ML20138H686
| ML20138H686 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/29/1997 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20138H676 | List: |
| References | |
| NUDOCS 9705070250 | |
| Download: ML20138H686 (18) | |
Text
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December 29, 1995
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3.6 PRIMARY SYSTEM 80mRY tasEs M AR kg LW 9 AGE h / ds A.
Thermal flmitatient i
The nactor vessel has been analyzed for thermal conditions encountered during heatup and cooldown operations conducted within the specified differential temperatures and rate lialts. Heatup and cooldown operations throughout plant 1tfe at unifors rates of 100*F/hr ware considered la the 1
tec:perature range of 100*F to 546*F and were shown to be-within the requirements for stress intensity and fatigue limits of Section II? of the ASME Botier and Pressure Yessel Code (1965 Edition). fae-ei..- __.
a.
4f-4eatup -"1 cle;urc reiiter. 64 30 sycl=2 Tw & sinum bein cyciss 4:54eu determined-as-ItT000 fer the-i m eter "c::c iHs
-eep::ted thet-the-reacter vessel will iMEt E.120 AeiL@.e; gggy I
- teeld" cycl?? d"$ the(n' J lifetis;&.-
~
The allowable number of reactor vessel closure bolt pre-loading cycles j
3.
Pressuriration Teaserature is 80. The allowable number of heatup and cooldown cycles during 4
bhe service lifetime is 260.
]
S.I.a Intervice Hydrestatic and Leak Tests Operating limits for the reactor vessel pressure and te:perature during normal heatup and cooldown, and during inservice hydrostatic and leak testing were established using 10 CFR 50 Appendix E, January 1992, and Appendix G, of the 1992 Addenda of Section XI of the ASME Boiler and Prassure Vessel Code. For the purpose of this analysis the reference temperature, j
RTm, of the reactor vessel material is based on the impact test data taken in accordance with' the requirements of the Code to which the reactor vassel was designed and manufactured. For the nactor vessel beltline region, a RTm of 137.2*F was calculated for 32.0 EFPY based on surveillance i
capsule 3CO', results. For the remainder of the reactor vessel, a RTm l
of +40'F was used as this is the maximum NDT temperatum (NDIT) pemitted j
by the reactor vessel purchase specification.
I Figure 3.6-1 establishes the minimum te@erature for hydrostatic and leak testing required by the ASME Boiler and Pressure Yessel Code,Section XI.
l Test pressures for inservice hydