggns95Einal.Bink,s. inlliam J. Dircks Executive Director - i
- for Operations Endlosures: See next page I
ENCLOSURE "A"
- i. - -
i
. t 2
Enclosures:
' 1. Meno 12/19/83 G. Sliter to i Distribution w/2 att
- 2. Mese 2/17/84 E. Sliter to Distribution w/4 att
- 3. Mene 2/16/84 R. Curtis to V. Senaroya. W. Butler, K. Kniel & V. h nan i w/ enc 1
- I s i I l i
i f i
- ENCLOSURE *A" 0
e
4, . I l
SUMMARY
OF PRELIMINARY RESULTS FROM THE EPRI 4 NEVADA TEST STATION PROGRAM a L , l The objec'.'ive tf the NTS test program was to study hydrogen burning in a ielatively large enciasure and to determine effects of the resulting e1viroreent on survivn1 of safety related equipment. The tests were perfomed in a spherical vessel of approximately 52 feet diameter. The vessel was l n r: ' .C., instrumented so that the most important parameters characterizing ) the hydrogen burn environment could be measured. The equipment exposed to the hydrogen burn was mounted on a special platfom located at the center of the sphere. It consisted of several types of pressure transmitters, solenoid valves, motor operated valves, limit switches, fan motors, ignitors - penetrations, RTD's and cables made by several different manufacturers. The tested equipment was operational during the tests with the exception of the cables which were spliced in closed loops and did not carry any current. All the equipment tested was new and unaged. The NTS test program consisted of 40 separ
~
tests. In 24 tests hydrogen was premixed with air and steam before ignit nd in 16 tests it was continuously ignited during the test. The test covered hydrogen concentrations of 5-13 v/o, steam concentration of v and rates of hydrogen in continuous injection tests of j.4 - 3.2 kg/ minute. Water sprays and mixing fans were used during some of the tests. The equipment was present in 15 tests (bot:s g pr3 mixed and continuous injection) inclut'ino tle tests with 13 v/o 4 hydrogen. M ch piece of equipment was exposed totseveral tests and was replaced only when (sd64 it exhibited signs of failure. In marw respects the conditions to which the
- equipment was exposed were more severe than the conditions anticipated to occur ,,
in an actual plant during a hydrogen burn. Note that +he r a hl m meenneament. , clocar loops located eenr the center of the test sphere is nnt tvoical e# 4% Tn a nuclear psa3w'he*a +ha e= M ac * *c renen 6 i v enclosed in ~ "4te 4n the THS tests, e went was located in a pl~ume of not gases generated by burning ' hydrogen and did not have heat sinks which would be present if the equipment were mounted on the walls of the vessel. . On the other hana the initia temperature of th equipm was lower than it would have been in a post-LOCA condition and because o the scaline effect enercy transfer to th,g amioment was lower. These factors tr5duced some degree of non-conservatisms t to the tests. l At present the recorded test data are still being evaluated and a relatively i limited amount of mostly qualitative results are available. In general, for the premixed tests with hydrogen concentrations equal to or below 10 v/o most J of the equipment did not exhibit significant signs of degradation and with few i exceptions did not lose their operational capability. This happened despite the fact that some temperatures exceeded post-LDCA qualification limits.
- Similarly in the continuous injection tests most equipment perfomed j satisfactorily. ,
3
** the hydrogen concentrations higher than 10 v/o the numbers of pieces of
.quipment e::'ibitiac a apparent mal.~ unctions increased although a majority of the equipment still performed satisfactorily. j [
ENCLOSURE "B" 6 u.e (ggNea
ftl $dMI6 W $* The most severe environment 6to which th67qu nt was exposed was a premixed S burning with and without water sprays at I v/oh)drogen. During these tests operability was significantly degraded. >heeouent examinatinne ----= w 4n-enc einn of water into f nstrument boxesf At this concentration cable insu-lation exposeo to hydrogen flames started to burn leaving its outer surface charred, but not exposing bare conductors. EPRI ascribed many it..trument fail-ures to improper installation and pointed out that most of the instruments un-
- derwent several tests,and so the damage might have been cumulative. ,
EPRI is planning to issue a preliminary draft of the data evaluation report in the middle of June 1984 This draft will contain only a partial evaluation. The complete evaluation of the data is expected tc be accomplished by the end of 1984. I 4 j i l l i 1 ENCLO3URE *B"
v- i
- t
, h NAY 161964 4
MEMORANDUN FOR: Those on Attached List i . 1 FROM: W. S. Famer, RES l
SUBJECT:
REPORT DF MAY 10 MEETING WITH EPRI AND ELECTAD-TEST TO REVIEW HYDROGEN BURN CABLE ' 1 POST-TEST EXAMINATION RESULTS
~
A meeting was held on May 10 at the Electro-Tes t Inc. test laboratories in San Ramon, California, to review the results of post-test examination of 4 cable and splice specimens which had been exposed to hydrogen burn tests in the NTS facility. Electro-Test had completed an AC withstand test. IR
- measurement and DC withstand test for 56 cable specimens, 3 Raychem Class IE splices and 5 amor splices. Still to be tested are the CONAX and j Westinghouse electrical penetrations and the remaining three 48 conductor '
Firewall III cable specimens (cable type 5). All MV power, LY power and control cable specimens tested to date were found to df splay adequate
- insulation resistance and could be expected to function. All four specimens of the Rockbestos IK8 coaxial cable (cable type 4) failed the electrical tests. In two instances the center conductor was exposed.
7 Individual Rockbestos coaxial cable specimens had seen premixed hydrogen , burns at 8, 10 and 13%. Also, one of the four Raychem XLPE coaxial cable specimens (cable type 9) failed the electrical tests. The failure of the cable type 9 coaxial cable specimen was identified as a short between one of the conductors and the braided shield. High voltage coaxial cables are normally used in conjunction with neutron-monitoring instrumentation. In a telephone conversation on May 16, Dr. George Sliter of EPRI stated , that EPRI had contacted Rockbestos and was told that Rockbestos IK8 coaxial cable had been found in 1980 tests to fail when heated due to differential thermal expansion of the conductor resulting in perforation of the j insulation and wire " kinking." These cables are no longer manufactured and are not considered Class IE. Why MP&L supplied them for the trsts is I unknown. _ The only Class 1E coaxial cable failure is. therefore. the = l sele of mrham coaxu s cable tvoe n wrncn saw a m on'iv hydrocin_ burn. i Tie other three Raycnem coaxial caDies passed the electrical tests. The l sample which failed will be destructively analyzed to see if the cable, had t a flaw or was out of specification. It was also reported in this
- conversation that Electro-Test had completed tssting of the remaining three i
48 conductor Firewall III Rockbestos cables (Type 5) and all had passed the electrical tests. The i ermanctor that failed out of 48 in tha Tvn= KR _ cable sample was noted to fail n i resen vely sow voltage and therefore
- may oe an incicauon of a void in tJoe insulation as menuisc1.ureo.
ENCLOSURE "C" l
EY 2 6 M4 Multiple Addressees 2 The cable specimens were clustered in seven stacks on the bench for visual
. inspection .(Figure 1). The first stack represented cables removed after test #5 (8% hydrogen premixed burn). The jackets on these cables showed
!, little or no damage. The second stack represented cables removed after I test f6 7.8%-hydrogen burn). Several of the jackets had splits. However, the surface of the jackets exhibited little or no charring. The remaining five stacks represented cables removed after test B, test 9, test II, test 14 and test 15 and, accordingly, saw 10% to 13% hydrogen burns. The cable jacket damage increased significantly as the hydrogen volume percentage in the burns was increased from 8 to 13%. It should be noted that the cable i jackets provide primarily for mechanical protection. The insulation underneath, which assures the electrical function, appeared unaffected in visual observations.
- L The 3 Raychem splices that were tested all passed the electrical tests and j are reported to have been made up as Class 1E in accordance with Rockbestos .
instructions. The 5 amor covered splices provided by Duke power all ! passed the electrical tests although the armor enclosures contained water I
- and appear to have leaked at the cover gasket. The use of metal enclosures l
) protects the Raychem wire splices inside- from the heat and flame of the ! i
'ydrogen burn and accordingly they were unaffected. However, this does not appear to be a design used extensively at other nuclear facilities.
In conclusion, except for the one Raychem coaxial cable Type 9 and the one ; conductor in the Firewall III sample Type SB, all cables passed the post ) test examination electrical tests and could be expected to function in the - event of a hydrogen burn at up to 13% hydrogen in the NTS facility. Those i < splices which EPRI hd identified as Class IE also passed the post tast examination electrica. tests and could he expected to function. The cable l Jackets displayed progressively increased damage as the percent hydrogen in l the burns increased from 8 to 135. However, the j,ackets provide primarily l l mechanical protection and appear to have protected the insulation underneath from damage. There does not appear to be any measurable I gradation in damage to the insulation with severity of hydrogen burn in the NTS tests. The results of the post-test examinetion of cables from NTS I appear similar to that from TMI-2 where jackets were also damaged but the i insulation was undamaged and the cables passed electrical tests. j Electro-Test conducted their tests in accordance with good accepted ; , practice. It was recommended that EPRI conside.r testing some of the coaxial and triaxial cables by frequency scanning for signals transmission degradation and attenuation. This would detect degradation that might not be apparent from the electrical tests. It would also be of interest to obtain physical prcpcrty (elongation and tensile strength) measurements of the insulation on some of the TMI cables supplied to NTS in order to compare their . property change with that measured at HEDL. A request for cable samples to be sent to HEDL will be made to EPRI. d
- - +-,. -,-vr-w -, , - y.r,-v-
l MAY I 61984 i j Multiple Addressees 3 The cables and splices were all believed by EPRI to be Class IE. However, documentation was not always provided by the supplying utilities. This documentation should be obtained or the utility be contacted and the categorization of the cables and splices confirmed. i The impohance of the coaximi cable failures has been downgraded by the recent (5/16/84) phone conversation with EPRI. It now appears that all Class IE cables have been shown to pass electrical function tests with the exception of one conductor in one of the four 48 conductor Firewall III cables and one conductor in a Raychem coaxial cable. Further tiestructi te " examination of both is to be conducted by EPRI to determine whether the ; failures were due to flaws in the cable as supplied at the factory. ' c;if.nti signed 8F: - William 5. Farner ! Electrical Engineering '
. Instrumentation & Control Branch l Division of Engineering Technology '
Office of Aclear Regulatory Research
Enclosures:
j
- 1. List of Cable Specimens (from EPRI QLR)
- 2. Cable Quick Look Conditions (from EPRI QLR)
- 3. Electro-Test Inc. Cable Test Procedures i
- 4. Electro-Test Inc. Cable '
Electrical Measurements '
- 5. Attendance List, Meeting on May 10, 1984 Distribution:
RL5 Files EEICB subj j EEICB rf 1 GArlotto i LShao i BMorris WSP EEICB/DET/RES W5 Farmer:md j 5//7 /84 l l
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7
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& cracks,' tapewrinkles, visibleT K9.e.d,$g;,
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l n*e.'RE M.QQM.7*!.'4-e'9hj!'3.$a. T.,-
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. TEST NO. OF i - E MIER OF .W." : :2. 2 . : DESCRIPTION ';.I.j l e TYPE 75PECNEN LAST EXPOSURE ' PRIOR EXPOSURES '-i".YE g.Qd$. ".;"c,? i.- . O
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ffe. Q'- '1 -i.g.:/ . . ST: 2 -33 s;@ .klaser tape visible .iFM' C'.-;*<;. isa "k :; tr:q.J s WW',;M.'ss nn .".sc j , .E;h r. r9cks,;.-WG4t.WneA - -4 f~i. Jf .. UA . .-' f % 11 ; A g l.5. io . 8 2,p .P;e metallic shield u et. .- cha bri-y,.~. :.i:.J ,.e//M .EE " 9 W . : s#f.3#,d/S cJe to visible many placekW
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scorch, 43500 crack '
- 5 yt... Scorck. cracks, metallic shield ' w#.}. '.
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Nan ~se thque-mm 'i flexible '
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wnsatre ensge"to d.T.Y: i yce .i.!f0T3:3th <iNd 7 4 w g,, . 7,- v. h* Q @"h%w .
~ gdsse astuR meu6 taterior' ihmamensieye. 'c'atiles
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}:14 Q i i Me .* 254 ,.". fd4f g 15 s Es visible dama -
1%{artal armor.~interfor'cabies ge"to flexible.r"_ '3
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- scorch Char = soot = on surface #;. IMervid*?fW discoloration ornot change in;g texture M 7M:. of surface W
- '.- / ash z .5 I .:.
' Cracks = tracks on sorface (depth discernible) . yw[.4%.M ..
j 'O. . Flakes = surface pieces of jacket removed. usually not through entire thicknes Eniges = local swelling ~of Jacket - @ ."+r-t W M & 6 i
~~_ _ . / Er' -
tskies = bulpes that form a repeating pattern ~.'.%$~
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ENCLOSURE 3 l i l 1 i I E_tectro-Test Inc. Cable Test Procedures 1
- 1. AC rated voltage withstand test at 60 cycles by immersing cables in water for 5 minutes.
- 2. IR test at 300 or 500 volts DC in accordance with IPCEA Standard.
- 3. DC voltage withstand test in water at 3 times AC test voltage for .
5 minutes in accordance with IPCEA S-19-81 Rubber-Insulated Wire Cable. i The ends of the cable insulation were cleaned, a heat shrink sleeve
~
put on and a guard circuit set up. Surface leakage current was negligible in water innersion tests to determine leakage current throug.*. the insulation. Multiconductor cable tests consisted of testing a single conductor and grounding the remainder. r t t t e r _a
,,u - - - -
, 7..
ENCLOSURE 4 Electro-Test Inc. Cable Electrical Measurements Cables: ,
- 1. AC rated voltage test l
- l' ma 56 cable specimens
- 5 na 5 cable specimens (4A, 4B. 4C, 4D & 9C)
Not yet tested 3 cable specimens (5A, SC & SD) l
- 2. Insulation Resistance (IR) at 300 or 500V DC l l
50 K megehm to 100 K megohm 49 cable specimens ' 1 K megohm to 49 K megohm 5 cable specimens 75 megehm to I K megohm 2 cable specimens
<1 megohen 5 cable specimens 4 i (4A, 4B, 4C, 4D & 9C) 1
- 3. DC voltage withstand test at 3X voltage j
< .5 ua 52 cable specimens
> .5 va but c 70 va 3 cable specimens
> 5000 va 6 cable specimens (4A, 4B, 4C, 4D, 9C & SB)
.- Note: l
- 1. Cable Type 5 specimen was under test. One of the 48 wires did not pass the DC voltage withstand test and had a leakage current of > 5000 va.
Splices: , Worst case values from tests of Raychem Class IE splices with cables attached: Cable Type AC IR DC Cable Description ma megehm ua 6A .15 100K .01 3 cond/10/1000V Rockbestos ) 10A .15 75K .04 4 cond/12/IDOOV Okonite i IIA .15 > 100K .035 3 cond/10/1000V Okonite j Worst case values for Duke supplied armor protected splices with cables attached: Cable Type AC IR DC Cable Description ma megohm va 21 .22 50K .02 1 cond/19/1000Y Brand Rex 22 .05 > 50K .02 4 cond/16/300Y Eaton 23 .04 > 50K >.01 2 cond/16/300V Brand Dex 24 .14 > 50K .01 3 cc '..' ' ' MY Nit? 25' .14 > 50K .03 3 ca. .s/ Av0V i.. corn'a - O
ENCLOSURE'~ S
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. . 7
'e THE PRODUCTION OF H9 IN LWR SYSTEMS DURING SEVERE ACCIDENTS The early production of H2 in LWR systems during a severe accident is mainly caused by the oxidation of the Zircaloy cladding by steam. This reaction produces one H2 molecula for every steam molecule reacted. The geometry of an undamaged (i.e., unmel'ted core) is very susceptible to oxidation by steam at temperatures above 2200 0 F (the current Appendix K limitation for LOCA events). The very high
- 1
- surface to volume ratio of the Zircaloy cladding allows sufficient contact i
area for the oxidation to occur. I 2 ] When an LWR undergoes a severe accident leading to loss of coolant, it i
- E y
- l. eventually reachef the stage at which the primary water is gradually boiled away leaving the core uncovered. For rapid blowdown events such as a large-break l
LDCA, accumulator injection will refill the core. If the lw pressure injection system then fails, the com water will boil off slowly at ambient pressure. . Boilaway times from the top of the core to the bottom are of the order of 20-40
; minutes depending on the exact sequence if no additional water is in.iected into :
the core by operator action (Note that at TMI-2 water was injected periodically ; by operr. tor action.) [ l ! State of knowledge calculations and direct experimental data (PBF Severe Fuel Damage Series I, Ex-Pile experiments at KfK in the FRG*) show that for an unattenuated boilaway only 15-25% of the Zircaloy cladding will be i fully oxidized prior to gross melting of the core (i.e., when temperatures l in excess of 5000 F are reached, and failure of the lower core support i 1 plate and lower plenum occur because of the high heating rate incurred.) After I j
! gross melting and core collapse into the reactor building lower cavity, the remaining unoxidized Zircaloy may undergo further oxidation depending upon
*Kernforschungszentrum-Karlsruhe in the Federal Republic of Gennany i !
i Enclosure "D" l 2 J r ---r- . ,
the availability of oxidant (steam) and the ability of the Zircaloy (present now with a. considerably lower surface to volme ratio within a large mass of molten U0 2 , Ir02 , steel, etc.) to come in contact with the steam. Very little infomation is available on the oxidation rate in such a system but experiments are planned to obtain the needed data. For accident sequences where an unattenuated boiloff is prevented by periodic water inNetions (bleed and feed), samthat more oxidation af the Zircaloy can be achieved prior to er without gross melting of the core. A situation similar to this occurred at TMI-2 where-45-50% of the cladding was oxidized. However, t is a limit to the total oxidation possible for this type of sequence aho. Maximum oxidation (i.e., maximum H2 production) will occur if the core is boiled down to certain water levels and held there by feed and bleed operations. However, the water level has i to be low enough for considerable oxidation but high enough to provide a sufficient steaming rate to preclude gross melting of the core. Calculations included by the Rogovin Study Group in their report on Three-Mile Island (Volme 2 Part 2, Appendix 11.8) illustrate the previous point. Figures 1A,18, & 1C show core temperatures achieved for uncoveries of 7-feet, 8-feet, and 9-feet respectively for the TMI-2 core. The individual curves on each plot indicate the axial location of the position from the top of the core. That is, 0 = top of core.1 = one foot down from top, 2 = two feet down from top, etc. All calculations assmed a boildown to the indicated level in t 4
l 20 minutes and holding at that level (presunably by bleed & feed) for an additional-60 minutes. To interpret L':csc curves in terms of the production i one must understand that very rapid oxidation of Zircaloy will not occur until the Zircaloy temperature exceeds 22000F. Using the kinetic equation for oxidation of Zircaloy given in NUREE/CR-0497. TREE-1280, Revision 2, one can calculate the percentage of cladding oxidized in one hour at various temperatures. Table 1 shows the results at three temperatures: 15000F, ~
~ -
0 1830 F and 22000F. Note that below 15000 only 2% oxidation will occur in one hour. Since t$e kinetics are a function of the square root of the time, only 2.8% oxidation will occur after 2 hours, etc. Therefore, for those regions of the core below say 15000F, it can be safely assumed that very little H2 will be produced fvless than 3%). Figure 1A shows, therefore, that for seven feet of the core uncovered for one-hour,v4 feet of the cladding is potentially totally oxizlizable. That is, a total H2 production of only 4/12 y 100 = 33%. Ecr eight feet of the core uncovered (Figure 1B), 5 feet of the core could be totally oxidized leading to a total possible H2 production from the clad of 5/12 x 100 = 42%. This latter case is considered to be close to what h actually happened at TMI-2. Note tha'. for the above two situations no part of the core exceeded 44000 F which is above the cladding melting of point but below the fuel (U02) melting point of fust over SD00 0F. Thus, no gross melting of the core would have taken place except f or some dissolution of 00 into the liquid 2ircaloy (in the upper two feet only). 2 Figure 1C shows the results for a nine-foot uncovery and hold. In this case 7/12 x 100 = 60% H2 would be producar. However, note th t 2.5 feet of R.a
- .7..
core will now reach the UO 2 melting p int and core collapse and gross melting ' l c111 begin. This large unst may cause failure of the lower plate and i a lower plenwn and lead to a core on the floor" event. That is, the core , would not be recoverable. Atote also that even this very severe event oxidized only : W60% of the cladding prior to core collapse. , i Therefore, it seems highly unlikely that an event can be constructed to P result in oxidation of more than about 50-60% of the cladding and still maintain i the core in a recoverable state (i.e., no gross melting and relocation). ; Experimental data to confirm such analyses will be performed on full-length 5 - fuel bundles by NRC in the EU reactor in Canada over the next year. Follow-on tests will be needed to confirm the calculations for various uncovery levels. Table 1. Percent Cladding Oxidized at Yarious < Taaperatures in One Hour
]
1 Temperature Present Clad (O )F, 0xidized { 0 2200 (1200 C) 20.5 o 1 1830 (1000C) 7.8 I 1 1500 (800'C) 2.1 j j i
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3 O , Accident Likelihood and Potential for " Arrested" Seoventes in PWRs with Laroc, Dry Containments Since the tire of the original staff proposals on interim hydrogen rulemaking (SECY B1-245A), a considerable amount of analysis has been performed with re-spect to the likelinood of severe accident sequerces in PWRs with large dry containments. A number of PRAs have been performed under the sponsorship of both the NRC and utilities, and have identified those sequences which contribute nest importantly to the respective estimates of the frequency of 1 core welting. Such infomation has been catalogued and studied as one task of , 1 the RES-sponsored Accident Sequence Evaluation Progrsm (ASEP). Based on this catalogue, it has been determined that transient-initiated accidents and small
- LOCAs are, in general, the most important types of sequences in most PWRs.
! itore specifically, ASEP identified five accident sequences which are generally important to the characterization of the core melt feequency in PWRs. These sequences are:
- Transient without early emergency core cooling
- Transient-induced LOCA without early emergency core cooling
- Small pipe break LOCA without early emergency core cooling
- Transient without reactor shutdown (ATWS)
T
- Transient without early emeroency core cooling and without contain- l L
sent heat removal. ENCLOSURE #El
y ._ ._ . 2 An important characteristic of these sequences (except ATWS) is that the loss of water _from,t,he reactor coolant systen is rather gradual, and thus the pro-cesces of corel uncovery, core damage, and gross core melting can be protracted. For this reason, there can be a significant potential for operator intervention to reste cooling of the fuel. In general, this racovery potential has not been in%6ed in published PRAs. Ora exception to this is the incorporation of the recovery of offsite electric power during a station blackout accident sequence. In this sequence, a recovery prchability (as a function of time) based on historical data is typically included. The potential benefit of accounting for operator recovery actions has been
~
evaluated as part of the NR',-sponsored IREP PRA of the Arkansas Nuclear One (AND) Unit I plant. In this study, faults which by themselves or in com-bination with other faults could lead to failures of important systems were l l categorized as recoverable or non-recoverable (e.g., an inadvertently closed valve was recoverable, a plugged valve was not). For each recoverable fault, a recovery location was identified and a critical time developed, based on how quidly the specific associated function had to be restored. The latter time was based on the onset of an " unacceptable" core conditien. For the IREP i i study, the unacceptable condition was specified as core uncovery, which for l transient events and small LOCAs occurred in roughly forty minutes. Using this data and central data on the probabilities of human restoration actions as. functions of time, a recovery probability was developed for each recoverable fault. , j j ! I i
,. c .
3 Sinct the publication of the AND IREP, this analysis has been extended to con- i l sider restoration tier correlated to an unacceptable core condition defined as onset of core' melting, rather than core uncovery. In effect, this extended the recovery time for small LOCAs and transients (including loss and subsequent I recovery of offsite power) from roughly forty minutes to a time on the order of seventy minutes. Ttble 1 displays the results of these analyses. i d 4
- In addition to these assessmer.ts cf recovery potential, the AND IREP also iden-
! tified two accident sequences in which it was possible that the operable core 4 1 ! cooling equipment might be adequate to prevent gross core melting, even though ! it would apparently not prevent core encovery. In Table 1. the last column (" alternative success criterion") includes this possible effect for the two j specific sequences. The table I results suggest several conclusions. With respect to operator
- recovery actions, it appears that, for this ANO case study:
- operator recovery acticns before core uncovery can have a significant
- effect on estimates of core melt frequency; and j l
4
- if recovery actions have not occurren by the time of core uncovery it is !
l uialikely that such actions will be taken before the onset of gross core i j melting. ; I I i ) I
' . . ,3 4
The second conclusion results from two factors. First, many recoverable faults - can be rectified very quickly, and, as such, are accounted for in the pre-core uncovery phast (thus the first conclusion). As time proceeds, however, the incremental changes in probability due to recovery actions decrease rapidly. That is, the recovery model corresponds to the intuitive concept that if the ~ operators have been unable to correctly diagnose and account for initial faults by the time the core is first uncoverred, then the likelihood of such actions by the time of gross core melting is relatively low. In addition, as such credit i for recovery. actions d creases the effect of recoverable faults, non-recoverable faults become increasingly more important and finally prevent any further gain. , Thus, as Table 1 shows, the only incremental reduction in probability occurs in s j the station blackout sequence (T(LOP)"KTiiTEE SPRM5 EE), due to the somewhat greater likelihood of offsite power restoration prior to core melting. The j inclusion of the probability reductica due to an alternative success criterion 1 1 for two sequences has a more pronounced effect, but the overall reduction stil. k ] r mains rather small.
- l i
In sumary, for this case study, it can be concluded that recovery actions are a significant factor in preventing severe core damage. However, such actions are most likely to occur before any damage has occurred (i.e., before core I uncovery). If the accident has progressed to the point of initial core 1 uncovery, then, under the assumptions used for this case study, it has rela-tively little chance of being arrested before gross core melting. 1 . 1 4 d
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TABLE 1(continued) .Il TERMS SLOCR 2? Small LOCA .4"-1.2") - SLOCA
;7J Small LOCA 1.2" - 1.7" -
1.7"-4")) Si.0CA , Small LOCA ; T Transient initiated by loss of offsite power T(ILDP)
, Transient initiated by failure of 4160 VAC bus A3 . .
TI , Transient initiated by failure of 125 VDC bus D01 4 Transient initiated by failure of 125 VDC bus D02 7 Til TI Transientwithallfront-linesystems(ECC,AFW,etc.) initially available i D'tf' Failure of emergency core cooling system in in.lection mode { ECT Failure of emergency core cooling system in ' recirculation mode ~ ' IPIULYT Failure of containment spray system
)
i 37RY Stuck-open safety / relief valve NW Failure of auxiliary,feedwater system RBC Failure of containment fan cooling system '
.[
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9 _ m c. ., . .. .. i
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^
$AICLOSUR[ 5 frMMGWD n gar Fwae d :
/v/Wo9' ,
l COMMISSION BRIEFING ON HYDROGEN CONTROL FOR MARK 111 BWRs 3 AND ICE CONDENSER PWRs . p E l l DECEMBER 10, 1964 j JiALC01.M l., 'ht.' 'T 443-7923 j DECEMBER 10, 1984 l
~
STATUS OF HYDR 0 GEN RULE . EXTEND $DCOMMENTPERIODEXPIREDAPRIL8,1982 I MAJOR COMMENTS RECEIVED
- 6QV5 DEFER TO SEVERE ACCIDENT RULEMAKING ggggAg/W 75% METAL-WATER REACTION IS T00 HIG REQUIREMENT FOR CONTAINMENT INTEGRITY T00
' RESTRICTIVE SHOULD NOT HAVE TO ANALYZE ALTERNATIVE SYSTEMS COLD SHUTDOWN IS INCONSISTENT WITH LIC. BASIS SCHEDULES ARE UNREALISTIC MEETING WITH COMMISSIONERS 11/09/83 RESPONDED TO COMMISSIONERS COMMENTS 12/28/83 INITIAL NTS TEST RESULTS JAN. 33 REQUESTED COMMISSIONERS DELAY DEC1SION 03/15/34 JULY 84
?
. POST TEST EXAMlHATIONS COMPLETED .
^
fos (" T63r 6cdt'tWA W WS oF CA9 c G fDST- % r GK4M/WA M ^'.5 oP Gi:?vifME W
'hlCOLM L. ERNST 443-7923 DECEMBER 10, 1934
'~
~
BACKGROUit0 FINAL RULE - PENDING CP/ML JANUARY 15, 1982 o H2 CONTROL - 100% METAL-WATER REACTION o CONTAINMENT PRESSURE 1 o SYSTEM AND COMPONENT FUNCTION o OTHER DEGRADED CORE ISSUES : FINAL RULE - MARK 1 AND 11 DECEMBER 2, 1981 o INERT CONTAINMENT o RECOMBINER CAPABILITY o HIGHPOINT VENTS l PROPOSED RULE - MARK III AND ICE CONDENSERS DECEMBER 23,1931 ) (COMMENT PERIOD i o H9 CONTROL - 75% METAL-WATER EXTENDED ON ;
' REACTION FEBRUARY 25, 1932) !
i o EQUIPMENT GUALIFICATION j . i o' ARALYSES 4 1lALCoLM L. ERNST , 443-7923 l DECEMBER 10, 1984 { 5
l
~
- l PRINCIPAL FEATURES OF R!llE APPLIES ONLY TO MARK 111 AND ICE CONDENSER PLANTS ,
o REQUIRES HYDR 0 GEN CONTROL SYSTEMS. 75% FUEL CLADDING - WATER REACTION NO LOSS OF CONTAINMENT INTEGRITY o FUNCTIONING 0F SYSTEMS At!D COMPONENTS DURING AND AFTER. HYDROGEN BURN HEEDED TO ESTABLISH AND MAINTAIN SAFE SHUTDOWN AND CONTAINMENT INTEGRITY LOCAL DETONATIONS INCLUDED.UNL SS UNLIKELY TO . OCCUR 4 o SUPPORTING ANALYSES o IMPLEMENTATION : . t SUBMIT IMPLEMENTATION SCHEDULE WITHIN 6 MONTHS MUTUAL AGREEMENT - LICENSEE AND NRC STAFF l MALCOLM L. ERNST
! 443-7923 DECEMBER 10, 1924
u ,~
~
a I i
SUMMARY
OF NTS TESTS PURPOSE Qgvgt
- 6M [W VAMPen0N ofM Amtcyrien o O G L.,5 4 o - ETiRii M LER SCALE TEST DATA ON Hg COMBUSTION o J CCESS EFFECTC OF 4C" ITER LCCATIGii Gn EUR"!N l o CTAI" CENERIC IiiFORisTiOW Gii EQUIFiiEiii RESF0iiSE /penov U" DER "YDROGE" EUR" CONDITIONS - ###d*# #
M6Afeg6 Gbvif /GSM LW4ta!& 03 SML , 1 DESCRTPT10N l Af12&C /+s foM66 Mod FA'fNr.3 \ l
) o HYDROGEN, COMB'GTION IN 52' D SPHERICAL VESSEL o 5 - 13 V/0 H2 (PREMIXED) o 40 V/0 STEAM o 0.4 - 3.2 KG/ MIN H (CONTINUOUS 2 INJECTION)
- o EQUIP!iENT EXPOSED TO MULTIPLE BURNS PRESSURE TRANSMITTERS I -
VALVES MOTORS CABLES
- PENETRATION ASSEMBLIES, ETC.
^ i o EQUIPMENT WAS OPERATIONAL EXCEPT FOR CA2LE
)
(cdct$rcA & RESULTS gggggygNS
- o CABLES EXTERNALLY DAMAGED DURING HIG ONCENTRATION !
TESTS i POST TEST EXAMINATION REVEALED ALL BUT TWO CABLES WERE FUNCTIONAL o ELECTRICAL PENETRATION ASSEMBLIES HAD 38 0F 52 LEADS j PASS POST TEST EXAMINATIONS NOT KNOWN IF FAILURES DUE TO SINGLE BURN OR CUMMULATIVE EFFECT OF MULTIPLE BURiiS ; 4 . l MALCOLM L. ERNST i 443-7923 DECEMBER 10, 1984 )
~
ST4FF' CONCLUSIONS CONCERNING NTS TEST RESULTS , o TESTS NOT DIRECT DEMONSTRATION OF EQUIPMENT SURVIVABILITY DUE TO DIFFERENCES IN SCALE AND CONDITIONS 2/20 CSALE TECT- 7ff #80 N i LOCATIONS OF EQUIPMENT
- -;' EAT LDAD5 nERE cGnER [Mo 845 0 # M EQUIPMENT NOT AGED eF CABLES NOT ENCLOSED IN C0i1DUITS [6&f664 3 ,
/V9 M / 4 9 M
- W/ !
CABLES NOT ENERGlZED M l WATER SPRAYS NOT AS DENSE ! o SOME TEST CONDITIONS CONSERVATIVE, SOME NONCONSERVATIVE l l MALCoLM L. ERusT 443-7923 DECEMBER 10, 1934
RESEARCH ACTIVITIES UNDERWAY o HCOG 1/4 SCALE DIFFISION FLAME TESTS FOR MARK 111 EARLY 1935 n SNL HORK : SMALL SCALE TESTS 7 ' ( EARLY 1986
- P._, MENTAL AND ANALYTICAL STUDIES BASED ON EARLY 1986 !
~ NTS GENERATED DATA TO DETERMINE STRESS ON ! EQUIPMENT l LOCAL DETONATION ANALYSES.FOR LARGE, DRY PWRs U2TE 1935 DIFFUSION FLAME MODEL DEVELOPMENT MID 1985 l l
- MALCOU4 L. ERNST l 443-7923 l DECEMBER 10., 1934 l
. ;~
r I l i i r STAFF RECOMMENDATIONS o APPROVE H2CONTROL RULE FOR MARK III BWRs AND ICE CONDENSER PWRs o I)EFER RULEMAKING FOR LARGE, DRY PWRs UNTIL COMPLETION OF RESEARCH EFFORT o STAFF WILL RECOMMEND WHETHER RULEMAKING FOR LARGE, DRY PWRs IS WARRANTED IN MID 1986 j l 4 l i-J l
)
'l 4 i :
MALvou4 L. ERNdi l 443-7923 DECEMBER 10, 1984 ;
l 1
~-
ffNCLUSIONS o RATIONALE FOR THE STAFF RECOMMENDATIONS IN SECY 83-357 , CODIFY PAST COMMISSION PRACTICE (SEQUOYAH - MCGUIRE) TMI EXPERIENCE 1 GREATER INHERENT CAPABILITIES OF LARGE-DRY CONTAINMENTS JUDGMENT THAT CONDITIONS DURING BURN WOULD BE NO WORSE THAN CONDITIONS DURING LOCA o NTS TESTS 7 CABLE CAST DO EGARDI UIPME ATION eMENT , o ANALYSIS OF NTS TEST SHOWS: f(GUIR l
- EQUIPMENT FUNCTIONAL, SOMEdCABLES AND ELECTRICAL pp3 PENETRATIONS NOT FUNCTIONAL
- TEST din unT s!ugLA7: n:AL Acc: peut rnNDITIONS 0 STAFF BELIEVES THAT NTS EXPERIMENTS DO NOT INVAllDATE RECOMMENDATIONS MADE IN SECY 33-357 i
MALCOLM L. EansT > 443-7923 DECEMBER 10, 1934
.. .: 1
. l l
i J i STAFF RECOMMENDATIONS ! o APPROVE H 2 CONTROL RUif 'OR MARK Ill BWRs AND ICE CONDENSER PWRs , , o STAFF WILL RECOMMEND WHETHER RULEMAKING FOR LARGE, DRY ; PWRs IS WARRANTED IN MID 1986 i. i 5 A i b 1 i
. t
.MALCOLM l. ERNsT 443-7923 i DECEMBER 10, 1984 !
g an , b eeYN e UNITED STATES E g NUCLEAR REGULATORY COMMISSION 5 a j WASHINGTON, D. C. 20555 k*****,/ July 3, 1985 J l MEMORANDUM FOR: Cecil 0. Thomas, Chief i Standardization and Special Projects Branch Division of Licensing FROM: Dennis M. Crutchfield, Assistant Director ' 4 for Safety Assessment Division of Licensing ,
SUBJECT:
PROOF AND REVIEW 0F THE CESSAR SYSTEM 80 TECHNICAL i SPECIFICATIONS The enclosed technical specifications (TS) for CESSAR System 80 Amendment 10 i (CESSAR 80) are being forwarded to you to be sent to Combustion Engineering - (CE) for proof and review. The letter transferring the enclosed technical ' specifications to CE should include the following: a request to return their comments by July 17, 1985; a request to justify by discussion any changes , requested to the enclosed TS and a request to provide the numbers in the ! CESSAR FSAR that should be in CESSAR 80 TS Tables 2.2-1, 3.3-2, 1.3-4 and 3.3-5. CE should be informed that the enclosed CESSAR 80 TS arE based on its i submittal and the staff's comparison of the NSSS part of the PVNGS-1 TS to the l Combustion Engineering STS Revision 3 as to whether differences were systems ; ! 80 differences or plant specific differences t The attachment provides an NSSS TS for the CESSAR System 80 standardized plant ! design. Because Falo Verde Nuclear Generating Station Unit One (PVNGS-1) is the first CESSAR 80 plant, the CrSSAR 83 (as opposed to BOP) part of the PVNGS-1 TS has beec, selected as a straw-man for an NSSS Standardized Technical Specification (STS) for CESSAR 80. Therefore, the attached proof and review ; CESSAR 80 TS for CE's review and comments are r marked-up copy of the NSSS ; part of the PVNGS-1 TS. Plant specific aspects of CESSAR B0 design in the ; i enclosed marked-up PVNGS-1 TS have either been annotated with references to the applicant's Safety Analysis Report (SAR) or deleted from the TSs. 1 i B507170544 B50703 I 7F ADOCK 05000470 l CF '
July 3, 1985 Mr. Jack Donohew, of ORBf 5 will be available during the proof and review period to answer any questions which arise. He is located in Room:310, i Phillips, and his telephone extension is 49-27176. < Even if CE has no comments and is in agreement with the technical
- specifications content, it is requested that a written response from CE to that effect be provided by the above specified date.
(
*p . l
- Dennis M. Crute field, ssis ant Director i for Safety Assessmen Division of Licensing
Enclosure:
CESSAR System 80 Amendment 10 Proof , 3 and Review Technical Specifications ! 3 cc: w/o enclosure a H. Thompson ; J. Zwolinski i j- S. Brown j 3 C. Moon ; P. Moriette ; . I 4 i I
- l 3
i 1 i i l 1 a' 3 a 3 A v
.]
July 3, 1985 a i Mr. Jack Donohew, of ORB #5 will be available during the proof and review period to answer any questions which arise. He is located in Room 310, Phil ips, and his telephone extension is 49-27176. Even if CE has no comments and is in agreement with the technical specifications content, it is requested that a written response , from CE to that effect be provided by the above specified date. I Dennis M. Crutchfield, Assistant Director 4 for Safety Assessment Division of Licensing 3
Enclosure:
CESSAR Systee 80 Amendment 10 Protf I and Review Technical Specificatir ns , i cc: w/o enclosure !
- H. Thompson
J.'Zwolinski ! 2 S. Brown i C. Moon
- l. P. Moriette
] DISTRIBUTION 1 JOEEERarDEFEe
! TSRG File !
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(. l/ K100!&RTfKlCCFi M) l TECHNICAL SPECIFICATIONS ~ n 1
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o UNIT NO. f DOCKET NO. A J i 1 I i , J i l t
INDEX DEFINITIONS l I i PRODF & Em CDP" . 10 DE F INITIO.N._S 1.1 ACTI0N...................................................... 1-l ! 1.2 AXIAL SHAPE INDEX........................................... 1-1 1.3 AZIMUTHAL POWER TILT........................................ 1-1 1.4 CHANNEL CALIBRATION..... ................................... 1-1 1.5 CHANNEL CHECK............ ................................. 1-1 1.6 CHANNEL FUNCTIONAL TE5T..................................... 1-2 1.7 CONTAINMENT INTEGRITY................................ ...... 1-2 . 1.8 COI1 ROLLED LEAKAGE.......................................... 1-2 1.9 CORE ALTERATION............................................. 1-2
? 1.10 DOSE EQUIVALENT I-131....................................... 1-3 g
1.11 E-AVERAGE DISINTEGRATION ENERGY............................. 1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME.................... 1-3 ' 1.13 FREQUENCY N0TATION.......................................... 1-3 1.14 GASEDUS RADWASTE SY5 TEM..................................... 1-3 @dE) 1.15 IDENTIFIED LEAKAGE.......................................... 1-3 ' 1.16 MEMBER (5) 0F THE PUDLIC..................................... 1-4 (M) i 1.17 0FFSITE DOSE CALCULATION MANUAL............................. 1-4 ( M>) 1.18 OPERABLE - OPERABILITY...................................... 1-4 , 1.19 O PE RAT I ON A L MOD E - M0D E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . s .... 1-4
-- 1.20 PHYSICS TEST 5............................................... 1-4 1.21 PLANAR RADIAL PD. KING FACTOR - F .......................... 1-4 1.22 PRESSURE BOUNDARY LEAPAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 ;
2.23 PROCESS CONTROL PR0 GRAM..................................... 1-5 (,6Cf) 1.24 PURGE - PURGING............................................. 1-5 (60?) 1.25 RATED THERMAL P0WER......................................... 1-5 i 1.26 REACTOR TRIP SYSTEM RESPONSE TIME........................... 1-5 1.,27 REPORTABLE EVENT............................................ 1-gy 1 4 5HUTDOWN MARGIN................... ........................ 1.ph51TE800NDARY.............................................. 1-6 (.bb) 1-6 14s0rTWARE..................................................
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3 i
PROOF & RIVlEW COPY INDEX f DEFINITIONS i SECTION PAGE
- 1. 0tIDIFICATION.............................................. 1-6 (,N[) !
- 1. SOURCE CNECK................................................ 1-6 (%P) 1-6 1.[M'TAGGEREDTESTBASIS.............................
- i. NEtMAt powEa............................................... 1-6
- 1. UNIDENTIFIED LEAKAGE........................................ 1-6 Nat51alc1Eo AntA........................................... i-6 c.isor) i.pVENTILATION 1.
EXHAUST TREATMENT SYSTEM 1-7 (&p)
. . . . .(&p) i . # van I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-7 ....
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INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PAGE SECTION 2.1 SAFETY LIMITS : 2.1.1 REACTOR C0RE............................................. 2-1 , 2.1.1.1 DNBR..................................................... 2-1 2.1.1.2 PEAK LINEAR HEAT RATE.................................... 2-1 2.1.2 REACTOR COOLANT SYSTEM PRE 55URE.......................... 2-1 , 2.2 LIMITING SAFETY SYSTEM SETTINGS r 2.2.1 RE ACT OR T RI P SET P0l NT S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.2.2 CORE FROTECTION CALCULATOR ADDRESSABLE CONSTANTS........... 2-2 t BASES . SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 uEACTOR C0RE............................................... B 2-1 P. 1. 2 REACTOR' COOLANT SYSTEM PRE 55URE............................ B 2-2 s
.a '
-2. 2 LIMITING SAFETY SYSTEM SETTINGS
- 2.2.1 REACTOR TRIP SETP0lNTS..................................... B 2-2 2.2.2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS........... B 2-7 PROD? & REVIEW Ei!
to lEssAth. AlsSS, y m: '/:= =:: : III
INDEX LIMITJNG COND1TXONS FOR OPERAT3ON AND SURVEXLLANCE REQUXREMENTS V PAGE SECTION 3/4.0 APPLICABIL1TY.................... ........................ 4 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS , 3/4.1.1 BORATION CONTROL 3A M SHUTDOWN MARGIN - Tc o l d > 210 % . . . . . . . . . . . . . . . . . . . . . . . SHUTDOWN MARGIN - T $2104........................ 3M 1-3 cold MODERATOR TEMPERATURE COEFFICIENT..................... 3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY................... 3/4 1-6 3/4.1.2 BORATION SYSTEMS . FLOW PATHS - 5HUTDOWN................................. 3/4 1-7 FLOW PATHS - 0,PERATING................................ 3/4 1-8 3/4 1-9
~
CHARGING PUMPS - SHUTD0WN............................. CHARGING PUMP 5 - 0PERATING............................ 3/4 1-10 3~ 3/4 1-11 BORATED WATER SOURCES - SHUTD0WN. . . . . . . . . . . . . . . . . . . . . . BORATED WATER SOURCES - OPERATING..................... 3/4 1-13' BORON DILUTION ALARM 5................................. 3/4 1- M [ 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA P051 TION.......................................... 3/41-/"3% POSITION INDICATOR CHANNELS - OPERATING. . . . . . . . . . . . . . . 3/4 1-25 POSITION INDICATOR CHANNELS - SHUTDOWN................ 3/4 1-26 4
., CEA DROP TIME......................................... 3/4 1-27 SH'ITDOWN CEA INSERTION LIM 17.......................... 3/4 1-28 REGULATING CEA INSERTION LIMIT5....................... 3/4 1-29 PR005 & RM 00'Y s>
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-72;'!C: = ; IV
1 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION _PAGE i 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE........................................ 3/4 2-1 ) 3/4.2.2 PLANAR RADIAL PEAKING FACTOR 5........................... 3/4 2-2, 3/4.2.3 AZIMUTHAL POWER TILT.................................... 3/4 2-3 3/4.2.4 DNBR MARGIN......................... ................... 3/4 2-5 3/4.2.5 RC5 FLOW RATE........................................... 3/42-/D1 , 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE.................... 3/42-)fm 3/4.2.7 AXIAL SHAPE INDEX....................................... 3/42-M2,. I 3/4.2.8 PRESSURIZER PRES 5URE.................................... 3/42-pfl3 3/4.3 INSTRUMENTATION . 4 _ -3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION...................... 3/4 3-1 . _~ 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM c- INSTRUMENTATION....................................... 3/4 3 k i 3/4.3.3 MONITORING INSTRUMENTATION , RADIATION MONITORING INSTRUMENTATION................. 3/4 3- [6CP) INCORE DETECTOR 5..................................... 3/4 3-fff7 j SEISMIC INSTRUMENTATION.............................. 3/4 3-#f.W(SOP) l METEOROLOGICAL INSTRUMENTATION....................... 3/4 3-pfpg N) ! REMOTE SHUTDOWN SYSTEM.M.mme.9 Mind...... 3/4 3-se+. i 10 0 POST-ACCIDENT MONITORING INSTRUMENTATION. . . . . . . . . . . . . - -
- 6YSTEMS
' 4 '
FIRE DETECTION INSTRUMENTATION....................... - gpy i j LOOSE-PART DETECTION INSTRUMENTATION................. 3/4 3% (.6Cf) ;
~' i RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMEN 3/4 3- (00P) l
- '*'+ 5 +Toesms o@T AT ION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4.4 REACTOR COOLANT SYSTEM j
; 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION
- STARTUP AN') POWER 0PERATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 HDT STAND 8Y............................................. 3/4 4-2 i HDT SHUTD0WN............................................ 3/4 4-3 COLD SHUTDOWN - LOOPS FILLED............................ 3/4 4-5 C0to SnUTo0WN - LOOPS MOT rIttro........................ u. 4-s l
10 cm35M(_1T.s-/osss _ _ - .. . v PRODF & RH COPY
INDEX LIMITING CONDXTION FOR OPERATf0N AflD SURVEILLANCE REQU3REMENTS s r SECTION PAGE 3/4.4.2 St.TuTY VALVES b. : 5HUTDOWN............................................. 3/4 4-7 ; OPERATING............................................ 3/4 4-8 3/4.4.3 PRE 55URIZER PRE 55URIZER.......................................... 3/4 4-9 AUXILIARY 5 PRAY...................................... 3/4 4-10 i 3/4.4.4 STEAM GENERATOR 5........................................ 3/4 4-11 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEM 5.............-............... 3/4 4-18 CfoF) OPERATIONAL LEAKAGE.................................. 3/4 4-19 3/4.4.6 CHEMISTRY............ ................................. 3/4 4-22' 3/4.4.7 SPECIFIC ACTIVITY....................................... 3/4 4-25 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SY5 TEM............................... 3/4 4-29 PRE 55URIZER HEATUP/COOLDOWN LIMIT 5................... 3/4 4-32 DVERPRE55URE PROTECTION SYSTEM 5...................... 3/4 4-33
" 3/4.4.9 STRUCTURAL INTEGRITY.................................... 3/4 4-34 I
i 3/4.4.10 REACTOR C00LAN' SYSTEM VENT 5..................e......... 3/4 4-35 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY IK1ECTION TANK5.................................. 3/4 5-1 l 3/4.5.2 3/4 5-3 ECCS SUBSYSTEMS - Teold350'F.........................
> l u 5.3 ECCS SU SYST EMS - T,,,, < 350 r. . . . . . . . . . . . . . . . . . . . . . . . . u 5-T j u4.5.4 RE RUE u MG mER T=. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . u 5-e !
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i IMDO: LIMITI"G CONDITIONS FOR OPERATION AND SURVilLLANCE REQUIREMENTS 3 j SECTION PAGE 3/4. 6 CONTAINMENT SYSTEMS l 3/4.6.1 PRIMARY CONTAltMNT l CONTAINMENT INTEGRITY.....s....... ................... 3/4 6-1 (3:)P) CONTAINMENT LEAKAGE.................................. 3/4 6-/f ( @ f) ; ! CONTAINMENT AIR LOCK 5................................ 3/46-ti( I NTERN AL P RE5 5URE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . k ) AIR TEMPERATURE..... .......................... 3/4 6-7f (@@ I- CONTAINMENT % 5TRUCTURAL INTEGRITY.............. 3/4 6-f,$. ( % i2 CONTAINMENT VENTILATION SY5 TEM....................... 3/4 6-f i ( N ) i 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS t S i CONTAINMENT SPRAY SY51LM............................. 3/4 6- [ - l 3/4 6-pr3 l N : E10MEF . h 5 crcWNC b ................ 5W h4 M ............ W 6- 85 GOD 3/46-Pf6 j 3/4.6/4 C TAINMENT ISOLATION VALVE 5............................ I ! 3/4.6.f y COMBUSTIBLE GAS CONTROL KYDROGEN MONITOR 5.................................... 3/4 6- (M ELECTRIC NYDROGEN REC 0 MINERS........................ 3/46-db(gof) j HYDROGEN PURGE CLEANUP SYSTEM.(8t0M.9tQ...... 3/4 6-h (g) WYIMt>C-FA MWOG, sysam 3M 10 (6V) ) w .(,.(, esTgemoo eea%st A\R cLEAeur sygram (.ofmoceL) .. . . iW+ 6-Io (600 l ( V4 7 VfKum 1 %usF vess (onlormL) g6-io (600 .. . ! 4.40TA\0MEr6 'ISOLATioOVM.VED Mb C.}%tEL. (0EQ ; ! 'fREssuRIERT160 sysWM (.cf* M) ! l - J d ' BO' W S $ M.. { ,. r ) S s s . $ 3 5 ygg Yk ; l ! l
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PR00F & REVH COPY INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS b SECTION - _PAGE l 3/4.7 PLANT SYSTEMS AuxiL sy ! 2 3/4.7.1 TURBINE CYCLE SAFETY VALVES . . . . . . . . . . . . . . . . . . . . 3/4 7-1 3/4 7-4 (M) r 2 7 FEEDWATER SYSTEM , . . . . . . . . . . . . CONDENSATE STORAGE TANK . . . . .
- g. . 3/4 7-5 ACTIVITY . . ......... . . . . . . . . . . . 3/4 7-s MAIN STEAM LINE ISOLATION VA yL ............
3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE 3/4 LIMITATION 10
~
3/4.7.3 COMPONENT COOLING WATER SYSTEM . . . . . . . . . . . . . 3/4 7-di d37) . _ 3/4.7.4 SERVICE WATER SYSTEM . . . . . . . . . . . . . . . . . . 3/4 7- wn (Bor>) 3/4.7.5 E"PC 0003: "0C u,JJ M,A{E, H,y,@,K , , , , , , , , 3f4 7.g,, ( pgg) > i 3/4.7.6 FLOOD PROTECTION . . g . . . . . . . . . . . . . . . . . 3/4 7- Wij (f30 D , 3/4.7.7 CONTROL ROOM EKERGEWCY 3 CLEANUP SYSTEM . . . . . . . . . . 3/4 7-Wil(60D 3/4.7.8 ECCS PUMP ROOM EXKAUST AIR CLEANUP SYSTEM . . . . . . . . 3/4 7- 4 (O O P) 3/4.7.9 SNUSBERS . . . . . . . . . . . . . . . . . . . . . . . . - 3/4 7-X t2 l 3/1.7.10 SEALED SOURCE CONTAMINATION . . . . . . . . . . . . . . . 3/4 7 M (BOP) 3/4.7.11 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM . . . . . . . . . . . . . 3/4 7- (6C>P) SPRAY AND/OR SPRINKLER SYSTEM . . . . . . . . . . . . . 3/4 7- M CO, SYSTEMS . . . . . . . . . . . . . . . . . . . . . . 3/4 7- w g & P)
~
- HALON SYSTEM I.....................
FIRE HOSE STATIONS . . . . . . . . . . . . . . . . . . 3/4 7-wg(G0F) 3/4 7- W) YARD FIRE MYDRANTS AND HYDRANT HOSE HOUSES . . . . . . 3/4 7 (M 3/4.7.12 FIRE BARRIER PENETRATIONS . . . . . . . . . . . . . . . . 3/4 7-gt N) 3/4.7.13 AREA TEMPERATURE M')NITORING . . . . . . . . . . . . . . . 3/4 7 MM 3/4.7.1 suT00WN COOL!nG SYSTEM g . . . . . . . . . . . . . . . . 3/4 7- # g t ; z f
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} PR00!& REEY CO?Y INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION _PAGE
@ 3/4.8 ELECTRICAL SYSTEMS l
3/4.8.1 A.C. SOURCES OPEPATING . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-1 (,8bf) , swuTDOWN . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-1 (60P) 3/4.8.2 D.C. SOURCES OPERATING . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-2 ($f) SsuTDOWN . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-2 (BeP) 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS [ OPERATING . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-3 (g) [ Ss0TDOWN .......... . . . . . . . . . . . . 3/4 8-3 (@) i 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT , PROTECTIVE DEVICE 5 . . . . . . . . . . . . . . . . . 3/4 8-4 C60P) -
~~
MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES . . . . . . . . . . . . . . . . . . . 3/4 8-4 (BOP. - 3/4.9 REFUELING OPERATIONS Q M ft.ll ft TU k 3/4.9.1 BORON CONCENTRATION . . . . . . . . f . . . . .3/4 . . 9-1 314.,.2 mTRUMwm0N.......... . . . . . . . . . . . . ,i4 ,-2
-, 3/4.9.3 DECAY TIME . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-3
. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS . . . . . . . . . . . 3/4 9-4 (GOP) 3/4.9.5 C0mVNICATIONS ,,m- - . . 4 . . . . . . . . . . . . . 3/4 9-4 (C30P) 3/4.9.5 ::rJ:'.": r:::U;; . . . . . . . . . . . . . . . . . . . .
.- 3/4 9-4 (,gpy 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING . . . . . 3/4 9-4 (,43ggj 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL . . . . . . . . . . . . . . . . . . . 3/4 9-5 LOW WATER LEVEL . . . . . . . . . . . . . . . . . . . . 3/4 9-6
(.6OD 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATibN SYSTEM . . . . . . . . 3/4 9-7 3/4.9.10 WATER LEVEL - REACTOR YESSEL t. gr-cd l "3/4.9.11 MATER LEVEL - STORAGE POOL . . . . . . . . . . . . . . . 3/4 9 M D 3/4.9.12 STORAGE POOL AIR CLEANUP ,5,vW , , , , , , , , , , , , , , 3/4 9 (609) I fu s t A u EMB L tE S. ., ... .. .. ... .. .. . .. .. ..W S ~~ I
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l SECTION - PAGE 3/4.10 SPECIAL TEt7 EXCEPTIONS - 3/4.10.1 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . 3/4 10-1 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS . . . . . . . . 3/4 10-2 3/4.10.3 REACTGR COOLANT LOOPS . . . . . . .g., g.g g.g.jgg? I 3/4.10.4 CEA POSITIOhh REGULATING CEA fi INSERT 10h 3/4 lv-* LIM 11 3/4.10.5 ""~~" " '"" ^" '"~" M*lT. (f ttP 6f> fop 6.
# 4 P#450.M.FM. J/4 10-6 s #
3/4.10.6 SAFETYINJECTIONTANKf*ZZUE: . . . . . . . . . . '. . . 3/4 10-7 3(4.10 7. S AFE1Y HMCCf8CM TAWK PGE%uRE . . . ... . . .
. 3[4 (0- 8 i :
3/4.11 RADI0 ACTIVE EFFLUENTS l 3/4.11.1 SECONDARY SYSTEM LIQUID WASTE DISCHARGES TO ONSITE ! EVAPORATION PONDS
~
CONCENTRATION........................................... 3/4 11-1 C80 P, . D05t.................................................... 3f4 n. m y LIQUID HOLDUP TANKS...................................., 3/4 n-y y( g gg; 3/4.n.2 GASEOUS EFFLUENTS , DOSE RATE............................................... 3/4 n-/2 (.669. DOSE - NOBLE GASES...................................... 3/4 H ,M1 ($ i DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE F0RM..................... 3/4 H-M Z. ( M , I GASE0US RADWASTE TREATMENT.............................. 3/4 H- MI. ( @ , EXPLOSIVE GAS MIXTURE................................... 3/4 H- N'l. @ g{ GAS STORAGE TANKS....................................... 3/4 H-dt (g) j 3/4.n.3 SOLID RADI0 ACTIVE WASTE................................. 3/4 H-MIS (60f,' s/4.n.4 TOTAL 00SE.............................................. 3/4H-Wh(h08 i 3/4.12 RADIOLOGICAL ENVIRONMENTAL DONITORING i 3/4.12.1 MONITORING PR0 GRAM...................................... 3/4.12.2 LAND USE 3/4 12-1 (k CENSUS.................................. ..... 3/412-41 @lf, 3/4.12.3 INTERLABORATORY COMPARISDN PR0 GRAM...................... 3/412-X,{ ($Q
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c55sAMO-A59S-575 - ' * = ' ' ""
INDEX BASES . 1 PAGE
.S.ECTION 3/4.0 APPLICABILITY............................ ............... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BOR* TION CONTR0L........................................ B 3/4 1-1 2
3/4.1.2 BORATION SYSTEM 5........................................ B 3/4 1-2 ' 3/4.1.3 POVAB LE CONTROL A55EMB LI ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/41-/3 , 3/4.2 POWER DISTRIBUTION LIMITS , 3/4.2.1 LINEAR HEAT RATE........................................ B 3/4 2-1 3/4.2.2 PLANAR RADIAL PEAKING FACTOR 5........................... B 3/4 2-2
- - 3/4.2.3 AZIMUTHAL POWER TILT.................................... B 3/4 2-2 ,
<. l 3/4.2.4 DNBR MARGIN............................................. B 3/4 2-3 < 3/4.2.5 RCS FLOW RATE........................................... B3/42-/$ 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE.................... B 3/4 2-4 ; 1 3/4.2.7 AXIAL SHAPE INDEX....................................... B 3/4 2-4 i I 3/4.2.8 PRE 55URIZER PRESSURE.................................... B 3/4 2-4 ) l . 3/4.3 INSTRUMENTATION j l 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGIMEERED SAFETY l FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 l l 3/4.3.3 MONITORING INSTRUMENTATION.............................. B3/43-/g @p}
% 3.4 lu m e a c p g g g ,,,, g n % g) ;
PR00FHFSEW COPY 9 wss&- wv Dsss-grs XI ,
- i
INDEX PRODI& REV;EiY CD?Y BASES V SECTION PAGE i 3/4.4 REACTOR COOLANT SYSTEM : l l 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCU~' TION........... B 3/4 4-1 3/4.4.2 SAFETY VALVE 5....................... ................ B 3/4 4-/'/., 3/4.4.3 PRE 550RIZER............................................. B 3/4 4-2 9 3/4.4.4 STEAM GENERATOR 5........................................ B 3/4 4-3 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE.......................... B 3/4 4-y3 3/4.4.6 CHEMISTRY............................................... B3/44-f3 ; 3/4.4.7 SPECIFIC ACTIVITY....................................... B 3/4 4-5 3 l 3/4.4.B PRES 5URE/ TEMPERATURE LIMIT 5............................. B3/44-fdf-
- ; 3/4.4.9 STRUCTURAL INTEGRITY....................................
B3/44-/6 1 3/4.4.10 REACTOR COOLANT SYSTEM VENT 5............................ B3/44-/7 3/4.5 E*4ERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANK 5.................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 E CCS SUB 5YST' CMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B3/45-{ 3/4.5.4 REFUELING WATER TANK.................................... B 3/4 5-3 1 h W i,6*W h. &.WW l 3/4.5 COM"fAINMENT SYSTEMS l 3/4.6.1 PAIMARY C0KTAINNENT..................................... 4 3/4 6-1 (60P) ; 3/ti.G 2 DEPRESSURIIATION AW CGO LING SYSTEM 5. . . . . . . . . . . . . . . . . . . . us.s "a:cai tos ct.swt>P sysur.M . . . . . -2. 3/4 6-3/4.s.yg, CONTAINMENT 150LAT10S VALVES............................ a
. SB 3/4 g/e/f,B 6-1 3/4.c.fg consusTIntE cAs co m at................................. B 3/4 s-ri. (Sopy M.I..A f@68ATiptc f%mEX HeosT A* cLsApoP
' SYSTEM (ofTioont) . .. . . . ,.
3 3q 4--i (bop,' , 4 3/%. T ' VPrt00M RErLusae VMdE5 ( OFDWAQ. 3 g z.--I L69P ((T~ IS x NTY%IM
*Wotom).
IEEX Dn.nnt ppaCp,LJ a. T : .v. , , p, p0py . BASES i 1 SECTION 9 AGE
- 3/4.7 PLANT SYSTEMS :
i l 3/4.7.1 'iURBINE CYCLE........................................... B 3/4 P1 B3/47-[ 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION. . . . . . . . . cammwsr B3/47-//.(.60P) 3/4.7.3 W COOLING WATER SY5 TEM.......................... 3/4.7.4 -
.................... B 3/4 7 (60I) 3/4.7.5 ULTIMATE HEAT 51NK...................................... B 3/4 7-/2, (60'f) 3/4 7.6 Ft oec '.ec3TerdTTor0 C -
; - r ;- m.......................... B 3/4 7-/p,, (M) 3/4.7.7 CONTROL ROOM ESSENTIAL FILTRATION SYSTEM..... .......... B 3/4 7-Jr4, (M)
Ecc.s 3/4.7.B seelnPUMP ROOM AIR EXHAUST CLEANUP SYSTEM................ B 3/4 7-/> @ 0 D
; 3/4.7.9 5NUBBERS................................................ B 3/4 7-/$ ( M 3/4.7.10 SEALED SOURCE CONTAMINATION............................. B 3/4 7-8 (6Cf) 3/4.7.11 FIRE SUPPRESSION SY5TEMS................................ B 3/4 7-[3 ( N ) .
i 3/4.7.12 FIRE-RATED A55EMBLIES................................... 03/47-M t'4..1.t AAE A WPe[RWTi.W2,E %Tod)M(, . . . . . . . 3/4.7.13 SHUTDOWN COOLING SY5 TEM................................. 31V+ 7-} u%0F) 30V B3/47-E3 _-- U_i :: P '^L ^;;T. "J " = ?:.' "!2M - ._ ? ? '* 2 0 3/4.6 ELECTRICAL POWER SYSTEMS
~
I 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and
; ONSITE POWER DISTRIBUTION SYSTEMS................ B 3/4 8-1 (. M )
3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................. B 3/4 8-/1 ( M j 3/4.9 REFUELING OPERATIONS 3/4.9.1 BOP.ON CONCENTRATION.................................. .. B 3/4 9-1 3/4.9.2 INSTRUMENTATION......................................... B 3/4 9-1 3/4.9.3 DECAY TIME.... ......................................... B 3/4 9-1 1 3/4.9.4 CONTAINMENT % PENETRATIONS....................... B 3/4 9-1 ($()9) 3/4.9.5 C0MUNICATIONS.......................................... B 3/4 9-1 (%
- =: ';;=: "
xIII
%3A(1N>Abss-5Ts l i
EX N00: & Wim C0W l BASES , i SECTION PAGE 3/4.9.6 r "".:2 .'/ n :4....................................... B 3/4 9-/$-(60f) y 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING. . . . . . . . . - B 3/4 9-/f, %Qf) 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION................ B3/4S-/J. , 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM................ B 3/4 9-/I (,, W ) 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL ........................................... B3/49-/7 3/4.9.12 5 S E .............. B 3/4 9-/J. (@) 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN......................................... B 3/4 10-l' l - 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT,
; INSERTION, AND POWER DISTRIBUTION LIMITS................ B 3/4 10-1 1
ig 3/4.10.3 REACTOR COOLANT L00PS................................... B 3/4 10-1 3/4.10.4 CEA POSITION. REGULATING CEA INSERTION LIMITS . ... AND REACTOR COOLANT COLD LEG TEMPERATURE................ B 3/4 10-1 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY........ B 3/4 10-1 > 3/4.10.6 SAFETY 10 ECTION TANKS.................................. B 3/4 10-2
,u . - - .m -,.m.,_ _
. g , q_ g 3/4.10./7SAFETYINECTIONTANKPRESSURE.......................... B 3/4 10-2 3/4.11 RADI0 ACTIVE EFFLUENTS
~
3/4.11.1 SECONDARY SYSTEM LIQUID WASTE DISCHMGES TO ONSITE l EVAPORATION P0NDS....................................... B 3/4 11-1 (60f) 3/4 11.2 GASEOUS EFFLUENTS....................................... B 3/411-4 (,609) 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................. B 3/4 11-/t @ D
- .v4.n. 4 TOTAt o05t.............................................. B 3/4 21-/g W )
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 (N) 3/4.u.2 tue uSE cExSuS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 u-/A (00P)
- 3/4.u.3 INTERgDRATORYCOMPARISONPR0 GRAM...................... B 3/4 12 g ( % )
7 ^ ' " " "N - 1- XIV c:essqQ)ss- 37.5
- A N
rerx PR00F & RMW COPY DESIGN FEATURES , SECTION PAGE 5.1 SITE . 5.1.1 SITE AND EXCLUSION BOUNDARIES........................... 5-1 , dY) 5.1.2 LOW POPULATION 20NE..................................... 5-1 (M) 1 a me . 1
,m
,,,,,.., er m .ee . m-n o..............-.....
_ _.._ ... _ ....... e m. _ v ,. j
- 5. 2 CONTAINMENT 5.2.1 5-1 CONFIGURATION...........................................
($0 h
- 5. 2. 2 DESIGN PRES 5URE AND TEMPERATURE......................... 5-1 g9) 5.3 REACTOR CORE +
5.3.1 FUEL A55EMBLIES......................................... 5-/.i -2 5.3.2 CONTROL ELEMENT A55EMBLIES.............................. 5-f,1, i . 5.A REACTOR COOLANT SYSTEM - 5.4.1 DESIGN PRESSURE AND TEMPERATURE......................... 5-f I 5.4.2 V0LUME.................................................. SO 5.5 METEORLOGICAL TOWER LOCA1 M ................................. 5-y f, (g) ; 5.6 FUEL STOR' AGE
,5.6.1 5.6.2 CRITICALITY............................................. 5-f1 (%
DRAINAGE................................................ 5-Jr1 (ggg) 5.6.3 CAPACITY................................................ 5-f1 (g ! 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-[g, ! I i
- >m .== ; em o CE3sAxLMsss-sTs a !
PROOF & RNIEW COPY ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY............................................... 6-1 (,$gp) 6.2 ORGANIZATION ' 6.2.1 0FFSITE.................................................... 6-1 (W) 6.2.2 UNIT 5TAFF........................................... ..... 6-1 ' gy 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP FUNCTION................................................... 6-h , hf)
;609)
CO*P05 m 0N................................................ 6-ef.
=
RESPONSIBILITIES........................................... 6-pf (g
- AUTHORITY.................................................. 6T1 050P) g REC 0RDS.................................................... 6-yJ.
6.2.4 SHIFT TECHNICAL ADVISOR........... 6-FJ - ( 09) 6.3 UNIT STAFF QUALIFICATIONS.................................... 6- hh 6.4 TRAINING..................................................... 6-/j, (hh@ 6.5 REVIEW AND AUDIT
~ , 6.5.1 PLANT REVIEW BOARD l FUNCT10N................................................... 6-yf @P)
COMPOSITION................ ............................... 6-M [) ALTERNATES................................................. 6# q MEETING FREQUENCY............. ............................ 6-7 [ @ : 6d 6-d (([68 QU0 RUM..................................................... 6-d (NI RESPONSIBILITIE5........................................... AUTHORITY.................................................. REC 0P.DS.................................................... 6-/[ l ; 6.5.2 TECHNICAL REVIEW AND CONTR0L............................... 6-$ hih Ib
- f_; urenr- fNTT 1 XVI MS$A f)369-STS 1
._ .. -_ . _ . _ _ = _ _ _ _ _
INDEX PROOF & RMEW COPY ! ADMIN 35TRATIVE CONTROLS 1 i PAGE SECTION 6.E. 3 wuCLEAR SAFETY GROUP i 1 FUNCTION................................................... 6-}d. (g , C0se0sITiON................................................ 6-ws gg 6- W h (B V ) CONSULTANTS......... ...................................... ] R m Ew..................................................... 6-xff, gp;
- Au o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-y1%. 3 AUTHORITY.................................................. 6-M. ( )
I REC 0RDS.................................................... 6- g g ( g @ {6 REPORTABLE EVENT ACT10N...................................... 6- 4 (109) ' 6.7 SAFETY LIMIT VIOLATION....................................... 6-[T, (k) 6.8. PROCEDURES AND PROGRAMS...................................... 6-[f,. (k) 6.9 REPORTING REQUIREMENTS . 6.9.1 ROUTINE REPORT 5............................................ 6-)Aff., ( W } STARTUP REP 0RT............................................. 6-If h { ANNUAL REP 0RTS............................................. MONTHLY OPERATING REP 0RT................................... 6-)fi, fW2 ! ANNUAL' RADIOLOGICAL ENVIRONMEN1AL OPERATING REPORT......... 6-yf1 6->rg @\ 08)h ) } SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT............. 6-)dg (%h f (ej
- 1. . 2 SPECIAt REPOR1s............................................ 6-g I
6.10 RE CORD RET ENT ! 0N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-)/ [ l l , 6.11 RADIATION PROTECTION PR0 GRAM................................ 6-)ff ( 6.12 NIGN RAoIATION AREA......................................... 6-iry. (sor; i 6-23 j 6.13 PROCESS CONTROL PR0 GRAM..................................... i
=a ur=y P:g) ::: , ryII l
% S-ST.s a
1 l l INbEX ADMINISTPATIVE CONTROLS SECTION PAGE I 6.14 0FFSITE DOSE CALCULATIOR MANUAL............................. 6-23 (% 6.15 MAJOR CHANGES TO RADI0 ACTIVE LIOUID, GASEOUS, AND SOLIO WASTE TREATMENT SYSTEMS........................... 6-24 (% 6.16 PRE-PLANNED ALTERNATE SAMPLING PR0 GRAM...................... 6-25 (r3Qh P100F& RECEW CD?Y . i
-o 1
l l
- i l
l
- raw .77 ( ""?!- XVIII
$ NSS .5 F5 l
INDEX PR00F & RMEW COPY LIST OF FIGURES PAGE 3.1-1 ALLOWABLE MTC M3 DES 1 AND 2............................ 3/4 1 / 3 ( M ) 3.1-2 MINIMUM BORA~ LED WATER VOLUME 5.......................... ~ 3/4 1 p l2 @ @
,-n r= LEasTri asumn:= ax:T mic:= r==..__- e/ :a i
ener nnuro t!w7r artro era nrui n?;u ?? py _ ,; ;o _ , _ 3.1-3 CEA INSERTION LIMITS V5 THERMAL POWER (COL 55 IN 5ERVICE)..................................... 3/4 1-31 @ P)
- . ;a n.* vucretinu t!- T p m_ vuseun eeuro _
*r erd 3frE) - 2/' 1 22 (c"LE! 0"T 3.2-1 DNBR MARGIN OPERATING LIMIT BASED ON COL 55 (COL 55 IN SERVICE)..................................... 3/4 2-h (60f) 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECT 10d
~
- CALCULATOR (COL 55 OUT OF 5ERVICE)...................... 3/4 2- M (@DP) 3.2-3 REACTOR COOLANT COLO LEG TEMPERATURE VS CORE POWER LEVEL.................................................. 3/42-)(gg
- 0.0 1 ZZ =ai" orc"T:e C:JT GARD Z C0;.:! _
ego ag m c;*7. t ; ; 7 3;g ;,,,,,,,,,,,,,,,,,,,,,,,,,,,,, ;f. y;; v 3.4-1 DOSE EQUILALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERKAL POWER VITH THE PRIMARY COOLANT SPECIFIC ACTIVITY
> 1.0 pCi/ GRAM DOSE EQUIVALENT I-131................... 3/4 4-3.4-2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 10 YEARS OF FULL POWER 21 j
~'
0PERATION.............................................. 3/4 4-K ( % 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST.............. 3/474 (60f) B 3/4.4-1 NIL-DUCTILITY TRANSITION TEMPERATURE INCREASE A5 A FUNCTION OF FAST (E > 1 MeV) NEUTRON FLUENCE (550*F IRRADIATION).................................... B 3/4 4- S (% 5.1-1 SITE AND D'CLUSION BOUNDARI ES. . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 l% 5.1-2 LOW POPULATION Z0NE.................................... 5-%2. M)
- m a m z w a .,.r.+..............=........../ m
-2:
6.2 1 0FFSITE ORGANIZATION................................... 6-q GOP) 6.2-2 ONSITE UNIT ORGANIZATION............................... 6-Q Mf)
- - 10 YEarr - ' XIX uswy /usss-s73
n . - .- .
,v FUEL c PP90F & R_ YEW 00PY '"ot g e s -y c.Ltr F5e M03E5 3,4' LIS N VTAT ES h r
PAGE 1.1 FREQUENCY N0TATION...................................... 1-KE' 1.2 OPERATIONAL M0 DES....................................... : 1-$ 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT j LIMIT 5.. ...................................... ........ 2-3' (,Bof) l 2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS........ 2-7 (.@@ W MONITORING FREQUENCIES FOR BACKUP BORON
! DILUTION DETECTION A - E 7- m
^
r^^5T^^^r/ ~
" ^ " ' ' *
.................... ~
3.1-1 FOR Kg f > 0.98............................. ........... 3/41-)(p 3.1-2 FOR 0.98 > Kg f > 0.97.................................. 3/41-4Q 3.1-3 FOR 0.97 > Kg f > 0.96.................................. 3/41-% i 3.1-4 FOR 0.96 > Kg f > 0.95.................................. 3/4 1.WJ.o i
- f 3.1-5 FOR K g f 5,0.95......................................... 3/4 1-) 62) 3.3-1 REACTDR PROTECTIVE Ih5TRUMENTATION...................... 3/4 3-3 3.3-2 lf REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES....... 3/43-M
' M4 :: E"Pi~! '" %;RC , ":^^Z , K 0;"'" WDt"e "TO -
- M L*? TE:!. ................w............. ...... ?/' 2-2:
4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE j REQUIREMENTS............................................ 3/4 3-)(l2
~.3-3 ENGIWEERED SAFETY FEATURES ACTUATION SYSTEM i
INSTRUMENTATION......................................... 3/4 3-%l7 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 3/4 3-)A2.$ M INSTRUMENTATION TRIP VALUE5............................. 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIME 5............... 3/4 3-%2.1 (.60PJ ) ., 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/43-)("M i 3.3-6 RADIATION MONITORING INSTRUMENTATION.................... 3/4 3p% (60P) 4.3-3 RADIATION MONITORING INSTRLW'MATION SURVEILLANCE l REQUIREMENTS............... . ......................... 3/4 3-)(% (30P'
- 3.3-7 SEISMIC MONITORING INSTRUMENTATION...................... 3/4 3-M 3g ( 609) 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENT 5............................................
i 3/4 3- d ( $0 9) 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............... 3/4 3-1)6M (.90F) 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION l SURVEILLANCE REQUIREMENTS............................... 3/4 3- W 59 (.r3er) l 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION ............. 3/43-kl 4,gg i w :mE - m , xx db$d s M$$ b3 i
j-.
'"o" PR005& RMN copy LIST OF TABLES -J PAGE ; 4.3-6 REMOTE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................... 3/4 3-M2 (% , 3.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION................ 3/4 3-% Q
- 4.3-7 POST-ACCIDENT MONITORING INSTRUMENTATION i SURVEILLANCE REQUIREMENTS............................... 3/4 3-M $
i 3.3-11 FIRE DETECTION INSTRUMENTS.............................. 3/4 3- M b (.80 h 3.3-12 LOOSE PARTS SENSOR LOCATION 5............................ 3/4 3- M [ (C3C)P) 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION......................................... 3/4 3- K N ( % 4 4.3-8 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-K Q g }
~ ~
4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED ,; OURING INSERVICE INSPECTION............................. 3/44-)( (60P) I 4.4-2 STEAM GENERATOR TUBE INSPECTION......................... 3/44-X ( 1 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES........ 3/44-X M a 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY........................ 3/4 4- % 4 4 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................ 3/44-%lb 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE 3/44-M ~ AND ANALYSIS PR0 GRAM.................................... l
~
j - 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM l WITHDRAWAL SCHEDULE..................................... 3/44-NEL., I 4.6-1 TENDON SURVEILLANCE - FIRST YEAR........................ 3/46-% (8Qh 4.6-2 TENDON LIFT-OFF FORCE - FIRST YEAR. . . . . . . . . . . . . . . . . . . . . . 3/4 6-k @ 3.6-1 CONTAINMENT ISOLATION VALVE 5............................ 3/46-M , 3.7-1 STEAM LINE SAFETY VALVES PER L00PS...................... 3/4 7-2 dOF)' i j i i i e
.... .. .or _ e/, - XXI
, "" N - S # 5
~
Wl9 T/.co Loot' OfccATict0 ugTH Fbog. h INDEX Q 6 N i h s 3 LIST OF TABLES
- h hk PAGE 3.7-2 MAXIMUM ALLOWABLE STEADY STATE POWER LEVEL AND MAXIMUM ~
VARIABLE OVERPOWER TRIP SETPOINT WITH INOPERABLE STEAM LINESAFETYVALVES.4.................................. 3/4 7-3 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM............................. 3/4 7-)( T 3.7-3 SPRAY AND/OR SPRINKLER 5YSTEMS.......................... 3/4 7- K lb ( M ) 3.7-4 FIRE HOSE STATION 5...................................... 3/4 7-M (% 3.7-5 YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE : HOU5ES.................................................. 3/4 7-%L (6(f) 4.8-1 DIESEL GENERATOR TEST SCHEDULE.......................... 3/4 8-)(d. (M) 3.8-1 D.C. ELECTRICAL SOURCE 5................................. 3/4 8-)(2. (83f) 4.e-2 BATTERY SURVEILLANCE REQUIREMENTS....................... 3/4 8-)( Z. (,8f)
/ ,
3.8-2 CONTAI MENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICE 5...................................... 3/4 8- Q N 3.8-3 WTOR-OPERATED VALVt5 THERMAL OVERLOAD PROTECTICN AND/OR BYPASS DEVICE 5................................... 3/4 8-$ (@f) 4.11-1 RADI0 ACTIVE LIQL'ID WA3TE SAMPLING AND ANALYSIS PROGRAM.. 3/411-g 4.11-2 RADIDACTIVE GASEOUS M STE SAMPLING AND ANALYSIS PR0 GRAM................................................. 3/411*[2. (k ' 3.12-1 RADIOLOGICAL ENVIR0 MENTAL MONITORING PROGRAM........... 3/4 12- Q g
?.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLE 5................................
3/4 12 Q ( @ P) 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL $ AMPLE a FI
- ANALYSI5................................................ 3/4 12- Q (f 1 B 3/4. 4-1 REACTOR VESSEL TODGHdE55. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4- (M) 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMIT 5. . . . . . . . . . . . . . . . . . . . . 5-Y3 gf
~ 5.7-2 PRE 550RIZER SPRAY MOZZLE UGAGE FACT 0R................... 5-X $ l 6.2-1 MINIMUM SHIFT CRLW COMP 051 TION.......................... 6-k (g i ' 6 O cGssA w mc == , -rnA)S$$-$33 a-. xx11 l
. ___.__________________1
1 1 l PR00F & REVIEW COPY , i I SECTION 1.0 DEFINITIONS I j > op 1 6 l l l I I i
'Q w ss q-osss-s rs l
I.0 DEFXNITSONS kbh hf The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. ACTION 1.1 ACTION shall be that part of a specification which prescribes remedial measurres required under designated conditions. AXIAL SHAPE INDEX 1.2 The AXIAL SHAPE INDEX shall be the power generated in the lower half of the core less the power generated in the upper half of the core divided by the sum of these powers. AZIMUTHAL POWER TIL'i - T q 1.3 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally i symmetric fuel assemblies.
~
CHANNEL CALIBRATION 1.4 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 which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
-CHANNEL CHECK i 1.5 A CHANNEL CHECK shall be the qua itative assessment of channel behavior during operation by observation. This determination shall include, where ,
possible, comparison of the channel indication and/or status with other I indications and/or status derived from independent instrument channels measuring the same parameter. J l
)
45DsifpSS-srs 11 l l t
"h DEFINITIONS i o
CHANNEL FUNCTIONAL TEST r 1.6 A CHANNEL FUNCTIONAL TEST shall be:
- a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify.0PERABILITY including alarm and/or trip functions.
- b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions,
- c. Digital computer channels - the exercising of the digital computer hardware using diagnostic programs and the injection of simulated i process data into the channel to verify OPERABILITY including alarm )
and/or trip functions. & J
- d. Radiological effluent process monitoring channels - thF . ANN:
at es o F r 1 pi , ST .ay e t al rfor d by ny s an st suc that ee e ntia ,[ ire .a n f .ct' na y st . The CHANNEL FUNCTIONAL TEST shall include adjustment, as necessary, of ~ the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy.
- CONTAINMENT INTEGRITY f 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, or
- 2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of i Specification 3.6.1.3, !
~ d. The containment leakage rates are within the limits of Specification )
' 3.6.1.2, ano 1
- e. The sealing mechanism associated with each penetration (e.g., welds, !
bellows or 0-rings) is OPERABLE. I CONTROLLED LEAKAGE ! 1.8 Not Applicable. j i CORE ALTERATION j 1.9 CORE ALTERATION shall be the movement or sanipulation of any component I within the reactor pressure vessel with the vessel head removed and fuel in l l the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position. WEL JRE $JT 1-2 MSSAd f11555-STS s l l
PR00F & RWlEW COPY DEFINYTIONS DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) which alone wou.d produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132, I-133,1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall.be those listed in Table III of TID-14844, " Calculation of Distance factors for Power and Test Reactor Sites." , T - AVERAGE DIf1NTEGRATION ENERGY 1.11 I shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum , of the average beta and gamma energies per disintegration (in MeV) for isotopes, ) other than iodines, with half-lives greater than 15 minutes, making up at 1 least 95% of the total noniodine activity in the coolant. N ENGINEERED SAFETY FEATURES RESPONSE TIME l l 1.12 The ENGINEERED SAFETY FEATURES RESPON5E TIME shall ba that time interval' from when the monitored parameter exceeds its ESF actuation setpoint at the
- channel sensor until the ESF equipment is capable of performing its safety - function (i.e., the valves; travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel '
generator starting and sequence loading delays where applicable. FREQUENCY NOTATION . 1.13 The FREQUENCY NOTATION specified for the perfomance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. GASEOUS RADWASTE SYSTEM g Mf an h yste design and i
%k alled 1.14 G U 'W 5 EM all r ce di ive seo ef uent yc ecti pri cool . sys ,
f ses om th prim ys tem d pr idin or dpF$y or . dup the p rpos of redu ing et al dios vity for to releas to t envir ent. IDENTIFIED LEAKAGE j
' 1.15 IDENTIFIED LEAKAGE shall be: !
i
- a. Leakage into closed systems, other than retetor coolant pump .
controlled bleed-off flow, such as pump seal or valve packing leaks I that are captured and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere from sources that are both f specifically 16cated 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 system.
% .v venvi f"@"1 1-3 dNk A1595-STS l
l
. _ _ - ____a
DEFIN1TIONS ! l MEMBER (5) 0F THE PUBLIC M , 1.16 .MB ( OF . P IC f all clu all erson who apt not ' occ .at ' , a y as cia d witA th lan T s cat ory .s no irici e o" es ft lic see, ts .tra ors, r ven ts. so e ude fror i ca gor- re rson ho ter e s' e to rvice quipe t.or or .e i d v ies Thi cate yd s in ude ersons who us port ns o the i. , r etr iona , occ .ati al, oth purp esnot/ assoc'ted th h pl . t. 0FFSITE DOSE CALCULATION MANUAL (ODCM) M , 1.17 e OF ITE 3E CA LATIO- NUAL all co in the rent me odolo and ram rs d in e cal ation offsit oses du o radi .tive ous nd l' uid e uents n the 1culati e of gase s and 1 uid eff sen' cai ing arm /tr setp nts, a in the nduct of e envi nmenta ra ologi mon' oring p ogram.
~
OPERABLE - OPERABILITY _ . 1.18 A system, subsystem, train, component, or device shall be OPERABLE cr 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 i 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 1.19 An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and cold leg reactor coolant temperature specified in Table 1.2. PHYSICS TESTS l 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and
'(of10CFR50.59,org3)otherwiseapprovedbytheCommission.1)describe PLANAR RADIAL PEAKING FACTOR - F g i 1.21 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane i average power density of the indiv'idual fuel rods in s given horizontal plane, excluding the effects of azimuthal tilt.
PRESSURE EDUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall. 6 6 iiiLi
/^"h 1-4 TSSA k, A)SSS-5)5
DF.FINIT10NS h h] PROCESS CONTROL PROGRAM (PCP) N 1.23 T PR tSS NTROL PROGF M sha cont the ovisions assure a the .IDIF CATI of w . ra activ weste resul in a was form wi i pre rti tha teet .e re aireme s of CFR P .t 61 and f low le 1 r ioa ive te d' posa sites. The P > shall identify ocess pa meters A nf1 .cin OLID CAT N such s pH, il con nt, H 2 O ntent, s ids con
- nt, ra of slidi cati agent o wast and/or ecessary dditive for each t e of ntic' ated aste, the ceptab boundary conditio > for th roce par eter shall b identi ed for ach wast type, b ed on 1 oratory scal and all-s le test g or perien . The PC shall so inclu an id tifi etion f condi ons th must satisfi , based n full- ale t stin , to sure th dewat ing of ead resin , powder d resins and fi .er slud s wil result volur s of fr water, the ti of dis sal, w' nin th imits of 10 CF, Part and of ow level adioact e waste ispos s' es. g
~
PURGE - PURGING OFA4bdWn 1.24 R rP WING 11 b .ie con . led cess disc g ing air 2 gas ror cor neme ' o ma' ain t gerat , pr ure, h .oity, c . ntr n, r ot r oper .ing .ditio in su am . er that eplacem 6 ai sr gas s re aired purif the co ineme .. RATED THERMAL POWER 1.25 RATED THERHAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of(380d) Ma't. y REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.
~'
, REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
SHUTDOW MARGIN . If0SEO 1.28 SHUTDOW KARGIN shall be the instantaneous amount of reactivity by which the reactor is subtritical or would be subtritical from its present condition assuming:
- a. No change in part-length control element assembly position, and l
- b. All full-length control element assemblies (shutdown and regulating) l are fully inserted except for the single assembly of highest i reactivity worth which is assumed to be fully withdrawn.
$,JT 1 1-5 r aw md - !
CES$f Uss.s -sTs l 1 I
J J SHIELD BUILDING INTEGRITY 1.23 5 .LD 4LD o INT aRITY s il exist nen: , j . ch ,oor i each ac ss openi is c sed ext , when th access op iing i eing us for no .a tr sit ent and exit then a 1 st on door sh .1 be el d,
. The s eld bui ing fil ation stem is PERABLE and l
. The sealing echanis associ ed with ach pen . ration .g., w ds, be ows or -rings)/sOPE LE.
M i i:
- PRODF& El3 CON 1
4 1 a' l l ( *# i l l i 1 1 %
< \
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1 DEF1N3T10NS I SITE BOUNDARY hW@ , o r e 50Fh'ARE 1.30 The digital computer 50Fh'ARE for the reactor protection system shall ne i the program codes including their associated data, focumentation, and procedures. ( @ SOLIDIFICATION ) 1.31 .ID IC sN . 11 the vers of r dioact* was from uid ! s .s a ..o eous nif iy di ibut , mon hic mmob" ed
)
l' ith fi ev sme shap , oun - by a able urfac , dis l j o ine al sides (fre stand'. ). SOURCE CHECK
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o e i* , l ji STAGGERED TEST CASIS ' 1 ~,3 A STAGGERED TEST BASIS shall consist of; .
- a. A test schedule for n systems, subsystems, trains, or other designated components obtaintd by dividing the specified test interval
- into n equal subintervals, and ,
i l b. The testing of one system, subsystem, train, or other designated j component at the beginning of each subinterval. j THERMAL POWER
- 1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the
- reactor coolant.
I UNIDENTIFIED LEAKAGE l 1.35 UNIDENTIFIED LEAKAGE shall be all leakage which does not constitute either IDENTIFIED LEAKAGE or reactor coolant pump controlled bleed-off flow. i s UNRESTRICTED AREA g @gdQ g i 1.36 .E C AR hall b y are at or b the S K0l'NDAR 3 ac s w h not ntrol y the ense r pu 'p 6 on I' 1 v al rom osur o radi n an dioacti< teri , or any ar w in e 51 BOUN Y use or re entialgg.erso or.ind i e ial,[instiutio , and/o recre onal purp6es. p ' J'f:
,%e dia .T 1-6 l CES5g A)S95-STS
DEFINITIONS , h hff VENTILATION EXHAUST TREATMENT SYSTEM k , 1.37
'T 10' EX s . TM SYST sha ea syste designe and !
in d r se eo rad dine ra activ ater in p cul . f1 ts yp ing v- ilat' . or nt e ust es thr h ch coal o rs d/ EP ilte for pur e of movi iodin or pa ic-1 es om ga ous e aust E eam . or t he r ase t he 2n ronm .. ch sys is t co idere 'o hav any ett noble s eff ents M E neer Saf y Feap .e (E atmo neri cleanu system are n con ' t be ' TILA ON EXKAUST LATMENT YSTE compo nts. i VENTING g, \ d 1.38 V INF ha be contr ed p ess of chargi air or s from conf'..eme to int tempe sure, essure umidit oncent . ion, or ' o er rat'.g to tion, such manner at te cement or gas n-+ ] rov' ed requ' ed dur" VENT G. Ven , used ' system mes, do no .
; i ya NTINF proces i - .7 d
1 i 'l a l l
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1
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) i k i \ f$ 1.1 63 -$3 1
DEFINITXONS TABLE 1.1 I FREQUENCY NOTATION NOTATION FREQUENCY 5 At least once per 12 hours. D At least once per 24 hours. I 1 W At least cnce per 7 days. !
' 'M A+ ':. . , u mes per mor.co -
_ .. intervais no greater _ tt. :; pp ouw . minem.m. - _ c ' ' " t h. , y . , . . . . M At ieast onc.) per 31 days. Q At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. .. l _ e n.3,
- s - < -- + n .nes reToato
- 5/U Prior to each reactor startup.
N.A. Not applicable. t d i l
~
- ??005& REnEllcopy 4
1 0
- " O Y:"N '
1-8 l WSS,4 -NSss -5TS
l l QEt s#
&WM p yp o .
l MAGE EVALUATION ,9<;og 4 %. l'. 9 % % f/* TEST TARGET (MT-3) / j6(4 #, 4 NN'N[O - e
'%o##gy *
** pg,%*A i
i ! l.0 bW M i
- bt T. E w
l,l M hb Mi i 1.25 l.4 i.6 i l 4 150mm * [ 4 6" > i ! /k f l*k *'%
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Ap :
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PROCF & REVlBf COPY DEFINITIONS , TABLE 1.2 l OPERATIONAL MODES i REACTIVITY % OF RATED COLD LEG THERMAL POWER
- TEMPERATURE (Tcold)
OPERATIONAL ODE CONDITICEy,ff I
> c.99 > 5% > 750'T
- 1. POWER OPERATION *
> 0.99 5 5% > 350*f !
- 2. STARTUP i
< 0.99 0 > 350*f
- 3. HOT STANDBY ,
I
< 0.99 0 350' > Tcold > 210*F
- 4. HOT SHUTDOWN
- 5. COLD SHUTDOWN < 0.99 0 5 210*f
~
0 5 135'F
- 6. REFUELING ** 5 0.95
- Excluding decay heat.
** Fuel in the reactor vessel with the vessel head closure bolts less then fully tensioned or with the head removed.
1 i l l i l I l l CESSA2@-osss-sg
~ ^^
. . . 7:' : "":7 - 1-9 1
___ _ _ _ _ _ _ _ _ _ _ _ _ ____ A
l l l l 1 i l i i= .i i l 1 o_ SECTION 2.0 SAFETY LIMITS
^
AND LIMIT /.NG SAFETY SYSTEM SETTINGS l
-e e*
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I
- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE N00:& RIM CON DNBR 2.1.1.1 The calculated DNER of the reactor core shall be maintained greater than or equal to 1.231.
APPLICAElt1TY: HODES 1 and 2. ACTION: Whenever the calculated DNBR of the reactor has decreased to less than 1.231, be in HOT STANDBY within I hour, and cot. ply with the requirements of Specifi-cation 6.7.1.
~
PEAK ~ LINEAR HEAT RATE 2.1.1.2 The peak linear heat rate (adjested for fuel rod dynamics) of the
~
fuel shall be maintained less than or equal to 21 kk'.it. APPLICABILITY: HODES I and 2. ACTION: Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the , fuel has exceeded 21 kW/f t, be in HOT STANDBY within I hour, and comply with the requirements of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia. APPLICABILITY: HODES 1, 2, 3, 4, and 5. 1 ACTION: MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HDT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and ccmply with the requirements of Specification 6.7.1. H0 DES 3, 4, and 5: Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, ano comply with the requirements of Specification 6.7.1.
~
b 2-1
SAFETY L3MITS AND LIM}T}NG SAFETY SYSTEM SETTINGS 2.2 LIMITIN3 SAFETY SYSTEM SETTINGS _ t REACTOR TRIP SETPOINTS S CON 2.2.1 The reactor protective instrumentation setpoints shall be set consistent I t with the Trip 5etpoint values shown in Table 2.2-1. l APPLICABILITY: As shown for each channel in Table 3.3-1. d ACTION: With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification . 3.3.1 untii the channel is restored to OPEr.ABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. 37 ::": 't;": ::' r*Li_".--e i nne r e e r e. e rai7-ie 5 5 2.2. ***e Protection Calculater Addressable Constants shall b ccordance with Table "' APPLICABILITY: As shown *e Protection C tors in Table 3.3-1. ACTION: 4 With a Core Protectie sculator Addressab'r Constan s conservative than
] the value sho - .. the Allowable Vtlue column ef Table 2.2- eclare the
( channel i e table and apply the applicable ACTION statement req ment of Sp ' . cation 3.3.1 until the channel is restored to OPERABLE status.
-s
~~
j ; I l l 1 l 1 i I 'O I l i I k
** L^ '!@f CB:A: us S__ONo!' ?S-STS. 2-2 l d
l )
" i g THl$ PAGE OPEM N;DIMG IUin. P; OF b TABLE 2.2-1 INFOR#WlOil P!O. 'i IC.' K PLICANT
.i>h a : REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 5 7 ;
l lp g l FUNCTIONAL UNIT TRIP SETPOINT All0WABLE VALUES a j l 1. TRIP GENERATION S
~
A. ocess M M Q* sT I ' Pressurizer Pressure - High Pressurizer Pressure - Low (,2.)
; ??"' - &
C !";*c.'0?- h
- ' "- p
; GZZ p>ia (Zi -Q }'
SteamGeneratorlevel-Lew(4) ; .- s,) h ;~.,,c) 4
. 5 team' Generator Level - High (9) _
,. - s ., , Jit- ; ^:.'" '^; K "c=>
Steam Generator Pressure - Low (3) ; ?!^ p;e {& * ; "!?;,;:
) %- M
. Containment Pressure - High ; 3. 0 p ' , % {2.21 2; Jyg l 7
w
- 7. Reactor Coolant Flow - Low (,"7)
- a. Rate (() '
,L""#2-75){') h ; 1 " I- {r}{73 Q
- b. Floor (4s) ; 50. 5 'O'T % ; , , . <.- w i v ) %
- c. Band Um) '
'? -~ '0?'7) M ' ^^
. ~,'0}'7) N
/p
. Local Power Density - High ' 'l ? Mf/ ^^ ^,,8'P 4^ C/ T i ^ ^ }
/ 9. DNOR - Lcw L ??! ,) g L ??1 O) 9et.
- 8. E core Neuta , Flux kM ht Variable Overpower Trip (70)
- a. Rate (S) -~!"'"/!,e'^^"" < IL " '-'- - """U i r'^L = " '0; W T::= ;<two."4a) g l
- b. CellinglN GRu ' !!! ~~ ^ ^^ff"
- L = w, 4
- : ^.Z u ; = m ;-) M
- c. Band (D ,,, _"",q_m^"""
;_ ^ ;, ,g j :^.^"ui.56HL K-N ,
_, m ,-m 16,
<ew E ' oaA (.ibP) 4
- GoP)
-M 'Ta_m.
- A(pD d s 5Ag .
IL k b n u m (nu.A (4,9) M We. t6cT)
1 ( THIS PAGE OPEN PENDlHG RECEIPT OF 3 g TABLE 2.2-1 (Ccntinued) INFORMATION FROM THE APPilCANT REACTOR PROTECTIVE INSTRUNENTATION TRIP SETPOINT LIMITS iA s FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
/2. Logarithmic Power level.* High (1)
?I
-B
- a. Startup and Operating :
m;;c . i m7 rua o ' ;nny w64 ,
- b. shutdown ,<-eaess-e+-44W+ <-e-sohfHet9g
; . .. .. . ~, . , , v ,. w C. Core Protection Calculator System
- 1. CEA Calculators Not Applicable Not Applicable
- 2. Core Protection Calculators Not Applicable Not Applicable D. $dpplementary Protection System C3
; y Pressurirtr Pressure - High ;""- yAo % ; ? p :-- -
% r II. RPS LOGIC h A. Matrix logic Not Applicable Not Applicable n
- 8. Initiation Logic Not Applicable Not Applicable y M
III. RPS ACTUATION DEVICES j , A. Reactor Trip Breakers Not Applicable Not Applicable
~
- 8. Menval Trip Not Applicable Not Applicable 9
4 %- AgLds sAR ;
TABLE 2.2-1 (ContinuQd) - REACTOR PROTECTIVE INSTRUMENTATf0N TRIP 52TP0]NT LIMITS I TABLE NOTATIONS ) (1) Trip may be manually bypassed above 10 *% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10 *% of RATED THERMAL POWER. . 4 (2) In MDDES 3-6, value may be decreased manually, to a minimum of 100 psia, as pressuriter pressure is reduced, provided the margin between the pres-surizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer . pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia. i (3) In M3 DES 3-6, value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and ' this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached. e 6 (4) % of the distance between steam generator upper and lower levil wide
- i range instrument nozzles.
(5) As stored within the Core Protection Calculator (CPC). Calculation of . the trip setpoint includes measurement, calculational and processor uncer-tainties, and dynamic allowances. Trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 1% of RATED THERMAL POWER. 4 Trie pprov DNBR mit is 1.231 which includes a partial rod bow pensity ) j pensa on. If he fuel b nup exceeds that for which an increased rod
- ow pen ty is uired, t DNBR limit sha be adjusted. n this ca e a j DNBR ip setp nt of 1.23 is allowed pro ded that the fference i com-1 pen ed by a increase i the CPC addres able constant RR1 as fol ws
i
~#
RB - RB o
- d (% POL)
RR1 new
= ERR 1 old U 100 d (1 DNBR) i where uncompensated alue of BERR ; RB is the uel rod R1 old is t bow nalty in % BR; RB ,is the uel rod bow nalty in % NBR alre y acc nted for in he DNBR limit; OL is the pc er operati limit; d d (% POL)/d (% C BR) is the abs lute value o; the most verse de ivative i of POL with res;det to DNBR. u ,
l PRODF & RIEW COPY
@g8b- Asss-srs 2.s l 1 .
l 2
TABLE 2.2-1 (Continued) l REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS l l TABLE NOTATIONS (Continued) l (6) RATE is the maximum rate of decrease of the trip setpoint. There are no l restrictions on the rate at which the setpoint can increase. FLOOR is the minimum value of the trip setpoint. BAND is the amount by which the trip setpoint is below the input signal unless limited by Rate or Floor. 1 Setpoints are % of 100% power flow conditions. j l (7) The setpoint may be altered to disable trip function during testing i pursuant ta Specification 3.10.3. l l (8) RATE is the maximum rate of increase of the trip setpoint. There are no i restrictions on the tete at which the setpoint can decrease. , CEILING is the maximut.. value of the trip setpoint. BAND is the amount by which the trip setpoint is above the input signal unless limited by the rate or the ceiling. l
~
(9) % of the distance between steam generator upper and lower level narrow range instrument nozzles. Qo) 87o d 3 RATO Thraemat Tows ( ! PR00F & HEW COPY Of s a k NN 9 _720 t ,.; LC" ; 6$ 2-6
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)
P100F & REl'EW COM : BASE 5 ) FOR , l I SECTION 2.0 , SAFETY LIMITS . . , AND LIMITING SAFETY SYSTEM SETTINGS 1 l 1 CEssh(<So -Nsss - STci,
,I , I PROD:& RESE# COPY . mu l The BASES contained in the succeeding pages summarize the -
- reasons for the specifications of Section 2.0 but in accord-ante with 10 CFR 50.36 are not a part of these Technical lo :- Sp.cifications.
f 1 I 1 -s i ! I f I 1 I l cEsARBb-Miss-sr3 ! i 1
4 PRODF & REW COPY !,1 2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS j BASES ;
- 2.1.1 REACTOR CORE ,
i The restrictions of these safety limits prevent overheating of the fuel cladding and possible c ;dding perforation which would result in the release of fission products to toe reactor coolant. Overheating of the fuel cladding , is prevented by (1) restricting fuel operation to within the nucleate boiling l regime where the heat transfer coefficient is large and the cladding surface ; i temperature is slightly above the coolant soturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kW/f t which will not cause fuel centerline melting in any I i fuel rod. l First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surf ace temperatu*e is only slightly greater than the coolant saturation temperature. 1 The upper boundary of the nucleate boiling regime is termed " departure from p_ nucleate boiling" (DNB). At this point, there is a sharp reduction of the
- heat transfer coef ficient, which would result in higher cladding temperatures
,; and the possibility of cladding failure. ! i - l Correlations predict DNB and the location of DNB for axially uniform and !. non uniform heat flux distributions. The local DNB ratio (DNBR), defined as-I the ratio of the precicted DNB heat flux at a particular core location to the A actual heat flux at thst location, is indicative of the margin to DNB. The minimum value of DNBR during normal operation and design basis anticipated operational occurrences is limited to 1.231 based upon a statistical combination ' of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit. "r '*"" ' L . J .:n >-9% =d 6 P -~rUr :1 n=nr r:. '4 g-m m m s .u u that <-- + - 4 : M,_ e 4
- 4 I j q en . -
., . - . s , = < - se of th. m .. ~ = -+- n u. j Second, operation with a peak linear heat rate below that which would '
cause fuel centerline melting maintains fuel rod and cladding integrity. i Above this peak linear heat rate level'(i.e., with some melting in the center), : fuel rod integrity would be maintained only if the design and operating : conditions are appropriate throughout the life of the fuel rods. Volam changes which accompany the solid to liquid phase change are signific.3t and i require accosmodation. Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting. Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit. To account for fuel rod tynamics (lags), the directly indicated linear heat rate is dynamically adjusted by the - CPC program. 3 Y E M ! S S TS B 2-1
SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS PR007& RM CD?Y l BA515,. J Limiting Safety System Settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and Limiting Conditions for Operation on DNBR and W/f t actgin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during nomal operation and design basis anticipated operational occurrences. 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this 5tfety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclidtes contained in the reactor coolant from reaching the containment atmosphere. *
-tor Coolant System components are designed 1974 Edition, ' .. !?M Addendum, of the or Nuclear Power P.lan
-a h
C Components which pemits a a e of 110% (2750 psia) of _ design pressure ety Limit of 2750 psia is there stent with q the tteria and associated code requirements. [ The entire Reactor Coolar.t System is hydrotested at 3125 psia to > demonstrate integrity prior to initial operation. ' 2.2.1 REACTOR TRIP SETPOINTS
~ '
The Reactor Trip Setpoints specified in Table 2.2-1 are the values at - which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System ; are prevented from exceeding their Safety Limits during nomal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in sitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than s
, the drift allowance assumed for each trip in the safety analyses. l The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Safety Limits of 1.231 and 21 W/f t, respectively.
Since then trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Yalues for these trips are therefore the same as the Trip Setpoints. l To maintain the margins of safety assumed in the safety analyses, the ! calculations of the trip variables for the DNBR - Low and Local Power Density - High trips include the measurement, calculational and processor uncertainties i and dynamic allowances as defined in CESSAR Systen 80 applicable system descriptions and safety analyses. k l M S S A*QPO 61-A mis y N 8 2-2 ! M V= I t 4
i ', ~N%
- 2. R R L SYS PRES RE i on' limi pmt ts
/
e ihte tri thi safe ity of he Re .cr oo nt 5 tem or, o rpre urita on a d th by pr ents r onu ides ontai ed i he re ctor ela from role aeof sp re. achingthee irper, [The reactor pressure vessel, piping, and pressurizer are designed to Section v Ill of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 11M (2750 psia) of the design pressure. The Reactor Coolant System valves and fittings, are designed to either Section III , i of the ASME Code or ANSI B 31.7 Class I, which permits a anximum transient ' pressure of 11M (2750 psis) of cosponent design pressure. See Applicant's g l FSAR for specific Code, Standard Editions, and Addenda. The safety limit of 2750 psia is therefort consistent with the design criteria and associated code i I (requirements._____ . . _ _ . _ _ , nt ri p 1 att . j '. ii
~
l PR005 & REREW C0?Y r 1 1 , 1 j ! l l. l f 1 i Amendment Number 9 E t-2 * **'E Il*
- SAFETY LIMlTS AND LIM 1T8NG SAFETY SYSTEMS SETTINGS BASES .
,i PROOF & REY!M CD)Y ! REACTOR TRIP SETPOINTS (Continued) a=" t- 4 :;;;; 5t'v e;n , i A n tM:Wg __w__t_k ::1:0:t4:- :' + 6 : f :nt p;cte.r,44c4Lgy"= H ; di::_:" ' . ; t': M rs.,;t
- Ci"-?M(") 0;p.;
- _ M; 21, ;r _ _ w i Manual Reactor Trip l l
4 The Manual reactor trip is a redundant channel to the automatic protective I l instrumentation channels and provides manual reactor trip capability. Variable Overpower Trip d A reactor trip on Variable Overpower is provided to protect the reactor d core during rapid positive reactivity addition excursions. This trip function ! will trip the reactor when the indicated neutron flux power exceeds either a 4 rate limited setpoint at a great enough rate or reaches a preset ceiling. The i m flux signal used is the average of three linear subchannel flux signals j originating in each nuclear instrument safety channel. These trip setpoints i; are provided in Table 2.2-1. (- i Logarithmic Power Level - High I d 4 The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in ,j the event of an unplanned criticality from a shutdown condition. A reactor i trip is initiated by the Logarithmic Power Level - High 1. rip unless this trip . ! is manually bypassed by the operstor. The operator muy manually bypass this l trip when the THERMAL POWER level is above 10 *% of inTED THERMAL POWER; this i bypass is automatically removed when the THERMAL POWER level decreases to
- 10 +% of RATED THERMAL POWER.
i j ' Pressurizer Pressure - High j The Pressurizer Pressure - High trip, in conjunction with the pressurizer i safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without l reactor trip. This trip's setpoint is below the nominal lift setting of the ! pressurizer safety valves and its operation minimizes the undesirable opera-l tion cf the pressurizer safety valves. a Pressurizer Pressure - Low 3 1 The Pressurizer Pressure - Low trip is provided to trip the reactor and j to assist the Engineered Safety Features System in the event of a decrease in j Reactor Coolant System inventory and in the event of an increase in heat
- " b 8 2-3 1 1
J
SAFETY. LIMITS AND LIMITING SAFiTY SYSTEMS SETTINGS ) BASES H00f & RE/!EW 4-9Y Pres <,urizer Pressure - Low (Continued) removal by the secondary system. During normal operation, this trip's set-point may be manually decreased, to a minimum value of 100 psia, as , pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure md this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint increases automatically as . pressurizer pressure increases until the trip setpoint is reached. The operator may manually bypass this trip when pressurizer pressure is below 400 psia. This bypass is automatically removed when the pressurizer pressure increases to 500 psia. Containment Pressure - High . 1 The Cortainment Pressure - High trip provides assurance that a reactor trip is initiated in the event of containment building pressurization due to a pipe break inside the containment building. The setpoint for this trip is identical to the safety injection setpoint. - Steam ':enerator Pressure i Low
- The Steam Generator Pressure - Low trip provides protection in the event of an in: rease in heat removal by the secondary system and subsequent cooldown of the rtictor coolant. The setpoint is sufficiently below the full load -.
operating ooint so as not to interfere with nomal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at less than l or equal to 200 psi; this setpoint increases automatically as steam generator pressure increases until the normal pressure trip setpoint is reached. Steas Generator Level - Low The Steam Generator Level - Low trip provides protection against a loss of feedwater (low incident and assures that the design pressure of the Reactor Coolant System will not be exceeded due to a decrease in heat removal by the secondary system. This specified setpoint provides allowance that there will be sufficient water inventory in the steam generator at the time of the trip to provide a margin of at least 10 minutes Mfore auxiliary feedwater is required to prevent degraded core cooling. Local Power Density - High The Local Power Density - High trip is provided to prevent the linear i heat rate (W/ft) in the limiting fuel rod in the core from exceeding the fuel ' design limit in the event of any i: Y W r anticipated operational occur-rence. The local power density is calculated in the reactor protective system utilizing the following information: g ID B 2-4
SAFETY LIM 315 AND LIMlTING SAFETY SYSTEMS SETT1HG5 BASES }' l 1 Local Power Density - High (Continued)
~
- a. Nuclear flux power and axial power distribution from the excore flux monitoring system; ,
- b. Radial peaking f actors from the position measurement for the CEAs;
- c. Delta T power from reactor coolant temperatures and coolant flow measurements.
The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and ctynamic compensation routines. These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs , when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the Peak Linear Heat Rate Safety Limit. CPC uncertainties
~
related to peak LPD are the same types used for DNBR calculation. Dynamic or*- - compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, lf and protection system equipment time delays. DNBR - Low The DNBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event i of 5';;- M anticipated operational occurrences. The DNBR - Low trio incorporates a low pressurizer pressure floor of NEL psia. At this pressure
)
a DNBR - Low trip will automatically occur. The DNBR is calculated in the CPC 1
- , utilizing the following information
- a. Nuclear flux power and axial power distribution from the excore <
., neutron flux monitoring system; !
i -
- b. Reactor Coolant System pressure from pressurizer pressure measurement;
- c. Differential temperature (Delta T) power from reactor coolant temperature ed coolant flow measurements; l d. Radial peakirg factors from the position measurement for the CEAs;
, e. Reactor coolant mass flow rate from reactor coolant pump speed; . f. Core inlet temperature from reactor coolant cold leg temperature measurements.
~
-4 Sag Ab. 'c.ww3 "5A1C ,
- q CESSAtto-psss-q m&-r_ s z-s l
--I SAFETY LIM 1TS AND LIMITING SAFETY SYSTEMS SETTINGS 1
- PR00.F & REVN 'Ohy BASES :-
- i r DNBR - Low (Continued)
The. DNBR, the trip variable, calculated by the CPC incorporates vario'us uncer-tainties and dynamic compensation routines to assure a trip is initiated prior i to violation of fuel design limits. These uncertainties and dynamic compens'a-l tion routines ensure that a reactor trip occurs when the calculated core DNBR
; is sufficiently greater than 1.231 such that the decrease in calculated core DNBR after the trip will not result in a violation of the DNBR Safety Limit.
CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling unce-tainties, and computer equipment processing : j uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays. 1 The DNBR algorithm used in the CPC is valid only within the limits
- indicated below and operation outside of these limits will result in a CPC '
_. initiated trip. 5 Parameter Limitino Value
- a. RCS Cold Leg Temperature-Low
- b. RCS Cold Leg Tempercture-High > @"f
<WF
- c. Axial Shape In6ex-Positive Not more positive than
- d. Axial Shape Index-Negative Not more negative than
- e. Pressurizer Pressure-Low > psia
- f. Pressurizer Pressure-High I psia ;
- g. Integrated Radial Peaking l
h. Factor-Low Integrated Radial Peaking Factor-High 5%
@h k
- i. Quality Margin-l.ow >$
Y gh a
)
-' Steam Generator Level - High pA m r .
- The Steam Generator Level - High trip _ ;
- : 'f: ' *: ; ^'--- ': 1_ l'-- --
J f - ~ -- ' n ._,..._ _ ::-- ; :_: . _ _ t u ^_ _ . '. E_ ' :
^
= ' ' ::"y I ^ ' ,,:I _.....
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- 7. Mi ".T 1 B 2-6 i
PROOF & RYEW COPY BASES i I Reactor Coolant Flow - Low M . The Reactor Coolant Flow - Low trip provides protection againkt a reactor coolant pump sheared shaft event and a '__ ; , - coastdown _ .,., m
.1__ C '__: : ..,
7.-.. . A trip is initiated when the pressure differential across the primary side of either steam generator decreases below i a variable setpoint. This variable setpoint stays a set amount below the pres-i sure differential unless limited by a set maximum decrease rate or a set minimum i value. The specified setpoint ensures that a reactor trip occurs ^_: ; , a.. ._ -- ,, ,, .- ~- . .._
. . -_ _. - . g
$ e Pressurizer Pressure - High (SPS)
The Supplementary Protection System (SPS) augments reactor protection against overpressuriretion by utilizing a separate and diverse trip logic from the Reactor Protection System for initiation of reactor trip. The SPS will i- initiate a reactor trip when pressurizer pressure exceeds a predetermined I 8 - value. P ss e- L.J 4 L% bm &A_ LBC
%a , p ~n bAm&fsAR, l
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. Pi!GOT a E i.'Bf CC0 4 :
1 l SECTIONS 3.0 AND 4.0 U- LIMITING CONDITIONS FOR OPERATION , 5 AND SURVEILLANCE REQUIREMENTS 1 ) l 4 M%Mfo-A)sSS-575 I
l l PR001 R&"BY CO?Y ) 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i 3/4.0 APPLICABILITY t LIMITING CONDITION FOR OPERATION i 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MDES or other ! conditions specified therein; except that upon failure to meet the Limiting ; Conditions for Operation, the associated ACTION requirements shall be met. i 3.0.2 Noncompliance with a specification shall exist when the requirements of I the Limiting Condition for Operation and/or associated ACTION requirements are . not met within the specified time intervals. If the Limiting Condition for 1 1 Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
- 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour, action shall be initiated to place the unit in a MODE in which the specification does not apply by placing
^
; it, as applicable, in:
, - i
- 1. At least HOT STANDBY within the next 6 hours, 1
- 2. At least HOT SHUTDOWN within the following 6 hours, and ;
- 3. At least COLD SHUTDOWN within the subsequent 24 hours.
i Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. This specification is not a.pplicable in MDE 5 or 6.
~
).0.4 Entry into an OPERATIONAL MODE or other specified condition shall not 1 be made unless the conditions of the Limiting Condition for Operation are met without relian;e on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIDKAL MODES as required to comply with ACTION statements. Exceptions to these requirements are stated in the individual specifications.
. 2 3/4 0-1
APFLICABXOTY t~ , s g., y. .
?isb-e;nOl d L.Wi UC i, SURVEILLANCE REQUIREMENT., ,
4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for 3 Operation unless otherwise stated in an individual Surveillance Requirement.
- 4.0.2 Each Surveillance Requirement shall be performed within the specified
, time interval w1 ' -
- a. A maximum allowable extension not to exceed 25% of the surveillance interval, and 1
- b. The combined time interval for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perfom a Surveillance Requirement within the specified time interval shall constitute a. failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated
- in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL MODE or other specified conditien shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified. i 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows: i a. Inservice inspection of ASME Code Class 1, 2, and 3 components and 4 inservice testing ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50,
-< Section 50.55a(g), except where specific written relief has been granted
- by the Commission purtuant to 10 CFR 50, Section 50.55a(g)(6)(i).
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Yessel Code and applicable Addenda shall be applicable as i
follows in these Technical Specifications: l I d S S6 M 5-$TS 3/4 0-2
APPLICABILITY % [h h;h SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued) ASME Boiler and Pressure .- Vessel Code and applicable Required frequencies Addenda terminology for for performing inservice ' inservice inspection and inspection and testing testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 164 days Yearly or annually At least once per 366 days
- c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing
~
activities. o- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
es* l h
REACTIVITY CONTROL SYSTEMS " 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL MUTDOWN MARGIN - T GREATER THAN 210'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MRGIN shall be greater than or equal to 6.0% delta A/k. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: l With the SHUTDOWN MARGIN 1ess than 6.0% delta k/k, ismediately initiate and continue boration at greater than or equal to 26 ppm to reactor coolant system '
; of a solution containing greater than or equal to 4CDD ppe boron or equivalent until the required SHUTDOWN MARGIN is restored.
j SURVEILLANCE REQUIREMENTS s
- 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal i
to 6.0% delta k/k: 1
- a. Within 1 hour after detection of an inoperable CEA(s) and at least once per 12 hours thereafter while the CEA(s) is inoperable. If the ,
inoperable CEA is immovable as a result of excessive friction or ' 1 sechanical interference or known to be untrippable, the above re-quired SHLfDOWN MARGIN shall be verified acceptable with an increased
-~
allowance for the withdrawn worth of the immovable or untrippable CEA(s). L
- b. when in MODE 1 or MODE 2 with K greater than or equal to 1.0, at i least once per 12 hours by vertNing that CEA group withdrawal is i within the Transient Insertion Limits of Specification 3.1.3.6.
- c. \
When in MODE 2 with K 1ess than 1.0, within 4 hours prior to ' echievingreactorcriff[alitybyverifyingthatthepredicted critical CEA position is within the limits of Specification 3.1.3.6. ! l l l , See Special Test Exception 3.10.1. ; / \ I l
$b Y ~~ Nsr - n ,$ $ $ - Q
^^'^ Vf* _
3f4 1 1 i
{ . REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (r:r.tinued) ;
- d. Prior to initial operation above 5% RATED THERHAL POWER aft'er each fuel loading, by consideration of the factors of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1J3.6.
- e. When in MODE 3 or 4, at least once per 24 hours by consideration of at least the following factors:
- 1. Reactor Coolant System boron concentration,
- 2. CEA position,
- 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 + 1.0% delta k/k at least once per s
.. 31 Effective Full Power Days (EFPD). This comparison shall consider at least
~
--- those factors stated in Specification 4.1.1.1.le., above. The predicted
!- reactivity values shall be adjusted (normalized) to correspond to the actual i~ core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.
i l l l
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I i i i a l. (EMAtFo-uniaA.X95 __7;;; .;aai 2
-5Ts 3/4 1-2 I
i
REACTIVITY CONTROL SYSTEMS 0f& kh h {opy SHUTD3W MARGIN - T LESS THAN OR EQUAL TO 210'F 1 l LIMITING CONDITION FOR OPERATION - l ! 3.1.1.2 The SHUTDOW MRGIN shall be greater than or equal to 4.0% deita Uk. APPLICABILITY: M')DE 5. i ACTION: ' With the SHUTDO=N MEGIN less than 4.0% delta Uk, immediately initiate and continue beration at greater than or equal to 26 ppm to the reactor coolant i system of a solution containing greater than or equal to 4000 ppm boren or 1 equivalent until the required SHUTDOWN MRGIN is restored. - ( d& I
- 4. SURVEILLANCE REQUIREMENTS j
4.1.1.2 The SHUTDOWN MRGIN shall be determined to be greater than or equal i i to 4.0% delta k/k: s
- a. Within 1 hour after detection of er in:perable CEA(s) and at least once per 12 hours thereaf tsr while the CEA(s) is inoperable. i 3
If the inoperable CEA is immovable as a result of excessive friction cr mechanical interference or known to be untrippable, the above required SHJTDOWN MRGIN shall be increased by an amount at least
- equal to the withdrawn worth of the immovable or untrippable CEA(s).
, b. At least once per 24 hours by consideration of the following factors:
- 1. Reactor Coolant Systes boron concentration,
- 2. CEA position,
- 3. Reactor Coolant Systes average temperature, 4 Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Sae.arium concentration.
d
~
- 3/4 1-3
, REACTIVITY CONTROL SYSTEMS -- I (Dt**b*P.u'*.,*.)L~:lnG.,..}
, iu,r .. L l M3DERATOR TEMPERATURE COEFFICIENT _
j LIMITING CONDITION FOR OPERATION l l 3.1.1.3 The moderate temperature coefficient (MTC) shall be within the area of Acceptable Operation shown on Figure 3.1-1. . APPLICABILITY: MODES I and 2"# ACTION: With the moderator temperature coefficient outside the area of Acceptable
- Operation shown on Figure 3.1-1, be in at least HDT STANDBY within 6 hours. ,
j i '[~ SURVEILLANCE REQUIREMENTS 5 4.1.1.3.1 The MIC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits. 4.1.1.3.2 The Mit shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:
- a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
- b. At any THERMAL POWER, within 7 EFPD after reaching a core average exposure of 40 EFPD burnup into the current cycle.
~'
- c. At any THERMAL POWER, within 7 EFPD after reaching a core average exposure equivalent to two-thirds of the expected current cycle end-of-cycle core average burnup.
l t h CWith Keff greater than or equal to 1.0.
#5ee Spetlal Test L.eption 3.10.2.
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REACTIVITY CONTROL SYSTEM 5
- MINIMUM TEMPERATURE FOR CRITICALITY O "4 -
h g' T: iwsf C.
,6u leh .s'.'.d i
j ; LIMITING CONDITION FOR OPERATION
- l 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tcold) !
shall be greater than or equal to 552*F. - i APPLICABILITY: MDDE5 I and 2#*. l ACTION: , 1
- With a Reactor Coolant System operating loop temperature (Teold) I'55 th""
552'F, restore T cold to within its limit within 15 minutes or be in HDT STANDBY within the next 15 minutes. 2 I ! I SURVEILLANCE REQUIREMENTS , j .- - I i 4.1.1.4 The Reactor Coolant System temperature (Teold) shall be detersined to , be greater than or equal to 552*F: ' 1 a. Within 15 minutes prior to achieving reactor criticality, and I b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant SystemgT ,3 ,is less than 557'F. l
-a 1
) i j - l
)
J { With K,g greater than or equal to 1.0. I cSee Special Test Exception 3.10.5.
\
, 3/4 1-6
REACTIVITY CONTPOL SYSTEMS l 3/4.1.2 BORATION SYSTEMS ___
=
FLOW PATHS - SHUTDOWN %hh . LIMITING CONDITION FOR OPERATION : l 3.1.2.1 As a sinimum, one of the following boron injection flow paths shall be OPERABLE: l
- a. If only the spent fuel pool in SpeciOcation 3.1.2.5a. is OPERABLE, a flow path from the spent fuel pool via a gravity feed connection and a charging pump to the Reactor Coolant System.
^
- b. If only the refueling water tar k in Specification 3.1.2.5b. is OPERABLE, a flow path from the refueling water tank via either a j charging pump, a high pressure safety injection pump, or a low pres-sure safety injection pump to the Reactor Coolant System.
[ APPLICABILITY: MODES 5 and 6. ACTION: I - With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 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, pover-operated, or automatic) in the flow path that is not locked, sealed, er otherwise secured in position, is in its correct position.
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l 4 4 M5ji&tGo - A49S- STS 3/4 1-7
REACTIVXTY CONTRDA SYST MS I i {Pf{ l FLOW PATHS - OPERATING I LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boren injection flow paths shall be OPERABLE: " a. p gra ty feed flow path from either the lefueling kater kank or the
\ pent ei kcol thro"gh CH-536 (RWT Gravity Feed Isolation valve) and a charging pum;, to the Reactor Coolant System, I'
b. ank through A gravity CH-327 feed Feed (RWT Gravity flow/ Safety pathInjection from System the kefueling I kter v
\ solation and a charging pump to the Reactor Coolant System,
- c. A flow path free either the fuelingkaterkankorthekpent uel '
hol throughKH-164 (boric Acid Filter Bypass Valve), utilizing ! gravity faed and a charging pump to the Reactor Coolant System. I APPLICABILITY: MODES 1, 2. 3, and 4.NCH- del (fgu,c kc.d FJAn, l ACTION: A4 W O i With only one of the above reqaired boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and berated to a SHUTD0h9 MARGIN equivalent to at least 6% delta k/k , at 21n*F within the 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. SURVEILLANCE REQUIREMENTS
~' 4.1.2.2 At least two of the above required flow paths shali be demonstrated
, DPERABLE:
! a. At least once per 31 days by verifying that each valve (manual, l 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 when the Reactor Coolant System is at i ~
normal operating pressure by verifying that the flow path required j by Specification 3.1.2.2 delivers at least 26 gpa for 1 charging pump and 68 ppe for two charging pumps to the Reacter conlant System-
~
- 4. 2.3 The ovisi s of cif ati 4.0 'ev t app 1 able fe entry .o de 3 r Mo 4 to erfo the urve la test g of 4
rovi dt is cifica n 4.1. .b testi fo d wi iin hour after chievi normal per ing essu in t rea or e olan sys . i 9lB6;M'h -Mns-STs
- m. . . , . . uui 4 3/41-8 i
I
REACTAVITY CONTROL SYSTEMS , 1 CHARGING PUMPS - SHUTDOWN PROD' & REYEW COPY l LIMITING CONDITION FOR OPERATION , 3.1.2.3 At least one charging pump
- or one high pressure safety injection pump or one low pressure safety injection pump in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERtBLE and capable of being powered from an OPERABLE emergency power source. ,
I APPLICABILITY: MODE 5 5 and 6. ACTION: With no charging pump or high pressure safety injection cump or low pressure safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving COF.E ALT"'ATIONS or positive reactivity changes. p i SURVEILLANCE REQUIREMENTS __ 4.1.2.3 No additional Surveillance Requirements other than those required by
- Speci fication.4. 0. 5.
-a j 1
I Whenever the reactor coolant level is below the bottor of the pressurizer in MODE 5, one and only one charging pump shall be OPERABLE, by verifying at least once per every 7 days that power is removed from the remaining charging pumps. l M0.".2 MCC M: - ..NSSS GT^ -STS 3/4 1-9 l l l l l
1 REACTXVITY CONTROL SYSTEMS s _ CHARGING PLMPS - OPERATIN3 l h LIMITING CONDITION FOR OPERATION 3.1.2.4 At least twe charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. AC110N: With only one charging pump OPERABLE, restore at least two charging pumps to OPERAELE status within 72 hours or be in at least H0f STANDBY and borated te / SHUTDOWN MARGIN equivalent to at Ieast 6% delta k/k at 210*F within the next , 6 hours; restore at least two charging pumps to OPERABLE status within the ne + ' 7 days or be in COLD SHUTDOWN within the next 30 hours. lM a SURVEILLANCE REQUIREMENTS _ 4.1.2.4 No additional Surveillanec' Requirements other than those required by Spet.ification 4.0.5. 4 CE1iS AftSO-loggg_ g
-- 5 "* - . . . 3/4 1 10
REACT]VXTY CONTROL SYSTEMS . _
- g. n s' ^~f~ , ,: ,,. ~
t,_.,-e
\
#'_'{p i BORATED WATER SOURCES - SHUT 00WN
}M LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be"' "
OPERABLE: , , GAS M
# "' '* Afr\ gee
- a. The spent fuel pool with: ed ) T.t-2.Q
- 1. A minimum borated water volume of ??,f' 'p"r:gn'd X q
- 2. A boron concentration of becween 4000 ppm and 4400 ppm boron, and
- 3. A solution temperature betwaen 60*F and 180*F.
- t. The refueling water tank with:
1. A minimum contained berated water volume of M,i::b kn and
- 2. A boron concentration of between 4000 ppm and 4400 ppm boron,
, and i ~
- 3. A solution temperature between 60*F and 120*F.
3 APPLICABILITY: MODESSf,and ACTION: With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water seurce is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water sources shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
~' , 1. Verifying the boron concentration of the water, and
- 2. Verifying the contained borated water volume of the refueling water tank or the spent fuel pool.
- b. At least once per 24 hours by vers 7ying the refueling water tank temperature when it is the source of borated water and the outside air temperature is outside the 60'F to 120'F range.
- c. At least once per 24 hours by verifying the spent fuel pool temperature when it d i the source of borater) water and irradiated fuel is present in the pool.
^
.. ,m ... M "" r " e A r '
s he ~ 3/4 1-11 i
i.
- 136*-6 * (40K) _..r___
i
,M,, ...
A .. _ _ . . _
, 2_ _.___
i 3 3,. .. Ox, .. _ . m _ _g _- . __ . _- _- # (. ,._. ng y
-=. g __, ' n- - ,
! ,3,. .$.. ,,0 x, _1 =i;7k _7.
'~
=
BE & EN MW i l li M- 3 (7.25K)- -4_
-s- .-
- - - - = f i
,,3.,...,,,,,
g.d. :WFb. '
/
__, ~ COLD $#f5 VOLUME 1 ;;._- . _ . _ r- -#-- . l 0 400 600 AVERAGE RE OR COOLANT SY T .. ' F i i i ! l I _ y, . 000 G AL. ( . OF) -.6%K
~
1 4 : ]: J = I i / 5%
- 573.744 G AL. (120 F1 0
Mi i - F l .' i I COLD SID L. PLUS uuE' j f 70% MAR . 525K EOU D IN TH l 3 RW j . QWT LEVEL - SPF) INSTRUMENT READING (1) 65% l i . 47 i ., ESF O L. P S MARGIN (2) , g ! ERA STEMPE URE,'F I l (1) THE LEV D YOL SHOWN ARE TH EFU THAT IN HE TANK WHIC 8 AND UME ABC l 15" OU FORYO X CON TIONS l (2) ' O D r; 5 A ONE S EO RCE ! CONTAIN INI. GA , (3) THis VOLUME is "LOUI DURIN DE6 l FIGURE 3.12 i MINIMUM SORATED TER VOLU ES l SGE \Lf% k -
~b 3/4 1-12 l
l l _ _ - _ _ _ _ _ . l
REACTIVITY CONTROL SYSTEMS ner o ' BORATED WATER SOURCES - OPERATING pN U h M {3 rt it' t l ! LIMITING CONDITION FOR OPERATION 3.1.2.6 Each of the following borated water sources shal; be SPIRABLE:
- a. The spent fuel pool with: .
- 1. A minimum borated water volume as specified in figure 3.1-9, and
- 2. A boron concentration of between 4000 ppm and 440.' ppm boron, and I 3. A solution temperature between 60*F and 180*F.
- b. The refueling water tank with:
- 1. A minimum contained borated water volume as specified in Figure 3.1-2, and j 2. A boron concentration of between 4000 and 4400 ppm of bor,on, and
, 3. A solution temperature between 60*F and 120*F.
APPLICABILITY: MODES 1, 2,0 3 ,4 and d. ACTION: 2
- a. With the above required spent fuel pool inoperable, restore the pool to OPERABLE status within 72 haurs or be in at least HOT STANDBY within the next 6 hours and borsted to a SHUTDOWN MARGIN equivalent t 3 to at least 6% delta L/L at 210", restore the above required spent fuel poo
- to OPERABLE status within the ne.t 7 days or be in COL.; SHUTDOWN j within the next 30 hours.
I b. With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next
.6 hours av,8 in COLD SHUTDOWN within the following 30 hours.
i SURVEILLANCE REQUIREMENTS l ., 4.1.2.6 Each of the above required borated water ararces shall be demonstrated OPERABLE: 2
- a. At least once per 7 days by:
- 1. Verifying the boron concentration in the water, and
- 2. Verifying the contained borated water volume of the water source.
- b. At least once per 24 hours by verifying the refueling water tank temperature when the outside air tesperature is outside the 60*F to 120'F range.
- c. At least once per 24 hours by verifying the spent fuel pool temperature when irradiated fuel is present in the pool.
}
'c-- c-- u v... r m a+ % y ya 3 PALD VERDE - UNIT 1 3/4 1-13 J
1
~
BORON DXLUTION ALARMS hnqu u g p"5 W ( LIMITING CONDITION FOR OPERATION l 3.1. 2. 7 Both startup channel high neutron flux alarms shall be OPERABLE. APPLICABILITY: MODES 3*, 4, 5, and 6. ACTION: '
- a. With one startup channel high neutron flux alarm inoperable:
1. Detemine the RCS boron concentration when entering MODE 3, 4, 5, or 6 or at the time the alare is determined to be inoperable. From that time, the RCS boron concentration shall be determined at the applicable monitoring frequency in Tables 3.1-1 through 3.1-5 by either boronometer or RCS sampling.** ,
- b. With both startup channel high neutron flux alams inoperable:
1 1. Determine the RCS boron concentration by either boronneter and i RC5 sampling ** cr by independent collection and analysis of two j RCS samples when entering Mode 3, 4, or 5 or at the time both
.T alares are determined to be inoperable. From that time, the
=
RCS boron concentration shall be determined at the applicable monitoring frequency in Tables 3.1-1 through 3.1-5, as applicable, '~ - by either boronneter and RCS sampling ** or by collection and analysis of two independent RCS samples. If redundant detemina- *
- f il tion of RCS baron concentration cannot be accomplished immediately, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the method for setemining and confirming RCS boron concentration is restored.
2. When in MODE 5 with the RCS level below the centerline of the hotleg or MODE 6, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one startup channel high neutron flux alare is restored to OPERABLE status.
- c. The provisions of Specification 3.0.3 are not applicable.
j -SURVE!LLANCE REQUIREMENTS J 4.1.2.7 Each startup channel high neutron flux alarm shall be demonstrated DPERABLE by performance of: 1 t i "Within I hour after the neutron flux is within the startup range following a reactor shutdown.
**With one or more reactor coolant pumps (RCP) operating the sample should be obtained from the hot leg. With no RCP operating, the sample should be
- {
obtained from the discharge Ifne of the low pressure safety injection (LPSI) pump operating in the shutdown cooling mode.
'(
4 CESSAR@-M9;s-ST!;
- . . ~ -. 4- 3/4 1-14 4
REACTIVITY CONTROL SYSTEMS "# v o
>ps.dsta.y5n3g CO?ui SURVEILLANCE REQUIREMENTS (Continued) ,
- a. A CHANNEL CHECK:
I
- 1. At least once per 12 hours. '
i
- 2. When initially setting setpoints at the following times: !
, a) One hour after a reactor trip.
b) After a controlled reactor shutdown: Within I hour after the neutron flux is within the startup range in MODE 3.
- b. A CHANNEL FUNCTIONAL TEST every 31 days of cumulative operation during shutdown.
m e N
- i. ,
I f i
-o 1
h TSSAf&->\)sss-srS
.-- - _ _ . , . . . . 3/4 1-15
TABLE 3.1-1 {}h h fg ki
--;= MONITORING FREQUENCIES FOR BACKUP ovnun DI L UT I ON DE T E C T I ON *i ^ Ty,,;7 ;;;, ; T ^ *': ^ ^ ' ?:C 4 I O'" 5:';; i.e a ;,:;; 7."2 '
" ^ ^ ': ?"i "E ! F OR K,f f > 0. 98 OPERATIONAL
-Number of Operatino Charging Pumps MODE O 1 2 3 3 12 hours I hour Operation not allowed 4 12 hours I hour Operation not allowed 5 RCS filled 8 hours I hour Operation not allowed 5 RCS partially
- drair,ca Operation not allowed -
~
- 6 24 hours 8 hours 4 hours 2 hours E
1 . a i i 1 i Tog SysTEMD !
- cycle l
- ~
l t i i i e N 3/4 1-16 : l l
TABLE 3.1-2 ((,4} g {.".{'g' {0 ' i M MONITORING FREQUENCIES FOR BACKUP BORON vitutavn
- DETECTION t: " "J':':0L CT ; :":^:; 0: ^ 7; ':: " L" I "i^_"*'
^ ^ "' ' " ^'
- i ^ M : FOR 0.98 > K,9f > 0.97 I
Number of Operatino Chargino Pumps - 3 OPERATIONAL
- MODE O 1 2 3 i 3 12 hours 2.5 hours I hour 0.5 hours
. 4 12 hours 2.5 hours 1 hour 0.5 hours 5 RCS filled 8 hours 2.5 hours I hour 0.5 hours 5 RCS partially dra!nedM 8 hours 0.5 hours Operation not allowed S 6 24-hours 8 hours 4 hours 2 hours i 4 ' tnt Tec.'m'cd Sggh 3.1.2.3 w3l ob 0 0 E c. e d)rw,uniM
'f*N' DE ame 1
i cah . I i i 4 l l l Q SYSTEM 60 Fb E,u
- g g G l c.yCLE S l
1 l C.EssAR82-#5%-Srs .. p. i.n h
- W4 1-U l
i
- TABLE 3.1-3 w b i Hi COM
! Jaqumman MONITORING FREQUENCIES FOR BACKUP BORON DILUTION l DETECTION Z ^ 7 - i a r^" 0^i^^'!': :: ^ ^ ^ ? 'E "' ~ ~' ' Q
/
j ^ ^C ^i" C *C ^ T;;<.,. ,G ~ T OR 0. 97 > K, , f > 0. 96 l Number of Operatina Charging Pumps - OPERATIONAL MODE O 1 2 3 1
- 3 12 hours 3.5 hours 2.5 hours I hour l 1
I ) 4 12 hours 3.5 hours 1.5 hours I hour 5 RCS filled 8 hours 3.5 hours 1.5 hours I hour i 5 RCS partially } drained t 8 hours I hour Operation not allowed
, f 6 24 hours 8 hours 4 hours 2 hours !
b 3.1.2 3 ud 0040 % b t% ore W l %s MOVE M.
- A ,
i i j, - 4 R~g SysTm 80
- EXTc.M)G K l
t p c.yCLE l. a
~
* ; 3/4 1-18 l
1 l TABLE 3.1-4 bhh h ?. Ebb h$((
- g - MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION _^2
^ '"
- T' C '; "' ' : ':: ; ' ': : ; i J" ',
-r '" ' ;? ;^.' T ;;: ^., ;; F OR 0.96 > K,,, > 0.95 Number of Operatino Charoino Pump _ss -
OPERATIONAL j MODE O 1 2 3 3 12 hours 5 hours 2 hours I hour i' 4 12 hours 5 hours 2 hours I hour j J l 5 RCS filled 8 hours 5 hours 2 hours I hour l 5 RCS partially l drainedV 8 hours 1.5 hours Operation not allowed. l l- 6 24 hours 8 hours 4 hours 2 hours
- i W- T e c.' Md {Ad m 3.t.23 ua0, ;
s chl NC l n,,6W
-k3 Mo3)G M um i&g MI 3
I l 1 l 4 l i ., s a l e l @( SV>tsm 80 i EXTtEt@GKM CYCLE f (?4^/B0-j = .u _ m::0SSS-STS ; - 2/4 2-1,
' 8 f. , TABLE 3.1-5 COPY l 3
--""M-~ MONITORING FREQUENCIES FOR BACKUP BORON DILUTION ;
' " " - " ' ' * ^ " ^ ^ - - - ' - ^ ~ ~ - - ^
DETECTION - ;4 y
- .'?
- ^ ^ .^ ^ ' O .' U ^7 : ^ 1. . ^^^^: : FOR K,ff 5 O.95 i
j e i OPERATIONAL Number f Operating Chargino f umps i l MODE O 1 2 3 3 12 hours 6 hours 3 hours 1.5 hours i 4 12 hours 6 hours 3 hours 1.5 hours . 5 RCS filled 8 hours 6 hours 3 hours 2.5 hours I i 5 RCS partially drainedt 8 hours 2 hours Operation not allowed ,)~ 6 24 hours 8 hours 4 hours 2 hours t-
- R 't 3 .1 2 3 l e j ONE i 4Mai ebW %sq w's M03E :
l oms Me. - i l i I a
~
i FUR sysTet4 CO I Ere@s) Fud_ ) CYCLE , i a i l i CSSS & N, ES-STS eff ;;;;,;. ^;;:' 3/4 1 20 i
REACTIVITY CONTROL SYSTEMS . . . _ S 3 /4.1. 3 M3VABLE CONTROL ASSEMBLIES i
- CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length (shutdown and regulating) CEAs, and all part-lengtn CEAs which are inserted in the core, shall be OPERABLE with each CEA of a given group positioned within 6.6 inches (indicated position) of all other CEAs in its group._ '
- ff't'-- it- :: '^.: -' 'A- ---t ': ;t' CE" L
;<;, &g 3 <..s .. .y , ;; ;,; ,,_;,;; ;y;_
7 ;, , ;,- ;;,
..r.
APPLICABILITY: MODES la and 2*. ACTION: 4
- a. With one or more full-length CEAs 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-4 ment of Specification 3.1.1.1 is satisfied within I hour and be in ji at least HOT STANDBY within 6 hours,
- b. With more than one full-length or part-length CEA inoperable or j(
misaligned from any other CEA in its group by more than 19 inches ' (indicated position), be in at least HOT STANDBY within 6 hours. 4
- c. With one or more full-length or part-length CEAs misaligned from any i
other CEAs in its group by more than 6.6 inches, operation in MODES 1 i and 2 may continue, provided that '-- 7 :: .; ::f _::d '- ::--f_ .c
- "'; : 2.'_
25 :-f :' within I hour the misaligned CEA(s) is ' either:
- 1. Restored to OPERABLE status within its above specified alignment requirements, or
- 2. Declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. After declaring the CEA(s) 2 inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided: ,
a) Within I hour the remainder of the CEAs in the group with the inoperable CEA(s) shall be aligned to within 6.6 inches of the inoperable CEA(s) while maintaining the allowable CEA sequence and insertion limits shown on Figure; 3.1-3 end--
==%ibne.; the THERMAL POWER level shall be restr (cted pur-surnt to Specification 3.1.3.6 during subsequent operation.
, "See Special Test Exceptions 3.10.2 and 3.10.4. i
. b 3/4 1-21 4
I LIMITING CONDITION FOR OPERATION (Continued) h ( P.A(1Artav I.l=T) ACTION: (Continued) i b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 : is determined at least once per 12 hours. : Otherwise, be in at least HOT STAUCBY within 6 hours. +
- d. With one full-length CEA inoper_able due to causes other than ment requirements, operation i(n MODES 1 and 2 may continue pursuant 1
to the requirements of Specification 3.1.3.6.
- e. With one part-length CEA inoperable and inserted in the core, operation may continue provided the alignment of the inoperable part length CEA is maintained within 6.6 inches (indicated position) of all other part-length CEAs in its group.
th t long CEAs erted eyond in rtion 1
/. surv 11ance ei er:
sting rsuan o Speci cation
.its, exc
.1.3.2, w for in 2 ho ,s l
1
. R . ore th part 1 th CEA to with their 11 ts, or
!i.
- 2. Reduce ERMAL1 ER to ss tha or equal t that fr ion of D THE . POWER hich is 11 owed by rt long CEA oup j po ion usi figur 3.1-2A.
SURVEILLANCE REQUIREMENTS 1 4.1.3.1.1 The poi,ition of each full-length and part-length CEA shall be determined to be within 6.6 inches (indicated position) of all other CEAs in its group at least once per 12 hours except during time intervals when one CEAC is inoperable or when both CEACs are inoperable, then verify the individual CEA j positions at least once per 4 hours.
, 4.1.3.1.2 Each full-length CEA not fully inserted and each part-length CEA which is inserted in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction at least once per 31 days.
1 4 4 4 CESARfo-$-sT5
%.u. 3/4 1-22 i
REACTIVITV CONTROL SYSTEMS n ; POSITION INDICATOR CHANNELS - OPERATING g h LIMITING CONDITION FOR OPERATION , 3.1.3.2 At least two of the following three CEA position indicator channels i shall be OPERABLE for each CEA: ;
- a. CEA Reed Switch Position Transmitter (RSPT 1) with the capability of determining the absolute CEA positions within 5.2 inches,
- b. CEA Reed Switch Position Transmitter (RSPT 2) with the capability of l determining the absolute CEA positions within 5.2 inches, and
- c. The CEA pulse counting position indicator channel. ,
I APPLICABILITY: DODE5 1 and 2. l 1 _' ACTION:
- 1 With a maximum of one CEA per CEA group ha.ving only one of the above required j CEA position indicator channels OPERABLE, within 6 hours either: ,
.. l
- a. Restore the inoperable position indicator channel to OPERABLE status, or
- b. Be in at least NOT STANDBY, or
- c. Position the CEA group (s) with the inoperable position indicator (s) at its fully withdrawn position while maintaining the requirements of Specifications 3.1.3.1 and 3.1.3.6. Operation may then continue provided the CEA group (s) with the inoperable position indicator (s) is maintained fully withdrawn, es. cept during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2, and each
., CEA in the group (s) is verified fully withdrawn at least once per
- 12 hours thereafter by its " Full Out" limit.*
1 SURVEILt.ANCE REQUIREMENTS 4.1.3.2 Each of the above required position indicator channels shall be determined to be OPERABLE by verifying that for the same CEA, the position indicator channels agree within 5.2 inches of each other at Isast once per 12 hours.
*CEAs are "ully withdrawn (Full Out) when withdrawn to at least 144.75 inchey i U 93 #p) . i
. _ _ _ _ _ I M_ " $ 7 5 3f4 1
REACTIVITY CONTROL SYSTEMS - hCiJ. & E BV C0FT POSITION INDICATOR CHANNELS - SHLITDOWN - LIMITING CONDITION FOR OPERATION i 3.1.3.3 At least one CEA Reed Switch Position Transmitter indicator channel shall be OPERABLE for each shutdown, regulating or part-length CEA not fully inserted. APPLICABILI"Y: MDDES 3*, 4*, and 5*. ACTION: With less than the above required position indicator channel (s) OPERABLE, , immediately open the reactor trip breakers. r
~~
SURVEILLANCE REQUIREMENTS a i
- 4.1.3.3 The above required CEA Reed Switch Position Transmitter indicator channel (s) shall be determined to be OPERABLE by performance of a CHANNEL ,
FUNCTIONAL TEST at least once per 18 months. With the reactor trip breakers in the closed position. t GSSMf0-!VS-sys
=== =.. m1dg .
l REACTIVITY CONTROL SYSTEMS t. CEA DROP TIME g gg g ; LIMITING CONDITION FOR OPERATION i 3.1.3.4 The individual full-length (shutdown and regulating) CEA drop time, ; from a fully withdrawn position, shall be less than or equal to 4 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90% insertion position with: .-
- a. Tcold greater than or equal to 552*F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2. ACTION:
~
- a. With the drop time of any full-length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE I or 2.
i. SURVEILLANCE REQUIREMENTS 7 4.1.3.4 The CEA drop time of full-itngth CEAs shall be demonstrated through measurement prior to reactor criticality:
- a. For all CEAs fo'410 wing each removal and reinsta11ation of the reactor vessel head,
- b. For specifically affected individual CEAs following any maintenance .
on or modification to the CEA drive system which co;1d affect the I drop time of those speific CEAs, and
- c. At Isast once per 18 months.
l
\
l l l I f cf#AfBet0PCS-ST5
.-.._ . . . . . - 3/4 1 y>
l
J REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT PRODF&fiarTJE7lj LIMITING COND] TION FOR OPERATION 0Y y) 3.1.3.5 All shutdown CEAs shall be withdrawn to at least 144.75 inche APPLICABILITY: MODES 1 and 2*#. ACTION: With a maximum of one shutdown CEA withdrawn to less than 144.75 inche , except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:
- a. Withdraw the CEA to at least 144.75 inche 13 or % s)
_ b. Declare the CEA inoperable and apply Specification 3.1.3.1. b i SURVEILLANCE REQUIREMENTS l 4.1.3.5 Each shutdown CEA shall be deterdined to be withdrawn to at least 144.75 inche (lP13 g
- a. Within 15 minutes prior to withdrawal of any CEAs in regulating ;
groups during an approach to reactor criticality, and l
- b. At least once per 12 hours thereafter.
-a R
See Special Test Exception 3.10.2. With K,ff greater than or equal to 1. I l l C.ES$put#D-WFss- ST5 r n- . . ._ _ " " 1 3/4 1-
i i 4 REGULATING CEA INSERTION LIMITS h a
- l 4
LIMITING CONDITION FOR OPERATION l 3.1. 3. 6 The regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits # shown on Figure 3.1-3* _ -. __.
- n- 4
- e ;t; r ;-- N ,,_ : 12 ~ A, ^/ . : m et 9 x^ '- -
y .CE A i insertion between the Long Tern Steady State Insertion Limits and the Trans-j tent Insertion Limits $ restricted to: 1 a. Less than or equal to 4 hours per 24 hour interval,
- b. Less than or equal to 5 Effective Full Power Days per 30 Effective j Full Power Day interval, and i c. Less than or equal to 14 Effective Full Power Days per 18 Effective
) Full Power Months. 1 APPLICABILITY: MODES 1* and 2*f. i ACTION: i i ~. a With the regulating CEA groups inserted beyond the Transient ' 1- Insertion Limits, except for surveillance testing pursuant to {- Specification 4.1.3.1.2, within 2 hours either: l l-
- 1. Restore the regulating CEA groups to within the limits, or
- 2. Reduce THERMAL POWER to less than or equal to that fraction of r RATED THERMAL POWER which is allowed by the CEA group position I
using Figurf 3.1-3.sasadadme
- b. wun U ;_gurating CEA grou,.3 inserted between the Long Tern Steady State Insertion Linits and the Transient Insertion Limits for intervals l greater than 4 hours per 24 hour interval, operation may proceed {
provided either: ' s
- 1. The Short Tern Steady State Insertion Limits of Figure 3.1-3 a
munistgmusuguese are not exceeded, or i
- 2. Any subsequent increase in THERMAL POWER is restricted to less
., than or equal to 5% of RATED THERMAL POWER per hour.
! "See Special Test Exceptions 3.10.2 and 3.10.4. WWith K,ff greater than or equal to 1.
**CEAs are fully withdrawn in accordance with Figde 3.1-3_ :,, ; !.1 '_ when thdrawn to at least 144.75 inchesOg A$p),
over cutback will cause wu ner (Case 1) Regulating Group 5 a egulating 4 and 5 to be dropped with no sequential l edditional Regulat s (Groups 1, 2, 3, a e 2) Regulating Group 5 or Regulating Group d with all er part of the remaining Regulating Gro , nd 4) being esquentially i j inserted. In e, the Transient Inse it and the withdrawal ' i segue pure 3.1-3 6 can be exceese to 2 hours e b ! [ r bid 374 3, 1 - -
- ~
7 ,1
?iD0F & E. ?[ic C .. -
i ' Followi1Df a reactor power cutback in which (1) Regulating Group 5 is 4 dropped or (2' Regulating Groups 4 and 5 are dropped and for cases (1) and (2) shou'.d the remaining Regulating Groups (Group 1, 2. 3. and 4) be sequentially inserted, the Transient Insertion Limit of Figure 3.1-3 can be exceeded for up to 2 hours. Also for cases (1) and (2), the specified overlap between Regulating Groups 3. 4 and 5 can be exceeded for up to 2 hours. ' i em t D* h d h i l 4 1 I k
s-REACTIVITY CONTROL SYSTEMS . PR007 & N 30PY / ACTION: (Continued) t__
- c. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for intervals '
greater than 5 EFFD per 30 EFPD interval or greater than 14 EFPD per 18 Effective Full Power Months, either: 1 Restore the regulating groups to within the Long Ters Steady 1. State Insertion Lir.its within 2 hours, or
- 2. Be in at least HDT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of .each regulating CEA group shall be determined to be r within the Transient Insertion Limits at least once per 12 hours except during
- time intervals when the PDIL Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours. The accumulated ii times during which the regulating CEA groups are inserted beyond the Long Term Steady State Insertion Limits but within the Transient Insertion Limits shall ..
be determined at least once per 24 hours. i
-a i
s CEESWiD-E95 -5T5 i 'f; ; ; unai 4 3/4 1 Y
- l
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- gm i J f ./ / / , / GROUP 2 7 90" / l l
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- l. l ? /\ j / D; / .' / ;
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/// l // Cnour 5 t '" /
4 J7 l /5HORT TERM Y_ / // ~ 2 r , , IN, SE,RTIO iii T a 7." E, , i y-c aour- 5 re-B 7'-r-r-r-y tr>wc
.r STEADY-:
y _ -
, f
~
STATE 05 LIMIT # ' ~ i iii !III /III t/ // , f f : I l E R
- e WRo S a
S o o a o I ! Iou er i i ! l l FIGURE 3.1-3 I I ! CEA INSERTION L; NITS VS THEIDEL POWER l _ '::L: , ;; 'in"' - Q$${N;?995-STS $GE APLRAsWf S$4 w 1-p +{ 1
3/4.2 POWER D15TRTBUT20N LIMITS 1-, ,:- -r -- . rc F[.') J '.: a .: di'. 61~.; 3/4 2.1 LINEAR HEAT RATE LIMITIN3 CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed 14.0 kW/ft. APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER. I ACTION: With the linear heat rate exceeding its limits, as indicated by either (1) the COL 55 calculated core power exceeding the COL 55 calculated core power operating limit based on kV/ft; or (2) when the COL 55 is not t>eing used, any OPERABLE Local Power Density channel exceeding the linear heat rate limit, within 15 minutes initiate corrective action to reduce the linear heat rate to wi. thin. the limits and either:
'_ a. Restore the linear heat rate to within its limits within I hour, or
- b. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL i POWER within the next 6 ho;;rs.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable. 4.2.1.2 The linear heat rate shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Crat herating Limit Supervisory System (COL 55) or, with the C01.55 out of service, by verifying at least once per 2 hours that the linear heat rate, as indicated on all OPERABLE Local Power Density channels, is less than or equal to 14.0 kW/ft. 4.2.1.3 At least once per 31 days, the COL 55 Margin Alarm shall be verified to ! actuate at a THERMAL POWER level less than or equal to the core power operating j limit based on 14.0 kW/ft. i a } 1 . CRMf#0 raws-5Ts
...~o.. . , ~ . . 3/4 2 1
4 POWER DISTRIBUTION LIMXTS 3/4.2.2 PLANAR RADIAL PEAKING FACTORS - F n g pyg @f$ L LIMITING CONDITION FOR OPERATION i 3.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F" ) shall be lesi than or equal to the PLANAR RADIAL PEAKING FACTORS (FC ) used in the Core Operating, Limit Supervisory System (COL 55) and in the Core Protection Calculators (CPC). APPLICABILITY: MODE I above 20% of RATED THERMAt POWER.* ACTION: C WithanF'jy exceeding a correspondirg F y, within 6 hours either:
- a. Adjust the CPC ?ddressaMe constants to increase the multiplier l applied to planar radial peaking by a factor equivalent to greater than or equal to F"y/F y and restrict subacquent operation so that a
~ C margin to the COL 55 operating limits of at least [(F"y/F,c, ) - 1.0)
.: x 100% is maintained; or 1 - .
- b. Adjust the affected PLANAR RADIAL PEAKING FACTORS (FC y ) used in the '
COL 55 and CPC to a value greater than or equal to th?. etasured PLANAR RADIAL PEAKING FACTORS (F"y) or ;
- c. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS
~'
A .2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F" ) obtained by using the incore deteetion system, shall be detemined to be less than or equal to the PLANAR RADIAL PEAKING FACTORS (FC ), used in the COL 55 and CPC at the ' following intervals:
- a. After each fuel loading with THERMAL POWER greater than 40% but
. prior to operation above 70% of RATED THERMAL POWER, and
- b. At least once per 31 Effective Full Power Days.
I "See Special Test Exception 3.10.2. ' I
~ !
4."3 ? M - L*!T 1 3/4 2-2 l l i i \
POWER DISTRIBUTION LIMITS I _ _. . n -,. ;- .vy 3/4.2.3 AZIMUTHAL POWER TILT - T g [ g h b hi.1 O I ( d I LIMITING CONDITION FOR OPERATION 3.2.3 Th' AZIMUTHAL POWER TILT (T ) shall be less than or equal to the AZIMUTHAL l q POWER ThT Allowance used in the Core Protection Calculators (CPCs). APPLItu31LITY: MODE 1 above 20% of RATED THERMAL POWER." ACTION:
- a. With the measured AZIMUTHAL POWER TILT determined to exceed the AI MJTHAL POWER TILT Allowance used in the CPCs but less than or equal to 0.10. within 2 hours either correct the power tilt or adjust the AZIMUTHAL POWER TILT Allowance used in the CPCs to greater than or equal to the measured value.
- b. With the measured AZIMUTHAL POWER TILT determined to exceed 0.10: ,
- 1. Due to misalignment of either a part-length or full-length CEA, ,
within 30 minutes verify that +5e Core Operating Limit Supervisory System (COL 55) (when COL 55 is being used to monitor the core power distribution per Specifications 4.2.1 and 4.2.4) is detecting the CEA misalignment.
- 2. Verify that the AZIMUTHAL POWER TILT is within its limit within 2 hours af ter exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 7 hours and verify that the Variable Overpower Trip Setpoint has betn reduced as vpropriate within the next 4 hours.
., 3. Identify an.. :orrect the cause of the out of limit condition
- prior to increasing THERML POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the AZIMUTHAL POWER TILT is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERML POWER.
l "See Special Test Exception 3.10.2. ( Casa 4e-to;ssg
**'? '!!T: -
- ""F b 3/4 2-3 ll
POWER DISTRIBUTION LIMITS r g SURVEILLANCE REQUIREMENTS 1 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. , 4.2.3.2 The AZIMdTHAL POWER TILT shall be determined to be within the limit above 20% of RATED THERMAL POWER by: ;
- a. Continuously monitoring the tilt with COLSS when the COLSS is OPERABLE.
- b. Calculating the tilt at least once per 12 hours when the COLSS is inoperable.
i
- c. Verifying at least once per 31 days, that the CULSS Azimuthal Tilt !
Alarm is actuated at an AZIMUTHAL POWER TILT -. . - _ _ - to the AZIMUTHAL POWER TILT Allowance used in the CPCs. _ 7(
- d. Using the incore detectors at least once per 31 EFPD to independently [
confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT. c I t i , TQ ! i l i i i
.I l
k N h D .T. __ _ - _ ....@_= S M 3/4 2-4 l l I
I
- POWER DISTRIBUTION LIMITS j-
! yn,TP.'T'""'!CCPV i , id ., i = - 3/4.2.4 DNBR MARGIN _ LIMITING CONDITION 40R 08ERATION 3.2.4 The DNBR margin shall be maintained by operating within the Region of Acceptable Operation of Figure 3.2-I rr'3.2-2, as applicable, = ' r:y e re-r' " N n ;.!- m ' ".;; W . :"TS: L 3-1 ,
~-~ ~ '
i APPLICABILITY:~ W DE l'above 70% of MTED THERMAL PDVER. ) ACTION: With operation outside of the region of acceptable operation, as i'ndicated oy
- either (1) the COL 55 calculated core power ex:eeding the COL 55 calculated core
- power operating limit based on DNBR; or (2) when the COL 55 is not being used, j eq any OPERABLE l'ow DNBR channel below the DNBR liuit, within 15 sinutes initiate er corrective action to restore either the DAER core power operating limi,t or
< - the Dh5 ARite within the limits and either: 3
- a. ^ Restore the DNBR core power operating limit or DNBR o wi in its limits within I hou , u
/ '
- - - b. Reduce THERMAL POVER to less than or equal to 20% ef RATED THERMAL POWER within the next 6 hours. -
- f SURVEILLANCEREQ'JIRQiNTS ..
> 4.2.4.1 The proviItens of Specification 4.0.4 are not applicable. l ,- 4.2.4.2 The DER shall be detemined to be within its limits when THERMAL ) POWER is above 205 of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Opersting Limit Supervisory System ! (00LSS) or, with the COLSS oct of servict, by verifying at least once per
- 2 hours that the DNBR margin, as indicated en'311 OPERABLE DNBR sargin i
channels, is within the limit shown on Figure 3.2-2. l j 4.2.4.3 At least once per 31 cays, the COL 55 Pergin Alarm shall be verified l 2
, to actuate at a THERMAL POWER level less than or equal to the core power l operating 11mit based :,is DNbR. ]
~ '
' - 4-2.4.4 Ti.e following DNBR or equivalent penaity factors shall be verified to be included in the COL 55 and CPC DNBR calculations at least once per 31 EFPD. ! M
. - Surnup (RTO) .- -. - -DNBI Penalty (E)*
0-10 O.5 j 20 - r- 1.0
- 20-30 2.0 i 30-40 3.5 i i 40-50~ 5.5 4 *The penalty for each baten will-be defamined from the batch's maximum burnup j j assembly and applied to the batch's maximum radial power peak assembly. A i single net penalty for COLSS and CPC will be determined
- from the penalties h associated with each batch accounting for the offsetting margirn due to th'e l lower radial power peaks in the higher burnup batches.
) en v%" - S ' 3/4 2-5 .
- (,E 99py @ Q - f y$$ - $~f$
l
l
^
FROBf & RPftW COPY FIGURE .2 1 f I ONBR ARGIN OPERATit' LIMIT B N COLSS (COLSS SERVIC i i i i i i 100 A--
/ - -
-t-
~
I 1 l t t t* f- T g . . . i
$ REGIO F _"
ACCE ABLE s OPE ATION
< B0 - - - - - - - - - . - - - - . i E / l 3C
/ .
EE ge / ,/ - - ,
. gy - l ; . - . . . . - .
~
l !~ b* / . ! i EO / ww h / REGh0N OF 1 ) e . UNACCEPTABLE ! g
/ OPERATION l as - /
s 8 '
/ l 5
g 20 -
/ -
y g - f 0' o ao / '
/ so' oo l powsR I PE wT or RAT /rsERu d
FIGURE 3.2-1 l DNBR MARGIN OPERATING LIMIT BASED ON COLS$ ' (COL 55 IK SERVICE) .
- SEE A/pttuu)T6 sAf i b 3/4 2-6
P~ -m_. _ 1 PM r. rr;~ ( - . i U 3.2- <m.,_.
!\ '
P Tl L B E CO P ECT ULA R . OLSS F SER , I J g an i i j -1
..._......-.-l, i .,:- _.. I -
i i .._,
- i. - t -
- ! I l .
I i. . -~
- i
.~ REGION OF ,
I# 7 l-' _l*
- i i .
I l
- - - ACCEPTAB i .:: .
%.-. _- - . . l -
l / 0.55 m NATION ,
)
i j i 2- . ; l . I
- L[ +
~
j j._ j : ', i- j i
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, i ! ( .05,0.51) i !'. .
I
;(.25, Ohti_. ._
..g _._ . . . . . . .
I f. i : . 0 .
- I -
I E :
- _. 4 _ _ .: ._. .__ ; . .. ... .. ..._.....-. ..
- g . _ . . _ . . , _
j - .- ! ~l l' , , I l 8 i
- 8 I i - l i .. .
g .._ : _. ; ._ - - ._ t .
.. . j - . .. ;_ . .. 4 _ _._ l ._. . .j --- i
._.g 8
0 46 !3 .. .s ,i , s
. t . l . . . s.
I 0.45 . j
~. 1 . .
b ,j__:._. - l
!'_-[-
e
- - i ' -h
-J:.,. : . REGION OF i 4 l . . . ,
j-- ,g i i : UNACCEPTABLE ;
'g ' '
E A-::-- ! . - - . 4 ] 3 .
..j. ... r, .
., l, l 1. OPERATION j 1 1,. - . - 3
- 00 -
I t- i l=.: . l_. : .i . . . . . l-. ._: . p; j.4..i:lu::: - $-- . --
+ , .
1 f -K 1-- . ;- . .e
- i .( .30,0.35) -
_ . . _ ._ o.
. . i i. .
I s_ . . .-. . .. . ., %. s
. ...j. : ; . . ,
O.35 ,
,- - : -- --- y-i i
. . _ _ _ . . . . . . . . ._. g. .. .
r , . i
- i .
s 1 i i
-~
- 0.30 0.3 0.3 4.2 4. 0.1 0.2 I
1 CORE AVERAGE A5l* i sm gurs.a r a e n me n a m ---^ . ma m menen i r G ed r a r N .- Mrv
.__w-- ^^= * -? _7 3.hd . .._.~-w. --- .v... m-.sy FIGURE 3.2-2 DNBR MARGIN OPERATING LINIT SASED ON CORE PROTECTION CALCULATORS (COL 55 OUT OF SERVICE) f am=.-, ~ . ws. &<w - *
-srs ser 3/4wouwn 2-7 me.
i
POVER DISTR 2BUT20N LIMITS . 3/4.2.5 RCS FLOW RATE P.00F1 1 d iF.W COPY , LIMITING CONDITION FOR OPERATION 3.2.5 The actual Reactor Coolant System total flow rate shall be greater,than er equal to 164.0 x 105 lbm/hr. APPLICABILITY: MODE 1. AtlION: With the actual Reactor Coolant System total flow rate determined to be less than the above limit, reduce THER.M L POWER to less than 5% of RATED THERMAL POWER within the e t 4 hours. - -
.. SURVEILLANCE REQ"IREMENTS
- ~
4.2.5 ihe actual Reactor Coolant System total flow rate shall be detemined
~
. -to be greater than or equal to its limit at least once per 12 hours. - ,
3
.a l
s l p- .. ~ ; c ., a . . :. .
. .. t ..
' .l- :'f..!.. ~ . 2. ._ . :.' . . ,
~ ^ ^ ~ ~ ~ ~
& Q^'..J } 0 - Z: N. g -$ 7eJ 3/4 2-8 -
ia ,
POWER DISTRfBUTION LIMITS _ 3/4.2.6 KEACTOR COOLANT COLD LEG TEMPERATURE (( g QD] [f LIMITING CONDITION FOR OPERATION
~
3.2.6 of TheMactor Opera /olant[lo %g [emperature Acceptable tion shown in Figure 3.2-3. g (T ) shall be withinXthe Are APPLICABILITY: PODE1*and2f y ACTION: Vith the reactor coolant cold leg temperature exceeding its limit, restore the temperature r/ within its Ifait within 2 hours or be in HOT STANDBY within the next 6 L xrs. _ SURVEILLANCE REQUIREMENTS
~
(
; 4.2.6 The reactor coolant cold leg temperature shall be detemined to be i within its limit at least once per 12 hours.
~
s i L
~
d. i c ,
*See Special Test Exception 3.10.4.
MMIO"M-STS
"".'0
. t" L ; wdii 3/4 2-9 l
" ,*s Y
- PE ~
C .: n-~ I.e. b . . ,_ , L ?,, . y
- l FIGURE 3.2 3 _.,,,_,_.,"
7 -~ REACTOR COOLANT COLD LEG TEMPER ATURE vs CORE POWER LEVEL ) : 580 i , , , , i i i i i 3 . 575 ! 570 J70 , I, 568 568 w I H5 . I g AREA OF ACCEPTABLE . OPERATION 562 . 560 /
, w l ~_ >-
O w 555 . a
- J i- D 22 I
8 550 , l 0 i. 540 . I e I i t I t g 1 1 i 1 j 10 20 30 40 50 SO 70 SO 90 100 i i ! CORE POWER LEVEL.E OF RATED T 1ERMAL POWER ' \ 1 j -, J 3 J J" 4 FIGURE 3.2-3 i j ItEACTOR COOLANT COLD LEG TDePERATURE VS CORE POWER LEVEL 1 1 i 4 i l
; 5? I 3/4 2-10
POWER DISTRIBUTION LIMIL
** j 3/4.2.7 AXIAL SHAPE INDEX l
LIMITING CONDITION FOR OPERATION n 3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:
- a. COL 55 OPERABLE
-0.28 i ASI i 0.28
- b. COL 55 0UT OF SERVICE (CPC) '
-0.20 i ASI i + 0.20 APPLICABILITY: MODE I above 20% of RATED THERMAL POWER *.
ACTION:
~ ~
- With the core average AXIAL 3HAPE INDEX outside its above limits, restore the core average ASI to within its limit within 2 hours or reduce THERMAL POWER to less than 20% of RATED THERMAL POWER within the next 4 hours. ,
'l SURVEILLANCE REQUIREMENTS 4.2.7 The core average AXIAL SHAPE INDEX shall be determined to be within its limit at least once per 12 hours using the COL 55 or any OPERABLE Core Protection ; Calculator channel.
-o l
4 .I i l "See Special Test Exception 3.10.2. h c4*Meo-mss % 3/4 1-11
POWER DISTRIBUTION LIMXTS .__. . 3/4.2.8 PRESSURIZER PRESSURE ((g*f b Sht bY
~
r LIMITING CONDITION FOR OPERATION 3.2.8 The press"riter pressure shall be maintained between l'815 psia and 2370 psia. APPLICABILITY: M0uis 1 and 2.* ACTION: With the pressurizer pressure outside its above limits, restore the pressure to within its limit within 2 hours or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS
~
4.2.8 The pressurizer pressure shall be determined to be within its limit at least once per 12 hours. t
*5ee Special Test Exception 3.10.5 l t
b . 5 -3/4 2-12
^
l
3/4.3 INSTRUMENTATION ,, _, 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION
/
'F.
itu i s o. I m g l.__._______.__ y _ __.__._.__._a ! LIMITING CONDITION FOR OPERATION , 3.3.1 As a minimum, the reactor protective instrumentation channefs and bypasses 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. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS _
~
4.3 1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and 4
~
CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown
. in Table 4.3-1.
,. l 4.3.1.2 The logic for the bypasses shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the
" Total No. of Channels" column of Table 3.3-1.
~'
4.3.1.4 The isolation characteristics of each CEA isolation amplifier shall i
' be verified at least once per 18 months during the shutdown per the following '
tests for the CEA position isolation amplifiers:
- a. With 120 volts A.C. (60 Hz) applied for at least 30 seconds across the output, the reading on the input does not change by more than ,
0.015 volt D.C. with an applied input voltage of 5-10 volts D.C. l
; 3/4 3-1 1 l
1
INSTRUMENTATION
- F SURVEILLANCE REQUIREMENT.i (Continued)
- b. With 120 eolts A.C. (60 Hz) applied for at least 30 seconds across the input, the reading on the output does not exceed 15 volts D.C.
- 4. 3.1. 5 The Core Protection Calculators shall be determined OPERABLE at least '
once per 12 hours by verifying that less than three auto restarts have occurred on each calculater during the past 12 hours. The auto restart periodic tests Restart (Code 30) and Normal System Load (Code 33) shall not be included in this l total. , 4.3.1.6 The Core Protection Calculators shall be subjected to a CHANNEL FUNCTIONAL TEST to verify OPERABILITY within 12 hours of receipt of a High CPC Cabinet Temperature alarm. 3 .(
- e O
p e i. l
)
~
Qf3Eest-Oss-5Ts _ -... -m. 4 3/4 3-2 , 4 L
^ ~
. :- S i e . 1 C:2 bp
\ TABLE 3.3-1 p.,
j REACTOR PROTECTIVE INSTRUMENTATION pJ,
- t M1HIMUM ry; 4 ;
CHANNELS APPLICA3tE - ! 10TAL NO. CHANNELS l M0005 40r11N FUNCTIONAL. UNIT OF CHANNELS TO TRIP OPERABLE i 1. TRIP GENERATION h J. l l A. Process
- 1. Pressurizer Pressure - High 4 2 3 1, 2 2
[ #
- 2. Pressurizer Pressure - Low 4 2 (b) 3 1, 2 2,3
% 3. Steam Generator level - Low 4/5G 2/5G 3/5G 1, 2 2,3
- 4. Steam Generator level - High 4/5G 2/5G 3/5G 1, 2 2,3
- 5. Steam Generator Pressure - Low 4/SG 2/5G 3/5G 1, 2, 3*, 4* 2,3
- 6. Containment Pressure - High 4 2 3 1, 2 2,3
- 7. Reactor Coolant Flow - Low 4/5G 2/5G 3/5G 1, 2 2,3 2,3 w 8. Local Power Density - High 4 2 (c)(d) 3 1, 2
" 4 2 (c)(d) 3 1, 2 2,3
- 9. DN8R - Low f 8. Excore Neutron Flux 1, 2 2,3
- 1. Variable Overpower Trip 4 2 3
- 2. Logarithmic Power Level - High #
- a. Startup and Operating 4 2 (a)(d) 3 1, 2 2,3 4 2 3 3*, 4*, 5* 8
- b. Shutdown a 0 2 3,4,5 4
?
C. Core Protection Calculator System ...
- 1. CEA Calculators 2 1 2 (c) 1, 2 6, 7 Core Protection Calculators 4 2 (c)(d) 3 1, 2 2,3,7 2.
- A bbi w n- e ,el
- l!: b l:d E Aw[ $i xw ** - .
- a cs
1 .; 1 - 1 l R TABLE 3.3-1 (Continued) ij g REACTOR PROTECTIVE INSTRtMENTATION l . j h ' MINIMUM CHANNELS APPLICABl.E
- TOTAL NO. CHANNELS FUNCTIONAL UNIT OF CHANNELS TO TRIP _ OPERABLE MODES __, ACTION
; l
; D. Supplementary Protection System I Pressurizer Pressure - High 4 (f) 2 3 1, 2 8 It II. RPS LOGIC V 3 1, 2 1
~ A. Matr.ix Logic 6 1 6 1 3 3*, 4*, 5* 8 r . -
Initiation logic 4 2 4 1, 2 5 B. 4 2 4 3*, 4*, 5* 8 y" ,. . , . . .. . , w,. III. RPS ACTUATION DEVICES , A. ReactorTr.ip Breaker 4 (f) 2 4 1, 2 5- . 8. 4 (f) 2 4 3*,4*,5*' - B. ,Manua) T, rip.... 4 (f) 2 4 1, 2 5 4 (f) 2 4 3*, 4*, 5* 8
. . lli e ' e . ,e M
., y- ;, t . .. . . - o
~ c3 re; y, . . , - y5 t} } (3*, 4 /
~-
.... ,; i
. . . Qo
!
- I. * *
.".I.3 Q.1
- a. -
* . 4 O
6 C"3
** U W
~
TABLE 3.3-1 (Continued) .- l X TABLE NOTATIONS
\-~
E3f & CiE br[ _jJ
. *With the protective system trip breakers in the closed position, the LEA :
drive system capable of CEA withdrawal, and fuel in t e reactor vessel. ' iThe provisions of Specification 3.0.4 are not applicable. . (a) Trip may be manually bypassed above 10 *% of RATED THERMAL POWER; ! bypass shall be automatically removed when THERMAL PC F is less than or j equal to 10 *% of RATED THERMAL POWER. (b) Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than er equal to 500 psia. (c) Trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass shall be autonatically removed when THERMAL POWER is greater than ! or equal to 1% of RATED THERMAL POWER. j ~-- (d) Trip say be bypassed during testing p rsuant to Special Test Exception 3.10.3.
'i. (e) See Special Test Exception 3.10.2.
(f) There are four channels, each of which is comprised of one of the four i
, reactor trip breakers, arranged in a selective two-out-of-four ,
configuration (i.e. , one-out-of-two taken twice) ACTION STATEMENTS ACTION 1 - With t%e number of channels OPERABLE one less than required by the Minimum Channels OPERULE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least ' HOT STANDBY within the next 6 hours and/or open the protective system trip breakers. l ' ACTION 2 - With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue i provided the inoperable channel is p1 meed in the bypassed or tripped condition within 1 hour. If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6.t The channel shall be returned to OPERABLE status no lat than during the next COLD SHUTDOWN. ) W A(ea4Adx34 ; I YXhYSN _ . . . . . . . - , , - 3/4 3-5 4
ULEMDDPl (CsettauGd)
~~
ACTION STATEMENTS , q
.g . .
Wit.h a channel process measurement circuit at affects _ j aultiple functional units inoperable or in test, bypass or trip all associated functional units as listed below: Process Measurement Circuit Fvattional Units - Y Q A g g M Lypassed/ Tripped' -
\
Variable Overpo o r (RPF) (Linear Power d local Power Dent (y - ligh (RP5) (Subchannel or Linear) DNBR - Low (RPS)
- 2. Pressurizer Pressure - High Pressurizer Pressure - Ligh (RPS)
(Narrow Range) Local Pewer Der.sity - High (RPS) DNBR s.ow (RPS)
- 3. Steam Generator Pressure - Steam Generator Pressure - Low Low Steam Generator Level 1-Low (ESF)
~
5 team Generator Level 2-Low (E5F)
- 4. Steam Generator Level - Low Steam Generator Level - Lew (RPS)
(Wide Range) Steam Generator Level 1-Low (ESF) Steam Generator Level 2-LowdEST) i 5. Core Protection Calculator Local Power Density - High (RPS) DNBR - Low (RPS) ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied:
- a. Verify that one of the inoperable channels has been bypassed and place the other channel in the tripped s.. ,
condition eithin I hour, and
- b. All functiona! 5'I s affetted by the bypessed/ tripped channel shall also be placed in the bypar. sed / tripped
~'
, condition as listed below:
rocess Meas rene Circuit Functional Unit Bypassed / Tripped MP
\.
. .inear 'ower
_.. M-->Variable Overpower (RPS) (Subchannel or Linear) Local Power Density - High (RPS) DNBR - Low (RPS)
- 2. Pressurizer Pressure - Pressurizer Pressure - High (RPS)
. High (Narrow Range) Local Power Dens DNBR - Low (RPS),}ty - High (RPS) t g
CW ^90 -N9Gs
'
- Si: T:' 1 - - -
g5 3f4 3-s
TABLE 3.3-1 (Continu9d) g gy l I ACTION STATEMENTS g
- 3. Steam Generator Pressure - Steam Generator Pressure - Low Low Steam Generator Level 1-Low (Esf)
Steam Generator Level 2-Low (ESF)
- 4. Steam Generator Level - Low Steam Generator Level - Low (RPS)
(Wide Range) Steam Generator Level 1-Low (ESF) Steam Generator Level 2-Low (ESF)
- 5. Core Protection Calculator Local Power Density - High.(RPS)
DNBR - Low (RPS) l STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL EUNCTIONAL TEST. Subsequent l I STARTUP and/or POWER OPERATION may continue if one channel is ' restored to OPERABLE status and the provisions of ACTION 2 are satisfied. ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
- ACTION 5 - With the number of channels OPERABLE one less than required ~
i
- by the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor trip breaker of the inoperable channel is placed in the tripped condition within I hour, otherwise, be in at least HOT STANDBY within 6 hours; however, the trip breaker associated with the inoperable channel may be closed for up to I hour for surveillance testing per Specification 4.3.1.1.
ACTION 6 -
- a. With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours, each CEA is verified to be within 6.6 inches (indicated position)
---P ':
of all. .ot,her _. . . . _ CEAs _ . mm in m _.its group._."_f^.:,.'.. u_ __ m, fry . s ,,..<.. , __
- b. With both CsACs inoperable '-f C^'M '- - '- . operation
-< aay cont'nue provided that:
Eithin 1 hour: Operation is restricted to the li shown in igure 3.3-1. The DNBR ma quired by
/ 5 fication 3.2.4 is ced by this I restr n when bo C's are inoperable and COLS5 eration.
1g rgin required by j b) The tin at Ra, Spe cation 3.2.1 is tained.
/ placed out c) he Reactor Power Cutback Sys '
of service. l l l i 1 C.E.iSSAit na n uronr .foem- NSSS -53 n 3/4 3 7 d
i gggp Pr100r"&.nB.[.'!!51;rf i
- 1. Within I hour the margins required by Specifications 3.2.1 and 3.2.4 are increased and maintained at a value equivalent to greater than or equal to 19% of RATED THERMAL POWER.
i
~ s
- e
- i
-s t
+
l
)
I l l
n a s r ., '.e :-i t 3:9;i TABLE 3.3-TTc6ntTnisid) 313 W r.; y ACTIDFF5TATERENTS Qf f Qj f
- .c. w e.
- 2. Within 4 hours:
S . m .2 . . n . . . s, .. . .6 3 :2
; . .r v - .
. .. , - :.r. -:: x.a) .. All full-length and part-length CEA groups are ,
= .
m n.n. , . :: withdrawn to and subsequently maintained at the ;
"*'
- M "testinig Full Out" position, pbrtuant excep@t br the 're during surveillanceireme' hts c.
-n r t.:
c ml ., - ' tion 4.1.3.1.2 or' for tohtre) when CEA group 5
.. may be inserted no further than 127.5 inches
. . .4 . withdrawn. .
,,.,A..g .
.,~~
h P The "RSPT/CEAC Inopera M ddressable constant .
-. , . .3 a v , , , - in the, CPCs. is se t- toy- . 1 - - -.__ __.-
- _: '. +* .,i/- dE49deenominoperathieMo$ AAg , l cy' The Cotr'oE !
n Element Drfh' Mhiin Control
'~ ' ' ' System (CED!9CS) ir placed ih' and kubsequently :
maintained in the "Standtif imodeexcept during
.. . . CEA group 5 motion pemitted by. a) above, when .
c the CEMCS way be operated in either the
- Manual-
. . Group" or " Manual. Indiyique-1." mode, i
- 3. .At .leasrt ence per 4 hours,+al1 full-length and par.t-l- - - - -
Iength CEAs are parified f.ully withdrawn except
. during surveillance testing pJrsuant ta Specification .
!. - . 4.1.3.1.2- or. during insention of CEA group 5 as = pemitted by. 2.a) above, timn vprify at least once ; ] -
., per 4 hours that the, insertad CEAs are aligned within -
.. 6.6 bettes (indicat.ed pos4 tion) of an other CEAs in -
j its group. . . . - e e i- ;. _ l
. Fo.110w ng a CEA pisalignment with.both CEAC.'s .r.:: ,,
. : . inop e and .COLSS..in operstdon, operation may .
i j - cont ue-provi d i. hat .withW 1 hour: ) 3-
- - ..?Y'"'The
- pohr NY utid' to' hW 'c'f hit pre-sisali ned -
, ,7 bd'M n'ot E Mu'ced' '"less than 5 of R iED THERMA P0'WER. WThir'p ' ' restriction replaces :
. : ;. .1 ~~. - : in power triction;nt 4:ication. 3.1 .1, \ ! ~# therwise. ec.ifi.cet. ion 3.-3 ,1 nnains icable. ;
- c. k;t both CEAC inoperab~le end. LSS out-of-s vice, '
i
. Of . rat, i of) may, ontinue. . p.rovide that:
c ... .
. . . . ... o
. : 1.3 -Within hogrk 1. .. .v
^ 2- ' '. ' *'.
.' 'a ) ' i exTs' ting CPC' sue of the C addressab e
...i.
t nst'aht IPERR1* ' multipied 1.19 and he j - "~" e'sulttng valve *4 re-entered nto the CP s. q . . , -
- ..f g . getdFP' 'Cuttia'ck Sys en is pla ed out i..
g ,,w, : - . : < c: ;
.v .: . <. i.v , .,;
cifica-t.ine , j
.g: . 7,j7,3.CdLS$
tion 3.274,' 'odt fFr$ doFierviceLie}tothis applical: 10 j operation. de of ( ;
; 3 9 f/j j.:f 1 TitT %Eiv '. % i
TABLE 3.3-1 (Continued) ~ . ;. n 3 -.r. I Inu 3 Q 1 r.i.h b 7.5.. I ACTION STATEMENTS _ __ j
. }
it .
- 2. Within 4 urs:
i 1 , a) I full leng and part length CEA groups are ! j withdrawn to nd subsequently maintained at the l i
~ -- - -
*Fuli Out" sition; except uring vurveillance j testing p suant to the re irements of Specifi- i cation .3.1.2 or for c trol when CEA group 5 l i may be nserted no furt than 127.5 inches withd wn.
. b) T "RSPT/CEAC Inop able" addressable constant l i the CPCs is se o be indicated t at both ;
EAC's are inope ble.
; / c The Control E ment Drive Mechan Control
- ' System (CE ) is placed in an subsequently l maintained, n the " Standby" e except during !
CEA group /5 motion pemit y a) above, when the CE 5 may be operated n either the " Manual
- 3 Group" r " Manual Indivi al" mode.
. t
,e *
- 3. At least nee per 4 hours, 11 full lengt nd part
i length EAs are verified 11y withdrawn xcept duriysurveillancetes ng pursuant t pecifica-tio 4.1.3.1.2 or duri g insertion of EA group 5 as itted by 2.a) ab e, then verif at least once i r 4 hours that t inserted CEA are aligned with' ,
.6 inches (indic ed position) all other CEAs ' t 4 its group. '
a ! i L . Following a sisalignmen with both CEAC' and COL 35 l inoperable, pera'tfon may ontinue provide hat within
! ..... ,1.. h.our: '.
i U'.The p' ' is reduced 45% of the pr misaligned pow !
' 'tist nee not be red ed to less than 0% of RATED T MAL l
! POWER. This power estrict n rep 1 es the power i restriction of ification .1.3 , otherwise 5 cifi-
, i ation 3.1.3.1 ins applicatile, i ! CTION 7 - With three or more auto restarts, excluding phriodic' auto restarts (Code 30 and Code 33), of one non-bypassed calcui tor I i e during a 12-hour interval, demonstrate calculator DPERABILITY
; by performing a CHANNEL FUNCTI0i4AL TEST within the'agxt 24 hours. l b; C. TION.8 _ r With. .; stumber.pf_9fERAblE channels one less than the Minimum l m Channels /0PERABLE requirement, restore en insperable channel i to OPERABLE status within 48 hours or open an affected reactor l
- tripibreaker Within the Mut he'u r.
db
- ** '? ~L .
hdE'"$N I bl4 2-
- v. , -
l i
- _ _ _ _ ~ _ . _ - . _ _ . _ -.- - .. . -. . . - . - . . _ _ - . _ _ - _ _ _
e . V 1 g \ TABLE 3.3-2 . REACTOR PROTECTIVE INSTRIINENTATION RESPONSE TIMES g jY FUNCTIONAL UNIT RESPONSE TI f l d I. TRIP GENERATION y
, . A. Process e "'
! 1. Pressurizer Pressure - High i $ seconds @
g Z i '
- 2. Pressurtzer Pressure - Low $ $ seconds
- 3. Steam Generator Level - Low i $ seconds j
] 4. Steam Generator Level - High 1 @ seconds -2
- 5. Steam Generator Pressure - Low 1 W seconds Eh
\A 6. Containment Pressure - High S O seconds %
- 7. Reactor Coolant Flow - Low i % second 5 @g
- 8. Local Power Density - High 4 R
[ a. Neutron Flux Power from Excore Neutron Detectors $ $ second*
< JGGS second**
- b. CEA Positions
- c. CEA Positions: CEAC Penalti Factor 7 (Aff second**
- 9. DNBR - Low O _
- a. Jeutron Flux Power from Excore Neutron Detectcrs < M second* I
- b. CEA Positions Cold Leg lemperature 7 JER$rsecond**
i #4 6 secondH 3 C3 c.
- d. Hot Leg Temperature $ WWTseconde#
second# M
- e. Primary Coolant Pump Shaf t Speed i y Reactor Coolant Pressure from Pressurizer < second#M f.
- g. CEA Positions: CEAC Penalty Factor 7 $2@ second** *
~
N 8. Excere Neutron Flux ,, g
- 1. Varleble overpower Trip i k second* g C3
- 2. Logarithmic Power Level - High y
- a. Startup and Operating $ #AP second* -'
- b. Shutdown <
~ g second*
'w d to lcm ovJ (.6ef')
$9 $ W33AR N
_ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ . _ - - . _ . _ _ _ _ _ _ _ - . - - _ s.:.-_ _- _a --- _ _ _ _ _ - _ _ _ -- _ _ . _ _ . _ _ _ _ . _ _ - -
i g TABLE 3.3+2 (C:nti.iued) 3 l . REACTOR PROTECTIVE INSTRtMENTATION RESPOMst TIME 1
. FUNCTIONAL UNIT RESPONSE TIME _..
. g
; C. Core Protection,041culator System %
i g
;. . CEA Calculators
~
Not Appilcable , 3 l I 2."CorePretectio[palcuNtorf',,.,, Not. Applicable' hf "P r l M 0. Supplementary Protectton Systee - g 5 $ second Presiur.lzerPypure,.,High h., II. RPS LOGIC ,
',I..{.v. . ,, . .
j
*A::
A. ,Matylx logic, Not Applicable Q m
- 8. Initiatly.i Log!" , .
Not Applicahle 3 N : III. RPS ACTUATICQ DEVICES ' .
- Not Applicable A. Reactor Trip.Breckers. ,
g B. Manual, Trip. Mot Applicable N Neutron detectors are exegt from response time testing. The response time of the twutron flux signal portion of the channel shall be measured from the detector output or from the input of first electronic component in channel.
#;-- ^ ^
- ' n -- ;f ^. _ !!;
^
.r , vi a m.. m . m.up -.", k O'" : : - . . :; : - - '" '
x
~
#The pulse transmitters measur'ing pug speed are exempt from response time testing. The response time shall be measured from the pulse shaper input, NResponse time shall be measured from the output of the resistance temperature detector (sensor). RTD response. time shall be measured at least once er 18 months. The measured response time sf the
/ g slowest RTD shall be less than or equal to a_ sm. 2
_ "'" it : x--* -* ?:7 e # "" 2.2. .ui s % - >: .. r -
. J ,... a da ' ' --?r seconds. "j a- * - - - * - '- ' ' e v ..J,ts .. h ;n:*--Le
- 2. . - -. _.. -; ; = -- e_o r e u arg ; g :
caree ne -
"'n ' l'- --*=1*8-c f--.;
en-i,
; 8-t
_ - ,tt c" ntcf c'- .. ^^ 4 c' '; t: f::':- ' :..;,:- ? ; c 2 ? ' ,-c ;- - ' ' * : s ; 7 -- - i m t = 'a t' - - -
^':: c r ;m._ = r - i . : t - ! ' --*
^
### Response time shall be measured from the output of the pressure transmitter. The transmitter response time shall be less than or equal to 0.7 second.
x "Wd PR00? & REV;EW COPY j
** The CEA position transmitteri,' art exempt from response time testing. The response time shall be measure'd ,
fr w the irpi o the CPC CEAC or signal isolator. , j (** Re5Ponse times are verified using.CPC Etspople Time Test 5sitware, and are for hardware delays only. )
- W$
$ p I l l i I I
e.r.* I'* p r.,n y m gv TABLd.3-2a 'd N 'I- I N 3 8 I CREAS S IN BERR BERR2,- DBERR4hERSUS i '. RTD' DELAY TilfE5 s, lE*. ', ERRO-- . - RR2 BERR4 , RTD DELAY TIME INCREASE NCREASE INCRE E (t) (%) (%) (%) ,
<8 sec . . -
. .0 0 0 870 ec < t < 10 sec 2. 5' 2.0 1. :.
- 10. see i: t:51 0 4ec 4.0 4.0 f.
. _' i
- :~ . . :: ;
3TE: BERR t m increase are not c mulativ . For ex ple, if he timej - consta t changes 10.0 t 5 13.0, om the va e of 8. e BERRO i crease rom its o 'ginal (t i 8.0 ec)
< t 5,10 0 see to he ran e
/
- valu is 6.0 not .5 + 6.0. -
=
3
. .- 1 L *
, h .
. T-r.
s
,, .' r. ,.
.S .
, Y %*
p s. .
- w ? s . -
t E : s. . ..' . : ; I i E 4-2
-? .. r -' '; ;- -- .
. t. ?* .- . . .
s . ,a
; ; C .,. t * *
.s
- ~ ;
. . .- r e
- . '. ;,. =
8 2 - -
.. ; .. . r:
. t -
- . 1 c.
W ! - i t. . .
~; ; . .
y . u
. . 3 . . .
l-
' g **e* :
;~* -
.f .. ,; .. ; . . , -
/ { tm
. .-. . C g .~.
O.
.e om.
m _- _m--p Jm n .g i --
. rikw 16 wu URJ4 A at 's a sw *#* *'
I l l i 1
' t l 4
""if 4. M r
- REACTOR PROTECTIVE INSTRUNENTATION SURVEILLANCE REQUIRENENTS i CHANNEL MODES IN WHICH I CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST REQUIRED
! iD FUNCTIONAL UNIT I I. TRIP GENERATION e r> 1 A. Process-g 1. Pressurizer Pressure - High S R M 1, 2 i
l-M 2. Pressurizer Pressure - Low S R M 1, 2 5 R M 1, 2 -
- 3. Steam Generator Level - Low Steam Generator level - High S R M 1, 2 4.
S R M 1, 2, 3* , 4'
- 5. Steam Generator Pressure - Low Containment Pressure - High R M 1, 2 5 6. S 1, 2 ' ~ "
Reactor Coolant Fiow - Low S R M Y 7. M, " (5) 1, 2
! 8. Local Power Density - High S 0 (2, 0;, ; (0, 5) ,
- 9. -DNBR - Low S Cir. ". 0 '0 m 5) "
*@ 1, 2 , j' "i"L % 4 9
i - i B. Excore Neutron Flux , 0 (2, 4), M (3, 4) lM 1, 2 ',. !
- 1. Verfable Overpower Trip 5 Q (4) ,
]l Logarithmic Power Level - High S R (4) M and S/U (1) 1,2,3,.4',1;
- 2. aad * '
- i ..~ q; . l C. Core Protection Calculator System .
S R N, R (6) '"' 1, 2 L- -
- 1. CEA Calculators
- 2. Core Protection Calculators S D (2, 4), R (4, 5) M (9), R (6) 1, 2 M (8), S (7) ,
i W ** Q Q,~ ' faR & Yf
" R M %W Il kgW keeS (&f) h 'Sm-. Cane.%ac% cAc,Dh W w s.o_. % I M see 1
ll, 1I 2 * *
- 5 5 5 5 .
H .. C , , , , IE 2
- 2
- HCD 4 4 4 4 WNE AR , , , ,
nO NLI ILU IQ 3 3 3 3 5eC 4 =@A'ss 8 X SEE , , , , EVR 2 2 2 2 2 - DR OU , , , , , . MS 1 1 1 1 1 -
) .
i ( ' u L LN A )0 ). / - EOT 1 - S NI S NTE ACT ( R 5 - T HN - N CU . , E F M M M M M M E R - I U Q .. E . R . E . _ C N . A
, L N
) L O d I LI . . .
e e E ET A. A.. A. A. u V NA . _ n R NR R N N N N - i U AB -
- t S HI
_. ' n o N CL A C O C ( I T . 1 A
- T L . . . .
_. 3 N E
. E NK A. A. A. A.
4 M NC 5 M N M M U AE E R HJ J L 1 CC B 5 . A N T I E ._ V I T
- me h
g _ C t i t s H 1 y 0 S - R P n e
\ o r
. R i u O t s s T c s r C e e e A t r k E o P S a R r c E e P r i C r y e g I B r z o V p
. i c L E a r i D i p
. t u g n r i n e s s o o N T r L i O T m e t I r T e r x a T o l I l P C i i A t a N p I r t U c u U p G t i T a n u O a n C e a _. L S L M I A R M A N S S O . P . . P . . I O R A 8 R A 8 T C . N . I . U I I F I I
}%
,}0 l
.l P
.l 'l l '
l TABLE 4.~3-1 (Contin 6ed) 3
'iABL N3TATfDNS
* - With reactor trip brealers'in the' closed position and3he'CEA drive j system capable of CEA withdrawal..and. fuel in.the reachrgessel.r. =-
l (1) - Each STARTUP or when pequiredhwit' the-reacto' trip breakfrs closed and the CEA drive system capable of rod withdrawal, if 'not performed j in the previous 7 days. g , (2) - Heat balance only (CHANNE FUNCTIONAL TEST not inJ1 udedL above 15%
! of RATED THERMAL POWER; adjus the /inear/pwer /evelt 0,s_CPC delta T g '
s power and CPC 3uclear power 2% S: to agree with the talarimetric i hrg, - caTcuration if bsolute dilference is greater than 2%. TuHng AHYSICS e TESTS, these da ly calibrations may bet suspended provided4hese;
; f calibrations are performed upon reaching each major test power p3ateau ,
i and prior to proceeding to the next major test power plateau. j i (3) - Above 15% of RATED THERMAL POWER, verify that the linear power sbb-
- channel gains of the excere ektectors are consistent with the vilues l
used to establish the shape annealing matrix elements in the Corm Protection Calculators. N 1 8 s i 4) - Neutron detectors may be excluded from CHANNEL CALIBRCIDN.
. - u .- -
sC m5) - Af ter each fuel loading and prior to exceeding 70% of 3t4TED THERMAL' '
- . POWER, the incore detectors shall be med to determins;tihe shapr '.
l! annealing matrix elements and the Core Protection Calculators shellf : 7 use these elements.
- 2 j ]
! (6) - This CHANNEL FUNCTIONAL TEST shall include the injection of sieciated process signals into the channel a's close to the sensorcas practicaible j to verify OPERABILITY ~ including alars and/or trip func,tions.
.5 ?m -
u j (7) - Above 70% of RATED THERMAL POWER, verify that the total' steady-statek
; RCS flow rate as indicated by each CPC is less than or equal tofthef.
l g ac_tual RCS total flow rate determined by either using the reacto.r g rooiant pump differential pressure instrumentatiolor by caloris'etric > y calculationgand if necessary, adjust the CPC addressable consta.rt flow coefficients such that each CPC indicated f1 w is less thaa.ct equal to the actual flow rate. The flow meastrrement uncertainti may be included in the BERR1 ters in the CPC and is equal to or greeter
~ ~
-i i than 4%.
- 48) Above 70% of RATED THERMAL POWER, verify that.the. total steady-state 4 RCS flow rate as indicated by each CPC is less than or equal to $he actual RCS total flow rate.6etermined by either using the reactor coolant cumn dif ferentral-presure in<trumentation and t ltrasonic i'
flowmeteradjustedpumpcurvisgor,catorimetreccsiculation (9) - The monthly CHANNEL FhMCT[0NAL.TE5J shall include 4verificat on that j the correct values of.addryssable zonstant;s are instatted in c:ach ! OPERABLE CPC per Specificatiorf2.2.2. . p y - : : e ) (10) - At least once per 18 montM aAd foiloAng saintenancept adjustment i of the reactor trip breakers .the CHANNELfUNCTIONAL TST shall 1 include independent verif katten of the undervoltsge .gpd shunt trips.
=. .- 1 i casMb M -sn
__ -. Ku.-g.e . ,. . ,n s j
l
~
INSTRUMENTATION g I 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety features Actuation System (E5FAS) instrumentation ) channels and bypasses 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.
~
APPLI'CABILITY: As shown in Table'3.3-3. ACTION:
- a. With an ESFAS instrumentation channel trip setpoint less conservative c than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION j - requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip
;. Setpoint value.
; With an ESFAS instrumentation channel inoperable, take the ACTION 4-
- b.
shown in Table 3.3-3. - - i SURVEILLANCE REQUIREMENTS 1 4.3.2.1 Each ESFAS instrumentation channel shall be, demonstrated OPERABLE by !
~
the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and Ci4AFNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2. , 1 4.3.2.2 The log 5c for the bypasses shall be demonstrated OPERABLE during the ' at power CHANNEL FUNCTIONAL TEST ef channels affected by bypass operation. 1 -< The total bypass function shall be demonstrated OPERABLE at least once per "18 months during CHANNEL-CALIBRATION testing of each channel affected by bypass operation. , i
~
4.3.2.3 2-The ENGINEEIED SAFE Y FEATURES RESPONSE TIM [ of vach ESFAS function s shall be demonstrated to-be within the lisit at 1. east once per 18 months. . Each; test shall knclude at least ene' channel per function such that all channels l are testfd at least 'once every N times 18 mo'nths where N"is the total number J of redunciant channels, in a specific ESFAS function as shewn in the " Total No,
' ^
i
- of Channels"-Column of Table.3.3-3.
z ,
=
- i
. s- I. .'.
~ .
= =- .- .-'
E E- N
- u . . . y.
i I5 - l .. - tE. - CBSP&kN9%-STS e >-)(
.c. ,-
I 4 iO TABLE 3.3-3 ENGINEFRED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION [ i MINIMUM [ TOTAL NO. CHANNELS CllANNELS APPLICABLE t .{ - 0F CHANNELS TO TRIP _ OPERABLE MODES ACTION E5FA SYSTEM FUNCTIONAL UNIT {
% 1. SAFETY INJECTION (SIAS)
W Sensor / Trip Units l A. M 1. Containment Pressure - High 4 2 3 1,2,3,4 13*, 14*
- 2. Pressurizer Pressure - Low 4 2 3 1, 2, 3(a), 4 13*, 14*
B. E5FA Systm logic R* 6 1 3 1,2,3,4 17
- 1. Matrix L e ic
- 2. Initiation Logic 4(c) 2(d) 4 1,2,3,4 12 Og 3. Manual SIAS (Trip Buttons) 4(c) 2(d) 4 1,2,3,4 12 C. Automatic Actuation Logic 2 1 2 1,2,3,4 16
- 11. CONTAINMENT ISOLATION (CIAS)
A. Sensor / Trip Units
- 1. Containment Pressure - High 4 2 3 1,2,3 13*, 14*,
- 2. Pressurizer Pressure - Low 4 2 3 1,2,3(a) 13*, 14*
B. ESFA System logic
- 1. Matrix logic 6 1 3 1,2,3, 17 2, Initiation Logic 4(c) 2(d) 4 1,2,3,4 12
! 1
; TABLE 3.3-3 (Continued)
, to ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION t '
i NINIMUN TOTAL NO. CHANNELS CHANNELS APPLICABLE
! !I E5FA SYSTEM FUNCTIONAL. UNIT OF CHANNELS 10 TRIP OPERABLE MODES ACTION A
i g I1. CONTAINMENT ISOLATION (Continued)
- 3. Manual CIAS (Trip Buttons) 4(c) 2(d) 4 1,2,3,4 12 -
W
^^
...m. " ^;"' (? 'r """ m ) *(c) ...
ooi u C, Automatic Actuation Logic 2 1 2 1,2,3,4 16 g III. CONTAIMMENT SPRAY (CSAS) y A. Sensor / Trip Units k Containment Pressure -- S High - High 4 2 3 1,2,3 13*, 14*
- 8. ESFA System logic
- 1. Matrix logic 6 1 3 1,2,3 17
- 2. Initiation Logic 4(c) 2(d) 4 1,2.3,4 12
- 3. Manual CSAS (Trip Buttons) 4(c) 2(d) 4 1,2,3,4 12 C. automatic Actuation logic 2 1 2 1,2,3,4 , , ,
16 e.'h .
,. . * , L .
g
^
; y#\
i* ,' , 1 t 3 TABLE 3.3-3 (Continued) E
- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION r
i ! . MINIMUM APPLICABLE c g TOTAL NO. CHANNELS CHANNELS i
,g ESFA SYSTEM FUNCTIONAL UNIT 0F CHANNEM TD TRIP e OPERABLE MODES ACTION
[d IV. MIN STEAM LINE ISOLATION (MSIS) 2
~ A. Sensor / Trip Units .
) (
l " Steam Generator pressure - 2/ steam 3/ stems 13*, 14*
'g 1. 4/ steam 1, 2, 3(b), 4 Low , generator generator generator *
- 2. Steam Generator level '
4/ steam 2/ steam 3/ steam 1, 2, 3, 4 1,3*, 14* High generator generator generator
- 3. Containment Pressure - High 4 2 3 1,2,3,4 13*, 14*
Y K ESFA System Logic y G.
- 1. Matrix logic 6 1 3 1,2,3,4 17
- 2. Initiation Logic 4(c) 2(d) 4 1,2,3,4 12
- 3. Manual MSIS (Trip Buttons) 4(c) 2(d) 4 1,2,3,4 12 C. Automatic Actuation Logic 2 1 2 1,2,3,4 10 t
13
.M o
*"r1 D
==
a c. C3 N M _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ ___ . - _ - . - - - - . _ - . - . . . . . . _ - . . - . - . . - - - _ - . . - ~ . - _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _ _ -
\ TABLE 3.3-3 (CCntinued)
]
l ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i MINIMUM f TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACil0M
, O ESFA SYSTEM FUNCTIONAL UNIT V. RECIRCULATION (RAS)
A. Sensor / Trip Units
<lg Refueling Water Storage f Tank - Low 4 2 3 1,2,3 13*, 14*
$ 8. ESFA System logic 3 1, 2, 3 17
- 1. Itatrik Logic 6 1
- 2. Initiation Logic 4(C) 2(d) 4 1,2,3,4 12 Y
X 3. Manual RAS 4(C) 2(d) 4 1, 2, 3, 4 12 ( C. Automatic Actuation Logic 2 1 2 1,2,3,4 16 VI. FEEDWATER (SG-1)(AFAS-1) A. Sensor / Trip Units Steam Generator #1 Level - 1 1. Low 4 2 3 1,2,3 :3 C) 13*, 14*
- 2. Steam Generator A 4 2 3 1, 2, 3 Q 13*, 14*
Pressure - SG2 > SGI p,,
- e ,
r ,j
.. t C"
.S :,
e ' ABL' -3 (C:rilinoed) i ENGINEERED SAFEif FEATURES ACTUATION SYSTEM INSTPUMENTATION l MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE L ESFA SYSTEM FUNCTIONAL UNIT OF CH4NNELS TO TRIP OPERABLE PODES ACTION
, VI. AUXILIARY FEEDWATER (5G-1)(AFAS-1) (Continued)
B. ESFA System Logic ig M 1. Matrix Logic 6 1 3 1,2,3 17
'Q 2. Initiatlon Logic 4(c) 2(d) 4 1,2,3,4 12 /
M [ C.
- 3. Manual AFAS Automatic Actuation Logic VII. p FEEDWATER (SG-2)(AFAS-2)
A. Sensor / Trip Units
- 1. Steam Generator d'2 Level -
Low 4 4(c) 2 2(d) 2 1 4 2 3 1,2,3,4 1,2,3,4 1,2,3 15 lb 13*, 14*
- 2. Steam Generator a Pressure - SG1 > SG2 4 2 3 1,2,3 13*, 14*
- 8. ESFA System Logic
- 1. Matrix Logic 6 1 3 1,2,3 17
- 2. Initiation Logic 4(c) 2(d) 4 1,2,3,4 12
- 3. Manual AFAS 4(c) 2(d) 4 1,2,3,4 15 C. Automatic Actuation Logic 2 1 2 1,2,3,4 16
- 8 VIII. LO55 0F POWEk (LOV)
A. 4.16 kV Emergency Bus Under-voltage (toss of Voltage) ' '" _Q , nu --_" g z- -1,4-?_ n a
, 14' B. 4.16 kV Emergency Bus under-voltage (Degraded Voltage) -
" ,_ Sgg,, Q; Q Sg
. u. -
.u . _
J__ s- 1
=o C1 na , &
IX. CONTROL ROOM ESSENTIAL FILTRATION - 5k
' " ,%i-; 'an E
25
I TABLE 3.3-3 (Continued) hf$i[ h f,((' ((-[ TABLE NOTATIONS (a) In ICDES 3-6, the value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia. (b) In MODES 3-6, the value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator I pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached. (c) Four channels provided, arranged in a selective two-out-of-four configuration (i.e., one-out-of-two taken twice). (d) The proper twc-out-of-four combination.
~
- The provisions of Specification 3.0.4 are not applicable.
" A'r %e la" 4.1 d '::i n., : f "-" ' ~ "* '
- ACTION STATEMENTS ACTION 12 - 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 STANDBY within the r> ext 6 hours and in COLD SHUTDOWN within the following 30 hours.
j ACTION 13 - With the number of channels OPERABLE one less than the Total j Number raf Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within I hour. If the inoperable channel is C bypassed, the desirability of maintaining this channel in the E- *
- bypassed condition shall be reviewed in accordance with '
apecification 6.5.1.6.tl The channel shall be returned to , OPERABLE status no late than during the next COLD SHUTDOWN. Mb With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all (' associated functional units as listed below. i l Process Measurement Circuit Steam Generator Pressure Low (b t
- 1. Steam Generator Pressure -
- iow Steam Generator Level 1-Low (ESF) ;
Steam Generator Level 2-Low (ESF)
- 2. Steam Generator Level Steam Generator Level - Low (RPS) '
(Wide Range) Steam Generator Level 1-Low (EST) Steam Generator Level 2-Low (ESF) k d I6 3f4 3 g
TABLE 3.3-3 (Centinu d) f,[ h hih b 4 ACTION STATEM' ENTS (Continued) ACTION 14 - With the number of channels OPERABLE one less than the Minimum ' Channels OPERABLE, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied: .i , l
- a. Verify that one of the inoperable channels has been bypassed }
and place the other inoperable channel in the tripped 1 condition within I hour.
- b. All functional units affected by the bypassed / tripped <
channel shall also be placed in the bypassed / tripped l condition as listed below: ' Process Measurement Circuit Functional Unit Bypassed / Tripped
- 1. Steam Generator Pressure -
Low Steam Generator Pressure - Low (b) Steam Generator Level 1 - Low (ESF) Steam Generator Level 2 - Low (ESF)
- 2. Steam Generator Level - Low Steam Generator Level - Low ~(RPS)
(Wide Range) Steam Generator Level 1 - Low (ESF} Steam Generator Level 2 - Low (ESFj STARTUP and/or POWER OPERATION may continue until the performance
- i. of the next required CHANNEL FUNCTIONAL TEST. Subsequent J~
STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERtBLE status and the provisions of ACTION 14 are satisfied.
- ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE I
status within 48 hours or be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours. l j ACTION 16 - With the number of OPERABLE channels one less than the Total ! 4 Number of Channels, be in at least NOT STANDBY within 6 hours i and in at least NOT SHUTDOWN witMn the following 6 hours; a however, one channel may be bypassed for up to I hour for i -e surveillance testing provided the other channel is OPERABLE.
- ACTION 17 - With the number of OPERABLE channels one less than the Mininum
- Number of Channels, restore the inoperable channel to OPERABLE i statos within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
E h hII .. =4 =- mb ow =r. W 5. -.
- affar #
_!.. T- _;-:; n., ,_-' : '-- T +-r """ I: i " ^" . l . . 1, ... . . ! . ... y . . .
- I: I 20 I : : ' ' ' ' "I ' ^ -
^# ; . .. .1 * . '. i I I I t i .. . . ^
- lu wyw i . . - .. st E -~ N O . 1; in C i
- Pff: ^, ~. . . ? ' ^.: .. . .u . ' . . . . . _ . - ' ' . . ;; " """^* "i t h i n t h*
~ -,
l '* " C .., " t_... 4 4 i TSSM:Mvs-srs 3,,3
,g l
-_ I
1 **- g 1
'O TABLE 3.3-4 b
E h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIOM TRIP VALUES
& E T *5 U,
TRIP SETPOINT All0WABLE VALUES ESTA SYSTEM FUNCTIONAL UNIT r I. SAFETY INJECTION (51AS) 5 O, z 5 4 I A. Sensor / Trip Units ,y
- 1. Containment Pressure - High i S psig 1 @ psig 8o l
1 $ psia III 3 I 2. Pressurizer Pressure - Low 1 M psia III _ B. ESFA System logic Mot Applicable Not Applicable gz Actuation Systems Not Applicable Not Applicable g C. II. CONTAINMENT ISOLATION (CIAS) gRm A. Sensor / Trip Units Qj l ! w 1. Containment Pressure - High i $ psig i S psig %o' A 2. Pressurizer Pressure - Low 1 O psia III 3 @ psia III Y ~~'A System logic Not Applicable Not Applicable B. Actuation Systems Not Applicable Not Applicable C. g p III. CONTAINMENT SPRAY (CSAS) i [ A. Sensor / Trip Units Containment Pressure High - High S O psig i $ psig , B. ESFA System logic Not Applicable Not Applicable Q l Actuation Systems Not Applicable Not Applicable gm C. IV. MAIN STEAM LINE ISOLATION (MSIS) ::j' i Sensor / Trip Units ( A. psla I3) l $ psla I3) 1$ -
- 1. Steam Generator Pressure - Low
- 2. Steam Generator Level - High 1 M NM(2) 5 g yg(2)
Containment Pressure - liigh i 4 psig i @ psig- j} 3. ESFA System logic Not Applicable Not Applicable u ! B. Actuation Systems Not Applicable Not Applicabic C.
9 i l , A ' TABLE 3.3-4 (Cen. .oed) w ENGINEERED SAFETY FEATURES ACTU75 ION SYS'.1 INSTRUMENTATION TRIP VALUES 2 ESFA SYSTEM FUNCTIONAL UNIT TRIP VALUES $ ALLOWABLE VALUES l j ,, V. RECIRCULATION (RAS) O D A. Sensor / Trip Units OO 1 Refueling Water Storage Tank - Low 1 9 % of Span M1 % of Span 1 $ l l j; B. ESFA System logic Not Appilcable Not Appilcable Oy I
- C. Actuation System Not Applicable Not Applicable I
h , VI. I
"'* *:"TT N FEEDWATE (SG-1)(4FAS-1) rn N g A. Sensor / Trip Units >,
h 1. 2. Steam Generator #1 Level - Low Steam Generator a Pressure - 16 WR 1 W psid I4} 1 @ WR 1 $ psid I4} fh 9%
.'> SG2 > SGI Z O
$ B. ESFA System logic Not App 11 cable Not Applicable u y C. Actuation Systems Not Applicable Not Applicable VII. "'~3:"Ti FEEDWATER (SG-2)( FAS-2) n h A. Sensor / Trip Units
- 1. Steam Generator #2 Level - Low 3 M WR I*} 1 M WR(*} __
- 2. Steam Generator a Pressure - 1 $ psid i W psid .o I SG1 > SG2 ESFA System Logic
- j:
B. Not Applicable Not Appilcable Actuation Systems Q C. Not Applicable Not App 11 cable no VIII. LOSS OF POWER 55:8 A. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) 2 1 N volts 1 & volts
'd B. 4.16 kV Emergency Bus Undervoltage
- M to @ volts 6 to 6 volts (Degraded Voltage) with a 35-second maximum time delay with a 35-second ,
maximum time delay Q-es
.IX. CONTROL ROOM ESSENTIAL FILTRATION 1 M pCi/cc i M pCi/cc
TABLE 3.3-4 (Continued) '((yh{h [h i TABLE NOTATIONS (1) In MODES 3-6, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pres-surizer pressure and this value is maintaintd at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip say be-manually bypassed below 400 psia; bypass shali be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia. (2) % of the distance between steam generator upper a.* lower level narrow range instrument nozzles. (3) In MODES 3-6, value scy be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached. , (4) % of the distance between steam generator upper and lower leves wide
,- range instrument nozzles.
M - % A k 5 -e rt $ S Ptt2.
- BJma e6%Ec6m 4
l l
-r 1 i
I i 4 (@M@-N5fs-sTs 3,43.g2+
I P1 03:& RSS COPY l l ENGINEERED SAFETY FEATURES RESPONSE TINEC RESPONSE TIME IN SECONDS i INITIATING SIGNAL AND FUNCTION
- 1. Manual
* \
a. l SIAC Safety Injection (ECCS) Not Applicable
- l l
Not Applicable %$@) Containment Isolation Containment Purge Valve Isolation Not Applicable (h@ ) 1
- b. CSAS
' Containment Spray Not Applicable (bop) a'
- c. CIAS Containment Isolatio Not Applicable (6Df) i
.am
- d. M515 Main Steam Isolation Not Applicable i ,
i )i e. RAS Containment Sump Recirculation Not Applicable (@op) g f
- f. FAS
'r '
Q; Feedwa;er Pu ps Not Applicab'e (60b) A i S C C.A5 h { Y[ W / i l 4 l ? l
- l
.....- **b] gf4 3 g-1
- _. . . -=
FABLE 3.3-5 (Continued) ?.132 & 'W'BY CD-)Y i . J., i ENGINEERED SArETY FEATURES RESPONSE TIMES 1 2NITI ATING SIGNAL AND FUNCTION RESPONSE TIME IN SECOND5 l
- 2. Pressuri:er Pressure - Low -
I
- a. Safety Injection (HPSI) 1 $ * /S* * .
- b. Safety Injection (LPSI) 1 $'/9 2
- c. Containment Isolation g, 4f/ #dr j
-:: ! :-t :* d u-ouen, valvet -
^^'l_'
] 2. v 5er LIAs ace ..s. ....o [ ? " '* " ; ---
- 3. Containment Pressure - High l l
Safety Injection (HPSI) i 9*/t** j a. u. b Safety Injection (LPSI) $ 4 * /p" Og n s Q. Q l' g 94 / 4f(
~
- c. Containment Isolation g3
. mC 4
,7 r u c i-..... s
" - "' _ . . . ;. ^
- i-~..n, 13 %. c -
7 ;-- xQ O
. . :- [***** j 2mc ,
- d. Main Steam Isolation 4 4/ 4t M- 8* l
- Z :g r 2
w i ! ' ) 5I . 1...su nb$v ,j ; ( '- l _ :. " :: ::- -: n : -'; . N - r 'r" z-m
- a. Z i
- e. Containment Spray Pump i F */d *
- OO j
- m*
0 l 8
- 4. Containment Pressure - High-High <h=
n. j
- a. Containment Spray < S * /S*
- m2 7
- 5. Steam Generator Pressure - Low
- a. Main Steam Isolation f A/ H 1
u-,e ... . . ~ ~ , , , , _ - . , . . :
- 3 ,
- 2. w(T c *r+ -1 nia , 9- { ~" -C' P " ;
l r ] 6. Refueling Water Tank - Low
- a. Containment Sump Recirculation i g */g * * !
l l 7. Steam Generator Level - Low ! l a. ater '"-t;-- O.:.4 - i g*/g*
- i A
- y a.. m i- y r.. h + rfre f ,-g --
- mu m u
- -w -
i a
-..m ,,r.,- m.,, .
I 374 3 g', 1 i CASSAMO-4!5 S - 9 75 l L-______ --
f -- __ TABLE 3.3-5 (Continued) (ly 2 pDn-e., lLeC..gjf hijfj ENGINEEREDSAFETYFEATURESRESPONSETIMES[ ~ % -- __ _J l INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS l
- 8. Steam Generator Level - High -
YM i
- a. Main Steam Isolation f m:: ..._ u ..n. ~n" r? W(" ::tri:. "':. $ I i,sC " _
i 9. Steam Generat:,r AP-High-Coincident With Steam Generator Level Low hm4/LF, W
- a. '. . Feedwater Isolation iF/F*
from the Ruptured Steam Generator
- 10. Control Room Essential Filtration Actuation 5 $ /1 405'* l
- 11. 4.16kVEmergencyBusUndervoltag8 (Degraded Voltage) g Loss of Power 90% system voltage $@ -
4.16 kV Ertergency Bus Undervoltage (loss of Voltage)2 ; l 12. Loss of Power iM THIS PAGE OPEN PENDING RECEIPT OF l INFORMATION FROM THE APPLICANT i y 1 l ) TABLE NOTATIONS l 1 - l '
- Diesel generator starting and sequence loading delays included. Response
- time limit includes movement of valves and attainment of pump or blower
, discharge pressure. l "" Diesel generator starting delays not included. Offsite power available.
; Response time limit includes movement of valves and attainment of pump or bicwer discharge pressure.
_ N'T! ;;' . ;; t:: t: d : + '1.1;;;; .,.;- H g ~ ~d' t i ... , .;1... u ru u ' -- :t4 .. ;:Mitic,; te ' S.E'o r re d
- i See A(gd4.M$ SA1C b hDCLYvNKs),
E M v n t.a. o f f M C 6 0 f )
=:
'e : : 1 u4 3M*7 6659MSb-Nss$-9r5 1
'" e, g TABLE t. 3-2 ENGINEERED SAFETY SEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMFMTS M'M $
{} CHANNEL M0005 FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE I i ESFA SYSTEM FUNCTIONAL UNIT CHECK CALIBRATION TEST 15 REQUIRED
- 1. SAFETY INJECTION (SIAS) y A. Sensor /Trfp Un!ts M 1. Containment Pre:Sure - High 5 R M 1, 2, 3, 4
- 2. Pressurizer Pressure - Low 5 R i' M 1,2,3,4 B. ESFA System logic
- 1. Matrix logic NA NA M 1, 2, 3, 4
- 2. Initiation Logic NA NA M 1, 2, 3, 4
- 3. Manual SIAS NA NA M 1, 2, 3, 4 _
C. Automatic Actuation logic NA NA M(1) (2) (3) 1, 2, 3, 4 % c:) II. CONTAINMENT ISOLATION (CIAS) M A. Sensor / Trip Units
- 1. Containment Pressure - High 5 R M 1, 2, 3 r
- 2. Pressurizer Pressure - Low 5 R M 1, 2, 3 h
- 8. ESFA System logic c:2
- 1. Matrix logic hts hA M 1,2,3/4 N L
- 2. Initiation logic G '8 M 1, 2, 3, 4
- 3. Manual CIAS M NA M J. 3, 4 um -3 es;5 -
w ;;;; - u - - _ i42, a I
( ... 1 { .
\
NJ: p TABLE 4.3-2 (Continued) c , g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS g7
= %? :
l M ' CHANNEL MODES FOR WHICH CHANNEL FUNCTIONAL SURVEILLANCE hc ' ESFA SYSTEM FUNClIONAL UNIT CHANNEL CHECK CAllBRATION TEST 15 REQUIRED _
- 11. CONTAINMENT ISOLATION (Continued) g C. Automatic Actuation logic NA NA M(1) (2) (3) 1, 2, 3, 4 III. COMTAINMENT SPRAY (CSAS)
A. *;ensor/ Trip Units R 1. Containment Pressure --
** High - High 5 R M 1, 2, 3 w
B. ESFA System logic
- 1. Matrix logic NA NA M 1, 2, 3, 4 ,
- 2. Initiation Logic NA NA M 1,2,3,4 ]
2
- 3. Manual CSAS NA NA M 1,2,3,4 .,j
.O C. Automatic Actuation Logic NA NA M(1) (2) (3) 1, 2, 3, 4
, , j r
. 8y
_.U ,
I
% \
TABLE 4.3-2 (Continued) + k x; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE :tEV IREMENTS CHANNEL M00E5 FOR WHICH
. CHANNEL CHANNEL FUNCIl0NAL SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT CHECK CAllBRATION TEST 15 REQUIRED g {
-' IV. MAIN STEAM LINE ISOLATION (MSIS) w A. Sensor / Trip Units k
g
- 1. Steam Generator Pressure -
low 5 R M 1,2,3,4
- 2. Steam Generator Level - High 5 R M 1,2,3,4 g 3. Containment Pressure - High 5 R M 1, 2, 3, 4
- 8. ESFA System logic
- 1. Matrix logic NA NA M 1, 2, 3, 4
- 2. Initiation logic MA NA M 1,2,3,4
- 3. Manual M515 NA NA M 1,2,3,4 3
c:3
'v1 C. Automatic Acutation Logic NA NA M(1) (2) (3) 1, 2. 3, 4 #
- xs E
8
e s Q j, t TABLE 4.3-2 (Continued) f , ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS N ; ;
' CHANNEL MODES FOR WHICH k CHANNEL FUNCTIONAL SURVEILLANCE CHANNEL CHECK CAllBRATION TEST 15 REQUIRED ESFA SYSTEM FUNCTIONAL UNIT 5
] V. RECIRCULATION (RAS)
A. Sensor / Trip Units
%1 Refueling Water Storage Tank - Low 5 R M 1,2,3 B. ESFA System logic Matrix logic NA NA M 1, 2, 3, 4 1.
NA NA M 1, 2, 3, 4
- 2. Initiation logic w
- 3. Manual RAS NA NA M 1,2,3,4
- C. Automatic Acutation Logic NA NA M(1) (2) (3) 1, 2, 3, 4 VI. E A1ER (SG-1)( -1)
A. Sensor / Trip Units @'.
'f .,e .s
- i. Steam Generator #1 Level -
Low 5 R M 1,2,3 m. r j,
- 2. Steam Generator f1 a Pressure SG2 > SG1 5 R M 1, 2,,3 , { '
F3
. .A! !
_ . _ _ .____m _ _ _ _ _ . . _ . _ _ _ . _ . _ . . . _.._______._.m_____________ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
p , hh
- TABLE 4.3 c (Continued) 1 O- ,
\
{ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH D>; CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE D( r ESFA SYSTEM FUNCTIONAL UNIT g CHECK _ CALIBRATION TEST 15 REQUIRED 1
, ; VI. FEEDWATER (SG-1)(F AS .) (Continued)
I! ! B. ESFA System logic D* *
- 1. Matrix logic NA NA M 1, 2, 3, 4 NA -NA M 1, 2, 3, 4
- 2. Initiation logic Manual AFA5 NA NA M 1, 2, 3, 4 3.
C. Automatic Actuation logic NA NA M(1) (2) (3) 1, 2, 3, 4 (~ VII. FEEDWATER (SG-2)( AS-2)
$ s' !
I J
~,
A. Sensor / Trip Units ( e *.
- 1. Steam Generator #2 Level - ':" <
.i ;
R* tme 5 R M 1, 2, 3
- 2. Steam Generator a Presstare SG1 > SG2 5 R M 1,2,3 l (, ;
- 8. ESFA System logic ;!
- 1. Matrix t.ogic NA NA M 1,2,3,4 ; '_ . j
- 2. Initiation logic NA NA M 1, 2, 3, 4 h"U
- 3. Manual AFAS v4A NA M 1,2,3,4 C. Automatic Actuation Logic NA NA M(1) (2) (3) 1, 2, 3, 4 VIII. LOSS OF POWER (LOV) hiL= Ws:P _ .
Ses R&qiXs s&z - Qs a ed o la ) , C _~?,'-^ t
% YO Q
TABLE 4.3-2 (Continued) TABLE NOTATION _ _ __ . _.m _ _ 4 4 (1) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS. ! (2) Testing of automatic actuation logic shall include energiza. tion / deenergization of each initiation relay and verification of proper operation of each initiation relay. - (3) A subgroup relay test shall be performed which shall include tM energization/deenergization of each subgroup relay and verification , of the OPERABILITY of each subgroup relay. "r';" ' -
+-d 6 ' -- :; i
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~
1 'C'LII!Ch (('!!"[0 Cf CT si ita,cs rd TOW Is TRAIN A TRAIN B ESF ACTUATION ESF ACTUATIO ' UNCTION DEVICE FUNCTION DEVIC ' ;- SIA A K10B SIA5 B 8 SIAS K409 SIAS B K409 1 j CIAS A K202 CIAS B K204 CIAS A K204 CIAS B K205 CIAS A K205 CSAS > K304 C5A5 A "4 M5f3 K305 , M515 A K3 5B K404 M515 A K404 AFA5 IB K113 i AFA5 1A K211 AFAS IB K211 AFAS 2A , K112 AFAS 2B K112 In the case of the followin elay which are tested during power operation,
- r- - - '---e a' :g
.m. . . m ,; x ,,,..,e , ; ,,m m . .s u m, ;
a bypassed or etc., which ill not preclu he relay from being tested but
, will not actuate t ocked out equipment a iated with the relay:
SIAS A 01 SIAS B K301 , SIAS A K410 51A5 B SIAS A K412 CIAS B K203
- CIAS A K203 CIAS B K210
- CIA K210 RAS B K104
- R A K104 RAS B K312 e i A5 A K312 RAS B K405 l 4 k RAS A K405 <
, AFAS 1A K113 ; x raw . wu. . 3/4 3-G.s9m-Mys-str 33
kahrn J PRODF & REEW COPY INSTRUMENTATION 3/4.3.3.1 3/4.3.3 M3NITORING INSTRUMENTATION Mb Trt4%Mk d-l RADIATION M3NITORING INSTRUMENTATION , 4 LIMITING CONDITION FOR OPERATION : SM 1 ~ i A ., , N . .. ., . .. . . . .... wn wring instrumentation cnanne a snown in ble 3.3-6 shall be OPERABLE with
- heir alarm / trip setpoints within t l s -ified limits.
APPLIC ITY: As shown in Table 3.3-6. ACTION: ;
. a. With a r 'ation monitoring channel alarm /tr setpoint exceeding
, the value s n in Table 3.3-6, adjust th etpoint to within the limit within ours or declare the cha el inoperable. l - ;
- b. With the number of nnels GPERAB one less than the Minimum j Channels OPERABLE req ement, e the ACTION shown in Table 3.3-6.
- c. The provisions of Specific ns 3.0.3 and 3.0.4 are not applicable. . . .
d i j SURVEILLANCE REQUIRE TS ' , m i l 4.3.3.1 ch radiation monitoring instrumentation channel shall 4 demo ated O RRABLE by the performance of the CHANNEL CHECK, CHAN CA 'ATION, anc CHANNEL FUNCTIONAL TEST operations for the MODES and a he ec. ncies shown Ir. Table 4.3-3. o 1 1 4 a T y W
.-me- ., . , - - - - - - - - _ . . . . .
INSTROMENTATXON W . INCORE DETECTORS , e r> .h . LIMITING CONDITION FOR OPERATION 3.3.3.2 The incore detection system shall be OPERABLE with:
- a. At least 75% of all incore detector locations, and l
- b. A minimum of two quadrant symmetric incore detector locations per corte quadrant. i An OPERABLE incore detector location shall consist of a fuel assembly contain;ng a fixed detector string with a minimum of four OPERABLE rhodium detectors or an OPERASLE movable incore detector capable of mapping the location.
APPLICABILITY: When the incore detection system is used for monitoring:
- a. AZIMUTHAL POWER TILT,
- b. Radial Peaking Factors, _
- c. Local Power Density,
,-- l
- d. DNB Margin.
ACTION:
- a. With the incora detection system inoperable, do not use the system for the above applicable monitoring or calibration functions. )
l . b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS y T4.3.3.2 The incore detection system shall be demonstrated OPERABLE: N
- a. By performance of a CHANNEL CHECK within 24 hours prior to its use if the system has just been returned to OPERABLE status or if 7 days or more have elasped since last use and at least once per 7 days thereafter when required.for monitoring the AZIMUTHAL POWER TILT, radial peaking factors, local power density or DNB margin ,
I
- b. At least once per 18 months by performance of a CHANNEL CALIBRATION l operation which exempts the neutron detectors but includes all i electronic components. The fixed incere neutron detectors shall be !
calibrated prior to installation in the reactor core. ; i Y 1,e. .:^^: $Z T==A
^
N, w,7 3/4 3-
! l l
~'
INSTRUMENTATION AC %EXeLoMML -
;Q.
- i SEISMIC, INSTRUMENTATION _ m- m _ _ _ . _
l h LIMITING CONDITION FOR OPERATION k@ SMQ611_1 WC %6M[g l l n-
----____..._:......___...,._a- ; ; ,,, , ,
j PERABLE. APPt PABILITY: At all times. l
- ACTION:
7 i l a. With e or more seismic monitoring instruments i erable for more than 3 s, prepare and submit a Special Repo to the Commission i i purs t. ant pecification 6.3.2 within the 10 days outlining , t tP,e cause of e malfunction and the plan or restoring the ' l g instrument (s) t PERABLE status.
- b. The provisions of Spe 'fications .3 and 3.0.4 are not applicable. I
~ ~
SURVEILLANCE REQUIREMENTS i UN - - A l % 5.3.3.3.1 Each of the abov eismic monitoring in userts shall be j demonstrated OPERABLE by .e performance of the CHAN CHECK, CHANNEL CALIBRATION and CHANNE UNCTIONAL TEST operations at th requencies shown in l
; Table 4.3-4. l
'i i 4.3.3.3.2 Each the above seismic monitoring instruments actu d during a seismic even greater than or equal to 0.02g) shall have a CHANNEL LIBRATION performed hin 5. days. Data shall be retrieved from actuated instr ts and j cnalyze o determine the magnitude of the vibratory ground motion. A tal i Repor hall be prepared and submitted to the Commission pursuant to Speci i ca n 6.9.2 within 10 days describing the magnitude, frequency spectrue, and i i, ultant effect upon facility featur== M netant to safety. N l l I i i l
}
l ll l 1 a h l 1
I t 3 ._. ! -YWin PROM & RBEW CO?Y i M - V4.3.3.3 Sakwt Wns~trtgIMMdrh ! ~id 3.5 I Eimumi.Wnfog:Mnw- ! %+b ) I w4.3-er same ng s' sumsnea. %&g ' i GrumeT3) p . ! A . e.3.4 h w gcaJLNe k$7 h h l l % Pt s .3 - P , M e b w b y4 k n t b g ' l i TvJtvurrev%Ce 7 e O &O ( f T h hs-sth Eb j 7 1 1
- - , _ - . - . . - _ _ _ . _ _ _ . ..~ _- .._ -. -
,\ INSTRUMENTATION y
--.} -
[ { [,(; .
- REMOTE SHUTDOWN SYSTEM 'UJ ATThwra fT"p,T10fD f
LIMITING CONDITION FOR OPERATION 1 3.3.3.5 The remote shutdown system di.......nt ;;it:i;:, ;:r . *ontrols,and I monitoring instrumentation channels shown in Table 3.3-9 shall ve OPERABLE. APPLICABILITY: 1 MODES 1[2{CN%k $ ACTION: i
- a. With the number of OPERABLE remote shutdown monitoring channels 4
less than required by Table 3.3-9, restore the inoperable channel (s)J to OPERABLE status within 7 days, or be in HOT _ next 12 hours. witron tne g g h ]l i b. Withoneormoreremoteskutdownsystem------'-'- - i eur control circuit noperable, restore the inoperable ni^_r ':} ' 1 circuit (s) to OPERABLE status :: in _ 7...: fur: c': ;: ;: ' ::i l 3
" :r.in f.? ' t'r', ,'fr
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fr C':::- ::' n ^.:..;. .. ,.-.. Jithi n 7 days , or be in HOT j . M within the next 12 hours. l l c. The provisions of Specification 3.0.4 are not applicable. l
% l NO 1
l SURVEILLANCE REQUIREMENTS 4.3.3.5 The)t$ note p$utdown fystamghall be demonstrated /g,h: I a. By performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6 for each remote shutdown monitoring instnmentation channel. j b. By operation of each remote shutdown system ft:_ _--- ' : 'tr' n - W control circuit including the actuated components at least once per 18 months. l I
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INSTRUMENTATION . nn - 'g
}_((ydte ?6 "I'~i ,.Y, W'Ji i }
POST-ACCIDENT MONITORING INSTRUMENTATION
~
LIMITING CONDITION FOR OPERATION
+
3.3.3.6 The post-accident monitoring instrumentation channels shown in ' Table 3.3-30 shall be OPERABLE. APPLICASILITY: M3 DES I and 3. , ACTION: ,
- a. With one or more accident monitoring instrumentation channels inoperable, take the action shown in Table 3.3-30. ,
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.6 Each post-accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIERA110N operations at the frequencies shown in Table 4.3-7. l N,
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h .[h h TABLE 4.;7-3.3.-{O ACTION STAlEMENTS l ACTION 29 - With the number of OPERABLE Channels ont less than the Required
- Number of Channels in Table 3.3-10, either restore the Inoperable Channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours.
- ACTION 30 - With the number of OPERABLE Channels one less than the Minimum l Channels OPERABLE in Table 3.3-10, either restore the Inoperable Channel (s) to OPERABLE status within 48 hours or be in at least l HOT SHUTDOWN within the next 12 hours.
l %"" 0; ":^2 0.. _ i.. vi G R^.LE C;mmma , vi. ^ ...i^
^L t-I Number of Channels, either restore the s E status l j s if repai without shutting down or ,
j . w u n. ci ication 6.9.2 within lowing the event out-lining the action taken, the cause of the ino -
+ and the
^~ - ' : ?: S f:: ^
n ^^:^. L '. ......._
- .. . .. ... , ^
- -. .,:t:
j ~ ACTION [- With the number of OPERABLE Channels one less than the Minimum 1 Channels OPERABLE in Table 3.3-10, either restore the inoperable k channel (s) to OPERABLE status within 48 hours if repairs are feasible without shutting down or: ' l Initiate an alternate method of monitoring the : : :-- 4p 1. ($)_ p .. J ......^..j; l 2. Prepare and submit a Special Report to the Commission pur- l j suant to Specification 6.9.2 within 30 days following the ! i event outlining the action taken, the cause of the inopera-
- bility and the plans and schedule for restoring the system j to OPERABLE status; and
- 3. Restore the system to OPERABLE status at the next scheduled t refueling. t J
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l l b $. 5555&5 Ebb 3g"$ ? E aaasasasaneasianus \ l 3/43-X
g .J a STRUMENTATION FIRE DETECTION RUM TATION EOth EMLuErJi MOUToOg i n LIMITING LONDITION FOR OPERATIQM SLM6\l14f0CE E%MMQg l
. M r , ?" " e it -ti:- ' : t --t:ti_. " , mu irii ,
2.,. ., .m. j etectic.n zone shown in Table 3.3-11 shall be OPERABLE. 4 PLICABILITY: #dhenever equipment protected by the fire detect 4on i erument , required to be OPERABLE. AC ]ON:
/ l i
j . With any, but not mcre than one-half the total i any fire zone l Function X fire detection instrument shown in ble 3.3-11 inoperable, ; j restore the inoperable instrument (s) to OPE. LE status within I 14 days or within the next 1 hour establis a fire watch patrol to t l_ inspect the zone (s) wit 5 the inoperable .strument(s) at least once j per hour, enless the instrument (s) is 1 ated inside the containment, i hen inspect that' containment zone at east once per 8 hours or l lI n'itor the containment air tempera re at least once per hour at . ! locations listed in Specificat n 4.6.1.5. ! t l f
- b. With more than one-half of the unction X fire detection instruments '
% in an fire zone shown in Tab 3.3-11 inoperable, or with any f Functi Y fire detection i truments shown in Table 3.3-11 inoperable, ;
i cr uith ny two or more a acent fire detection instruMnts shown in ; Table 3. 11 inoperable, ithin 1 hour establish a fire watch patrol l' to inspect the zone (s) ith the inoperable instrument (s) at least once per he , unless he instrument (s) is located inside the o containment, hen i ett that containment zone at least once per ! l 8 hours or mon or he containment air temperature at least once ) _g per hour at the cations listed in Specification 4.6.1.5.
- c. The provision of cifications 3.0.3 and 3.0.4 are not applicable.
l l -Y , i btLANCE ItEDUI , ENTS m I j i
- I i
4.3.3.7.1 En of the above required fire tection instruments which are : l : accessible ring plant operation shall be ok nstrated OPERABLE at least once L TEST. Fire detectors l per 6 mont by performance of a ChWHEL FUNCTI : which are et accessible during plant operation s 1 be demonstrated OPERABLE l ach COLD SHUTDOWN l J by the rformance of a CHANNEL FUNCTIONAL TEST duri exceed 3g 24 hours unless performed in the previous 6 eo
- s. l l
ssociated l 4.3 .7.2 The NFPA Standard 720 supervised circuits supervis l ! h the detecter alares of each of the above regwired fire detec
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@sTRVmENTATtW j 3/4' . 3 . 4 - O r ~ 2 -
O FTURBINE OVERSFEED PROTECTION A LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS . 3/4.3.4 Turbine Overspeed Protection See Applicant's SAR. 9 d
- e 4
34
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1 i 3/4.4. REACTOR COOLANT SYSTEM s_ 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION g* g g] Q STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION _ . , 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each , loop shall be in operation. l APPLICABILITY: MODES 1 and 2.* ACTION: With less than the above required reactor coolant pumps in operation, be in at least HOT STANDBY within I hour.
~
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\
SURVf1LLANCE REQUIREMENTS _ 4.4.1.1 The above required reactor c',olant loops shall be verified to be in 3 operation and circulating reactor coolant at least once per 12 hours.
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- . - m REACTOR COOLANT SYSTEM 3-r g OQg hj-[
t,- HOT STANDBY LIMITING CONDITION FOR OPERATION i j 3.4.1.2 The reactor coolant loops listed below shall be OPERABLE and at least cne of these reactor coolant loops shall be in operation." , 1
- a. Reactor Coolant Loop 1 and its associated steam generator and at
' 1 east one associated reactor coolant pump.
- b. Reactor Coolant Loop 2 end its associated steam generator and at least one associated reactor coolant pump.
): APPLICABILITY: MODE 3. 1 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. With no reactor coolant loop in operation, suspend .*11 operations l \q involving a reduction in boron concentration of the 3eactor Coolant j System and immediately initiate corrective action to return the i required reactor coolant loop to operation.
i i SURVEILLANCE REQUIREMENTS l i i 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying
; correct breaker alignments and indicated power availability.
j : p A4.1.1. 2 At least one reactor coolant loop shall be verified to be in j operation and circulating reactor coolant at least once per 12 hours. 1 l 4.4.1.2.3 The required steam generator (s) 9 hall be determined OPERABLE l verifying the secondary side water level to be > 25% indicated wide range ] level at least once per 12 hours.
\
*All reactor coolant pumps may be deenergized for up to I hour provided i (1) no operations are permitted that would cause dilution of the Reactor l
Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature. - E 6 3/4 4-2
REACTOR COOLANT SYSTEM
. a 5suT00. PROD:& REV!M COPY LIMITING CONDITION FOR OPERATION 1 _
i : { 3.4.1.3 At least two of the loop (s)/ train (s) listed below shall be'0PERAPLE .
- and at least one reactor coolant and/or chutdown cooling loops shall be in - i eperat!on." ;
- a. Reactor Coolant Loop 1 and its associated steam generator and at i least one associated reactor coolant pump,** l
- b. Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump,"* t I c. Shutdown Coolir.g Train [ ;
! d. ShutdownCoolingTrain[ )(
~
- APPLICABILITY
- MODE 4.
j~ ACTION: M a. With less than the above required reactor coolant and/or shutdown cooling loops OPERABLE, issnediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; ; ' if the remaining OPERABLE loop is a shutdown cooling loop, be in
- COLP SHuTDOW within 24 hours.
- b. With no reactor coolant or shutdown cooling loop in operation, suspend
, all operations involving a reduction in boron concentration of the l Reactor Coolant System and immediately initiate corrective action to, i l' return the required coolant loop to operation. 1 .f ; 1 t
*All reactor coolant pumps and shutdown cooling pumps may be deenergized i for up to I hour provided (1) no operations a u 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.
p4144t
**A reactor coolant pump shall not be started with one or more of the Leactor Coolant Svstem cold leg temperatures less than or equal to.3l4*F during y M coolcown. 7 'F during heatup, unless the secondary water temperature )(
(sateration erature corresponding to steam generator pressure) of each
. team generator is less than 100'F above each of the Reactor Coolant System cold leg temperatures.
Nhb %R 3/4 4-3
REACTOR COOLANT SYSTEM ,
.. . r n HOT SHUTDOWN i {C
- f h [ES dO O,,j . l SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), if not in operation, shall be i
determined to be 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 the secondary side water level to be > 25% indicated wide range level at least once per 12 hours. 1 4.4.1.3.3 At least one reactor coolant or shutdown cooling loop shall be
! verified to be in operation and circulating reactor coolant at a flow rate greater than or equal to 4000 gpm at least once per hours.
)- 1 .- D
- c I
1 rp - _ e ~ 1 h M M M ST-5TS
*th. ... un6. . 3/4 4 4
a REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED =h0BF&R. T W@ i ! LIMITING CONDITION FOR OPERATION , F l 3.4.1.4.1 At least one shutdown cooling loop shall be OPERABLE and in [ operation *, and either: .i 1 ,
- s. One additional shutdown cooling loop shall be OPERABLE #, or - ;
- b. The secondary side water level of at least two steam generators
- shall be gr*,t;r than 25% indicated wide range level. l
- APPLICABILITY
- MODE 5 with reactor coolant loops filledN . :
ACTION: l I a. With less than the above required loops OPERABLE or with less than , ! the required steam generator level, immediately initiate corrective ! l action to retuin the required loops to OPERABLE status or to restore ! l the required leyel as soon as possible.
- b. With no shutdown cooling loop in operation, suspend all operations q involving a reduction in boron concentration of the Reactor Coolant *
! System and immediately initiate corrective action to return the , , . required shutdown cooling loop to operation. ly . ! SURVEILLANCE REQUIREMENTS , t f l, 4.4.1.4.1.1 The secondary side water level of both steam generators when l N 1 required shall be determined to be within limits at least once per 12 hours. ; l 4.4.1.4.1.2 At least one shutdown cooling loop shall be detemined to be in ! operation and circulating reactor coolant at a flow rate of greater than or l equal to 4000 gpa at least once per 12 hours. j i
#Dne shutdown cooling loop may be inoperable for up to 2 hours for i
' + ,__ surveillance testing provided the other shutdown cooling loop is OPERABLE
- and in operation.
44t4% ! l NA reacter coolant pump shall not be started with one or more[of the Reactor : Coolant System cold leg temperatures less than or equal to 316'F during M coot 6own, or@'F during haatup, unless the secondary water temperature l saturation temperature corresponding to staan generator pressure) of each i steam generator is less than 100*F above each of the Reactor Coolant System i cold leg temperatures. , 4
*The shutdown cooling pump may be doenergized for up to I hour provided ;
! (1) no operations are pemitted that would cause dilution of the Reactor !
- Coolant System boron concentration, and (2) core outlet tamperature is maintained at least 10'F below saturation tamperature.
Okb ' SM i _ 3/4 4-5 i {
I F REACTOR COOLANT SYSTEM _ COLD SHUTDOWN - LOOPS NOT FILLED , PR00F & RBf3 COM ,
)
LIMITING CONDITION FOR OPERATION
. i 3.4.1.4.2 Two shutdown cooling loops shall be OPERABLE # and at least one ,
shutdown cooling loop shall be in operation." APPLICABILITY: MODE 5 with reactor coolant loops not filled. , 1 ACTION-~
- a. With less than the above required loops OPERABLE, immediately initiate corrective action return the required loops to OPERABLE status as soon as possible, d
- b. With no shutdown cooling loop in operation, suspend all operations involving a reduction in b een concentration of the Reactor Coolant i System and immediately initiate corrective action to return the l required shutdown cooling loop to operation ~.
Y SURVEILLANCE REQUIREMENTS 4.4.1.4.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpa at least once per 12 hours. l 71 ,; 805E shutdown cooling loop may be inoperable for up to 2 hours for surveillante testing provided the other shutdown cooling loop is OPERABLE and in operation.
* \
The shutdown cooling pump may be de-energized 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.
)
43E!!5 mL ERarf.ii >-A555-S L T5 3/4 4-6 , l l l
REACTOR COOLANT SYSTEM
~# ' . - U-.),\ '
3/4.4.2 SAFETY VALVES k nn. . - h h so,ii.E l SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer code safety valve shall be 0PERABLE with a lift setting of 2500 psia 1 1%.* _ APPLICABILITY: K)DE 4.
. ACTION:
- a. With no pressurizer code safety valve OPERABLE, imediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation.
- b. The provisions of Specification 3.0.4 may be suspended for up to 12 hours for entering into and during operation in M3DE 4 for purposes of setting the pressurizer code safety valves under ambient (HOT) conditions provided a preliminary cold setting was made prior to heatup.
7 SURVEILLANCE REQUIREMENTS _ 4 4.4.2.1 No additi'nal Surveillance Requirements other than those required by Specification 4.0.5. hQ
'The lift setting pressure shall correspond to ambient conditions of the valve i at nominal operating temperature and pressure.
lr'
@SSAM$0-OSCRr-STS
- FE L .'ETCE unii 1 3/4 4-7
REACTOR COOLANT SYSTEM OPERATING l [ h fgh J i i. LIMITING CONDITION FOR OPERATION
- 3.t.2.2 All pressurizer code safety valves shall be OPERABLE with a lift sottitig of 2 LOO psia 1 1%.*
APii t C*BI'. ?TY. MODES 1, 2, and 3. ACTION: With one pressurize. 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 HOT SHtJTDOWN within the following 6 hours with the shutdown cooling system suction line relief valves aligned to provide everpressure protection for the Reactor Coolant System.
. I i- ..
SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional Surveillance Requirements other than those required by i Specification 4.0.5. I it. ," ~~ l 1 i GThe lift setting pressure shall correspond to ambient conditions of the valve at nominal operr. ting temperatare and pressure. l l l 4 ~ I GJRSS/4ttfkA)S99-STc _**'a "E " E - L* F 1 # 3/4 4-8
I REACTOR COOLAWT SYSTEM 3/4.4.3 PRESSURIZER E , ,, _ { ? tj Q% PRESSURIZER i
/ :
\
LIMITING CONDITION FOR OPERATION E 3.4.3.1 The pressurizer shall be OPERABLE with: 2 9'-
" dy " =*
- water w ai ' ;~"-- n.. . . ; e:' t r 2N . .; : . . . . ^
?... ':20 :2i: '*}-d j _-= r ;t;. , . ..e
. . . : . . ; l ;' ! ; ; . L ,.. . . ._ .,..' ^ ;;t c.; : .. ^.;; ? : d
...; .. ;.. 6 6.s w vue, vi y. ... :... h;.t: : ::::M : :" i: ' ;_
" :.... '..G . ., ;..c, ;;; w _
y : ;; 'c;: 0u 1r w - ::n ;,..,,,, ,,,_ ;,,,; , y.,,,,, APPLICABILITY: MODES 1, 2, and 3. ACTION: I- a. With only one group of the above required pressurizer hreaters ' OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HDT STANDBY within the ncxi 6 hours and
- in HOT SHUTDOWN within the following 6 hours.
. y Y b. With the pressurizer otherwise inoperable, restore the pressurizer to e OPERABLE status within I hour, or be in at least HOT STANDBY with the
- reactor trip breakers open within 6 hours and in NOT SHUTDOWN wft.hin the following 6 hours.
SURVEILLANCE REQUIREMENTS i 4.4.3.1.1 The pressurizer water volume shall be determined to be within its l limits at least once per 12 hours. 1
.T #
4 1.2 The capacity of the above required groups of pressurizer heaters l shall be verified to, be at least 150 W at least once per 92 days. [ 4.4.3.1.3 The emergency power supply for the pressurizer heaters shall be demonstrated DPERABLE at least once per 18 months by verifying that on an ] Engineered Safety Features Actuation test signal concurrent with a loss-of-offsite power: I
- a. The pressurizer heaters are automatically shed from the amergency a power sources, and i
- b. The pressurizer heaters can be reconn:<cted to their respective buses manually from the control roos.
s W V l.
9 E t N PR00F & ?HH COPY
~
- a. A steady state water voluse less than or equal to 585 indicated level ft.),an (gd cu.ft)butgraaterthan275indicatedlevel.(44gcu.
c,% ; i
- b. At east two groups of pressurizer heaters capable of being powered f IE buses each taving a nominal capacity of at least 150 kw.
l n 3- > 9 i e y
)
REACTOR COOLANT SYSTEM kiCh[ h [{d,0.E bhi AUXILIARY SPRAY
}
LIMITING CONDITION FOR OPERATION 3.4.3.2 Both auxiliary spray valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With only one of the above required auxiliary spray valves OPERABLE, restore both valves 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 none of the above required auxiliary spray valves OPERABLE, restore at least one valve to OPERABLE status within the next
_ 6 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. x SURVEILLANCE REQUIREMENTS
\-
4.4.3.2.1 The auxiliary spray valves shall be verified to have power available to each valve every 24 hours. 4.4.3.2.2 The auxiliary spray valves shall be cycled at least once per 18 months. T- - ) i i NNi@N4TS sf4 4 13
i REACTOR COOLANT SYSTEM 3/4.4.4 STEAM GENERATORS N00F & EU UM i LIMITING CONDITION FOR OPERATION
.~ i 3.4.4 Each steam generotor shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4. i ACTION: ) With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasting T cold ab ve 2109. S*.iRVEILLANCE REQUIREMENTS i 4.4.4.0 Each steam generator shall be demonstrat2d OPERABLE by performance of ' the following aupented inservice inspection program. 3 4.4.4.1 Steam Generator Sample Selection and Inspection - Each steam generator 4- shall be oeterminec OPERABLE during shutcoen by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. l l 4.4.4.2 Steam Generator Tube Sample 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 perfonned at the ; l frequencies specified in Specification 4.4.4.3 and the inspected tubes shall be verified acceptable per the acceptance crittria of Specification 4.4.4.4. The tubes selected for each inservice inspection shall include at least 3% of the total neber of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except: l 4_ m. Where experience in similar plants with similar water chemistry
" indicates critical areas to be inspected, then a+. least 50% of the tubes inspected shall be from these critical areas.
I b. The first sample of tubes selected for each inservice inspection l
- (subsequent to the preservice inspection) of each steam generator shall include
l 1 c46essuro.-ass-sy3 3/4 4-11
- ~ ~
7 .. REACTORCOOLANTSYSTq P",,f u *L9. P,a'- }3[ -, SURVEILLANCE REQUIREMENTS (Continued) j g i l
- 1. All nonplugged tubes that previously had detectable wall penetrations (greater than 20%). ?
l
- 2. Tubes in those areas where experience has indicated potential ,
problems. i
- 3. A tube inspection (pursuant to Specification 4.4.4.4a.8.) shall ;
be perfctmed 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 tuce shall be selected and subjected to a tube inspection. ,
- c. The tubes selected as the second and third samples (if required by I
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. l (i*
- 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: ! Category Inspection Results i 4 C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes 1 are defective. ! 76 ,_ C-2 One or more tubes, but not more than 1% of the . total tubes inspected are defective, or between
~
- l j 5% ard 10% of the total tubes inspected are degraded tubes. l 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 j to be included in the above percentage calculations. l l
'}
- CG%%A40'N%S-ST5
- . . . . -- u< <-u
REACTQR COOLANT SYSTEMS 1 SURVEILLANCE REQUIRJ4ENTS (Continued)
, l 4.4.4.3 Inspection Frequencies - The above required inservice inspect, ions of 4 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 calender months of initial crit-
- icality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months af ter the previous inspection. If two consecutive inspections following service under AVT conditions, wat including the preservice inspection, .
result in all inspection results fallieg into the C-1 category or if two consecutive inspections demonstrate that p m iously observed degradation has not continued and no addittoral degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months. i
- 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
_- into Category C-3, the inspection frequency shall be increased to at
- least once per 20 months. The increase in inspection frequency (t shall apply until the subsequent inspections satisfy the criteria of .
" Specification 4.4.4.3a.; the interval may then be extended to a . maximum of once per 40 months.
- c. Additional, unscheduled inservice inspections shall be perfomed on t each steam generator in accordance with the first sample inspection i
specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1. . Primary-to-secondery tubes leaks (not including leaks originetir.g from tree-to-tube sheet welds) in excess of the limits of Specification 3.4.5.2.
- 2. A seismic occurrence greater than the Operating Basis e
, 4~
- w Earthquake. ,
! 3. A loss-of-coolant accident requiring actuation of the engineered safeguards.
?
! 4. A main steam line or feedwater line break. l
- l cessuro-pss-m
* * ' T ^.C - i" ? Y 3/4 4-13
1 REACTOR COOLANT SYSTEM 1"' er
- p1gi & PKEn ,
~
SURVEILLANCE REQUIREMENTS (Continued) 4.4.4.4 Acceptar e Criteria i
- a. As used in this Specification .
- 1. Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specificatians. Eddy-current testing indications below 20% of the nainal tube wall thickness, if detectable, may be considered ;
i as imperfections. j
- 2. Decradation scans a servievinduced cracking, wastage, wear, or
- general corrosion occurring on either inside or outside of a tube.
- 1. Degraded Tube means a tube containing imperfections greater than or equU to 20% of the nominal wall thickness caused by 1 degradation.
1
. 4. E Degradation means the percentage of the tube wall thickness 1
- P - affected or removed by degradation. )
1
- 5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.
- 6. Pluacing Limit means the imperfection depth at or beyond which l the ttne shall be removed from service and is equal to 40Kr-- % i of the nominal tube wall thickness.
- 7. Unserviceable de w ibes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Er.rthquake, a loss-of-coolant
~
accident, or a steam line or feedwater line break as specified 6, in 4.4.4.3c. , above.
- 8. Tube Inspection means an inspection of the steam generator tube j from the point of entry (hot leg side) completely around the ,
U-band to the top support of the cold leg. l
- 9. Preservice Inspection Jeans an inspection of the full length of j
, each tube in each steam generator performed by eddy current ; techniques prior to service to establish a baseline 1 l 1 d l YN t,. h %YL .Vt\ Q4C,N1, Q N iJnrs d ) MSNM955-575 m$4 bw[tAA2 A 1. l?_1 b Or l
' ~
REACTOR COOLANT SYSTEM -
~
-~~~"'.-
,,v._. .
i p .,
~
t W - =
" ~~___.
SURVEILLANCE REQUIREMENTS (Continued) l l I condition of the tubing. This inspection was perfomed prior to the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections. i
- 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.
4.4.4.5 Reports
- 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 6.9.2.
I b. The complete results of the steam generator tube inservice inspection s. shall be submitted to the Commission in a Special Report pursuant to p Specification 6.9.2 within 12 months following completion of the inspection. This Special Report shall include:
- 1. Number and extent of tubes inspected. j
- 2. Location and percent of wall-thickness penetration for each indication of an imperfection.
- 3. Identification cf tubes plugged. i
- 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 6.9.2 within 30 days and prior to JK. ~ _
~
resumption of plant operation and shall provide a descrig ; ion of investigations conducted to detemine cause of the tabe degradatier, and corrective seasures taken to prevent recurrence. l l f
-m 44"&Mf0-Ms%-yrs,
.- m. . 3/4 4-15 i
i i l
TABLE 4.4-1 % h hi J MINIMUM NUMBER OF STEM GENERATCRJ 10 BE l I INSPECTED DURING INSERVICE INSPECTION The inservice inspection may be limited to one steam generator on.s rotating schedule encompassing 6% of the tubes _if the results of the first' or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steam generators seey be found to be more severe than those in other steam generators. Under such circumstances the sample sequence chall he modified to inspect the mos*, severe conditions. i p A Q i a [ 5
- s w
f 1 l T
""/_: fran: :: ; 3/4 4-16
W r . .- s
s, TABLE 4.4-2
'h N i
\
f6 STEAltl GENER ATOR TUBE INFECTION r 3RD SAfWIPLE fMSPECTION 2ND SAARPLE INSPECTION 1ST SARSPLE INWECTION g Action Required Result Action Required Action Required Result
'D Semple Site Result ii N. A. N. A. N. A.
C-1 None M. A. D eft A minimum of S Tthes per S. G. None N. A. N. A. < C-2 Plug defective tubes C-1
@I i
C-1 None end W additionel Plug defective tubes 2S tubes in this S. G. C-2 and inspect additW C-2 Plug defective tubes 45 tubes m this S. G. Perform action for C-3 C-3 result of first semple Perform action for - C-3 C-3 result of fF:st N. A. N. A. Y sempk i C-3 Inspect sH tubes in AHotM U this S. G., plug de. S. G.s are None N. A- N.A-factive tubes and C-1 inspect 25 tubes in Some S. G.s Perform action for N. A. N. A, each other S. G. C-2 but no C-2 result of second _ edditional se m ple
- Footifiestion to NRC S. G. are .
3 pursuant to $50.72 C-3 ' -
.)
(bil2l of 10 CFR AMtionel inspect all tubes in j I, 1, Port 50 S. G. is C-3 each S G. and plug
- defective tubes. I '.
' Notification to NRC N. A. N. A. ! .- 3
~
pursuant to 550.72 .; d {kL:j
~
(b)(2) of 10 CF R Part 50
*7 i
N Where N is the twmber of steem generators in the unit, and n is the number of steem generators insp ,
']y n during en inspection l- - - - - - - - _ _ _ _ __ _ __ . _ . _ _ . . _ . _ _ . _ _ _ ._ _
REACTOR COOLANT SYSTEM
' ~ ' - ~ ~
~
3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE . ,r~- > 1 LEAKAGE DETECTION SYSTEMS I
/
! LIMITING CONDITION FOR OPERATION ,
- "'E2 . '.% ni h :.,; ".. m . CM.... C,;ttr '?" y st artinn svsters shall e OPERABLE:
- a. A containment atmosphere particulate radioactivity monitorin ystem,
. The containment sump le. vel and flow monitoring system, ;
- c. he containment atmo'ephere gaseous radioactivity no toring system.
. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: I i-
- l With only two of the ab required leakage de ction systems OPERABLE, j-_
operation may continue for to 30 days pr ded grab samples of the ' y containment atmosphere are o ined and yzed at least once per 24 hours
- 5' when the required gaseous and/o arti ate radioactivity monitoring system st HOT STANDBY withi1 the next 6 hours
! g is inoperable; otherwise, be in a ng 30 hours. and in COLD SHUTDOWN within the f l SURVEILLANCE REQUIREMENTS
' ~
4.4.5.1 The 1 age detection systems shall be demonstr d OPERABLE by:
- a. ontainment atmosphere gaseous and particulate monito
- systea performance of CHANNEL CHECK, CHANNEL CALIBRATI d CHANNEL !
70 FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, l i
\.b Containment sump level and flow monitoring system performance of CHANNFt FilTRDATTrut at immet anr e noe lg ocnths.
l e e l 3cEm spbns 2 DL S&Ntudb SM
-y I
Cess 7 ^. ; amn. /bi%NSS$ MT1 - STS 3/4 4-18 i
i 1 l l . 4 REACTOR COOLANT SYSTEM - OPERATIONAL LEAKAGE b a I ( l LIMITING CONDITION FOR OPERATION 1 l 3.4.5.2 Reactor Coolant System leakage shall be limited to: ,
- s. No PRESSURE BOUNDARY LEAKAGE, . l
- b. I gpm UNIDENTIFIED LEAKAGE. y (9&)
l c. I gpe total primary-to-secondary leakage through all steam generators, F j and M gallons per day through any one steam generator, j ]
- d. 10 ppm IDENTIFIED LEAKAGE from the Reactor Coolant System, and j l e. I gpm leakage at a Reactor Coolant System pressure of 2250 1 20 psia !
from any Reactor Coolant System pressure isolation valve specified I in Table 3.4-1. E_ APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: I- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor
. Coolant System pressure isolation valves, reduce the leakage rate to wn.ninklimitswithin4hoursorbeinatleastNOTSTANDBYwithin the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
2
- c. With any Reactor Coolant System pressure isolation valve leakage j greater than the above limit, isolate the high pressure portion of tne affected system from the low pressure portion within 4 hours by j use of at least one closed manual or deactivated automatic valve, 4 or be in at least NOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
)
- d. With RCS leakage alarmed and confirmed in a flow path with no flow
<4 - rate indicators, commence an RCS water inventory balance within I hour to determine the leak rate. I SURVEILLANCE REQUIREMENTS i j 4.4.5.2.1 Reactor Coolant System leakages shall be demonstrated to be within j sach of the above limits by: ! a. Monitoring the containment atmosphere gaseous and particulate
- radioactivity monitor at least once per 12 hours.
I b. Monitoring the containment susp inventory and discharge at least l once per 12 hours. i 1 i
? 5N CK h $hR
~ON s/4 4-19
" W8 REACTOR COOLANT SYSTEM X [ ${[ bbrq
.w I
SURVEILLANCE REQUIREMENTS (Contirued) ) 1 E
- c. Performance of a Reactor I;oolant System water inventory balance at l least once per 72 hours.
- d. Monitoring the reactor head flange leakoff system at least once per I 24 hours. >
4.4.5.2.2 Each Reactor Coolant System pressure isolation valve specified in ; Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within ; its limit
- a. At least once per 18 months, i b.4 Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been ;
performed in the previous 9 months, j i ; i - c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, !
- d. Within 24 hours following valve actuation due to automatic or 3
manual action or flow through the valve, ,
\
_ . .f- "e _' ^f. : - ?2_n_t' _n ': '. 'a.~. ' ,M a ecl -- r r - _ "_ : _ . : , b . 6 .~,. ' . ,
)
5 i The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 - 3 or 4. i l 1 4. - l 4 l A.2 m f '"'
^4 ... . ---
'-'h" T_
' ' ? * * *"!. h *. *:!,* L; '. ;,hi:f ' -._....,...w
;..' .m.
- " ::: T;".":..'T...' --....
.__......:-m...... ....-_ _ _ _ _ . . _ _ . .
i lM~5,pAstBO-AhiS$-g "t'^ . M i 'f:7 . 3/4 4-20 1 I l
i . P30r & REVlEW COPY TABLE 3.4-1 i !t REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES 1 j VALVE DESCRIPTION
~
! iGT l', as a A t.rin.K
- 1) % tnnD 'n "".. ^. .un -
- 2) SW ,
- 3) SM innp 9a or/et r"::<
- 4) M ... ., n ,. o 4 nun
- 5) M . te" 1". ',;: ;nas l i
- 6) g LRT 2 ^ n nton - .
2." ::: ; nan l 7) @ ' ^ ^ *
- 8) @ '.^^P -^ ^^i ncen i
- 9) .425t97 ;""? 1^. ^,; nind i vnton - ;
l _
- 10) 4j>f!l8V - LZT 2 3; :::."I:" :::::: i
_,,,e,e..._ .- ,, --- . . . ~ i -
.., .-,. or tn .. . . n. , . . . . . -
{_ ,e e,e_,,.-
.<,4
--- --. --a4 .-...nre rurn-6 v- .
l !. - 12'
- - - LOT I .:P ' :.~ itan nhancuLAUUh i.nr.Lw l
- \-
,,s
'. '. n,!:
xo ^ ; wvr 4 m m, ___..,,,,,,,,..,,mm,
..... . no . . _ _ _ - - - - . . . - -
r y ,. ,, l
.._ ._ me, .
,,, ... _ .....,........,n.wu mm ..~m 1' . . , , .v . au Luvr z nr
). ,,s ,.- _
- - , , . .. . . . . y e m, . . - - . --- - . . . ~
vaa, w., . . ..._ __ A nww . wn ...m l ' -- emn, w t w irt s
. 173 sia.,run, a,p -
.. . s .-".,.n .. .. .. __ _ __ _ _
, sn eve-nvgram a > ano , cuirrrvw enm tw v ent ivrew _
,os , ---- -,_m, ,mn,,w 3 m my,,m e , ,. _. ,. . .- .. r-
. . wvr 2.....-... _____ _
enne,ne , cat iTTM f ens r , e m e- , e ", , , -
~r c uitr m a, i
i l ~I N w y :.:.:..:. _. ,,n,.e.. :.. a , = ..;.. _ . c:--m--e-J.; i . 4 me. a mim. s
%co-. -
r-e-a , _ _ _ . , fl. stage es great than 1.0 but 1 s than equal 5.0 - are i the ! consid d accepta e if the atest me ured r?t has no excee I rate termined previou st by a amount t redt s the rgin i be en measu leakage te and maximus - rsiss e rat of 5.0 l by 0% or gre er.
- 2. akage ra s greater han 1.0 but les than o equal 5.0 re i
i considere unaccep e if the atest me red ra excee d the r e ! detersi d by the evious t by an unt t reduce the es n I betwe measured eakage r and the inum ruissib rate 5.0 by 5 or grea r. j 3. L age rate greater n 5.0 gps consi red un captab . k
# Sea Ag*flids SAR j M 5/4 4-21
\ MSAgro-AlSSS-STS
,,---T - av-- r _ -
* ----------,--r
,_~_
.5. 1h f,,b REACTOR COOLANT SYSTEM .g{}[hh g 3/4.4.6 CHEMISTRY ,
LIMITING CONDITION FOR OPERATION i l 3.4.6 The Reactor Coolant System chemistry shall be maintained within the ~ limits specified in Table 3.4-2. APPLICABILITY: At all times. i ACTION: MODES 1, 2, 3, and 4: , s. With any one or scre chemistry parameter in excess of its Steady , State Limit but within its Transient 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.
- b. With any one or more chemistry parameter in excess of its Transient M- Limit, be in at least HOT STANDBY within 6 hours and in COLD ,
SHUTDOWN within the ic11owing 30 hours. At All Other Times: 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 less than or equal to 500 psia, if applicable, and 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 prior to increasing the pressurizer pressure above 500 psia or prior to q proceeding to MDE 4. j i
~SUD ILLANCE REQUIREMENTS l
4.4.6 The Reactor Coolant System chmistry shall be determined to be within the limits by analysis of those pe Aters at the frequencies specified in Table 4.4-3. 1
#M,@. -NGSS-575
. - . _ _ _ . . _ 3/4 4 22 a i
TABLE 3.4-2 REACTOR COOLANT SYSTEM
-Pil00F & REVIEW COP CHEMISTRY ,
STEADY STATE TRANSIENT LIMIT LIMIT ; PARAMETER
$ 0.10 ppm 5 1.00 ppm DISSOLVED OXYGEN CHLORIDE $ 0.15 ppm 5 1.50 ppm FLUORIDE i 0.10 ppm 5 1.00 ppm
" Limit not applicable with Teold less than or equal to 250'F.
O1- .o ( d' h 3 cws:*ro
"?
- "E'_^: C .:ssem
.. 3/4 4-23
- ;- it j fp ,n - ng ^J eC(
h f (1 a b {-a. {.I - TABLE 4.4-3 _ REACTOR COOLANT SYSTEM i CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS , [ SAMPLE AND ' t PARAMETER ANALYSIS FREQUENCY l DISSOLVED OXYGEN At least once per 72 hours , t CHLORIDE At least once per 72 hours
- FLUORIDE At least once per 72 hours
)
."Not required with Tcold less than or equal t 250*F if ~
I. 'd y
< l l
l l 1 1 e I r 'S' 3/4 4-24
REACTQR COOLANT sys1EM 3/4 4 7 SPECIFIC ACTIVITY gg* g )) gy
- LIMITING CONDITION FOR OPERATION 4
3.4.7 Se specific activity of the primary coolant shall be limited to: , a. Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and
- b. Less than or equal to 100d microcuries/ gram.
J APPLICABILITY: MOLES 1, 2, 3, 4, and 5. j ACTION: j MODES 1, 2, and 3*: u
- a. With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the allowable i limit (below and to the left of the line) shown on Figure 3.4-1, j operation may continue for up to 48 hours provided that the
- i cumulative operating time under these circumstances does not exceed jj \;- 800 hours in any consecutive 12 month period. With the total ..
cumulative operating time at a primary coolant specific activity i greater thar 1.0 microcurie / gram DOSE EQUIVALENT I-131 exceeding ; l 500 hours in any consecutive 6 month period, prepare and submit a
< Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the number of hours above this limit. The i provisions of Specification 3.0.4 are not applicable.
- b. With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line i shown on Figure 3.4-1, be in at least NOT STANDBY with Tcold less i than 500'F within 6 hours.
1 T- c. With the specific activity of the primary coolant greater than j 4 100d microcuries/ gram, be in at least HOT STANDBY with Teold I than 500'T within 6 hours. I l With Teold greater than or equal to 500*F. 1 1 1 J 1 I i GM .MW5 3f,4.,3
1 I I F?.gBFE'nG'ViC07i l l REACTOR COOLANT SYSTEM _, i I
)
1 LIMITING CONDITION FOR OPERATION (Continued) I ACTION: (Continued) :
- i i
l MODES 1, 2, 3, 4, and 5: , d. With the specific activity of the primary coolant greater than , j . I afcrocurie/ gram DOSE EQUIVALENT I-131 or greater than l f I1004 microcuries/ gram, perfom the sampling and analysis require- i f ments of item 4.(a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. A Special Report shall be orecared and submitted to the Commission pursuant to Speci-4 " ^ - *- ^ "'---+^- ' '--- f ication
" r1:7 P.; 9. .Q~ o"", ;n. "..m".
6.;.1. 5 - C":', C- : F. ':- nnx * :-9, :- f
- !" : ' , .t ~ d : nt 5 M . : M . n . . .-2 , '.' . . ': . ; ' n : ^: ;; ' . t. . , E . .
I M............m. O.". ." E This report shall contain the results N- of the specific activity analyses together with the following l infomation:
- 1. Reactor power history starting 48 hours prior to the first if sample in which the limit was exceeded, l' l l
- 2. Fuel burnup by core region, '
i , 3. Clean-up flow history starting 48 hours prior to the first T j j sample in which the limit was exceeded. / ; t j 4. History of degassing operation, if any, starting 48 hours j prior to the first sample in which the limit was exceeded, and i The time duration when the specific activity of the primary
~
5. I coolant exceeded 1 microcurie / gram DCSE EQUIVALENT I-131. i j *[ - ! 'N j SURVEftLANCE REQUIREMENTS i 4.4.7 The specific activity of the primary coolant shall be detersined to be within the limits by performance of the sampling and analysis program of i Table 4.4-4. 1 l I 4 ! I 4 ! u l GF55/gne%A56-5T5 -- - 3f4 4 26 i i ! 4 I l l
t l4 b \ TABLE 4.4-4 h PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN %Ai!CH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED 8
- 1. Gross Activity Determination At least once per 72 hours 1, 2, 3, 4
- 2. Isotopic Analysis for DOSE 1 per 14 days 1 EQUIVALENT I-131 Concentration V 3. Redlochemical for E Determination 1 per 6 months
- 1
- 4. Isotopic Analysis for Iodine (a) Once per 4 hours, 1#, 2#, 3#, 4#, 5#
Including I-131, 1-133, and I-135 whenever the specific activity exceeds ~ 1.0 pCi/ gram, DOSE R
- EQUIVALENT l-131 '
i t or 100/E pC1/ gram, and O (b) One sample between 2 and 6 hours following 1, 2, 3 m a THERMAL POWER y change exceeding 15% ., of the RATED THERMAL POWER within a 1-hour ' period. One sagle is sufficient if plant has Q gone through a SHUTDOWN or if transient is Q coglete in 6 hours.
-o<
__ L
,, L
# Until the specific activity of the primary coolant system is restored within its limits.
- Sample 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.
]
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. ... .,.. i .. ,,,i , . , i
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- x. ... 6 6 . 4 .
, i.
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a - p g . . . , . . . ... p
. g , .. . . ... ...
- g . . ., . . . ...
w .. , . .. , . . . .
~ 200 * * * * *
. g .
'UNACCEPTAB LE
. , . .x .... 6 . , i OPERATION . ..,
u . .
*. \ . ** . . , ' ,* .... ..
g , , ... . ... 1 ... ,e, . i.., . . . ,
! . . . .... . . i g ... . r . , . .6 v. . ..
. . x. i . .. . . ,i g . . i . . . . . . . . ,. n . . . . . . . . . . p .s g gg
- . ... . 1 . . . . . .... . . , '
- q . . . . .
. 1 e . . . . . i . i4
. , ,... 1.., . .. , ,. . ..
.... ....
- 6 . ,4 , i wi, .6 . .... . i s l . ,; , .i .... . ., xi , , . . i . . . .
. ,.,, , i. .. ...
iii . . . . t- -
.. . . .. , ,... .y . .., .. i . .
- a
- . . .. ...> ,.. t .i. . i i . . . .
..i . . . . . . .
. . . .i , 6 . . . , . .. .x . . . . . i I g 100 4 . . . .. , , , . . ... , ,,, g.., . ,
, . .. .... . . . ... ii., s.. . . . . . ,
, . i ,e . . . 6 .6 , .. 6 1. . . . i
. , .i . . . . .. .
, . . . . . is. .. , . .
.... .... , . . * .. . .s i .... . . .
j l . . . . . . . . i... . . e gi . . . . 5 . EcuTis'LE : . !!.: :
. . . ,. . . . . . I '. k ..
5, OPERATION
> . . . 3. . ..., .i . .... i i i . .. . ... i
] g s 3 6 .
...e e t 6 .
4 . ii.
,i l q g . . . 4 i . 6 . .
. . . . , . l . .t ,t .
T, . . 1 1 . . ,, i . . , i . . . . . .,. - ..., , , . .
~NR
! . , . . . . . . . . . 6 . . . . . . .. .
, . . .., , i ; . . , . i i i .,i. . .
i i !
- 5 .. . . . +.- . . . -> i . .* - . ' + ..
l l 0 d i 20 30 40 50 80 70 30 90 100 ) a ! PERCENT OF RATED THERMAL POWER ; i d l FIGURE 3.4-1 l DOSE EQUIVALENT I-131 PRIERY COOLANT SPECIFIC ACTIVITY LIMIT YERSUS PERCENT OF RATED THERPut POWER WITH THE PRIE RY COOLANT SPECIFIC -- ACTIVITY > 1.0 pCUGRAM 00SE EQUIVALEKT I-131 N 3/4 4-28 C-E5*.#.2-MiES -575 l
1 l l REACTOR COOLANT SYSTEM . i 3/4. 4. 8 FRE55URE/ TEMPERATURE LIMITS l~, W. .
- l. REACTOR COOLANT SYSTEM / ... -
=>
) L_'. - : LIMITING CONDITION FOR OPERATION l s i j 3.4.8.1 The Reactor Coolant System (except the pressurizer) temperature and i pressure shall be limited in accordance with the limit lines shown on i Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak j and hydrostatic testing with: l 10 0 "; id k ,t;;;:- g a. A maximum heatup rate of )C"F " ^^^-
. -: th 4
ct. ; ';.. th:......
;r ;;m.,;?uor t; 0;ger m , ,...hour.
- y. . .my.
itt :,th:..: =;;'t "Cf :n,.;,eh;.. ; y.e 7 . - m . m ,
, ; . _7_ _ . , ;m. __ . y. . r . ;
t 86 0 ' " l
-- b. A saximum cooldown rate of 3C'F per hour,_it' 'C: .... .. .. v...t.:;
h.: ;:... .. . , .' 1: F *r '0*" ;:- t: r c't' ^CE r M ':; t r;.,. L., . m .. ^....- 200"" hi S:: t': : :;r:' t: !!!"", ;.r.: 1 . l 200"' 7.. ;~...m ^;; ;;'t hg ^.. ,...;... ,,..t. ;i.: '20*". _
- c. A saximum temperature change of 10'F in any 1-hour period during l f.- inservice hydrostatic and leak testing operations g g g Q ,, Q l W
APPLICABILITY: At all times
- h ACTION:
I With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perfom an engineering evaluation tu l determine the effects of the out-of-limit condition on the structural
- integrity of the Reactor Coolant System; detemine that the Reactor Coolant 1 System remains acceptable for continued operations or be in at least NOT '
j STANDBY within the next 6 hours and reduce the RCSg T ,y and pressure to less i -than 210*F and 500 psia, respectively, within the following 30 hours. l l N l i SURVEILLANCE REQUIREMENTS
~~
l l l l 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be detemined to be within the limits at least once per 30 minutes during system i heatup, cooldown, and inser< ice leak and hydrostatic testing operations. l 4.4.8.1.2 The reactor vessel acterial irradiation surveillance specimens
- shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR Part 50 Appendix H in accordance with the l
schedule in Table 4.4-5. The results of these examinations shall be used to ! update Figure 3.4-2. I ! "See special Test Exception 3.10.5. : , N 3/4 4-29 l (.4%S AR. 6'b-Mus- sq;, - 1 1
FIGURE 3.4-2 i [g[ h I o. 0 YRS. j i_..._3==:it.!(1) 13c oF,1500 PSI ! 2:.9 & " '" . ~ ' - ~ ~ ! l I:_ f (2) 2058 F,1450 PT -i --- F,500 ,A . 3000 _..__...j____(3) ! --:-=.. _ __ ___. . . s _ _ (4) ggoF,47 IA I ~ '. . 27:i'. r :;..._- _'.'.: ...5.. ,5_ __.. 5) ( 157'F, PSIA !.~4 .. , l PSIA I ~.=;r.p5;;.;j- = ..
-- . sig:. L :=_=:(6) 194,F(7) 264, ,p-
. .- 1 ,500
... J~PSIA l
,500 PSI A t' l
- .:: . . . =. . .. . ... .. ._: ..r.=.___ .=. . . .: .- . ,I = :==_. ; (S )
- r. ,
F,500 PSIA l
- - - ~-2::li:E-i.:-bli.- .. ; T 22"(9) .:-/s.
= :-$ , l l g . . . . _ : : = - - a nc c- :: .:: . ..: . c i:. -
t-
= ..: . . . - = . _ = . . _
j
=r..:_..._..___--
u.
..._.....r.-
. :- . := f.. ._ , . _ . : .=.
~
- n_;.
g- _ * ,_ _.
,r~
------=~-) ~=' ':C = }
. ..: . .. := -.:.__ p. ..: i
*w . . . . _ . . . . . ..
- : :- w " ;. ..:iF
.?i -
- 1
+
=
7 _..- . . . . . . . . -i ..
- .. = .:
4 .:. : .y , = .:. - . ~_ ::.:f. . ... = - g _ s. : r.
. --. . . .. . . r
- 4. . - -- := , g .-- . r -- - '"
$ .,,;;y: . t.=
g a i. W - n: p: u-.. C;;==_
~
i . . . . . . . . . . . s._.. _. . ,- d_ . . T'Hr Zw - .." i ;:. . g -- .>:. - -= w - h :-.. g . :.g:..::-- :Eu;- . .
.w t i a .
g r-- ,w ..
- {i. .:::.
t -
- -- c ~ g .
a !_f..' 8:;- '... .-- = w
- 1500 -
^ [(1) -
P :- :-'
~
< ..e- .
. (2) - - . = . . - _ . .:: .._-- . .
) M -- -
- r. .- : -- :;- .- .
.. . i j
j
-. .:-.).u/j:i.fi.:.~
'.."='."4 '- M p;; -. ._ir:.i , ~ ~
l i Y . ."$M _IE..r.: a bi58.~" ~
.. i. i ~ f
) 3n00 f.. F ~-i)[ * - - - - . d!+f fi;P :- - -:- . e -
\
i _ _.a ._.
~
- f. . u,. .
; f .. -
- =
- et , ~.
- -- - ..-..._... . .
- - - - - - =
".~.....
I' 1
'\. (3)
.J75) (61 "-- {7) : :- 2 ' ,(8) 3 (9) ...._ ._ . . . . .
{..20 o R j l g : _ . . .:: . . . . . .. ,
=
1 .. -
.H TUP ,I ~ ", E . . . - ....m 10 *lHR 6_t u: . : -_ . .. .
s
- COOLDOWN - ~'"- 'wg-,..._.,.._- . - .; . .. ,. .- .
, L _.. .
. - = J
. i
=--
t, g J-- . ; ;. ._;?. . _ _ . -- .* 1 , i i i iiiiii, i i i i i i i i i i i t-100 200 300 400 000 M hrrw- - _- - , i FIGURE 3/4 3.4-2 i ! RCS PRES $/ TEMP LIMITS (0 - 10 YRS) FULL POWER OPERATION j l %6E R99LMhW1fS& !, G... 8%_tta .M..t95-5Tg_ .
=-
s, h ' TA8tE 4.4-5 g 3:
. REACTOR SSEL m TERIAL SURVEILLANCE PROGR g . wg m RWAL SCK W E!
/
e< : ' LEAD
; CAPSULE VE55EL tocAtiog FACTOR w!THORAwALTIME(EFPY1 l; WMER _.
g 38' 1.5 " '" 8
- I Standby 43' 137' '
4- 5 I standby 4 142 230' 1. 5 12 - 15 5 i 310' 18 - 24 6
- d t
= $
h h} -
\ P,
\ ,
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' { ,.
- 0
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REACTOR COOLANT SYSTEM 1 PRESSURIZER HEATUP/COOLDOW LIMITS 1 LIMITING CONDITION FOR OPERATION i 3.4.8.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup rate of 200'F per hour, and
- b. A maximum cooldown rate of 200*F per hour.
6 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; perfors an engineering evaluation to detemine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer
, y remains acceptable fnr continued operation or be in at least HOT STANDBY
- within the next 6 hours and reduce the pressurizer pressure to less than
- ij 500 psig within the following 30 hours.
SURVEILLANCE REQUIREMENTS i < 4.4.8.2.1 The pressurizer temperatures shall be determined to be within the
- limits at least once per 30 minutes during system heatup or cooldown.
- 4.4.8.2.2 The spray water temperature differential shall be detemined l
for use in Table 5.7-2 for each cycle of main spray with less than four reactor coolant paups operating and for each cycle of auxiliary spray T , operation. l 'N a T 4 9 4 C.G*.4MSo-M959 -gT3
- 3/4 4-32
i - REACTQR COOLANT SYSTEM j j
,-% p a?
r a ._> J Je I 'i OVERPRESSURE PROTECTION SYSTEMS .. - LIMITING CONDITION FOR OPERATION f [M I 3.4.8.3 Bothkhutdownkoolinghstem(SCS suction line relief valv'es with ! lift settings of less than or equal to psig shall be OPERABLE and aligned ' to provide overpressure protection for the Reactor Coolant System. APPLICABILITY: When the reactor vessel head is installed and the temperature , of one or more of the RCS cold legs is lese than or equal to: Mt l a.(4M*Fduringcooldown
- b. 'F during heatup i
.' ACTION: With one SCS relief valve inoperable, restore the inoperable valve to i i a. J OPERABLE status within seven days or reduce Tcold to less than 200*F 2 and, depressurize and vent the RCS through a greater than or equal to
~ Do not start a g _ W square reactorinch vent coolant (s) within pump the next if the steam eightsecondary generator hours. water temperature f.: is greater than 100*F above any RCS cold leg temperature.
l i b. With both SCS relief valves inoperable, reduce Tcold to less than 200*F
- and depressurize and vent the RCS through a greater than or equal to 5 g_ % s,quare inch vent (s) within eight hours. Do not start a reactor
! coolant pump if the steam generator secondary water temperature is greater than 100'F above any RCS cold leg temperature. . c. In.the event either the SCS suction line relief valves or an RCS vent (s) are used to mitigate an RCS pressure transient, a Special
~
lg .. Report shall be crepared and submitted to the Commission pursuant to
-apecification 6.9. ithin 30 days. The report shall describe the circtestances init a ing the transient, the ef fect of the SCS suction
- line relief valves or RCS vent (s) on the transient and any cortretive action necessary to mvent recurrence.
1
- d. The provisions of Soecificetion 3.0.4 are not applicable. i SURVET pWCE REQUIREMENTS _
4.4.8.3.1 Each SCS suction line relief velve iihall be verified to be aligned i to provide overpressure protection for the RCS once every 8 hours during
- a. Cooldown with the RCS temperature less than or equal topF.
l 1;. iteatup with the RCS temperature less than or equal ta j 4.4.8.3.2 The SCS suction line relief valves shall be verified OPERAB'E with j the required setpoint at leett once per 18 months.
+w Sme RFMesm"$ 5PR-. l l MetMPERDe-etTPt 3/4 4-33 3
CF5s4f.P0 -ASSy-M l 1 l 3
REACTOR C00LAN" SYSTEM ! 3/4. 4. 9 STRUCWRAL_ICEGRITY {py h h h * [
)
) LIMITING CONDITION FOR OPERATION J . 3.4.9 The structural integrity of ASME Code Class 1, 2. and 3 components , rhall be maintained in accordance with Specification 4.4.0. . APPLICABILITY: ALL MODES ACTION: ! I a. With the structural integrity of any ASME Code Class I component (s) ' not conforming to the above requirements, restore the structural . 1 integrity of the affected component (s) to within its limit or
- 1solate the affected component (s) prior to increasing the Reactor ,
1- Coolant System temperature more than 50'F above the minimum ; ! temperature required by NDT considerations. i l- b. With the structurai integrity of any ASME Code Class 2 component (s) , j- not conforming to the above requirements, restore tM structural ' integrity of the affected component (s) to within its Ifnit or j lu isolate the affected component (s) prior to increasing the Reactor l l Coolant System temperature above 210*F. ' l t c. With the structural integrity of any ASME Code Class 3 component (s) T I , not conforming to the above requirements, restore the structural ' f I l integrity of the affected component to within its limit or isolate i
- the affected component from service. l 1 l
- d. The provisions of Specification 3.0.4 are not applicable.
, SURVEILLANCE REQUIREMENTS l
l e.~f. -:-, i F .~4.4.9 In addition to the requirements of Specification 4.0.5, each reactor i* coolant pump flywheel shall be inspected per the recommendations of Regulatory , j Position C.4.b of Regulatory Guide 1.14, Revision , ^" " -- T . CC . ' 1 l I l , l l h ' IM S bk M M
- 4 * .1 a a egad 3 C49ji@70rA#55-575 l[ ~ ~ . . . . _ . -4 3/4 4-34 ll i
~
REACTOR COOLANT SYSTEM l 3/4.4.10 REACTOR COOLANT SYSTEM VENTS fnn.2 9 GTs %- ( N .. jg;t G*'
~
l 1 l LIMITING CONDITION FOR OPERATION i 3.4.10 Both reactor coolant system vent paths from the reactor vessel head shall be OPERABLE and closed. i APPLICABILITY: MODES 1, 2, 3 and 4 ' ACTION:
- a. With only one of the above required reactor vessel head vent paths OPERABLE, restore both pates to OPERABLE status within 72 hours or be >
SHUTDOWN within in at the following least NOT STANDBY or. within the next 6 hours and
- b. With none of the above required reactor vessel head vent paths OPERABLE, restore at least one path to OPERABLE status within
! the next 6 hours or be in at least HOT STANDBY within the next
, 6 hours and in)$ SHUTDOWN within the following% hours. y ' t.oL3 L30 SURVEILLANCE REQUIREMENTS 4 r (- 4.4.10 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months - by:
- s. Verifying all manual isolation valves in each vent path are locked in the open position.
- b. Cycling each vent through at least one complete cycle from the control room.
b- c.' verifying flow through the reactor coolant system vent paths l during venting. : 1 1 i 3/4 4-35
).
I .- 3/4.5 EMERGENCY CORE C00LINC SYSTEMS (ECCS) Ti
"~~
l 3/4.5.1 SAFETY INJECTION TANKS I LIMITING CONDITICM FOR OPERATION I i 3.5.1 Each Reactor Coolant System safety injection tank shall be OPERABLE with: l
- a. The isolation valve key-locked open and to the val've removed, i
- b. A contained borated water level of betwe 802cubicfeet)and*72 Rep (1914cubicfeep j c. A boron concentration between and 4400 ppm of boron, and j j d. A nitrogen cover pressure of between 600 and 625 psig. l
, e. Nitrogen vent valves closed and power removed.**
; f. Nitrogen vent valves are capable of being operated upon restoration of power. i APPLICABILITY: MGDESIk2*,3,*t,and4*t. X ACTION:
l_
, a. With one safety injection tank inoperable, except as a result of a j i _ closed isolation valve, restore the inoperable tank to OPERABLE ]' status within I hour or be in at least HOT STANDBY within the next p
6 hours and in HOT SHUTDOWN within the following 6 hours. ,
- b. With one safety injection tank inoperable due to the isolation valve 4
being closed, either immediately open the isolation valve or be in ; at least HOT STANDSY within I hour and be in HOT SHUTDOWN within j the next 12 hours. SURVEILLANCE REQUIREMENTS i 4.5.1 Each safety injection tank shall be demonstrated OPERABLE: I a. At' least once per 12 hours by: 2 1. Verifying the contained borated water volume and nitrogen l cover pressure in the tanks is within the above limits, and ) @ s- tWith pressurizer pressure teater M than or equal to O psia. When pressur-
; izer pressure is less than psia, at least three scfety injection tanks [
sust be OPERABLE, each with a minimum pressure of 254 psig and a maximum ! pressure of 625 psig, and a contained borated water volume of between 0% i narrow range (corresponding to 60% wide range indication or 1415 cubic feet) } and 72% narrow range indication (correspondSng to 81% wide range indication or 1914 citic feet). With all four safety injection tanks OPERABLE, each tank shall have a sinimum pressure of 254 psig and a maximum pressure of 625 psig, i and a contained borated water volume of between UK narrow range (corresponding . to 39% wide range indication or 962 cubic feet) and 72% narrow range indica-l tion (corresponding to 81% wide range indication or 1914 ceic feet). In ! ! ICDE 4 with prgurizer pressure less than 430 pais, the safety injection tanks may be isolated.
"See Special Test Exceptions N 3.10 8f b
** Nitrogen vent valves may be cycled as necessary to suintain the required nitrogen cover pressure per Specification 3.5.1d.
N 3/4 5-1 MSNMO"M""UI
i EMERGENCY CORE COOLING SYSTEMS p;g & W,3 QPY g SURVEILLANCE GQUIREMENTS (Continued) +
- 2. Verifying that each safety injection tank isolation valve is
- open and the nitrogen vent valves are closed.
- b. At least once per 31 days and within 6 hours after each solution '
level increase of greater than or equal to 7% of tank narrow rang . level by verifying the boron concentration of the safety injection tank solution is between and 4400 ppe. X l c. W Ti$ At least once per 31 days when the RCS pressure is above Me psi [L by verifying that power to the isolation valve operator is removed. g ' i
- d. At least once per 18 months by verifying that each safety injection J -
tank isolation valve opens automatically under each of the following i conditions: )2 1. When an actual or simulated RCS pressure signal exceeds g 515 psia, and , , ,
- 2. Upon receipt of a safety injection actuation (SIAS) test signal. I tA4
- e. At least once per 18 months by verifying OPERABILITY of[RCS-SIT g '
differential pressure alars by simulating RCS pressure > 715 psia with SIT pressure < 600 psig.
)
/
i
- f. At least once per 18 months, when SITS are isolated, by verifying l the SIT nitrogen vent valves can be opened.
- g. At least once per 31 days, by verifying that power is removed from )
- the nitrogen vent valves.
- f , .
)
l 1 U l
} l 3/4 5-2
EMERGENCY CORE COOLING SYSTEMS
,P]FM
., , P()fp'.9 onNa
.. w. t ,
- u. .- -
3 /4. 5. 2 ECCS SUBSYSTEMS - T old GREATER THAN OR EQUAL TO 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooiing System (ECCS) subsystems shail be OPERABLE with each subsystem comprised of: I
- a. One OPERABLE high pressure safety injection pump,
- b. One OPERABLE low pressure safety injection pump, and
- c. An independt-3 perRABLE flow path capable of taking suction from the refueling water tank on a fafety A jection /ctuation f gnal i and y automatically transferring suction to the containment sump on a fecirculationjettuation pipnal. 7 APPLICABILITY: DODES 1, 2, and 3*.
_- ACTION: (i* a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least H3T STANDBY within the next 6 hours and in HOT SHUTDOWN within the following g 6 hours. A b. In the event the ECC5 is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared ond submittad to the Commission pursuant to Specification 6.9.2fwith' n 90 cays des-cribing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usa,ge factor for each affected injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70. dN' O , CenbLs 1 I l l I
"With pressurizer pressure greater than or equal to psia.
cgngn wns , ,_, 1 1
i v hhf h . o ) EMERGENCYCORECOOLINGSYSTEE ~ ~ ~ ~ ~ *S I
- SURVEILLANCE REQUIREMENTS ._
) . 3 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying that he following valves ,
are in the indicated positions with the valve L ., M - g ^. : i
- Valve Numbe Valve Function Valve Position
- 1. 1. HOT LEG INJECTION 1. SHUT 51h604 ~
?- - e P "" ;;l Z. ;;G ; S::T:::: 1. anvi ,
2/. 51 $ 509 Y HOT LEG INJECTIONSHUT h
- ... :: ,. .07 G im h!!C, 4. ma i -
1~ b. At least once per 31 days by: ' ?i: * !# 1. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise i i secured in position, is in its correct position, and 1 2. Verifying that the ECCS piping is full of water by venting the l accessible discharge piping high points. ; i
< 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 containmei.!. sump and cause restriction of the
- pump suctions during LOCA conditions. This visual inspection shall
. be performed: V' jN -- 1. For all accessible areas of the containment prior to j establishing CONTAINMENT INTEGRITY, and i
- 2. For all the affected areas within containment at the completion >
of containment entry when CONTAINMENT INTEGRITY is established.
- d. At least once per 18 months by:
hv1 653 bb
- e% m $,Yu ad y~?
i 4a.& cEsAeJee-vosss-sTs s %
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M 1 EMERGENCY CORE COOLING SYSTEMS 7.%..*I',,,..nr=-*' l -
' k . . ' N' I k L. ~~
i SURVEILLANCE REQUIREMENTS (Continued) I
- 1. A visual inspection of the containment sump and ve ifying that the subsystem suction inlets are not restricted by debriscand ,
that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.
- 2. Verifying that a minimum total of ". b: f;J. ' ..' e '
G;^ ' . , I' { -"' - { * -] -y
...m... . . . , . , ,
;'::;.' Q .td....7m
.m.. .
.6. mm.6 i neo 1 if; , ; ;, : -+sa mu- .-7--.-..*_ 7 :: 7 7 ; ;< n ntt . n nn, n
- ' T^,' i.... . T J w....--h ;L;t *; ;.. .....v ., ..so m ..m.m.,
' - 1.; L.;^,waiiva vi
. - " M.;.t;d at: ';; tN
.., : _ ; d ; ; Et L. . .
- ""' , -d ; " r f t'-.
- ~;: 1 2 . : n i . . " :~.... -
. . . . . ; :: 7 - " : - 15 . . . ,.f
'S-
- e. At least once per 18 months, during shutdown, by:
- 1. Verifying that each automatic valve in the flow path actuates to its correct position on IAS{andRAS$testsignalg.
, k
- 2. Verifying that each of the following pumps start automatically J
upon receipt of a Mfetyjhjectionjuttuation/est/ignal: !
- a. High/ressureAfety/njectionpump.
- b. Lowfessure/afetyMjectionpump.
f ,
- 3. Verifying that on a Acirculation Atuation Jdst gignal, the containment sump isolation valves open, the HPSI, LPSI and :
CS pump minimum bypass recirculation flow line isolation valves l and combined SI mini-flow valve close, and the LPSI pumps stop. ) 1 ] f ~~ f. By verifying that each of the following pumps develops the indicated differential pressure at or greater than their respective minimum allowable recirculation flow when tested pursuant to Specifica-I tion 4.0.5:
- 1. High pressure safety injer. tion pump greater than or equal to 1761 psid.
j 2. Low pressure safety injection pump greater than or equal to 165 psid. MN b8 - b NS O { i ex h tL M downrnvydt_ l "n vowi
/**e E m sp+
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CEs%Go- Nsss -375
l . ' ** .~. ,# . l
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)
EMERGENCY CORE COOLING SYSTEMS ' ) SURVEILLANCE REQUIREMENTS (Continued) l 1 .
- g. By verifying the correct position of each electrical and/or mechanical l
' position stop for the following ECCS throttle valves: i
- 1. Within 4 hours following completion of each valve stroking l i
operation or saintenance on the valve when the ECCS subsystems are required to be OPERABLE.
' 2. At least once per 18 mont .
LPSI System ) Hot Leo Injection Valve humber Valve Number J 306, ' 1.
~
- 1. SI 615, SI 51 604
- 2. 51 625, SI 307 2. SI 609 l
- 3. SI ' 635 3. SI 321 .
- 4. 51 645 SI 331
- h. By performing a flow balance test, during shutdown, following
. completion of modifications to the ECCS subsystems that alter the subsystee flow characteristics and verifying the following flow rates: HPSI System - Single Pump The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to K gpa. ,3 g LPSI System - Single Pump
%W r
- 1. Injection Loop 1, total flow equal to 20^0 _y10:0 gpa Injection Legs IA and IB when tested individually, with i r,
.s 2.
the other leg isolated, shall be within y gpa oT each 7N other.
- 3. Injection Loop 2, total flow equal to '^00 5 ^^ gpa MY
- 4. Injectic,n Legs 2A and 2B when tested indivi ually, with the ib other leg isolated, shall be within y gpa of each otner. i Simultaneous Hot Leo and Cold Leo Injection - Sin 21e Pump
- 1. Not Leg, flow equal to p gpa O ;
j 2. Cold Leg, flow equal to i gpa 7W l ama-se
- .n 3/4 5-6
, "I %'t l l PPiODF &PREW CD / vmgm Ifsrs 1 S I - (. i 7 # 7I-61l.
- 7. SI-(.21 # 5r-424 i
.i 3. sr-43 7 # sr-4%
- 4. SI-447 ) sr -44/. ;
- _ l if -
li-I I 1 i ., J f i ) i i' i i
l EMERGENCY CORE COOLING SYSTEMS . p y' 8[ {f , 3/4.5.3 ECCS UBSYSTEMS - T ;old LESS THAN 350*F _ - - - - - - - - - LIMITING CONDITION FOR OPERATION
=
3.5.3 As a minimum, one ECCS subsystem comprised of the followins shall te OPERABLE:
- a. An OPERABLE high pressure safety injection pump, and !
- b. An OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signal.
i APPLICABILITY: ~ 3 *W MOD (4 ACTION: l- :
- a. With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem
-_ to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 bours.
- 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 5.9.2 within 90 days
/ describing the circumstances of the actuatiren and the total accimulated actuation cycles to date. The current value of the usage
> factor fc each affected safety injection sezzle shall be provided in this Special Report whenever its value exceeds 0.,0. l SURVEILLANCE REQUIREMENTS l . 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable surveillance requirements of Specification 4.5.2. ; 1 l l 1 f
# WdA y nne_.Aa*YA'""176D M .
4F3sM&-MSSS-s75 3,. , l . l
PROD:& PsiiW COPY EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER TANK
+
LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank (RWT) shall be OPERABLE with:
- a. A minimum borated water volume as specified in Figure 3.1-2 of Specification 3.1.2.5, g g' g .
- b. A boron concentration between 4000 and 4400 ppe of boron, and 4
- c. A solution temperature between 60*F and 120*F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: j~ With the refueling water tank inoperable, restore the tank to OPERABLE status
- s. within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- SURVEILLANCE
- :EQUIREMENTS 4.5.4 The RWT shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
j 2, 1. Verifying the contained borated water volume in the tank, and l 2. Verifying the boron concentratian of the water.
- b. At least once per 24 hours by verifying the RWT temperature when the (outside) air temperature is outside the 60*F to 120*F range.
6
; 3/4 5-8
- i
i t gg 3/4.6 CONTAINMENT SYSTEMS . 3/4.6.1 PRIMARY CONTA]NMENT .l CONTAINMENT INTEGRITY 2 i j LIMITING CONDITION FOR OPERATION Mf@ 6MkYE)M MM/N hdl42E MEiJT5 q
^ :.1.1
. .. a wn ma nn i Anatbu n sna 61 ne maintainea.
APPLICABILITY: MODES 1, 2, 3, and 4.
, ACTION:
ithout primary CONTAIPMENT INTEGRITY, restore CONTAINMEN NTEGRITY within i hour or be in at least HOT STANDBY within the next 6 urs and in COLD l 5 ' DOWN within the following 30 hours. I SUR LLANCE REQUIREMENTS ), jt. jy 4.6.1.1 imary CONTAINMENT INTEGRITY s .11 be demonstrated: 1 i a. At st once per 31 days verifying that all penetrations
- not ;
F $ capabi f being closed OPERABLE containment automatic isolation 1 N valves a required to e closed during accident conditions are f closed by v 'es, bl d flanges, or deactivated automatic valves l 1 secured in the p tions except as provided in Table 3.6-1 of Specification 3. . } . By verifying at each ontainment air lock is in compliance with j the requi nts of Spec cation 3.6.1.3. I . , p. After ch closing of each pe ration subject to Type B testing, i
~'
, exce containment air locks, if ened following a Type A or B l te , by leak rate testing the seal l l rifying that when the measured leaka th gas rateatfor P,these seals is49.2 psig and l I added to the leakage rates detemined pur nt to Specifica-l tion 4.6.1.26. for all other Type B and C pe rations, the combined leakage rate is less than or equal to 0.60 L,.
l 1 Except valves, blind flanges, and deactivated automatic valves wh are located inside the containment and are locked, sealed, or othervi i secured in the closed position. These penetrations shall be verified I closed during sach COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.
- 1 T
\ CG599410
=_ rns-N$s -ST5 W: : '~l
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4 CONTAINMENT SYSTEMS g
- g {] p ;
I l 3 /4. 6. 2 DEPRESSURIZATION AND COOLING SYSTEMS kh*h hygpryyn,) ! CONTAINMENT SPRAY SYSTEM ( LIMITING CONDITION FOR OPERATION
~
3.6.2.1 Two independent containment spray systems shall be OPERABLE with each i on-ay system capable of taking suction from the M on a containment spray cctuation signal and automatically 1ransferring suction to the containment Each spray system flow path from i sump on a recirculation actuation stgnal.an OPERABLE shutdown cooling heat exchanger. J the containment Lump shall be v '
- APPLICABILITY: A T1,2,5,and4.I l ACTION:
i j 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 i j -- the next 6 hours; restore the inoperable spray system to OPERABLE status ! i within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. I a
=
SURVEILLANCE REQUIREMENTS l!
- 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE
At least once per 31 days by verifying that each valve (manual, l
/ a. l
,A power-operated, or automatic) in the flow path is positioned to take l suction from the M on a containment spray actuation (CSAS) test signal. l l
- b. Byverifyingthateachpumpdevelopsanindicateddifferential pressure of greater than or ual to till psid at greater than or l
j equal the minimum allowabl e ulation flowrate when tested g pursuant to Specification . .S. '
- c. At least once per 31 days by verifying* that ' " '
the
' ' system piping is ./dg.
^ - - - ~ ~ -
} ' fm>ul l o f wa te r '_ ", '_ '_ a ' " ' -"-' g .
. . . v a, ., ,.._.
,s
\ M, l
- d. At least once per 18 months, during shutdown, by:
j Verifying that each automatic valve in the flow path actuates j 1. to its correct position on a Abntainment Aray Atuation (CSAS) K j $
;M and fecirculation Aftuation (RAS) gest pignal.Mst flgnal, !
1
- 2. Verifying that won a Acirculationgctuation .
l the containment sump isolation valves open and that a 5 j recirculation mode flow path via an OPERABLE shutdown cooling l J heat exchanger is established. l ) l ;
= m . .a. . . m n z ' m - z - -" . ...
5 ( d 3/4 6- j S j u
CONTAINMENT SYSTEMS .0 & EvW CON l SURVEILLANCE REQUIREMENTS (Continued) l I i
- 3. Verifying that each spray pump starts automatically,dn : : : ' t; _ ;
fr.;;:t'- ::tr:t':: ' " fi ; . . . : a containment spray actuation l (CSAS) test signal. -
- e. At least once per 5 years by performing an air or smoke flow test I through each spray header and verifying each spray nozzle is ,
! unobstructed. d 5 d i l l 4 - , t d l l e i ! i i l a ~ J 1 i i I i
)
,i IN b w+ 1 v7-
i S y @ [Tt W CDNTAINMENT SYSTEMS gg g hh hh
- ..m.-
C'^'^L SYSTEM ll LIMITING CONDITION FOR OPERATION / Sm M AAh 3.6.2.2 C'_ [ ste shall be OPERABLE with: , ! The ' E .. .A ! a. An spray chemical addition tam. containing a level of between 90% 1 and 100% (816 and 896 gallons) of between 33% and 35% by weight N2 "4 1 solution, and ( l
- b. Two spray chemical addition p eps each cepable of adding N,H4 solution from the spray chemical addition tank to a containment sprly system 1 pump flow.
! APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: h Q, With the '- N : b : :' /ystem inoperable, restore the system to OPERABLE
,- status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the iodine removal system to OPERABLE status within the next 48 hours
]j .j - or be in COLD SHUTDOWN within the following 30 hours. .. l J l SURVEILLANCE REQUIREMENTS . y .- . 4.6.2.2 WY N" The . J : _ stem 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:
Y , I. Verifying the contained solution volume in the tank, and i j - l 2. Verifying the concentration of the N H24 solution by chemical j analysis. . c. By verifying that on recirculation flow, each spray chemical addition j pump develops a siischarge pressure of 100 psig when tested pursuant , to Specification 4.0.5 l d. At least once per 18 esths, during shutdown, by l 1. Verifying that ase.i automatic valve in the flow path actuates toitscorrectpositionona/ontainment/prayjictuation(CSAS) A
/estfignal,and
- 2. Verifying that each spray chemical addition pump starts automatically on a CSAS test signal.
I %IM ,M-4TS 3f, ) 4
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) ,
- e. At least once per 5 years by verifying each solution flod r te from the following drain connections in the '-d' c :- " . ystem:
- 1. ::" "25:
N^ . ....;.;: 0.63 1 0.02 gpe. i [2. I:: .;;, i l
' A. . . ; ;: '- 0. 63 1 0. 02 ppe.
4% ; [ ( M s14n,4_ N .ut:m) ( . I .l E s 1Y i i i l t l
.~s 4
i
< +
i 3 I i i l 4 Cff?-NRSS~STS u< s-f( i I I
i s -
,t Id to J
i - l LIMITING C1NDITION FOR OPERATION AND SURVEILLANCE REQUIREMEhis I i' OK -__MN__wDV
~-
- 1/4.5,2.3_ ,_
___m_____- _____ __ g 4 l
- ===.=. y% .
MM g a ,,, m > D i i m j* i l l l . J ; l l . t 7 i 5 ; i , i I , i , I i i i l "Nka 3/4 6-
CONTAINMENT SYSTEMS
~'
i} ef- pa;r e rnTN P, k em - U k "-'" ' ' ' 4: i 3/4.6./ CONTAINMENT ISOLATION VALVES F - -~ i i LIMITING CONDITION FOR OPERATION
- 3. The containment isolation valves specified in Table 3.5-1 shall be OPE ABLE with isolation times as shown in Table 3.6-1. -
APPLICABILITY: MODES 1, 2, 3, and 4. ! ACTION:
- 1. With one or more of the isolation valve (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, r
j ;J :
- b. Isolate each affected penetrathn within 4 hours by use of at least l one deactivated automatic valve secured in the isolation position,"
or 4 ) . c. Isolate the affected penetration within 4 hours by use of at least one closed manual valve or blind flange;* or J d. Be in at least H3T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ' 1 j SURVEILLANCE REQUIREMENTS 4 4.6.3.1 The' isolation valves specifie-1 in Table 3.6-2 shall be demonstrated I j OPERABLE prior to returning the valve to service af ter maintenance, repair, or replacement o p r work is performed on the valve or WAQ its associated M actuator, control, i
%A
- eg } 4.6.3.2 Each isolation valve specified in f-" '
- : "; , ", -- ' C :' Table 3.0-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
1 a. M er c.ak Verifying that on s TE, _f'! r- ??^* test signal, each isolation , I valve actuates to its isolation position.4--3d ; i' b. M. I Verifying that on a Containment Radiation-High test signal, 4 H- y i e^"r'--nt p-g i a; at. a w Lm.. ...'.;':r p ;it':- ( p) n "The inoperable isolation valve (s) may be part of a system (s). Isolating the l
'affected penetration (s) may affect the use of the system (s). Consider the tech- i
- nical specification requirements on the affected systes(s) and act accordingly.
M hb - M A4LO kd I.L.C D O . enn yenne_ _.m - armg&-m-m 3/4 6-[ l
#I l ~~
r '3y (0@! j Lus aps op n e a sa.,W W w uc/ M j
=1 n acund +0 b-Al l
( k D 1 1 l M N E4ns, \ i: T i
\
i lE i 1 , l l i 4 i ) .~l , j 1 1 1 1 4 4 l i ; ! I l i j
i CONTAINMENT SYSTEMS
,(( h h _,
l i l t. ) SURVEILLANCE REQUIREMENTS (Continued) i ! i 4.6.3.3 The isolation time of each power operated or automatic valve of l $ .!::^.i: : ", " :-f C r' Table 3.6-1 shall be determined to be within its Itmit ! when tested pursuant to Specification 4.0.5. : l l M .I.: ^ L C ::' _:7... .,.. ::d '- !::^_'r- 0 :' '9'r ? El il' R l demonstra DERABLE pursuant to 10 CFR 50, Appendix J, with the ex lon of j those check valv etnoted as 'Not Type C Tested." 4.6.3.5 The isolation valv ecified in Sections r G of Table 3.6-1 s shell be demonstrated OPERABLE as tred b & ation 4.0.5 and the i Surveilvance Requirements associated w ose Limiting Conditions for i Operation pertaining to each valv system ich it is installed. Valves. I secured ** in their actuated ion are considere erable pursuant to this !_ specification. I j 4.6.3.6 T nual isolction valves specified in Section H of Ta 6-1 j- shal demonstrated OPERABLE pursuant to Surveillance Requirement 4. . - a
- pecification 3.6 1 1 aM I
l 4 l 1 ,.,, I i J i i 1 A
- A
--/
f jy1 v e y or o C ..';. s evented f = nnie ma*4nul -----+4-a - 1 d . l M55Mro-= ~, -MPGS-STS u u4 s-pfg i
1 Y: I
~ .. J CONTA1 mEkl 150L T ON VALVES i
l ? i MAXIMlJM l l ACTUATION 3l ' VALVE PENETRATION TIME , 1 NUMBER NUMBER FUNCTION (SECONDS) . n. m..~,,, .ni av en u un u. A Ab)
\ -UV 023 9 Containment radwaste sump pump to 30 i LRS holdup Lank l RDB-UV 9 Containmen1 racNaste sump pump to 5 l LRS holdup tank j RDB-UV 407 9 Containment radwaste sump post- 5 -
-- accident sai pling system ,
SGB-HV 200 11 owncomer fahdwaterchemi 1 ! ! - ection )j E SGB-HV 201 12 Downc injection fledwate chemical 1 N SIA-UV 708 23 Containee c e sump to post- 5 ( accident pling stem ) j HCB-UV 044 25A Cont mo cor (inlet) nt air radios vity 12 J .
! HCA-W 045 25A ontaini ent air radioactivity 12 '
- nonitor (inlet)
HCA-W 046 25B Containnhntairradioactivity 2 i
- -( noniter q outlet) ,
l 5 i HCA-W 047 5B Containme't air radioactivity 12 1 monitor (4 utlet) GAA- 002 29 N, to stem i generator and reactor 10 i drain tank GAA-W 001 30 N, to SI t inks 10 G l NSW##-MisS
? "'l C- ; Z. . -
SP5 3/4 6
a; . t . I\ _ m e _; ; i alve location ESF Required Montsum Valve Penetration Number Velve Nuuter
%%5fD - ~ ,C x Relative to Actuation ontainsent Signal _
Post-Accfdent Valve Postt1on , Actuation _ Time (sec) A. Neuetely Actosted Valves 11 5I-331 Mot leg Injection Valve tside None Open 10 12 5I-321 Mot'I.egInject1onValve tside None Open 10 13 SI-616,617 14 SI-626, 627 Nigh Pressure Cold , 15 SI-636, 637 Leg Injectton Valves Du side SIAS Open 10 , 16 SI-646, 647
- a. 17 SI-615 3 18 SI-625 Low Pru sure Cold 9 Ig SI-635 Leg Injection Valves Outs SIAS Open 10
% 20 SI-645 y 23 SI-673 Containment sump 1 solation Inst RAS Open 20 8 valve m SI-674 Containment semp isolation Out 1 RAS Open 20 f?'
valve ::==> 24 SI-675 Containment sump 1 solation valve I de RAS - Open 20 bCC3 m SI-676 Conta1sument soup isolation s1de RAS Open 20 p
- valve g .
c2 , E 27 5I-690 Shutdown Cooltag Warump itside None Open or Closed 30 < q,# bypass valve t: 51-656 shutdown Coolfng isolation tside None Open or closed , 80
-l valve
[ , 51-654 Shutdown Cooling Isolation valve Inside None Open or Closed 80 l
- ___._.___._m. _ . _ _ . . _ _ . _ . _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ . . . . * - _ ___.m. _. ___.... .m - _ e-. .-=...-
l TABLE 3.6-1 (Centinu:d) nn CONTAINMENT ISOLATION VALVES g3;Op i .upp.n; er,may4r
- b"i 6 j i _ _ _
I MAy tw w ACTUATION ALVE PENETRATION TIME UMBER NUMBER FUNCTION (SECONO l r A. CONTAINMENT ISOLATION (CIAS) (Continued) HPA-UV 01 35 Containment to hydrogen recombiner 12 HPA-UV 0 ?3 35 Containment to hydrogen recombin 12 HPA-UV 01 4 35 H2 control system 5 HPB-UV 00 ? 36 Containment to hydrogen ret iner 12
}
HPA-UV 005 38 Containment to hydrogen r ombiner 12 HPB-UV 004 36 H recombiner return t containment 12 l (inlet) IV HPA-UV 023 38 H2 cc trol system 5 l HPB-UV 006 39 H recombiner turn to contLinment 12 N (inlet) CHA-UV 516 0 Letdown li from RC loop 2B to 5 regenerat ve heat exchanger and J letdown eat exchanger F CHB-UV 523 40 Leto line from RC loop 2E to 5 rege rative heat exchanger and i let wn heat exchanger f
# CHB-UV 924 40 tdown line to post-accident 5 ampling system I SSB-UV 201 42A ressurizer sample surge line 5 r sample surge line SSA-UV 204 42A Press 5 SSB-UV 202 42 Pressurizer samp rge 'ine 5 SSA-UV 205 42B Pressurizer sample surge lin 5 SSB-UV 00 42C Pressurizer sample surge line 5 SS UV 203 42C Pressurizer sample surge line 5
# J k
b-- hhh 3/4 6-
t Q
---*ashs.a.4 ? F-m) 00[f g, YCom we Locatt ESF Required Maxinus Valve Penetratton lative t Actuation Post-Accident Actuation Number Valve lhs6er m~ @6 .m.
. ta1puen 51gnal Valve Pos1 tion T1me (sec) 2tt SI-691 Shutdown Cooling Warsup tside None Open or Closed 30 bypass valve SI-655 Shutdown Cooling isol: tion ist None Open or Closed 60 valve "
SI-653 Shit'down Cooling isolation In i None Open or Closed 80 valve l 29 51-682 Safety Injection Tank fill ins SIAS Closed 5 and drain isolation valve '
. 40 CH-523 CVCS Letdown Line Outs de CIAS Closed 5 4 2 CH-516 Isolation Valves Inst . CIAS/SIAS Closed 5 CVCS Charging Line Iso- Out 1 W
f 41 CH-524 Mone Open or Closed 5 rn a lation Valves 43 CH-505 Reactor Coolant Pump Con- Ou 1 CIAS Closed 5 s j i CH-506 trs11ed Bleedoff Contain- In 1 CIAS Closed 5 sment Isolation Valves 44 1-560 Reactor Drain Tank I ide CIAS Closed 5
'H-561
- Section Isolation Valves si CIAS Closed 5 j 45 CH-580 Reactor Makeup Water Supply is1 CIAS C1csed 5
- Isolation Valve to the ROT 57 CM-255 Seal Injectfon Containment tst None Open or Closed 5 Isolation Valve Q
f 8. Menval Valves
~-.g h,, 29 51-463 Safety Injection Tank Fill tside Ihme Closed ' Not Applicchie
- and Drain Isolation Valve j .
i
TABLE 3.6-1 (Continued) < -- g {II] *[fY COWTIINMENT ISOLATf0N VALVES htA Y T MitW
/A ACTUATION TIME f VALVE PENETRATION NUMBER NUMBER FUNCTION (SECONDS) i'
- n. w niAIMMr.N1 nut.A 4 WM nAAs>
(Continued) HA-UV 560 44 Reactor Drain tank to pre-holdup 5 ion exchanger CHB-U 1 44 Reactor Drain tank to pre-holdup 5 fon exchanger ! CHA-UV 580 Makeap to reactor drain tank 5 - 4 CHA-UV 715 45 Makeup to reactor drain to post- 5 accident sampling syster GRA-UV 001 52 vent to WG surg ank 12 U RDT ve to WG ge tank 10 GRB-UV 002 52
- WCB-UV 63 60 Nomal chil water to containment
- ACU (inle 10 WCB-UV 61 61 Nom chilled water containment AC (outlet) 10 WCA-UV 62 61 Normal chilled water to cont ~ nt ACU (outlet) 10 )
! \ 1
./
sl l ( f - Valve exempt from Type C testing I i bY?T. .
, - - . _ -. 3yz y-[
l
.T ,,
\\
Aa=* t em 9 Am - ~ h A
; Ive Locatt ESF Nequired Menlaus Valve Penettetlen Num6er Valw Nuuter fg:d.Mr lative t tafnmen< _
Actuation 5ienal pest-Accident _ Valve position Actuation Tfue (sec) _ 41 CN-393 CVCS Charging Line Inside None Closed Not Applicable CM-854 Isolation Valves tside pone Closed Not Applicable i C. Check Valves , , , , 11 51-533 Not Ley Injectlen Line I 1 None Open Not Applicable 12 SI-523 Isolet'on Valve 13 SI-113
.s-14 51-123 Mi$ Pressem Cold leg In None Open Ilot Applicable 15 51-133 Injection Line Isolation l 16 $1 144 Valve 17 SI-t14 18 5I-124 Lee Pressere Injection In e None Open Not Applicable 19 5I-134 Line Isolatten Valves -
28 SI-144 41 06-431 CVCS Chart9 81 Line I 1 flone Open or Closed Not App 11 cable CH-433 Isolation Va'ves - 45 CH-494 IIeecter Nebesp IInter Supply side None Closed Not Applicable Isolation Valve to the ROT
,, 57 CM-835 Seal Injection Containment nside None Open or Closed flot Applicable
- Isolation Valve d(k ...
4,N Sh ft' i f * . p//
,,A .
4,',/ l
i i CONTAINMENT SYSTEMS ; 3/4.6.4 COMPL'5TIBLE GAS CONTROL HYDROGEN MONITORS 1 gui &E]i : j L - 1 LINITING CONDITION FOR OPERATION M) l_@MUAdg dlfh~-fp)[
- l l
. l l
O : . ..e nwo n. 6unts umie nt nyarogen monn.or. .ns i a se vn.KAstE. MPLICABILITY: MDDES 1 and 2.
- i ACTION:
- a. With hydrogen monitor inoperable, restore the 14 erable monitor !
to OPE status within 30 days or be in at le HOT STANDBY l
, within the ne hours. I
- l. ~ b. With both hydrogen so ors inoperable estore at least one monitor' to OPERABLE status withi 7? hours e in at least HOT STANDBY l j ,
within thc. next 6 hours. 1
- a. c. With one hydrog:n monitor perab and the hydrogen monitor in
@ the Post Accident Sampi System OPE E, the provisions of i i Specification 3.0.4 not applicable. ie l I JS JRVEILLANCE REQUIREM l i q % - - ! I
- J N ydrogen monitor shall be demonstrated OPERABLE by the rformance l4 6.4.1 Ear
.l c FaC ,, CHECK at least once per 12 hours, a CHANNEL FUNCTICMAL TES t N tast ce per 31 days, and at least once per 92 days on a STAGGERED TEST SIS j by forming a CHANNEL CALIBRATION using sample gases containirs a nominal:
, s. One vol me percent hydrogen, balance nitrogen.
l b. Four volume percent hydrogen, balance nitrogen. ! x 8 ! l J i i
- d E JSAf M -A)p55
- 6 3/46-g 12.
i
- [_105Egli G
- f a r30."# F l 3/4.4.tgr Camswfigx.C,as.&M !
l & #,(a.M'\ NSCins - j 3/4.6,.g.2. E kdnic, k dcenSdwts v l 74.6, E. 3 W3 ?9 Obg Yam
- c. Copt mu 1 l ; A 6.EV Lpp &Qdrvm 7e.yucriuitww Awn Ex eve lC 34.6.6 l A hugwm0$ prod) .
l
%GT O a c u.t w n A b d & s ( s 1 l 0 4 l % 4,. 8' seamdonp Gm%mwtGA !
!"- Ty Cobsmeet,0giunwJt) ! Ya.4.2 ;, h 4 Sd % Aa.c % Sb as.4.t.z 6dA 641%dyg i M.s:s E t3 Swh %
- b e st,1 l
TYb ! I
~ ,
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE 00:& E N M SAFETY VALVES [ . LIMITIM CONDITION FOR OPERATION 3.7.1.1 All main steam safety valves shall be OPERABLE with lift settings as specified in Table 3.7-1. APPLICABILITY: MODES 1, 2, 3, and 4*. I'PhM J. ACTION:
~
- a. With both reactor coolant loops and associated steam genersters in operation and with one or more** main steam safety valves inoperable per steam generator, operation in MODES 1, 2, and 3 may proceed
- provided that within 4 hours, either all the inoperable valves cre restoted to OPERABLE status or the " _.. Li lG. trip setpoin s ti* nduced per Table 3.7-2; otherwise, be in at least HOT STANDBi within the next 6 hours and in COLD SHUTDOWN within the following :
30 hours.
- b. Operation in MODES 3 and 4* may proceed with one reactor coolant loop and associated steam generator in operation, provided that there are no more than four inoperable main steam safety valves associated with the operating steam generator; otherwise, be in i COLD SHUTDOWN within the following 30 hours. I
- c. The provisions of Specification 3.0.4 are not applicable. l
.g SURVEILLANCE REQUIREMENTS 5
4.7.1.1 No additional Survef11ance Requirements other than those required by i Specification 4.0.5. l l a Until the steam generators are no longer required for heat removal. The maxtos mober of inoperable safety valves on any operating steam generator is four (4).
- b 3/4 7-1
Q I
\\
; 6 TABLE 3.7-1
' l STEAM LINE SAFETY VALVES PER LOOPS h
F. b
. bI VALVENLNSER[ MINIMUM RATED CAPACITY **
LIFT SETTING (11%)* f 5 S/G No. 1 S/G No.__2 wg a. 4GG46 M pt 5GG 46 M 64 125(psig s.s -^0 lb/hr
~ . , . ,
h~b. SGG46H&-' 3GE-Pm MMIT- 125[psig Ib/hr
- c. 4GG46 M 79- 40E-fmMIS9 1290 psig Th.M ,332 lb/hr
- d. SOETST"TT .4GE-fmMMMP= 1290 psig N,,1 0
. . .,.,,.M I b/hr
- e. 19999P579 .4Gfr.p64-966 - 1315 psig I lb/hr o
. 95t3eoq y f. 40E=9mMif5 4GE4mMISP 1315 psig a., , .,3v Ib/hr
. L 1315 psig
% ooo ., , .7.,e I b/hr
- g. 5G4 957"576 982-fm MIS 6 I
- h. SGG4mMpfP EGG 46 H 69 1315 psig $801)g"Ib/hr
. _ . , . , ,5'
- ygg ,C
- 1. 5GE4mMp99= 4GG4E M 94 1315 psig .,,.,A lb/hr {
J. ter acu a*2 $N 1315 psig , tb/F- ;
.}
i
'. ,a1 \,
I
*The lift setting pressure sha11 corresrond te d 'ent conditiota at the v:41ve at nominal operating temperature txl pressure.
Q-
** Capacity is rated at 1(f t sett4g +.5 accwelation. ,
t L r{% P.)* Y fR
- O -
A E k smf ._ t- N Nwk . hh6 (x ' k
- [Ee
.; RODE &RBUI E0l i
- ma
$ a5 35 Gs i >.
\ : a I; s lg lo 0 9 n iN ,h b $ N i g 3 bE o d" PO I
..l i'
~
A
. J3 gir *~
~ ..
~
i ; E.o ES 'I w my *EE 1
- i
- It -
l~"a E6 ? {0 5 l > 5 , E 5 lf, 5 $! I s- In rr
# u*E g gm a *E
*Wi E ~ ~ "*
E 1 W i 5 <* 'a EC 40
*3 C65 Ado-res;w
- .ow w wa - ;^^;;T . - 3/4 7-3 l
4
t I a . 4 !
- PLANT SYSTEMS
- h. ., l,, h6 ..'.de O, e, La g .yf i
Ai j i AUXILIARY FEEDWATER SYSTEM i _ , _ _ . l LIMITING CONDITION FOR OPERATION M $L)MNM@ ll) $ k k ' 4 *, L s a+ 1..e, 1 e' :: 7,; 7,7,1;; g ,; 7 7.-+ar mur m arv f== 4=+ '
,_ J
- and associated flow paths shall be OPERABLE with
- '
- a. Two feedwater pumps, each capable of being powered from sepa te OPERABLE emergency busses, and
! b. One feedwater pump capable of being powered from an OP BLE steam j supply system. l AFPLI ILITY: MODES 1, 2, 3, and 4*. i ACTION: , 1
- a. Wi one auxiliary feedwater pump inoperab , restore the required
- auxi ary feedwater pumps to OPERABLE sta s within 72 hours or be ! , in at ast HDT STANDBY within the next hours and in HOT SHUTDOWN .
j , within following 6 hours. l.. b. With tus au liary feedwater pumps noperable be in at least HOT iC STANDBY withi 6 hours and in HOT HUTDOWN within the following j, 6 hours. ;
- j i j c. With three auxilia feedwa r pumps inoperable, immediately initiate corrective action to sto at least one auxiliary feedwater pump l
to OPERABLE status as as possible. 4 SURVEILLANCE REQUIREMENTS Q 4.7.1.2 Each auxiliary fee ater pump sha be demonstrated OPERABLE:
- a. At least once r 31 days on a STAGG TEST BASIS by:
-r 'yb 1. Testi the turbine-driven pump and motor-driven pumps i pur ant to Specification 4.0.5. The p isions of Specifica-
, t n 4.0.4 are not applicable for the turb e-driven pump for i try into MDDE 3. l 2 Verifying that each valve (manual, power-operate , or automatic) in t'w flow path that is not locked, sealed, or et ise secured in position, is in its correct position.
- 3. Verifying that all manual valves in the suction lines from he
- primary AFW supply tank (condensate storage tank CTE-701) to 1 each AFV pump, and the manual discharge line valve of each AFW l pump are locked, saaled er otherwise secured in the open positio i
i I j N'm ^.: 6 .- .~. . .. . . xp W nd ':: '-+ = 1 ) No .Ifar '
'F5 s/4 7-4 l -
l: 3 _M o
- PR00F &RIEW COPY I
= j 3/c 7.i.a f A m 3: g e % sig m ( gor fp&com% sAR i a 1 i I i : ! !.i . l 1 1 r i j 1 4 : TI
- s i
i I l i
/ l l
l _. PLANT SYSTEMS l "
- CONDENSATE STORAGE TANK h, n O er.ne
- n. 'e'- P'.3 l
._ J LIMITING CONDITION FOR OPERATION i
N , 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a levirl f
~
! of at least feet (300,000 gallons). f %gu i APPLICABILITY: MODES 1,2,3gand4.*# [jg) I ACTION: ( ' With the condensate storage tank inoperable, within 4 hours either: c a. Restore the CST to OPERABLE status or be in at least HOT STANDBY ; l l withinthenext6hoursandinHOTSHUTDOWNwithinthefollowing(4 6 hours, or
- b. Demonstrate the PERABILITY of the Pr- _: :.-- _ _ _ _ . __ sr I \
d backup supply to e __: r feedwater pumps and restore the condensate storage tank to OPERABLE status within 7 days or be in at J E. least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN with a 1 OPERABLE shutdown cooling loop in operation within the following 6 hours. i i SURVEILLANCE REQUIREMENTS i 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLE at least
- once per 12 hours by verifying the level (contained water volume) is within,its I limits when the tank is the supply source fo he
- r"; feedwkter pumps 7 t.
4.7.1.3.2 The '
-" all be demon trate ERABLE at l ' least once per 12 hours whenIrver it ._x tr ::':; t s the supply source for the auxiliary feedwater pumps by verifying: A g i
.a. That the W5 _I T : to the :r " ' N eed l
system isolat o og _ e (3 0 00 galio'ns i "Until the staan generators are no longer required for heat removed.
#Not applicable when cooldown is in progress.
l
**SaaA$ W %spR. -
i
- oAa.eht e l r NY 3/4 I7-/[
e M P_l).NT. SYSTEMS .. ACTIVITY . FF i'"' * - ( , ,,1 . .___ _. _ LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be - less than or equal to 0.10 microcurie / gram DOSE EQUIVALENT I-131. APPLICABILITY: MODES 1, 2, 3, ant 4. ACTION: With the specific activity of the secondary coolant system greater than 0.10 microcurie / gram DOSE EQUI,VfLENT I-131, be in at least HDT STANDBY within
- 6 hours and in COLD SHUTDDWN within the following 30 hours.
4 SURVEILLANCE RF'0lREMDiTS 4.7.1.4 The specific actitity of the secondary coolant system shall be ^ detemined to be within the limit by performance of the sampling and analysis program of Tabl e 4.7-1. i
.t f
I i
=
4 ( C =E ==: SSA - trTW O - N.s99 P 5 s/47-Q
a TABLE 4.7-1 00F & win @ SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY ) 5 AMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS ' l AND ANALYSIS FREQUENCY
- 1. Gross Activity Determination . At least once per 72 hours
- 2. Isotopic Analysis for DOSE (a) 1 per il days, whenever
- EQUIVALENT I-131 Concentration the on is activity determina- l tion i
- 'icates iodine con-centrat hus greater than 10% ,
of the allowable limit. l (b) 1 per 6 months, whenever the - gross activity detersination indicates iodine concentra-tions below 10% of the allowable limit. j
.~/
j i I
~
r , Jf 5 3/47-g
PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION l l
~
l 3.7.1.5 Each main steam line isolation valve shall be OPERABLE. , , APPLICABILITY: MDDES 1, 2, 3, and 4. 1 i ACTION: i MDDE 1: With ose main steam line isolation valve inoperable but open, POWEP, OPERATION 4 may continue provided the inoperable valet is restored to OPERABLE status within 4 hours; otherwise, be in at least MDDE 2 within the next 6 hours. MDDES 2, 3, and 4-4 With one main steam line isolation valve inoperable, subsequent operation in
- . MDDE I, 3, or 4 may proceed provided
- ,
l3. 'T
- a. The isolation valve is maintained closed.
- b. The provisions of Specification 3.0.4 are not applicable.
) Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
i SURVEIL (.ANCEREQUIREMENTS i 4.7.1.5.1 Each main steam line i solation valve shall be demonstrated OPERABLE by verifying full closure within 4.6 seconds when tested pursuant to ; e g Specification 4.0.5. i %
- i. 4.7.1.5.2 The provisions of Specification 4.0.4 are not applicable (or entry i into MDDE 3 or MDDE 4 to perform the surveillance testing of Specification
- 4.7.1.5.1 provided the testing is perforined within 12 hours after achieving
- normal optrating steam pressure and nomal operating temperature for the secondary side to perfore the test. 1 9 [
@%DV5EF-5W 3f4 7/1
~~~~
t l PLANT SYSTEMS { _ . . t ATM05PHE ALVES at$ b hk
^ --
)
LIMITING CONDITION FOR OPERATION i g' $/h $ 1N>D 3.7.1.6 )>( atmospheric dump valve / shall be OPERABLE. . { ] APPLICABILITY: MODES 1, 2, 3, and 4.* ! ACTION: yg i L. With less than atmospheric dump valv6per steam generator OPERABLE, restore
- the required atmospheric dump valve $to OPERABLE status within 72 hours; or be in ,
i at least HOT STANDBY within the next 6 hoursh MtfT"$S\!T")04,4,)Q With k j b.asus I 4 hg ( W. l - SURVEILLANCE REQUIREMENTS .
,, ,. n , n 4
pn=::en.W M Q@( 4.7.1.6 Each atmospheric dump va hall be demonstrated OPERABLE: R-
- a. At least once per 24 hours by verifying that thetn'M : ;: ,- "-'
*:' .... ,.... "??"'"
, b. Prior to startup following any refueling shutdown or cold shutdown ] i of 30 days or longer, r["y that all valves will open and close fully. I ,E-i l" 4 S e. A Lw(A SArt 'nce e s% l l bgN.E.%t. dvweSp'&M dawy i l l i
*When steam generators are being used for decay heat removal.
l i OET5&r&O-m ... :- =:::A457-N Di.a s/4 7-
% w -- w ---... - 4 ,m.,w . n _ 4 ,4, .p S, m ,,
__I l _ LwiC N00F &RBS E
=
l D. ( $ O/h] O L OC 4- M N) D j i i stubwA Ww xaus) ro oee.saets l WoT STAtoD3y ubb 6 'neuns omd
- - s e Shut'hWO m&M r re , NQ
,401 AR.
l
$ devis. $demid- Mna Use_ cd- bna Fe.d GI c
- anus, um s, ga..
- g. .. ..
} i l l l
PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATre PR0lif & REi2 CC LIMITING CONDITION FOR OPERATION , 3.7.2 The tempe ture of the secondary coolant in the steam generators shall be greater than F when the pressure of the secondary coolant in the steam generator is greater than ypsigN APPLICABILITY: At all times. ACTION: With the requireeents of the above specification not satisfied:
- a. Reduce the steam generator pressure te less than or equal to ;
psig within 30 minutes, and
- b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam Z generator. Determine that the steam generator remains acceptable g fcr co r e.ntinuedg opration prior to increasing its temperatures above s SURVEILLANCE REQUIREMENTS Mk [
4.7.2 The pressure in the econdary side of the stenal generators shall be determined to be less than psig at least once perd hours when the temperature of the secondary coolant is less than F. g
~
ea* n e-kW k hff Dicw$$p SAR [ csse := %e20-Nc;-my
=:T :
2/4 74 L
i , j PLANT SYSTEMS MN &RIN M j 3/4.7.3 ;',L.:n COOLING WATER SYSTEM
^ ) i LIMITING CONDITION FOR OPERATION M $ Q M @ l,,l J I N @ b h d l $ g g O
, . . , . . , n . . . . . - . .-.r..-.... .......... suue1ng we w a vvo . . .. . . - l ERABLE. l APPL AE J: MODES 1, 2, 3, and 4. , ACTION: i With only one ser.tf al cooling water loop OPERABLE, restore least two loops to OPERABLc tatus within 72 hours or be in at least STANDBY within the next 6 h rs and in COLD SHUTDOWN within the lowing 30 hours. 1 P 1 SURVE!!!ANCE REQUIREMENTS f
.. 4.7.3 At least two essential cool wate cops shall be demonstrated i F1 ERABLE:
j y
- a. At least once per 31 days y ver ing that each valve (manual, power-operated, or aut tic) serv ng safety-related equipment '
M that is not locked, led, or otherw secured in position, is in its correct pos ion. l b. At least once r 18 months during shutdown, verifying that )
, each automa valve servicing safety-related e poent actuates i
\ ect position on an SIAS test signal.
~
- to its e J
l
- c. At I st once per 18 months during shutdown, by verify that the J
1 n N' es ntial cooling water pumps start on an SIAS test signa j # d At least once per 18 months during shutdown, by verifying tha ach
- valve (manual, power-operated, or automatic) servicing safety- ted
- equipment that is locked, sealed, or otherwise secured in position, is in its correct position.
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i , 1 PLANT SYSTEMS 3/4.7.9 SNUBBERS & LIMITING CONDITION FOR OPERATION 3.7.9 All hydraulic and mechanical snubberi shall be OPERABLE. The -only l snubbers excluded from this requirement are those installed on nonsafety-
- related systems and then only if their failure or failure of the system ,
- en which they are installed, would have no adverse eff ect on any safety- l related system. APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MDDES. 1 ACTION: With one or more snubbers inoperable on any system, within 72 hours replace or - restore the inoperable snubber (s) to OPERABLE status and perfors an engineering cvaluation per Specification 4.7.99. on the attached component or declare the attached system inoperable and follow the apreopriate ACTION statement for ; ' ~ that system. j SURVEILLANCE REQUIREMENTS 4.7.9 Each snubber shall be demonstrated OPERABLE by performance of the ! following augmented inservice inspection program and the requirements of ; Specification 4.0.5.
- a. Inspection Types l As used in this specification, type of snubber shall mean snubbers 3
of the same design and manufacturer, irrespective of capacity. -
- b. Visual Inspections
, Snubbers are categorized as inaccessible or accessible during reactor <
i -' - 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 atter 4 months but within 10 months of commencing POWER i OPERATION and shall include all hydraulic and mechrnical snubbers. If all snubbers of each type are found OPERABLE dwing the first inservice visual inspection, the second inservice visual inspection of that type shall be performed at the first refueling outage. Otherwise, subsequent visual inspections of a given type shall be performed in accordance with the following schedule: I t;(-ssA*20-N, _ee-- sssT5 af4 7pg T
PLANT SYSTEMS r i
- i PP.00F&%h-- C0i i 4
SURVEILLANCE REQUIREMENTS (Continued) - t No. of Inoperable Snubbers of Each Type Subsequent visual per Inspection Period Insnection Period *# O 18 e3nths t 25%~
- 1 12 months c 25% ,
. 2 6 months 1 25% ) 3,4 124 days t 25% , 5,6,7 62 days 1 25% 8 or more 31 days 1 25% r
- c. Visual Inspection Acceptance Criteria Visual inspections shall verify that: (1) there are no visible indica-1 tions of damage or impaired OPERABILITY and (2) attachments to the
- foundation or supporting structure are secure, and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage j are secure. Snubbers which appear inoperable as a result of visual
- inspections may be detemined OPERABLE for the purpose of establishing a the next visual inspectior. interval, provided that: (1) the cause of ;
) ~- the rejection is clearly established and remedied for that particular
- snubber and for other snubbers irrespective of type on that system i
- i that may be generically suspectible; and (2) the affected snubber is
- I functionally tested in the as-found condition and detemined OPERAPLE i per Specifications 4.7.9f. When a fluid port of a hydraulic snubber
! is found to be uncevered, the snubber shall be declared inoperable and T '
- cannot be detemined OPERABLE via functional testing unless the test -
l is started with the piston in the as-found setting, extending the } piston rod in the tension mode direction. All snubbers connected to
- an inoperable common hydraulic fluid reservoir shall be counted as
4 inoperable snubbers. Snubbers which appear inoperable during an area post maintenance inspection, area walkdown, or Transient Event Inspec-tion shall not be considered inoperable for the purpose of establishing the Subsequent Visual Inspection Period provided that the cause of the inoperability is clearly established and remedied for that particular
., snuher and for the other snubbers, irrespective of type, that may be
- -' ; peu rally susceptible.
i d. Transient Event Inspection
- An inspection shall be performed of all hydraulic and mechanical snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as detemined from a review of operational data and a visual inspection of the systems j within 6 months following such an event. In addition to satisfying 4
*The inspection interval for each type of snubber on a given system shall not be lengthened more than one step at a time unless a generic problem has been 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. s I
l FThe provisions of Specification 4.0.2 are not applicable. d i u n vrew rr7- 3/4 7 t
.M%44t>-AM9S-STS .
0 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) PROSF & tM cnPY the visual inspection acceptance criteria, freedom-of-motion of
- sechanical snubbers shall be verified using at least one of the i following
- (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. f
- e. Functional Tests ,
! During the first refueling shutdown and at least once per 18 m'onths thereafter during shutdown, a representative sample of snubbers shall be tested using one of the following sample plans. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional AcMnistrator shall be notified in i
1 writing of the sample plan selected 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 *est. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.9f., an additional 10% of that type of sn"bber shall be functionally tested until no ) more failures are found or until all snubbers of that type have l.- been functionally tested; or
- 2) A representative sample of each type of snubber shall be func-p tionally tested in accordance with Figure 4.7-1. "C" is the total
- number of snubbers of a type found not meeting the accepttnce
" requirements of Specification 4.7.9f. The cumulative number of ( snubbers of a type tested is denoted ty "N". At the end of each
- day's testing, the new values of "N" knd "C" (previous 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" 3 ! j l region, testing of snubbers of that type may be terminated. J When the point plotted Ites in the " Continue Testing" region, additional snubbers of that type shall be tested until the point i falls in the " Accept" region or the " Reject" region, or all the 1 -/ ~ snubbers of that type have been tested. Testing equipment failure during functional testing may invalidate that day's test-i ing and allow that day's testing to resume anew at a later time, l providing all snubbers tested with the failed equipment during i the day of equipment failure are retested; or
- 3) An initial representative sample of 55 snubbers shall be functionally tested. For sech s u bber type which does not meet the functional test acceptance criteria, another sample of at
- Isast one-half the size of the initial sample shall be ttsted 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 from 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 {
-: , m sia,pg
( i PLANT SYSTEMS p] , r. , SURVEILLANCE REQUl#EMENTS (Continued) l I as the snubber is tested. If the point plotted falls on or j 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 , i j " 4 cept" region or all the snubbers of that type have been > tested. 1 1 The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and i reviewed before beginning the testing. The review shall ensure as far as practical that they are representative of the various configur-ations, operating envirennents, range of size, and capacity of i snJbbers of each type. Snubbers placed in the same locations as : b snubbers which failed the previous functional test shall be retested 4 at the time of the next functional test but shall not be included in j the semple plan. If during the functional testing, additional ! j sampling is requirtd due to failure of only one type of snubber, the ; ] functional testing results shs11 be reviewed at the time to determine , J _ if additional samples should be limited to the type of snubber which j[ has failed the functional testing. !? f. Functional Test Acceptance Criteria i l The snubber functional test shall verify that: a Activation (restraining action) is achieved within the specified
\
/
, 1) l range in both tension and compression; ; j 2) Snubber bleed, or release rate where required, is present in both tensi a and compression, within the specified range;
- 3) For mechanical snubbers, the force required to initiate er l maintain motion of the snubber is within the specified range j y in both directions of travel; and l 4) For snubbers specifically required not to displace under
- continuous load, the ability of the snubber to withstand load
! without displacement. : i 1 Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be j correlated to the specified parameters through established methods. I
- g. Functional Test Failaire Analysis An engineering evaluation shall be made of each fail tre to meet the
- functional test acceptance criteria to determine the cause of the '
i failure. The results of this evaluation chall 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. I l
^
.. 4-. .b 3/4 7y 7r :
1 PLANT SYs1 EMS *7 ., . .
. c. . . _ . _
, l l 6 . .s - - - - - -
= - -
SURVEILLANtfREOUIREM1NTS(Continued) ) _ _ _ _ \ For c e snubbers found inoperable, t.n engineering evaluation shall I be pet.'ornea on the cegenents to which the inoperable snubbers are i attached. The purpose of this engineering evaluation shaill be to ! determine if the cei.ponents to which the inoperable snubbers are I attached were adversely affected by the incperability of the snubbers i in order to ensure that the component remains capable of meeting the l designed service. j 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 evalu-ated and if caused by manufacturer or design deficiency all snubbers of the same type subject to the same defect shall be functionally
; tested. This testing requirement shall be independent of the require-j ments stated in Specification 4.7.9e. for snubbers not meeting the functional test acceptance criteria.
- h. Functional Testing of Repaired and Replaced Snubbers ,
Snubbers which fail the visual inspection or the functional test ' .- acceptan:e criteria shall be repaired or replaced. Replacesent snubbers and snubbers which have repairs which might affect the lE functional test result shall be tested to meet the functional test l* critaria beft,re installation in the unit. These snubbers shall L hava set the acceptance criteria subsequent to their most recent service, and the functional test must have been perforzad within x 4 12 months before being installed in the unit. l
- 1. Snubber Seal Replacement Program !
The service life of hydraulic and mechanical snubbers shall be monitored to ensure that the service life is not exceeded between i l j surveillance inspections. The maximum expected service life for ] various seals, springs, and other critical parts shall be determined j and established based on engineering information and shall be )
- -, extended or shortened based on monitored test results and failure j "5 .
history. Critical parts shall be replaced so that the maximum l service life will not be exceeded during a peri W when the snubber ! is required to be DNRABLE. The parts replacements shall be docu-mented and the documentation shall be retained in accordance with
- Specification 6.10.2. ,
l i i
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i i l l PROD? & REV!EW COPY
- )
i l 10 ~ 1 g i 8 7
? EJECT j l
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- 4 g 3
! CONTINUE / i l 2 TESTING , l 3 l , 2 / - l l ~l - - ACCEPT i 1 r
- A
! 0 10 20 30 4C 50 80 70 80 90 100 i N i i FIGURE 4.7-1 SAMPLING PLAN FOR SNUB 8ER FUNCTIONAL TEST l I .
- )
u _. I.....
. b 3 3f4 7 l i
i l
l PLANT. SYSTEMS I 3/4.7.10 SEALED SOURCE CONTAMINATION PROCF & RM COPY < LIMITING CONDITION FOR OPERATION M S O @ E l L1 A (\) $ b O Ol f E /4 E W l t a . . . - i v. . w tacn sea m source containing racToamvr 6 . . . . w . m. . . . . . ~ . . . i 100 microcuries of beta and/or gamma emitting material or 5 microcuries of 1pha emitting material shall be free of greater than or equal to 005 microcurie of removable contamination. APP CABILITY: At all times. ACTIO ] a. With a sealed source having removable contamina on in excess of the ' j- ove limit, immediately withdraw the sealed urce from use and either: : I 1. Decontaminate and repair the sealed urce, or i
--) ~~"" 2. pose of the sealed source in ccordance with Commission "
1 j Re lations.
- JJg jo. The provisi s of Specificatio 3.0.3 and 3.0.4 are not applicable.
l37 NJ UR) EILLANCE REQUIREMENTS I 4 10.1 Test Requirements Ea sea 1ed source shall be testea for leakage l\ an.1. c/or contamination by: j 4 a. The license or ! l b. Other ons specifically autho
- ed by the Commission or an i Agre t State.
l .:::The test met d shall have a detection sensitivit f at least j 0.005 mic rie per test sample. 4.7.10.2 est Frequencies - Each category of sealed sou o (excluding startup ources and fission detectors previously subjected < ore flux) shall be te ed at the frequencies described below. l
- a. Sources in use - At least once per 6 months for all sealed rces l containing radioactive material:
- 1. With a half-life greater than 30 days (excluding %drogen 3),
j and
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i PLANT SYSTEMS l-
- I ww.frp 0 p NGU L{f}}
) 3/4.7.13 SHUTDOWN COOLING SYSTEM lL-' WUI C it- Id - l l ! LIMITING CONDITION FOR OPERATION i i 3.7.13 Two independent shutdown cooling subsystems shall be OPERABLE, with l each subsystem comprised of: : l , j i
- a. One OPEAABLE low pressure safety infection pump, and ,
j J b. An independent OPERABLE flow path capable of taking suction from the l ' RCS hot leg and discharging coolant through the shutdown cooling I heat exchanger and back to the RCS through the cold leg injection lines. APPLICABILITY: MODES 1, 2, and 3. / ACTION: j a. With one shutdown cooling subsystem inoperable, restore the inoperab1'e subsystem to OPERABLE status within 72 hours or be in at laast HOT
- STANGBY within I hour, be in at least H0T SHUTDOWN within the next j 6 hours and be in COLD SHUTDOWN within the next 30 hours and continue j~ action to restore the required subsystem to OPERABLE status.
- I b. With both shutdown cooling subsysters inoperable, restore one j subsystem to OPERABLE status within I hour or be in at least HOT P STANDBY within I hour and be in HOT SHUTDOWN within the next 6 hours
! and continue action to restore the required subsystems to OPERABLE j l' status. i l
- c. With both tdown cooling subsystems inoperable and both reactor l l
coolant loops inoperable, initiate action to ttstore the required subsystems to OPERABLE status. ' l . SURVEILLANCE REQUIREMENTS 4.7.13 Each shutdown cooling subsystes shall be demonstrated OPERABLE: lg.
'e At least once per 18 months, during shutdown, by establishing .
j a. l l shutdown cooling flow from the RCS hot legs, through the shutdown { cooling heat exchangers, and returning to the RCS cold logs. 1
- b. At least once per 18 months, during shutdown, by testing the automatic and interlock action of the shutdown cooling system connections from j the RCS. The shutdown cooling syst= urtion salwe= ="11 e open- #eiR-l when RCS pressiste is greater than psia. The shutdown cooling i j systes suction valves located outs containment nhall <: lese auto - g i
- matica11y when RCS pressure is greater than @ ps' a. Tte shutoown i'
- cooling system suction valve located inside Ghtainment shall close j automatically when RCS pressure is greater than 700 psia. ,
! E M N : N N 4 T'S 3/4 7- . 8 l i ! I
2 ! 3/4.8 ELECTRICAL POWER SYSTEMS j 3/4.8.1 A.C. SOURCES hh b i i 4 OPERATIN LIMITING CONDITION FOR OPERATION k h h}} I h[ l { 7
-n3 , Ae .# S,*ynmayae g;; e :g g r m -- c =ha ' i:
i' OPERABLE: , l a. Two physically independent circuits from the offsite transmiss , network to the switchyard and two physically independent cir its ;
- from the switchyard to the onsite Class IE distribution s em, and l
- b. Two separate and independent diesel generators, each h:
l
- 1. Separate day fuel tank with a minimum level o 2.75 feet 1
(550 gallons of fuel), and i 2. A separate fuel storage system with a a num level of 80% l 1 (71,500 gallont of fuel), and : A separate fuel transfer pump. [1 ! APPLICABILIT MODES 1, 2, 3, and 4. ACTION: > - a. With eith an offsite circui or diesel generator of the above l
~
required A. 1ectrical p e zources inoperable. demonstrate the j 15 OPERABILITY of e remaini A.C. sources by performing Surveillance 4
- Requirements 4.8. .la. nd 4.8.1.1.2a.4 within I hour and at i , least once per 8 ho hereafter: restore at least two offsite Y ! circuits and two die enerators to OPERABLE status within 72 noun i ,
or be in at least ST Y within the next 6 hours and in COLD SHUTDOWN within followin 30 hours.
- b. With one offst circuit and o itsel generator of the above g'I required A.C electrical power so es inoperable, demonstrate the bd OPERABILITY f the remaining A.C. so ces by perfoming Surveillance Requirene s 4.8.1.1.la. and 4.8.1.1. 4. within 1 hour and at least o e per 6 hours thereafter; resto at least one of the l
inope d3e sources to OPERABLE status with hours ter be in at i le NOT STANDBY within the next 6 hours and COLD JHUTDOWN l w hin the following 30 hours. Restore at least offs ua circuits nd two diesel generators to OPERABLE status withi 2 hours from j 4 the time of initial loss or be in at least HOT ST within the j !a next 6 hours and in COLD SHUTDOWN within the following hours. l 1
- c. With one diesel generator inoperable in addition to ACTION or b.
I ! above, verify that: ! All required systesis, sesystaas, trains, components, and de es 1. 3 that depend on the remaining OPERABLE diesel generator as a ' l source of amargency power are also OPERABLE, and
- 2. When in QODE 1, 2, 3, or 4*, the steamrdriven auxiliary feed pump i l
- OPERABLE.
l If these conditions are not satisfied within 2 hours, be in at least j HDT STAICBY within the next 6 hours and in COLD SHUTDOWN within the ; 4 ' following 30 hours. r ^? P r - ^r '. ; t. ,_. ::; M ":- M - - " j N 1/4 0-1 l CEsSM10-nssS gg .
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l 3/4.9. REFUELING OPERATIONS 1 l 3/4.9.1 BORON CONCENTRATf0N
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LIMITING CONDITION FOR OPERATION l i J l 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling hall be maintained uniform and sufficient to ensure that the more restrictive 4 cf the following reactivity 1 conditions is met: g 1 a. Either a K,9f of 0.95 or less, or l
- b. A boron concentration of greater than or equal to 2150 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 boratior) at greater than or equal to l 40 gpm of a solution containing > 4000 ppm boron or its equivalent until li K is reduced to less than or equal to 0.95 or the boron concentration is rINoredtogreaterthanorequalto2150 ppm,whicheveristhemore .
restrictive. { SURVEILLANCE REQUIREMENTS 4.9.1.1 The. more restrictive of the above two reactivity conditions shall be detemined prior to: l l a 3 Removing or unbolting the reactor vessel head, and j < 1
~ I
- b. Withdrawal of any full-length CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel. I 4.9.1.2 The ron concentration of the Reactor Coolant System and the refueling , shall be detemined by chemical analysis at lesst once per j 77 hours. I ' i
% \ <
"The reactor shall be maintained in LODE 6 whenever fuel is in the reactor :
vessel with the reactor vessel head closure bolts less than fully ',ensioned or with the head removed. .- j cpAf h u{e7{ _ nu 7 sssato-mss-mg/g g.g
P.EFUELING OPERS.TIONS t 5 j
}f4.9.2 INSTRUMENTATION J _
LIMITING CONDITION FOR OPERATION l 3.9.2 As a minimum, two source range neutron flux monitors shall be OPEPIBLE i and operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room. i APPLICABILITY: MODE 6. ACTION:
- a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involvirg CORE ALTERATIONS or
- positive reactivity changes.
- b. With both of the above required moni', ors inoperable or no.t_ - .
operating, determine the boron conc'entration of the Reactor Coolant -
- System at least once per 12 hours.
;{* ,
SURVEILLANCE REQUIREMENTS . 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE
-by performance of: ;
- 5 a. A CHANNEL CHECK at least once per 12 hours,
- b. .A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start !
of CORE ALTERATIONS, and I
- c. A CHANNEL FUNCTIONAL TEST at least once per 7 days. ;
I
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! 6 . f 3/4 9-2 g nww wbnyb ~WAI E 1
REFL.ELING OPERATIONS 3(4.9.3 DECAY TIME ' .93 & as COM I I LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subtritical for at least nours. , L APPLICABILITY: During movement of irradiated fuel in the reactor pressure , vessel. ACTION: I With the reactor subtritical for less than tours, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. J SURVEILLANCE REQUIREttEh75
- 8 4.9.3 The reactor shall be determined to have been subtritical for at least
- hours by verification of the date and time of subtriticality prior to 4 movement of irradiated fuel in the reactor pressure vessel.
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4 c:Cn*#1o -fossi-sTs 3/4 9-3
i REFUELING OPERATIONS j
& OW 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS MITING CONDMION FOR OPERATION N 6\\bC)MN@ bWEM(({g, 1
1 ^ *
- 10. ' " _
'...., r.,m..uons snali ce in - wiiv. og suws.
] I a. The equipment door closed and held in place by a minime:n of foar bolts,
- b. A minimum of one door '.n each airlock is closed, and j c. Each penetration providing direct access from the contai ent atmosp iere j o the outside atmosphere shall be either:
1 F 1. Closed by an isolation valve, blino flange, or anual valve, o y . 2. Be apable of being closed by an OPERACLE omatic containme purg valve. - 2 N
- Al PLICABILITY: During r ALTERATIONS or movement 'tradiated fuel with n l; tt e containment.
l; AC IION:
- Wi th the requirements of the above sp iff tion not satisfied, immediately I
i st spend all operations involving CORE A ATIONS or movement of irradiated f6 el in the containment building. , l j SL RVEILLANCE REQUIREMENTS !J ~
- 4, 9.4 -Each-ef the abov equired containment building pe rations shall be .
f di termined to be eitt in its closed / isolated condition or able of being se osed by an OPER/I automatic containment purge valve within hours prior l t i the start of i at least once per 7 days during CORE ALTERATI or moveme L o ' irradiated el in the containment building by: ' I a. erifying the penetrations are in their closed / isolated condit i or b he containment purge valves per the applicab rtions of Specificu..: '** i 4
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- 7 1 -.
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-- 1 5 I LIMITING CONDITION FOR OPERATION I
Y 3.9.8.1 At least one shutdown cooling loop shall be OPERABLE and in crcr&ifon.* APPLICABILITY: MODE 6 when the water level above the top of the reactor pressure vessel flange is greater than or equal to 23 feet, a ACTION: With no shutdown cooling loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay hest load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate correctivt j action to return the required shutdown cooling loop to OPERABLE and operatint status as soon as possible~, Close all containment penetrations providing - direct access from the containment atmosphere to the outside atmosphere w'. thin - i - 4 hours. i SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one shutdown tooling loop shall be verified to be in i l _ operation and circulating reactor coolant at a flow rate of greater than or i 1
- equa14o 4000 gpa at least once per 12 hours. -
i n l 4 l l 1 ^ l l
*The shutdown cooling loop may be removed from operation for up to 1 hour per l i 8-hour period during the performance of CORE ALTERATIONS in the.yicinity
^'
of i the reactor pressure vessel hot legs n J ';; ::: :'" r: ^.. .
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REFUELING OPERATIONS 1 LOW WATER LEVEL 1
'] h LIMITING CONDITION FOR OPERATION e._
3.9.8.2 Two independent shutdown cooling loops shall be OPERABLE and at least one shutdown cooling loop shall be in operation.* APPLICABILITY: MODE 6 when the water level above the top of the reactor
- pressure vessel flange is less than 23 feet. j ACTION:
- a. With less than the required shutdown cooling loops OPERABLE, <
immediately initiate corrective action to return the required loops to OPERABLE status, or to establish greater than or equal to 23 feet. of water above the reactor pressure vessel flange, as soon as j possible. i
- b. With no shutdown cooling loop in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately d- initiate corrective action to return the required shutdown cooling ;
1 loop to operation. Close all containment penetrations providing 1 direct access from the containment atmosphere to the outside . atmosphere within 4 hours. SURVEILLANCE' REQUIREMENTS 1
- j
~ 4.9.8.2 At least one shutdown cooling loop shall be verified to be in g operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm at least once per 12 hours. !
i l i
*The shutdown cooling loop may be removed from operation for up to I hour per 8-hour period during the perforsance of CORE ALTERATIONS in the' vicinity of the reactor pressure vessel hot legs n fr' ; ;. ..'": :: * " ' ; ' erre f- 7 Fe G9sMto-Nsss-m*
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ryuELINGOPERATIONS " is enc y n rr,.-'t'pff' l 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM ,,
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) LIMITING CONDITION FOR OPERATION h hob 6 LLklDT ' Ulkd7kMy i . i 1 7.0 I .' , m w f.; . . . ' ; _ , ; . ; I . ; . . l . . . v. . ,y a wn. a s.s i e ur UrtMAblE. LITY: During CORE ALTERATIONS or movement of irradiated fuel w .. i n the containment. ACTION: With the containment pu valve isolation system inoperable close each of I the containment purge pene tions providing direct acces rom the containtnent l atmosphere to the outside atm ere. The provisions Specification 3.0.4 are not applicable. ] - 1 . .- - i, i i
- SURVEILLANCE REQU . ENTS 1 x i
4.9.9 containment purge valve isolation system shall be demonstrate OPER . withir, 72 hours prior to the start of and at least once per 7 days j _d ng CORE ALTERATIONS by verifying that containment purge valve isolation i . curs cn manup.1 initiation and on CPIAS. - l 1 i = j w u co h m 4_p g ragm b j .
- n 3/4 9-i 1
I l REFUELING OPERATIONS 1 3/4.9.10 WATER LEVEL - REACTOR VE5SEL 1 ig @f$ FUEL ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.9.10.1 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange. APPLICABILITY: During movement of fuel assemblies within the reactor pressure vossel when either the fuel assemblies being moved or the fuel assemblies seated within the reactor pressure vessel are irradiated. ACTION: ! With the requirements of the above specification not satisfied, suspend all ,
' operations involving movement of fuel assemblies within the pressure vessel. ; !I j SURVEILLANCE REQUIREMENTS .
4.9.10.1 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. ;
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F REFUELING OPERATIONS F ., nevJ CEAs .y'" { { f d; bb j L-LIMITING CONDITION FOR OPERATION 3.9.10.2 At least 23 feet of water shall be maintained over the top of the fuel seated in the reactor pressure vessel. APPLICABILITY: During movement of CEAs within the reactor pressure vessel, when the fuel assemblies seated within the reactor pressure vessel are irradiated. ACTION: With the requirements of the above specification not satisfied, suspena all operations involving movement of CEAs within the pressure vessel. E SURVEILLANCE REQUIREMENTS ' 4.9.10.2 The water level shali se 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 CEAs. e f * - - 1 i 1
. J l
N 3/4 9-N ;
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I REFU.EUNG OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL I p g{h - LIMITING CONDITION FOR OPERATION 5 % 3.9 11 At least 23 feet of water shall be maintained over the top of irra , diated fuel assemblies seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool. ACTION: With the requirement of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the feel storage areas and restore the water level to within its limit within 4 hours. 'y 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. 1 e - =
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- ocase Poot air tLEAca>F SysTm REFUELING OPERATIONS 3/4.9.12 TL':L %:L :7.. _:: !"!f/. Y:"':=:^^ TCGn l
I r,t.. I . P t @,,'ii L' ? I '- LIMITING CONDITION FOR OPERATION f* - l
.I .
i o.,. m ... ..wwwaent Tuel building essential ventilation systems sna t i be i ERABLE.
- AP ICABILITY
- Whenever irradiated fuel is in the storage pool.
) ACTIO ith one fuel building essential ventilation system i verable, fuel a. ement within the storage pool or crane operation th loads over the torage pool may proceed provided the OPERABL uel building j essen 1 ventilation systes is capable of bein owered from an . j OPERABL mergency power source. . Restore the soperable fuel _ building sential ventilation system to OP LE status within ~ 7 days or s end all operations involvin vesent of fGe7 within _ the storage p or operation of the f handling machine over the Z storage pool.
$ b. With no fuel buildin ssential v ilation systes OPERABLE, suspend ..
. all operations involvin ovene of fuel within the storage pool or I crane operation with loads the storage pool until at least one ! fuel building essential van ion system is restored to OPERABLE status, i
! I j c. The provisions of Spet ication 3.0.4 e not applicable.
l . SURVEILLANCE REQUIREMENTS
, m l
~
4.9.12- The above quired fuel building essential ventilatio ystems shall' { g be demonstrate ERABLE:
- a. least once per 31 days on a STAGGERED TEST BASIS by in isting, from the control room, flow through the HEPA filters and c oal ,
adsorbers and verifying that the system operates for at least ) 15 minutes, l i l ! b. At least once per 18 months or (1) after any struuural maintenance . on the HEPA filter or charcoal adsorber housings, or (2) following i' painting, fire, or cl M esi release in any ventilation zone i j ,
.memig with t he sys+= by: - l
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3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN hu LIMITING CONDITION FOR OPERATION i 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be l suspended for measurement of CEA worth and shutdown margin providad reactivity j
- equivalent to at least the highest estimated CEA worth is available for trip l insertion from OPERABLE CEA(s), or the reactor is subtritical by at least'the l
reactivity equivalent of the highest CEA worth. APPLICABILITY: f MODES 2 [3* e m M erT d i ACTION:
- a. With any full-length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, famedi-ately initiate and corttinue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 4000 ppm ,
i boron or its equivalent until the SHUTDOWN MRGIN required by ' i Specification 3.1.1.1 is restored. l~ '
- b. With all full-length CEAs fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and j.!.
l continue boration at greater than or equal to 40 gpa of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. l{ l I SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length and part-length CEA required either ,s partially or fully withdrawn shall be detemined at least once per 2 hours. ; l 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full ;
- insertion when tripped from at least the 50% withdrawn position within 24 hours l
prior to reducing the SHUTDOWN ERGIN to less than the limits of Specification 3.1.1.1. , F _ j O 4.10.E3 H an in MDDE 3 the reactor shall be detemined to be )( <
- If suberitical by at least the reactiv ty equivalent of the highest estimated CEA i i % worth or the reactivity equivalent of the highest estimated CEA worth is avail-
! able for trip insertion from OPERABLE CEAs at least once per 2 hours by con- . sideration of at least the following factors: l t i a. Reactor Coolant System boron concentration, I f b. CEA position, i c. Reactor Coolant System average temperature, I d. Fuel burnup based on gross thermal energy generation,
- e. Xenon concentration, and .
j
- f. sam rt cone.ntration. _.
l l
" Operation in ICDE 3 & shall be limited to 5 conseevtiver hours.
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SPECIAL TEST EXCEPTIONS
- p. . 7 1
3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT IW 4 RYiON, N D f W . t j POWER DISTRIBUTION LIMITS , , _ LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and
- b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below. ,.;.4
~
APPLICABILITY: MODES 1 and 2. ACTION: With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 are suspended, either:
- a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or 4
- b. Be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS
-m-i 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.5, 3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE require-i ment of I.C.1 (CEA Calculators) of Table 3.5-1 are suspended and shall be verified to be within the test power plateau.
4.10.2.2 The linear heat rate shall be determined to be within the limits of . Specification 3.2.1 by monitoring it continuously with the Incore Detector 4 Monitoring Systes pursuant to the requirements of Specifications 4.2.1.3 and : 3.3.3.2 during PHYSICS TESTS above 20% of RATED THERMAL POWER in which the l requi'rements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, l
. 3.2.3, 3.2.7, or the Minimum Channels 05ERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 are susp nded.
s
@ 6,N k. 0 ~'N 3/4 10-2 sm_ . . . . . _ . . . . . .
e SPECIAL TEST EXCEPTIONS 3/4.10.3 REACTOR COOLANT LOOPS I g l I LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specification 3.4.1.1 and noted requirements o'f Tables 2.2-1 and 3.3-1 may be suspended during the performance of startup PHYSICS TE5TS, provided:
- s. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and
- b. The reactor trip setpoints of the OPERABLE power level channels are set at less than or equal to 20% of RATED THERMAL POWER.
- c. Both reactor coolant loops and at least one reactor coolant pump in cach loop are in operation.
APPLICABILITY: During startup PHYSICS TESTS. , ACTION: 1p
*' With the THERMAL POWER greater than 5% of RATED THERMAL POWER or with less than '
the above required reactor coolant loops in operation and circulating reactor coolant, immediately trip the reactor. J. SURVEILLANCE REQUIREMENTS 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 startup PHYSICS TESTS. i - 4.104.2 -fach logarithmic and variable overpower level neutron flux monitoring i -v channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior
# to initiating startup PHYSICS TESTS.
4.10.3.3 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. ! I I i 1 > 9 3/4 10-3 l
i _3 SPECIAL TEST EXCEPTIONS - - j i .- . 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE i LIMITING CONDITION FOR OPERATION , 3.10.4 The requirements of Specifications 3.1.3.1,3.1.3.6and3.2.6majbe l suspended during the performance of PHYSICS TESTS to determine the isothermal i temperature coefficient, moderator temperature coefficient, and power coefficient l provided;tM "...'t' ~' ca :19::t'^a ' '> 1 =*= == 4 *+ ' d rd d:t:-i a d == ,
. . . . z g t . .e z. e :- ~
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- APPLICABILITY
- MODES 1 and 2. ;
ACTION: With any of the limits of Specification 3.2.1 being exceeded while the , , requirements of Specifications 3.1.3.1,3.1.3.6and3.2.6aresuspended,either:..g
'; a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or !
l b. Be in HOT STANDBY within 6 hours. , E ' I SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined at least once per hour during l 4 PHYSICS TESTS in which the requirements of Specifications 3.1.3.1, 3.1.3.6 l i and/or 3.2 6 are suspended and shall be verified to be within the test power i plateau. - j 4.10.4.2 The linear heat rate shall be determined to be within the limits of ! Specification 3.2.1 by monitoring it continuously with the Incore Detector 4
.- Monitoring System pursuant to the requirements of Specification 3.3.3.2
' f' . during PHYSICS TESTS above 20% of EATED THERMAL POWER in which the requirements of Specifications 3.1.3.1, 3.1.3.6 and/or 3.2.6 are suspended. hnt 3 , l - j ) I PALO VERDE - UNIT 1 3/4 10-4 l 1 l
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l - 1 l 1 i_ PRODF & g3 COM SPECI AL TEST EXCEPTIONS { j 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.5 The minimum temperature and pressure for criticality limits f Spect-l i fications 3.1.1.4 and 3.2.8 may be suspended during low temperature PHYSICS ! TESTS to a minimum temperature of 300*F and a minimum pressure of 500 psia l provided: ! a. The THERMAL POWER does not exceed 5% of RATED THERMAL POVER. l' j b. The reactor trip setpoints on the OPERABLE Variable Overpower trip , i channels are sit at i 20% of RATED THERMAL POWER, and 2,
- c. The Reactor Coolant System temperature and pressure relationship is l'
maintained within the acceptable region of operation required by Specification 3.4.8 except that the core critical line shown on ;
,- Figure 3.4-2 does not apply. '
4
, .- APPLICABILITY: MODE 2*.
if ~ ACTION: . F
- a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, l
) 1 ! immediately open the reactor trip breakers. ( [, b. With the Reactor Coolant System temperature and pressure i i relationship within the region of unacceptable operation on
- Figure 3.4-2, immediately open tne reactor trip breakers and restore the temperature-pressure relationship to within its limit within 1
30 minutes; perform the engineering evaluation required by j Specification 3.4.8.1 prior to the next reactor criticality. ! ~ ! - - SURVEILLANCE REQUIREMENTS 3
- ' 4.10.5.1 The Reactor Coolant System toeperature and pressure relationship i
shall be verified to be within the acceptable region for operation of l l Figure 3.4-2 at least once per hour. l i 4.10.5,2 The THERMAL POWER shall be determined to be i E of RATED THERMAL l POWER st least once Per hour. i 1
- 4.10.5.3 The Reactor Coolant Systes temperature shall be verified to be
! greater than or equal to 300'F atleast once per hour. \ 4.10.5.4 Each Logarithmic Power Level and Variable Overpower channel shall be l
- subjected to a CHANkEL FUNCTIONAL TEST within 12 hours prior to initiatig low l
temperature PHYSICS TESTS. i - \ l i l ) *First core only, prior to first exceeding BK RATED THEIMAL POWER. b "N 1 gf410j-(, i :
l - j SPECIAL TEST EXCEPTIONS _ _ _ . 3/4.10.6 SAFETY INJECTION TANKS ;, j . i LIMITING CONDITION FOR OPERATION l 3.10.6 The safety injection tank isolation valve requirement of j Specification 3.5.la. may be suspended during partial stroke testing of the
- low pressure safety injection check valves ($1-114, 51-124. 51-134, 51-144) q provided
1 i a. That power to the isolation valve is restored and the SIAS signal is { not overri den. 1 j b. Only one isolation valve at a time is closed during the testing for l _ no longer than I hour. ! c. ThatthevalveIskeylockedopenedwithpowerremovedbefore-thenext - _ isolation valve is closed. - i:: APPLICABILITY: l While partial stroke testing of the low pressure injection check valves during i norstl plant operation. l ACTION: 4 If the requirement of Specification 3.5.la. was suspended to perfors the ;
- Specification 3.10.6 partial stroke test and if any of the Specification 3.10.6 '
i requirements are not met during the Specification 3,10.6 partial stroke testing,
- the Limiting Condition for Operation shall revert to Specification 3.5.1 and the j _3.5.1 ACTION shall be applicable.
-7 -
] ' SURVEILLMCC REQUIREMENTS \ l 4.10.6.1 A valve alignment shall be performed within 4 hours following j completion of testing to verify that all valves operated during this testing 1 are restored to their normal positions and that power is removed to the SIT
- ' Isolation valves.
1 1 41 1 1
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i I SPECIAL TEST EXCEPTIONS fy 3/4.10 SAFETY INJECTION T
~
LIMITING CONDITION FOR OPERATION 3.10. The safety injection tank (SIT) p-essure of Specification 3.5.1d, may k be su[spended,for low temperature PHYSICS TESTS provided: l
- a. The THERMAL POWEft does not exceed 5% of RATED THERMAL POWER; a
- b. The SITS have been filled per Specification 3.5.lb. and pressurized a to 175 to 225 psig below the RCS pressure _. - -- -
254 psig; i c. All valves in the injection lines from the SITS to the RCS are open i and the S!Ts are capable of injecting into the RCS if there is a j . decrease in RCS pressure. APPLICABILITY: MODES 2
. 3 ACTION:
]
! If all the SITS do ot meet the level and pressure requirements of Specification 3.10 restore all the SITS to meet these requirements or be in HOT STANDBi d'.hin 6 hours and be in HOT SHUTDOWN within the following 6 hours.
- SURVEILLANCE REQUIREMENTS 4
.A.10 .1 The THERMAL POWER shall be determined to be less than 5% of RATED THE L-POWEit at least once per hour during low pressure PHYSICS TESTS.
q . 54.10f.2 Every 8 hours verify: ) a. All the SITS levels meet the requirements of Specification 3.5.lb. J ]
- b. All the SITS pressures meet the requirements of Specification 3.10.S.
I c. The valve alignment from the SITS to the RCS has not changed. l l l
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3/4.11 RADI0 ACTIVE EFFLUENTS l 3/4.11.1 SECONDARY SYSTEM LIQUID WASTE DISCHARGES TO ONSITE EVAPORATION PONDS CONCENTRATION i
- LIMITING CONDITION FOR OPERATION k N D ..
i f\ }
}
l Y .1 2 N ::n. M ut*- ' ._ ...;^.i.. -.^.....". .. . iww Trom secondar-
- sys liquid waste to the onsite evaporation ponds sha be limited to th
! lower t of detectability (LLD) defined as 5 x 10 7 Ci/ml for the pr' - cipal gamm itters or 1 x 10 8 pCi/m1 for I-131. l i~ APPLICABILITY: 5 1, 2, 3, and 4. l I d ACTION: j When any secondary system liq waste di arge pat concentration i- determined in accordance with th utve ance requ ments given below exceeds the specified_LLD, divert that disc e pathway the liquid radwaste system )p - without delay.. i SURVEILLANCE REQUIREMENTS n . .A 4.11.1.1.1 Radioactive liquid stes collected in the e cal Waste neutralizer
- tank shall be sampled and a yzed prior to their batchwise charge to the onsite evaporation pond accordance with the sampling and ana is program
- specified in Table 4 1.
! - 4.11.1.1.2 W the concentration of radioactive material in the chemica
~
- waste ne ~-
~
izer tank exceeding the specified LLD, samplu and analyze other l .7 , seco y system discharge pathways in accordance with the sampling and analysis
- - ran specif kd in Table 4.11-1.
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- 3/4.12 RAD 10LDGICAL ENVIRONMENTAL MONITORING 1
- t T I l 3/4.12.1 MONITORING PROGRAM %n ui k h i h _1 LIMITING CONDITION FOR OPERATION h @ .)D(d 6V'MTF d)RE ME.tJR 1 2. n . ; . . . . .. . . u . r . . . . : . .,. _ . . . . .. . .. . . . .v ,. . # a m " '- ~m+ ~
1 as specified in Table 3.12-1. l j A ICABILITY: At all times. j ACTION:
- a. W the radiological environmental monitoring progt not being 1 condu d as specified in Table 3.12-1, prepare and ubmit to the j Commiss' in the Annual Radiological Environnen Operating Report '
i required b ecification 6.9.1.7, a descriptio f the reasons for
" not conducti the progree as required and t lans for preventing l a recurrence.
a l# b. With the level of r c oactivity as the sult of plant effluents in l ::. an environmental samp%g medita at a ecified location exceeding ,
- 7 the reporting levels of 3ble 3.12- en averaged over any calendar I i quarter, prepare and subal to the onnission within 30 days, pursuant i to Specification 6.9.2, a in eport that identifies the cause(s) '
i for exceeding the limit (s) an efines the corrective actione to be F taken to reduce radioactive nts so that the pote:tial annual i
- f dose
- to A MDSER OF THE P IC is ess than the calendar year limits of Specificatio-- 3.11.1.P 3.11.2. , and 3.11.2.3. When more than i one of the radi'.. .lides n Table 3.1 are detected in the sampling i medium, this repcrt s be submitted -
concentration /0 , concentration (2) + ***> 1*0 i reporting lev G) reporting level ) - l l
~ - When radionuel s other than those in Table 3. are detected and i are the resu of plant effluents, this report sha be submitted if l
\ the potent I annual dose
- to A MDBER OF THE PUBLI s equal to or greater an the calendar year limits of Specificatio 3.11.1.2,
! 3.11.2 , and 3.11.2.3. This report is not required if he seasured i leve of radioactivity was not the result of plant eff1 s; however, l in h an event, the condition shall be reported and desc ' d in the ual Radiological Environmental Operating Report. ! l I
- c. With silk or fresh leafy vegetable samples unavailable from one i more of the sample locations required by Table 3.12-1, identify locations for e,btaining replacement samples and add than to the ra -
logical environmental monitoring program within 30 days. The specif I
*The methodology and parameters used to estimata the potential anns1 dose tod
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- FOR ,
a .- 1 SECTIONS 3.0 AND 4.0 .d '
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS em an. m - m$ w m m e j l I CEvAReo-oc,5 -sts ! X \ l
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, NOTE 5 The BASES contained in the succeeding pages sununarize !
the reasons for the specifications of Sections 3.0 and 1 4.0 but in accordance with 10 CFR 50.36 are not a part i of these Technical Specifications. 1 _ i I l i ( C655A@-foss5%
I' 3/4.0 APPLICABILITY I g 1 BASES l . . The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4. 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is j provided to delineate specifically when each specification is applicable. ,
- 3.0.2 This specificatico defines those conditions necessary to constitute
, compliance with the terms of an individual Limiting Condition for Operation '
_ and associated ACTION requirement. ) 3.0.3 This specification delineates the measures to be taken for 4 _- circumstances not directly provided for in the ACTION statements and whose i occurrence would violate the intent of a specification. For example, Specifi- "s. cation 3.6.2.1 requires two contcinment spray systems to be OPERABLE and
'- provides explicit ACTION requirements if one spray system is inoperable.
Under the terms of Specification 3.0.3, if both of the required containment spray systems are inoperable, within I hour seasures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours, in at leust HOT SHUTDOWN within the following 6 hours, and in COLD SHUTDOWN in the . subsequent 24 hours. , 3.0.4 This specification provides that entry into an OPERATIONAL MDDE or other specified applicability condition must be made with (a) the full coer plement of required systems, equipment or components OPERABLE and (b) all i ether parameters as specified in the L viting Conditions for Operation being
-met without regard for allowable deviations and out of service provisions y - contained in the ACTION stategents.
- - i The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded.
! Exceptions to this specification have been provided for a limited number l of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the j appropriate specifications. ; l . . n. - b.- . 4 N 8 3/4 0-1
! M00F &RBqC0( I BASES 4.0.1 This specification provides that surveillance activities necessary 1 to ensure the Limiting Conditions for Operation are set and will be. performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable. Provisicas for additional surveillance ' activities to be performed without regard to the applicable OPERATIONAL MODES < cr other conditions are provided in the individual surveillance requirements. Surveillance requirements for Special Test Exceptions need only be performed j when the Special Test Exception is being utilized as an exception to an i individual specification. l, 4.0.2 The provisions of this specification provide allowable tolerances for perforsing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational
< flexibility because of scheduling and perforsance considerations. The phrase i -
"at least" associated with a surveillance frequency does not negate this 1 allowable tolerance value and permits the performance of more frequent _ .
surveillance activities. _ i The tolerance values, taken either individually or consecutively over three test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond ' that obtained from the nominal specified interval. 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting
)/ '
! Conditions for Op ration. Under these criteria, equipment, systems, or l components are assumed to be OPERABLE if the associated surveillance activ- : l ities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems, or , _ components OPERABLE, when such items are found or known to be inoperable
- . although Ltill meeting the surveillance requirements.
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BASES I 4.0.4 This specification ensures that the surveillance actiEities assoc Mted with a Limiting Condition fo- Operation have been performed within the specified time interval prior to entry into an OPERATIONAL MDDE or other - applicable condition. The intent of this provision is to ensure that surveil-lance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Con.11 tion for Operation. Under the terms of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance
- activities must be performed within the stated surveillance interval prior to
' placing or returning the system or equipment into OPERABLE status. t _ 4.0.5 This specification ensures that inservice inspection of ASME Code l Cuss 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and l , 3, pumps and valves will be performed in accordance with a periodically updated l 2 version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda ) .. as required by 10 CFR 50.55a. Relief from any of the above requirements has jf- been provided in writing by the Commission and is not a part of these Te ;inical Specifications. f This specification includes a clarification of the frequencies for performing 4 the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification
' is provided to ensure consistency in surveillance intervals thoughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing i
activities. Under the terms of this specification, the more restrictive requirements 1 of the Technical Specifications take precedence over the ASME Boiler and l 7 ; Pressire )fissel Code and applicable Addenda. For example, the requirements of
- Specification 4.0.4 to perfom surveillance ar.tivities prior to entry into an ,
OPERATIONAL MDDE or other specified applicability condition takes precedence ! over the ASME Boiler and Pressure Vessel Coc'e provision which allows pumps to i 2 be tasted up to I week after return to u rmal operation. And for example, j ) the Technical Specification definition af CPERABLE does not grant a grace j
- period before a device that is not capable of perfoming its specified function j i is declared inoperable and takes precedence over the ASME Boiler and Pressure l Vessel Code provision which allows a valva to be incapable of performing its j specified function for up to 24 hours before being declared inoperable. ;
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I b d 2/4.1 REACTIVITY CONTROL SYSTEMS PROOF & RNiM COPY I BASES j 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOW MARGIN I A sufficient SHUTDOW MRGIN ensures that (1) the reactor can be made , subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within accept-able limits assuming the insertion of the regulating CEAs are within the limits of Specification 3.1.3.6, and (3) the reactor will be maintained sufficiently subtritical to preclude inadvertent criticality in the shutdown condition.
, SHUTDOW MRGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tcold. The most restrictive j condition occurs at EOL, with (cold at no load operating temperature, and is i- - associated with a postulated steam line break accident and resulting uncon- --
trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOW ERGIN of 6.0% delta k/k is required to control the reactivity transient. 1 Accordingly, the SHUTDOW MRGIN requirement is based upon this limiting condition and is consistent with the criteria used to establish the power i dependent CEA insertion limits and with the assumptions used in the FSAR i- Safety Analysis. With Teold less than or equal to 210'F, the reactivity transients resulting fros uncontrolled RCS cooldown are minimal and a 4% Ak/k SHUTDOW ERGIN requirement is set to ensure that reactivity transients resulting from an inadvertent single CEA withdrawal event are minimal. n _ i 'I 5 3/4.1.-i.3 -MODERATOR TEMPERAi~JRE C0 EFFICIENT (MTC) l The limitations on moderator temperature coefficient (NIC) are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cycle. The surveillance requirements for measurement of the NTC during each fuel cycle are adequate to confirs the NTC value since this coefficient changes slowly due principally to the reduction d in RCS boron concentration associated with fuel burnup. The confirmation that
- the measured NTC value is within its limit provides assurances that the coef-ficient will be maintained within acceptable values throughout each fuel
- cycle.
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i ll l REACTIVITY CONTROL SYSTEMS % hhh BASES l
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- 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made cr'itical with the Reactor Coolant System cold leg tesperature less than 552'F. This limitation i is required to ensure (1) the moderator temperature coefficient is within its
; analyzed temperature range, (2) the protective instrumentation is within its l
normal operating range, and (3) to ensure consistency with the FSAR safety analysis. 3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components ree,uired to perfom this function include (1) borated water sources, (2) charging pumps,
- (3) separate flow paths, and (4) an emergency power supply from OPERABLE diesel generators. The nominal capacity of each charging pump is 44 gpa at its dis- .,
charge. Up to 16 gpa of this may be diverted to the volume contro3-tank via - the RCP control bleedoff. Instrument inaccuracies and pump perfomance uncertainties.are liatted to 2 gpa yielding the 26 spa value. U-With the RCS temperature above 210'F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional 4 capability in the event an assumed failure renders one of the systems inoper- ' able. Allowable out-of-service periods ensure that minor component repair or l corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period. j The boration capability of either system is sufficient to provide a f' l SHUTDOWN MARGIN from expected operating conditions of 4% delta k/k after xenon decay and cooldown to 210'F. The maximum expected boration capability require- _ ment occurs at EOL from full power equilibrium xenon conditions and requires j ,_ ;",?^? gat 4cns of 4000 ppa borated water from either the refueling water tank
, or the spent fuel pool. ,
i ) With the RCS temperature below 210'F one injection system is acceptable without single failure consideration on the basis of the stable reactivity 4 condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection systes becomes inoperable. The mstrictions of one and only one operable = charging pumo whenever reactor coolant level is below the bottom of the pressur-izer is based on the assumptions used in the analysis of the boron dilution event. The baron capability required below 210'F is based upon providing a 4% ' delta k/k SHUTDOWN MARGIN after xenon decay and cooldown from 210'F to 120*F. This condition requires , allons of 4000 ppe borated water from either the refuelinr water tank or t spent fuel pool. 4 We - ! *% 5ea Ag51AME SA8- ,l
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REACTIVITY CONTROL SYSTEMS f t BASES l BORATION SYSTE*.S (Continued) The values of water volumes, temperatures, and boron concentrat' ion l ' in the refueling water tank are provided to ensure that the assumptioris used in the initial conditions of the LOCA safety Analysis remain valid. The OPERABILITY of one boron injection system during REFUELING ensures ; that this system is available for reactivity centrol while in MODE 6. ) With the RCS temperature below 210'F while in MODES 5 and 6, a source of borated water is required to be available for reactivity control and makeup for ; losses due to contraction and evaporation. The requirement of 33,500 gallons ' j of 4000 ppm borated water in either the refueling water tank or spent fuel pool ensures that this source is available.
~
mits on contained water volume and boron concenteu on recirculated i also ensure a een 7.0 and 8.5 '
, w within containment after a LO
! ; ton on i o ine an . ett of chloride and cau 1 j' ' s ens and components.
- 3/4.1.2.7 BORON DILUTION ALARMS The startup channel high neutron flux alarms alert the operator to an F inadvertent boron dilution. Both channels must be operating to assure detection of a boron dilution event by the high neutron flux alams. If one b or both of the alams are inoperable at any time, the bases for ACTION t l
i statements are as follows:
- a. One startup channel high neutron flux alare not operating: i l ~
_- With SD]y one startup channel high neutron flux alam OPERALLE while in ; l _ EDE 3r 4, 5, or 6, a single failure to the alare could prevent detection of boron dilution. By periodic monitoring of the RCS boron concentration
- f-j 5 by either boronometer or RCS sampling, a decrease in the boron concentra- This tion during an inadvertent th,ron dilution event will be observed.
provides alternate methods of detection of boron dilution with sufficient time for tamination of the event before complete loss of SHUTDOWN MARGIN j and return to criticality.
- b. Both startup channel high neutron flux alares not operating:
When both startup channel high neutron flux alams are inoperable, there Therefore, l d is no means of elarming on high neutron flux when suberitical. either simultarmous use of the boronneter and RCS sampling or independent l collection and analysis of two RCS samples to monitor the RCS boron con-centration provides alternate indications of inadvertent boron dilution. 1 This will allow detection with sufficient time for termination of baron j dilution befom complete loss of shutdown margin and return-to criticality. l j' l MMMM-@SS-sTS , ,p. ,
l 4 i l 1 j i y' 1 REACTIVITY CONTROL SYSTEMS 1 i PR00F &W BASES }
/
3/4.1.3 MOVABLE CONTROL ASSEMBLIES - The specifications of this section ensure that (1) acceptable power l distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) the potential effects of CEA misalignments are limited to ccceptable levels. The ACTION statements which pemit limited variations from the basic requirements are accompanied by additicral restrictions which ensure that the e-iginal design criteria are met. ! The ACTION statements applicable to a stuck or untrippable CEA, to two or more inoperable CEAs, and to a large misalignment (greater than or equal to
, - l' inches) of two or more CEAs, require a prompt shutdown of the reactor since oither of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or unteippable -
- CEA, the loss of SHUTDOWN MARGIN. '
For small misalignments (less than 19 inches) of the CEAs, there is (1) a yl small effect on the time-dependent long-term power distributions relative to those used in generating LCOs and L555 setpoints, (2) a small effect on the i
' available SHUTDOWN MARGIN, and (3) a sma".1 effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with small -
misalignments of CEAs pemits a 1-hour time interval during which attempts i may be made to restore the CEA to within its alignment requirements. T: a 1-hour time limit is suf ficient to (1) identify causes of a misaligned CEA, ,
)
(2) take appropriate corrective action to realign the CEAs, and (3) minimize l the effects of xenon redistribution. . j The CPCs provide protection to the core in the event of a large )
- _ misalignment (greater than or equal to 19 inches) of a CEA by applying
, .f , appropriate penalty factors to the calculation to account for the misaligned, ' - CEA. However, this misalignment would t.:ause distortion of the core power i distribution. This distribution may, in turn, have a significant effect on (1) the available SHUTDOWN MARGIN, (2) the time-dependent long-term power l i distributions relative to those used in generating LCOs and L555 setpoints, ; ! and (3) the ejected CEA worth used in the safety analysis. Therefore, the i ACTION statement associated with the large misalignment of a CEA requires a prompt realignment of the misaligned CEA. i The ACTION statements applicable to misaligned or inoperable CEAs include ! i i requirements to align the OPERABLE CEAs in a given group with the inoperable f CEA. Conformance with these alignment requirements briscs the core, within a j short period of time, to a configuration consistent with that assumed in generating LCO and L555 setpoints. However, extended operation wit'. CEAs l l significantly inserted in the core taay lead to puturbations in (~,) local i burnup, (2) peaking factors, and (3) available SHUTDOWN MARGIN which are more adverse than the conditions assumed to exist in the safety analyses and LCO
)
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REACTIVITY CONTROL SYSTEMS ,,,-,- an-ar , rnos, b Ett.mU e l a BASES MOVABLE CONTROL ASSDELIES (Continued) i l and LSSS setpoints detemination. Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing. Operability of at least two CEA position indicator channels is required to detemine CEA posit.as and thereby ensure compliance with the CEA alignment , l and insertion limits. Tm CEA " Full In" and " Full Out" limits provide an l additional independent mean. for detemining the CEA positions when the CEAs i l are at either their fully inserted or fully withdrawn positions. Therefore, l' the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.
" - CEA positions and OPERABILITY of the CEA position indicators are required 4 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 ( ij ~ LCOs are satisfied. The maximum CEA drop time restriction is consistent with the assumed CEA 1 drop time used in the safety analyses. Measurement with Tcold greater n an or equal to 552*F and with all reactor coolant pumps operating ensures that the i. measured drop times will be representative of insertion times experienced 5 during a reactor trip at operating conditions. Several design steps were employed to accommodate the possible CEA guide tube wear which could arise from CEA vibrations when fully withdrawn. Specifically, a programmed insertion schedule will be used to cycle the CEAs
- between the full out position (" FULL OUT" LIMIT) and 3.0 inches inserted over j J the gel gele. This cycling will distribute the possible guide tube wear j -< over a larger area, thus minimizing any effects. To accommodate this
" .prograaned insertion schedule, the fully withdrawn position was redefined, in
- some cases, to be 144.75 inchepr gre ter. {(,8 TQ9)]
1 The establishment of LSSS and LCos requires that the expected long- and short-ters behavior of the radial peaking factors be deteraired. The long- . ) ters behavior relates to the variation of the staatty-state radial peaking factors with core burnup and is affected by the amount of CEA insertion ; )l assianed, the portion of a burnup cycle over which such insertion is assumed and the expected power level variation throughout the cycle. The short-tem behavior relates to transient perturbations te *,he steady-state radial peaks : due to radial menon redistribution. The magn.'.udes of such perturbations
- depend upon the expected use of the CEAs during anticipated power reductions
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t j i _ REACTIVITY CONTROL SYSTEMS Qg1g Qf ' L__,._ _
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l IS ~ - i MOVABLE CONTROL ASSE4 LIES (Continued)
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and load maneuvering. Anal.vs re performed based on the expected mode of operation of the NSSS (bast oad maneuvering, etc.) and from these analyses CEA insertions are determined and a consistent set of radial peaking factors ' defined. The Long Tern Steady State and Short Tem Insertion Limits are deter-oined based upon %e assumed mode of operation used in the analyses and provide a means of preserving the assumptions on CEA insertions used. The limits speci-fied serve to limit the behavior of the radial peaking factors within the bounds l determined from analysis. The actions specified serve to limit the extent ofThe l radial menon redistribution effects to those accoasnodated in the analyses.
- Long and Short 'sers Insertion Limits of Specification 3.1.3.6 are specified for '
the plant which has been designed for primarily base loaded operation but which has ,the ability to accommodate a limited amount of load maneuvering.
; m T - 4.rt Tu .etion Limits of Specification 3.1.3.6 and the " -
3 j . Insertion Limits of Specification 3.1.3.b ensure that (1) the minimum 5 l ip RGIN is maintained, and (2) the potential effects of a CEA ejecti cccide are limited to acceptable levels. Long-ters operation at the ran-e effects sient Ins ion Limits is not pemitted since such operation could r distribution which could invalidate assumptions ed to deter-en the core p cine the behavio f the radial peaking factors. The PVNGS CPC an LSS systems are responsiLle for t safety and monitoring functions, respectively, the reactor core. COLSS son ors the DNB Power o help the operator main-Operating Limit (POL) and va us operating parameter tain plant operation within the siting conditions an or operation (LCO). Operat- l the event Anticipated Operational l ing within the LCO guarantees that r trip in time to prevent un-Occurrence (A00), the CPCs will prov a rea receptable f_uel damage. . .a The COLS$ reserves the Required Ov powe rgin (ROPM) to account for the A00 for the PVNGS plants. Loss of Flow (LOF) transient which i 5),he limit i the moni ng function is perfomed When the COLSS is Out of Service ( conjunction with a echnical Specification via the CPC calculation of DNBR nich restricts the reac r power sufficiently C005 Limit Line (Figure 3.2-2
; to preserve the R0PM.
with the CEAC The reduction of t CEA deviation penalties in accordan has been j (Control Element Ass involved y calculator) sensitivity reduction pro deviation perfomed. This" t setting many of the inward single not be f penalty factors 1.0. An inward CEA deviation event in effect wou CPC accompanied by w application of the CEA deviation penalty in either DNB and LHR near Heat Rate) calculations for those CEAs with the reduc penalty f ors. The protection for an inward CEA deviation event is thus j
! account for separately. _
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i l REACTIv1TY CONTROL SYSTEMS { k h. ! BASES
- .i MOVABLE CONTROL ASSEMBLIES (Continued) p ies,two
' ^
n inward CEA deviation event occurs, the current CPC T algoritrst, a static l
- pena - ors to each of the DNB and LHR calculations. e second, a xenon !
penalty facto , applied upon detection of the event. ion of time after the redistribution pen is applied linearly as in degradation f a e inward CEA deviation event-CEA drop. The expecte d is accounted for in two - for which the penalty facto "*s been i * -
- M '.. ......
. one margin degrada- 8 The R0PM reserved in COLSS it nd xenon redistribution penalties l ; ;' ... - wination of th a; n accordance with the curve in exceeds the reserved R0PM, wer reduct In addition, the p ngth CEA maneuvering is
- l .
Figure 3.1-2B is requir.e with Figure 3.1-2A to justi duction of the PLR - i restricted in accor deviation penal actors.
. idered The tech ' al specificatio~n pemits plant operation if both CEACs are ,
ino le for safety purposes after this period. l
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l ) The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA l- " Insertion Limits of Specification 3.1.3.5 ensure that 1) the minisum SHUTD06 ' i MARGIN is maintained, and 2) the potential effects of a CEA ejection accident are limited to acceptable levels. Long term operation at the Transient Insertion Limits is not pemitted since such operation could have effects on l the core power distribution which could invalidate assumptions used to deter-mine the behavior of the radial peaking factors.
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i 3/4.2 POWER DISTRIBLITION LIMITS I
- BASES "
! 3/4.2.1 LINEAR HEAT RATE - The limitation on linear heat rate ensures that in the event of a t0CA, j the peak temperature of the fuel cladding will not exceed 2200*F.
- Either o7 the two core power distribution monitoring systems, the Core i Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate
- does not exceed its limits. The COLSS perfoms this function by continuously monitoring the core power distribution and calculating a core power operating i
limit corresponding to the allowable peak linear heat rate. Reactor operation l at or below this calculated power level assures that the limits of 14.0 kW/f t ' are not exceeded. The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alars is annunciated in the event that the 3 core power exceeds the core pow" operating limit. This prt vides adequate
- - margin to the linear heat rate operating limit for nomal steady-state opera-
) tion. Nomal reactor power transients or equipment failures which do not [ is require a reactor trip may result in this core power operating limit being exceeded. In the event this occurs, COLSS alams will be annunciated. If the event which causes the COLSS limit to be exceeded results in conditions which I approach the core safety limits, a reactor trip will be initiated by the Reactor - Protective Instrumentation. The COLSS calculation of the linear heat rate ' includes appropriate penalty factors which provide, with a 95/95 probability / l confidence level, that the maximum linear heat rate calculated by COLSS is ! conservative with respect to the actual maximum linear heat rate axisting in 1 -
- the core. These penalty factors are detemined from the uncertainties l t l assoctatechrith planar radial peaking measurement, engineering heat flux i
i u uncertainty, axial densification, software algoriths modelling, computer
- processing, rod bow, and core power measurement.
1 i Parameters required to maintain the operating limit power level based on l ' linear heat rate, margin to DNB, and total core power are also annitored by the CPCs (assuming minimum este power of 20% of RATED THEIMAL POWER). The 20E i RATED THEIMAL POWER threshold is due to the neutron flux detector system being inaccurate below 20% core power. Core noice level at low power is too large to obtain usable detector readings. Therefore, in the event that the COLSS is
- not being used, operation within the lie'ts of Tipure 3.2-2 can be maintained 4
by utilizing a predetemined local pow doratty margin and a total core power i limit in the CPC trip channels. The above listad uncertainty and penalty factors plus those associated with the CPC starte test acceptance criteria are also included in the CPCs. _ t b 8 3/4 2-1 1 i
POWER DISTRIBUTION LIMITS MDF & US C0W i BASES t. 3/4.2.2 PLANAR RADIAL PEAKING FACTORS Liciting the values of the PLANAR RADIAL PEAKING FACTORS (F C) used in the l COLSS and CPCs to values equal to or greater than the measuredJLANAR RADIAL PEAKING FACTORS (F ") provides assurance that the limits calculated by.COLSS ' ~ and the CPCs remain valid. Data from the incore detectors are used for
, determining the measured PLANAR RADIAL PEAKING FACTORS. A minimum core power I
at 20% of RATED THERMAL POWER is assumed in determining the PLANAR RADIAL PEAKING FACTORS. The 20E RATED THERMAL POWER threshold is due to the neutron l ilux detector system being inaccurate below 20% core power. ) ' Core noise level at low power is too large to obtain usable detector readings. 'The periodic ' J sur.*eillance requirements for determining the measured PLANAR RADIAL PEAKING FACTDRS provides assurance that the PLANAR RADIAL PEAKING FACTORS used in COLSS and the CPCs remain valid throughout the fuel cycle. Determining the l measured PLANAR RADIAL PEAKING FACTORS after each fuel loadits prior to exceeding 70% of RATED THERNM. POWER provides additional assurance that the l core was properly loaded. > 3/4.2.3 AZIMUTHAL POWER TILT - T, - The limitations on the AZIMUTHAL POWER TILT are provided to ensure that
~
design safety margins are maintained. An AZIMUTHAL POWER TILT greater than 0.10 is not expected and if it should occur, operation is restricted to only , i those conditions required to identify the cause of the tilt. The tilt is i normally calculated by COLSS. A minimum core power of 20% of RATED THERMAL ' POWER is assmed by the CPCs in its input to COLSS for calculation of l AZIMUTHAL POWER TILT. The 20% RATED THERMAL POWER threshold is due to the - 1 neutron flux detector system being inaccurate below 20% core power. Core > noise level et low power is too large to obtain usable detector readings. The survcilla ) equirements specified when COLSS is out pf service provide an : acceptable sians of detecting the presence of a steady-state tilt. It is necessary to explicitly account for power asymmetries because the radial 1 peaking factors used in the core power distribution calculations are based on an untilted power distribution. , i
" l
- , The AZIMUTHAL POWER TILT is equal to (P;gjg/Puntilt) 4.0 who m l
AZIMUTHAL POWER TILT is measured by assuming that the ratio of the power ) at any core location in the presence of a tilt to the untilted power at the ; location is of the form: l Pgggg/Puntilt = 1 + T, g cos (8 - So) l where: j T, is the peak fractional tilt amplitude at the core periphery g is the radial normalizing factor
~6i ' s the azimuthal core location to is the azimuthal core location of maximum tilt i
a WWF?%S e sis 2-2 I . i
POWER DISTRIBUTION LINITS , ,. , p, BASES > ;
.- - l l AZIMUTHAL POWER TILT - T ,(Continued)
P tilt/Pynggjg is the ratio of i.he power at a core location in the presence i of a tilt to the power at that location with no tilt. . l . . . 1
% ia ninm. N.C TRT m ..- in sne m.. .. J ;. .: .. =
1 a .m a r ar- ---,g.~. =..m=.vna a _ , f 3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative enveltm of operating conditions consistent with the safety j analysis assumption. and which have been analytically demonstrated adequate to j maintain an acceptable minimum DNBR throughout all anticipated operational i occurrences, of which the loss of flow transient is the most limiting. Opera- ) tion of the core with a DNBR at or above this limit provides assurance that an l acceptable minimum DNBR will be maintained in the event of a loss of flow
; transient. .
] i , Either of the two core power distribution monitoring systems, the Core j - Operating Limit Supervisory System (COL 55) and the DNBR channels in the Core l Protection Calculators (CPCs), provide adequate monitoring of the core power i : - distribution and are capable of verifying that the DNBR does not violate its i limits. The COL 55 performs this function by continuously monitoring the core i i power distribution and calculating a core operating Itait corresponding to the ! allowable minimum DNBR. Reactor operation at or below this calculated power
- level assures that the limits of Figure 3.2-1 are not violated. W COLS5
](r calculation of core power operating limit based on DNBR includes appropriate penalty factors which provide, with a 95/95 probability / confidence level, that I the core power limits calculated by COLSS (based on the minimum DNBR Limit) is l l conservative with respect to the actual core power limit. These penalty factors l l are determined from the uncertainties associated with planar radial peaking l j asasurement, engineering heat flux, state parameter esasurement, software j algorithm modelling, computer processing, rod bow, and core power measurement. l- - Parameters required to maintain the margin to DNB and total core power are etso sanitored by the CPCs. Therefore, in the event that the COLS5 is not
- ., being used, operation within the limits of Figure 3.2-2 can be maintained by J "g utilizing a predetermined DNBR as a function of AXIAL SHAPE IEEX and by
! monitoring the CPC trip channels. The above listed uncertainty and penalty i factors are also included in the CPCs which assee a minimum core power of 20% i of RATED THERMAL POWER. The 20% RATED THERMAL POWER threshold is due to the ! neutron flux detector system being inaccurate below 2GK core power. Core l j noise level at low power is too large to obtain usable detector readings. ' ! h DNBR penalty factors listed in Specification 4.2.4.4 are penalties used to accommodate the effects of rod bow. The amount of rod bow in each ! assembly is dependent upon the average burnup experienced by that assembly. Fuel 1 assemblies that incur higher average burnup will experience a greater magnitude ! of rod bow. Conversely, lower burnup assemblies will experience less rod bow. j - N penalty for each batch required to compensate for rod how is determined from i a batch's maximum everage assembly burnup applied to the batch's maximum inte-1 grated planar-radial power peak. A single not penalty for CDLES and CPC is then l , determined from the penalties associated with each batch, accouRting for the off-J setting margins due to the lower radial power peaks in the higher burnup batches.
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POWER. DISTRIBUTION LIMITS ' ppli BASES t _ ._ . - 3/4.2.5 RCS FLOW RATE --i This specification is provided to ensure that the actual RCS_ total, flow rate is maintained at or above the minimum value used in the safety analyses. , 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in the safety analyses, f 3/4.2.7 AXIAL SHAPE INDEX _ ~ This specification is provided to ensure that the actual value'~of t'Ee core _ ~ ' average AXIAL SHAPE INDEX is maintained within the range of values used in the _ safety analyses. I_
?
3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the ectual value of ' pressurizer pressure is maintained within the range of values used in the safety analyses. , O
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l _ j 3/4.3 INSTRUMENTATION ,,p p**' h[tg.: ta adIU i SASES { l 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLMENTATION
- .h . . l The OPERABILITY of the reactor protective and Engineered Safety Features Actuation Systems instrumentation and bypasses ensures that (1) the associated j Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to perwit a channel to be out of service for testing i
or maintenance, and (4) sufficient system functior.a1 capability is available from diverse paran.eters. i l The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design 5 for the protection and sitigation of accident and transient conditions. The ) integrated operation of each of these systems is consistent with the assumptions l j used in the saf:ty analyses.
- _. Resporse time testing of resistance temperature devices, which are a part of the rea
- tor protective system, shall be performed by using in-situ loop l, current t'est techniques or another NRC approved method.
l Any modifications which are made to the core protection calculator soft- l f ware (including changes of algorithes and fuel cycle specific data) shall be lf,
- performed in accordance with SF" 7. r_ r- "W 'Oc '.72. . '" _.,, " c _ l
{- 4 a cru_sa g a,=;;>- -- m e g 7 m ya, =;;>-er a' r r ^f.f MRC WV\ ' approved procedure on CPC software modifications. A t The design of the Control Element Assembly calculators (CEAC) provides reactor protection in the evert one or both CEACs become inoperable. If one i CEAC is in test or inopershh,, mirication of CEA position is performed at , l 1 east every 4 hours af the second CEAC fails, the CPCs in conjunction with j _ T1 ant Tt chnise pecifications will use DNBR and LPD penalty facters and j increaser' LelR and LPD margin to restrict reactor operation to a power level
,f that -ill ensure safe operation of the plant. If the margins are not 5 maintained, a reactor trip will occur.
t T. Z i .. #i . 6 6im i. - i . . L . ; ^.: ' a...... 6...i i l lfo urement uncertainties. Therefore, the actual RCS total flow l L detersi reactor coolant pump differential pressure ntation or j by caleriastric ca does not have to be con ve1y compensated for measurement uncertainties. , An analysis was done to specif r l wel below which an addi- ! tienal power reduction is un ry even if there CEA afsalignment vith j CEACs out of service. This s the completion of the CEAC's Out of Service (005) This ] anal aproves ANPP Unit 1 Cycle 1 power capability from about 75 to 3 L
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l ._ I INSTRLMENTATION i P!!007 & RBpN a0J1 , - ! BASES j REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRLMENTATION (Continued)
- n. . . . m . < - 2 . . _ 4 ,, , o _ % ... .. .,_,.
l __ __ s 1 isalignment occurred from this power level. The power penalty fact at I would odate changes in radial peaks and one hour xenon redistri ion thai l would occur . re were a CEA misalignment with CEACs out of ser e. The , quotient of the r and the CEA misalignment Power Penait actor is the j maximum power (50% power which DNBR SAFDL violation wil ccur enn if i there is a CEA misalignment fr L conditions. Below s power. extra j thersal margin will be available to plant. Thus r CEA mi u lignment, power reduction below this limiting unne s ary. The lowest core power for a POL was ca d to be 70% of rated power. 3 this was based on the following worst C fluid itions. - _ High Temperature : 580'F Low Pressure - . 1785 psia - - - ! ASI : -3. Under raction: 0'.865 ); low : 95% of full flow i - Hioh Radial Paak - L M ( * - ' P ' ' '" * ; ^^ .' ; _ 0;~ " . , . ] _ I -- f The surveillance requirements specified for these systems ensure that the ] overall system functional capability is maintained comparable to the original i design standards. The periodic surveillance tests performed at the minimum 3 l frequencies are sufficient to demonstrate this capability. / I The measurement of response time at the specified frequencies provides J assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times ~
, indicated as not applicable. The response times in Table 3.3-2 are made up of j the time to generate the trip signal at the detector (sensor response time) and i .,
the time for the signal to int.errupt power to the CEA drive mechanism (signal l ; or trip delay time). The response times are taken from the sequence-of-events j Tables in Section 15 of CESSAR. !
1 j Response time may be demonstrated by any series of sequential, overlapping, { j or total channel test seasurements provit'ad that such tests demonstrate the total channel response time as defined. Jensor response time verification may be demonstrated by either (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times. 3/4.3.3 MDNITORING INSTRUMENTATION I a f 3/4.3.3.1 RADIATION W NITORING INSTRUMENTATION D M
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) _3/4.3.3.2 INCORE DETECTORS
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! The OPERABILITY of the incere detectors with the specified minimum comple- ' ment of equipment ensures that the measurements obtained from use,cf this system accurately represent the spatial neutron flux istribution of the reactor core. 3/4.3.3.3 SEISMIC INSTRUMENTATION
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capability PERABILITY of ' the seismic instrumentation ens 0rbs to promptly determine the .ff!T~~ and evaluate the respons featu o a setssic event { ity is required to permit compar ant to safety. This capabil-3 the design basis for th asured response to that used in l y to determ pursuant to Ap a* shutdown is required ; with the 10 CFR Part 100. The inst
.dations of Regulatory Guide 1.12. "Instrumentat consistent - h-April 1974 as identified in the PVNGS FSAR. l T
j! 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION gh gg l'. cient meteorologica ' of the meteorological instrumentation ensures ' W doses to the public as a result of lable for estinatin l a - j- materials to the atmosphere. This radioactive l j ( for initiating protective d to evaluate the need ! public and is otect the healt ety of the ! ! "Onsite recommendations of Regulatory .23 l rograms," February 1972. Wind speeds less the PH asured by the meteorological instrumentation. 3/4.3.3.5 REMOTE SHLTTDOWN SYSTEM '3 M S W td % E M S ) _ , The OPERABILITY of the remote shutdown systasfensures that sufficient i capability is available to permit safe shutdown anif maintenance of HDT STANDBY of the facility from locations outside of the control room.
-J
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is required in the event control reon habitability is lost and is consistentThis capability ) with General Design criterion 19 of 10 CFR Part 50. The parameters selected to be monitored ensure that (1) the condition of the reactor is known, (2) conditions in the RCS are known i are available for residual heat removal. (4) a source,o(3) the steam generators l f water is available for
- makeup the RCS. to the RCS, and (5) the charging system is available to sakeup water to
\ i 1 The OPERABILITY of the remote shutdown system insures that a fire will not precluds achieving safe shk h . The remote shutdown systas instrumenta-tiog control M circuits ! necessary la alfainate I effects of the fire and allow operation of instrumentation, control and power i circuits required ta achieve and maintain a safe shutdown conditten are independent the reactor. of areas This where a fire could damage systems normally used ta shutdown capabilit i[ l and Appendix R to 10 CFR 50.y is consistant with General Design Critarion 3
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. ~ _ .
B 3/4 3-3 l GSSAe90-NSSS47s
INSTRUMENTATION I FK " QM & m.M l o s BASES
- REMOTE SHUTDOWh SYSTEM (Continued) j _ _ vu .s - - -- - - - .
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l ! 3/4.3.3.6 POST-ACCIDENT MONITORING INSTRUMENTATION ! i The OPERABILITY of the post accident monitoring instrumentation ensures j that sufficient information is available on selected plant parameters to monitor j and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Ters Recommendations." '
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l _ v.w3- . . i f . The Subcooled Margin Monitor (SW ), the Heat Junction Thersocouple (NJTC), , ; and the Core Exit Thermocouples (CET) comprise the Inadequate Core Cooling (ICC) > instrumentation required by Itas II.F.2 NUREG-0737, the Post TMI-2 Action Plan. l' The function of the ICC instrumentation is to enhance the ability of the plant : operator to diagnose the approach to existance of, and recovery from ICC. j l Additionally, they aid in tracking reactor coolant inventory. These instruments are included in the Technical Specifications at the request of MRC Generic Letter 83-37. These are not required by the accident analysis, nor to bring the plant to Cold Shutdown. l 5 In the event more than four oensors in a Reactor Vessel Level channel are inoperable, repairs may only be possible during the next refueling outape. This is because the sensors are accessible only after the missile shield and reactor vessel head are removed. It is not feasible te repair a channel 2 except during a refueling outage when the missile shis1d and reactor vessel head are removed to refuel the core. If both channels are inoperable, the channels shall be restored to OPERABLE status in the nearest refueling out-age. If only one channel is inoperable, it is intented that this channel be
- restored to OPERABLE status in a refueling outage as soon as' reasonably j qssible.,
7 9 4.313. 8 FIREFETECTIONINSTRUMENTm0N
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y INSTRtMENTATION y7on -., g/d 8ASES I . 4 3 51 , a
/ ~ . _ . _
FIRE DETECTION INSTRLMENTATION (Continued) E t A^.m d ... ..~ . e ..: u
^
i , 2;; ' ' : , .m . u . t' :t - m . v j ytes will reduce the potential for damage to safety-related equipment is a tegral element in the overall facility fire protection progr . Fira ectors that are used to actuate fire suppre systees represer i a more critic important component of a plant's protection program tha n detectors that are lied solely for earl warning and notification. Consequently, the minimum r of oper fire detectors must be greater. I ! The loss of cietection ility fire suppression systees, actuated ty l fire detectors, repre a significant de ation of fire protection for ar y area. As a resul te establishment of a fire ch oatrol must be initiate: at an earlier ge than would be warranted for the of oetectors that prc - vide only ly fire warning. The establishment of freq fire patrols in 1 he baffee areas is required to provide detection capability un he inoperabl e
~
lin unentation is restored to OPERABILITY. l 2 - The fire zones listed.in Table 3.3-11, Fire Detection Instruments, or n.ww.... .o m ..m , - . .. - r m- a rm 3/4.3.3. LOOSE-PMtT DETECTION INSTRLMENTATION [ . - .G^_^.^ "_ !"' ' W 1r r: : :-t f:t ::t ' r ' : t r - .^r ' n u.. . .. " ^ - primary suffir. lent capability is ava H! loose metallic f sys. ten and avoid or sitigate damage to prima nents. The allowable out-of-service times an u rements are cons th the egu atory Guide 1.133, " Loose-Part IV.ection Progr j - p, w ey cm. . - e,e4pe.w. w . N t w annetars." May 1901. 3/4.3.3. RADIDACTIVE GASEOUS EFFLUENT IONITORING INSTRLMENTATION ~ W ve. - ,-- - ........ . , -_. -
.. . a.. a . . .a n -
7 - ntrol, A app 1Icab1 E the releas'e [ofladtoactive'natedals Ii pale ~ou0 ts e actual or potential releases of gaseous effluents. The al 'p set-
- point these instruments shall be calculated and adjus accordance with
. the methodo nd parametars in the ODCM te ensure t alarm / trip will
! % occur prior to exce. > *ne 11 sits of 10 CFR Part . This instrumentation also includes provisions o toring (and c sllir.g) the concentrations of j fe potentially explosive pas r5xtures- 005 RADWASTE SYSTEM. The OPERA-sILITY und use of this instrumentati 1 stent with the requirements of i General Design Critaria 60, 63, of Appe A to 10 CFR Part 50. l
- There are two separ radioactive gaseous effluen itoring systems: ,
the low range eff) nitors for normal plant radioactive sous effluents
- and the high r effluent monitors for post-accident plant tive gaseous i e ffluents. Iow range monitors operata at eil times until the tration l 4, e? red tvity in the effluent becomes toc 'igh A during post-accident ce tions.
gh range monitors only operate when the concentration of radioactivity (,9 71 tAmafnuna+ <=: W n 4:' t '- N 5 n ; - M+**= $ k
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j CESSA(f 0 - N%5-5T5 I i t
1 JM ' l ! PROBE &RBL #I l ' W4.3.4 41@8G CVERSPEG WoTE/TrotJ l i Y Sfo . i i 1 n - 4 . 1 l D# p i - a i - 1 I
3/4.4 REACTOR COOLANT SYNEM~ BASES pg % jj 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and I associated resctor coolant pumps in operation, and maintain DNBR above 1.231
- ) during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires j that the plant be in at least HOT STANDBY within 1 hour. ,
) In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations
- require that two loops be OPERABLE.
j In MDDE 4, and in MDDE 5 with reactor coolant loops filled, a single i reactor coolant loop or shutdown cooling loop provides sufficient heat removal I capability for. removing decay heat; but single failure considerations require that at least two loops (either shutdown cooling or RCS) be OPERABLE. Thus, l if the reactor coolant loops are not OPERABLE, this specification requires that two shutdown cooling loops be OPERABLE. - In MODE 5 with reactor coolant loops not filled, a single shutdown cooking
" loop provides sufficient heat removal capability for removing decay heat; but s' ingle failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops j ; , be OPERABLE.
g The operation of one reactor coolant pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification, and produce
~
gradual reactivity changes during boron concentration reductions in the Reactor f
- f Coolant System. A flow rate of at least 4000 gpm will circulate one equivalent Reactor Coolant System volume of 12,097 cubic feet in approximately 23 minutes.
The reactivity change rate associated with boren reductions will, therefore, ! be within the capability of operator recognition and control. l W restrictions on starting a reactor coolan g
- pump n MODES 4 and 5,1 with one or more RCS cold legs less than or equal to 'F during cooldown or )(*F i
during heatup are provided to prevent RCS pressure transients, caused by energy ; j additions free the secondary systas, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients l 3 and will not exceed the limits of Appendix G by restricting starting of thq i M RCPs to when the secondary water temperature of each steam generator is less i C than 100'F above each of the RCS cold leg temperatures. , ! O i 3/4.4.2 SAFETY VALVES l The pressurizer code safety valves operate to prevent the RCS from being l j pressortzvo atme its Estety Ltrit af 2750 psia. Each safety valve is designec to relieve a minimum of 460,000 lb per hour of saturated stens " " l eetys4wt. The relief capacity of a single safety valve is adequate to relieve K any overpressure condition which could occur during shutdown. In the event y that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. 4 0 4 h-NyMcM6. sAE !( j MsA%m=.NM-sTs
.___ -.. p 3/4 4-1 3
1
REACTOR COOLANT SYSTEM PR00F & LYEW COM BASES SAFETY VALVEL (Continued) ; During operation, all pressurizer code safety valves must' k OPERABLE to ' prevent the RCS from being pressurized above its Safety Limit of 2750 psia.
' The combined relief capacity of these valves is sufficient to limit the system pressure to within its Safety Limit of 2750 psia fol!owing a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no i'
reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., 9: : ': r direct reactor trip on the loss of turbine) and also assuming no operation f the steam dump valves. Demonstration of the safety valves' % lift CAR.d 4 *Me settings M% % CL-will occuronly during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code. 3/4.4.3 PRESSURIZER An OPERABLE pressurizer provides pressure control for the Reactor Coolant - Systen during operations with both forced reactor coolant flow and with natural-circulation flow. The minimum water level in the pressurizer assures the i pressurizer he'aters, which are required to achieve and maintain pressure control, remain covered with water to prevent fait ee, which could occur if the neaters were energized uncovered. The maximum water level in the pressurizer ensures that this parameter is saintained within the envelope of operation assumed in the safety analysis. m The maximum water level also ensures that the RCS is not a hydraulically solid systes and that a steam bubble will be provided to I i accommodate pressure surges during operation. The steam bubble also protects > the pressurizer code safety valves against water relief. The requirement to I ~ verify that on an Engineered Safety Features Actuation test signal concurrent ! with a loss-of-offsite power the pressurizer heaters are automatically shed l from the energency power sources is to ensure that the non-Class lE heaters do not reduce the reliability of or overload the emergency power source. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability to control Reactor Coolant System pressure and establish and - g maintain natural circulation. ! The auxiliary pressurizer spray is required to depressurize the RCS by cool- i ing the pressurizer steam space to permit the plant to enter shutdown cooling. The auxiliary pressurizer spray is required during those periods when normal pressurizer spray is not available, such as during natural circulatioie and during l the later stages of a normal RCS cooldown. The auxiliary pressurizer spray also i distributes boron to the pressurizer when normal pressurizer spray is not avail-able. Use of the auxiliary pressurizer spray is required during the recovery i from a steam generator tube rupture and a small loss of coolant accident. l.
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1 , \ REACTOR COOLANT SYSTEM --- - ene &o nqn-m r.w r(Jr !s F.ntil but t l< Basis i i 3/4.4.4 STEAM GENERATORS . g The surveillance requirements for inspction of the steam generato,r tubes ensure that the structural integrity of this portion of the RCS will be main-i tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection ! of steam generator tubing is essential in order to maintain surveillance of ] the conditions of the tubes in the event that there is evidence of mechanical ' damage or progressive degradation due to desi:n, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that correctiv* ; i measures can be taken. - l l b ! The plant is expected to be operated in a manner such that the seconoary { coolant will be maintained within those chemistry limits found to result in 1 negligible corrosion of the steam generator tubes. If the secondary coolant l 1_ - chemistry is not maintained within these limits, localized corrosion may likely i j; result in stress corrosion cracking. The extent of cracking during plant opera-j- : j tion would be limited by the limitation of steam generator tube leakage between l the primary coolant system and the secondary coolant system (primary-to-secondary j leakage = 0.5 gpa per steam generator). Cracks having a primary-to-secondary j( <f leakage less than this limit during operation will have an adequate margin of ) safety to withstand the inads imposed during normal operation and by postulated , accidents. Operating plants have demonstrated that primarv-to-secondary leakage i ) of 0.5 gpa per steam generator can readily be detected by 'adiation monitors of j steam generator blowdown. Leakage in excess of this limit will require plant i shutdown and an unscheduled inspection, during which the leaking tubes will be j located and plugged. ! Wastage-type defects are unlikely with proper chemistry treatment of the l , ,7 secondary-coolant. However, even if a defect should develop in service, it will i "
' be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Staan generator tube inspections of 1
operating plants have demonstrated the capability to reliably detect degradation { that has penetrated 2GK of the original tube wall thickness. ' ] Whenever the results of any steam generator tubing inservice inspection
- fall into Category C-3, these results w' 11 be promptly reported to the Commis-i sfon pursuant to Specification 6.9.1 prior to the resumption of plant gperation.
) Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional ed @-current inspection, and revision of the Technical Specifications, if i necessary. '. 'D B 3/4 4-3 l
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4 REACTOR COOLANT SYSTEM gg g gg gy l BASES l 3 /4. 4. 5 REACTOR COOLANT SYSTEM LEAKAGE kqk ' Q{ 3/4.4.5.1 LEAKAGE DETECTION SYSTEMS l vide1 b = : = u._ u a : :. w =: ..,... _ , ...:. . = ..u.. 1
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l nd detect leakage from the reactor coolant r oundary, Containment sump monitoring the sump level increase i
- prict to the sump being pumped down, a at the equivalent of I gpm leakage into the sump. Theie on systems are sendatie" af *- . l.6vry uuide 1.45, Reactor Coolant Pressure* with the recom-d a M - e:- ; n ,n ;, s on o eakage 3/4.4.5.2 OPERATIONAL LEAKAG_E t
{ _ Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can Be rRiuced - to a threshold value. A threshold value of less than I gpm is sufficiently - 1 I low to ensure early detection of additional leakage.
- 7 The 10 ppm IDENTIFIED LEAKAGE limitation provides allowances for a 1
limited amount of leakage from known sources whose presence will not interfere ) with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
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j The surveillance requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of valve ' failure and consequent intersystem LOCA. Leakage from the RCS pressure isola- t tion valves is IDENTIFIED LEAKAGE and will be considered as a portion of the j allowable . limit, i 1' The total steam generator tube leakage limit of 1 gpa for both steam generators ensures that the dosage contribution from the tube leakage will be limited to less than Part 100 guidelines for infrequent ano liatting fault fU
- events. The 1 gpa limit is consistent with the assumptions used in the anal-
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event of a main steam line rupture or under LOCA conditions. l PRESSURE BOUNDARY LEAKAGE of any angnitude may be indicative of an impend-l ing failurit of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOW.
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i n f j REACTOR COOLANT SYSTEM pmceA"s] (h t BASES i 3/4.4.6 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure tb corrosion I of the Reactor Coolant System is minimized and reduces the potential'for Reactor l Coolant System leakage or failure due to stress corrosion. Maintaining j the chemistry within the Steady State Limits provides adequate corrosion 1 protection to ensure the structural integrity of the Reactor Coolant System ] over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration
- levels in excess of the Steady St. ate Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on 5
the structural integrity of the Reactor Coolant System. The time interval p permitting continued operation within the restrictions of the Transient Limits ! provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits. The surveillance requirements provide adequate assurance t5at j - concentrations in excess of the limits will be detected in sufficient time to I
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take corrective action. i )i . l 3/4.4.7 SPECIFIC ACTIVITY !r , The limitations on the specific activity of the primary coolant ensure ' !\ that the resulting 2-hour doses at the site boundary will not exceed an appro-l priately small fraction of Part 100 limits following a steam generator tube i rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpa and a concurrent loss-of-offsite i electrical power. The values for the limits on specific activity represent '
- limits based upon a parametric evaluation by the NRC of typical site locationsjg i _These values are conservative in that specific site parameters of the " ' " A
! . site, Apch_A5 site boundary location and meteorological conditions, were not l ., considered in this evaluation. ) The ACTION statement permitting POWER OPERATION to continue for limited l time periods with the primary coolant's specific activity greatar than )l 1.0 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit ! shown on Figure 3.4-1, accosmodates possible iodine spiking phenomenon which l ] may occur following changes in THERHAL POWER. Operation with specific activity ' j levels exceeding 1.0 microcurie /gias W5L EQUIVALENT T-T31 Dut WIThin the ) limits shown on Figure 3.e+1 aust be restricted to no more than 800 hours per year (approximately 10E of the unit's yearly operating time) since the activity j levels allowed by Figure 3.4-1 inctme the 2-hour thyroid dose at the site i boundary by e facter of up to 20 following a jostulated steam gererator tube j rupture. The reporting of cumulative operating time over 500 hours in i arty 6 month consecutive period with greater than 1.0 microcurfe/ gras 00SE ! EQUIVALEWT I-131 will allow sufficient time fer Commission evaluation of the ! circtestances prior to reaching the 800-haur limit. !( i * +
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f SPECIFIC ACTIVITY (Continued) J Reducing T l eold to less than 500'F prevents the release of activity should j a steam generator tube rupture since the satur tion pressure of the primar'y coolant is below the lift pressure of the atmospheric steam relief valves. ) The surveillance requirements provide adequate assurance that excessive specific i activity levels in the primary coolant will be detected in sufficient time to l take corrective action. Infomation obtained on iodine spiking will be used j to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. i 3/4.4.8 PRESSURE / TEMPERATURE LIMITS .
" All components in the Reactor Coolant System are designed to w1UistaEd _
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the effects of cyclic loads due to system temperature and pressure changes. i These cyclic loads are introduced by nomal load transients, reactor trips. and startup and shutdown operations. The various categories of load cycles
,; usert for design purposes are provided in ChLpters 3 and 5 of the FSAR. During ii startx and shutdown, the rates of temperature and pressure changes are limited I
so as rat to exceed the lieft lines of Figure 3.4-2. This ensures that the j maximus @cified heatup and cooldown rates are consistent with the design ' q assu'ptions and sat'sfy the stress limits for cyclic operation. , During heatup, the themal gradients in the reactor vessel wall produce , Wr.a1 stresses which vary from compressive at the inner wall to tensile at the outer wall. These themal induced comp essive stresses at the inner wall l tend to alleviate the tensile stresses int;>ced by the internal pressure. .1 l At the outer wall of the vessel, those theren1 stresses are additive to 1 the pressure inducett tensile stresses. The magnitude of the thermal stresses at l l ' either location is dependent on the retn of heatup. Consequently, each heatup i ! I rate of interest must be analyzed on an individual basis for both the inner and j outer wall. i The heatup and cooldown limit curves (Figured 3.4-b ' 't are com- )( I i pesite curves which were prepad by determining h most conservative ca.se, a with either the inside or outside wall controlling, for arty heatup or cooldown ! rates of up to 100*F per hour. Tha heatup and cooldown curves were prepared ] based upon the east limiting value of the predicted adjusted reference tempera-tureattheendoftheserviceperiodindicatedonFiguref3.4-2and4,4=4. y j *
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l REACTOR COOLANT SYSTEM BASES Pang & RNEW WY PRESSURE / TEMPERATURE LIMITS (Continued) 1 RT The reactor vessel materials have been te ,.ed to determine their initial an an increasedinresultant ti$$T; fast neutron (E greaterReactor than 1 MeV) opera-the RT upon the fluence and NI. Therefore, an adjusted reference temperature, based ' I' idual element content, can ~se. predicted using
" Effects of Residual Elements on Predicted Radiat Materials."
The heatup and cooldown limit curves Figur 3.4-2 M 2." include predicted adjustments for this shift in RT ; X i j NDT at he end of the j applicable service period, as well as adjustments for possible errors in the pressure and temperature sensing instruments. ', The actual shift in RT periodicallyduringoperatiHRTofthevesselmaterialwillbeestablished by r. moving and evaiusting, in accordance with ASTM E185-73 surveillance s ar.d Appendix H of 10 CFR 50, reactor vessel material irrad the core area.pecimens installed near the inside wall of the reactor vessel in Since.the neutron spectra at the irradiation samples and { for a sample can be applied with confidence to the ad i - reactor vessel; delta RT j determined from the surveillance capsule is differe calculatIOTdelta RT for the equivalent capsule radiation exposure. NDT 4 j The pressure-teeperature limit lines shown on Figure 3.4-2 for reactor I to assure compliance with the minimum temperatur to 10 CFR Part 50. ' specimens are removed and examined to determine changes properties. m l Figure 3.4-2 based on the greater of the following:The results ( { the actual shift in reference for 4 j as determined by fapact testing, r (2) the pre icted shift in reference tagerature for the limiting weld {U, and plate as determined by RG 1.99 " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." e The maximum RT for all Reactor Coolant System pressure retaining ( i i materials, with the$1ception of the reactor pressure vessel, has been C dete sined to be 40'F. The Lowest Service Temperature limit Gy i RT ! since Article Mi-2332 (Stamer Addenda of 1972) of s Dased upon this i Sobr and Pressure Vessel Code requires the tionLowest III of the ASME Service Tem <, RT } j, prIlu+re pressure of 3125must psia. be limited to a maximum of 20% of the sys _ 3 4 , 1 ~ a : --- - i - n . w
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tor L.........., ; _.O "" y; j The number of reactor vessel irradiation survefilance specimens and the
- frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
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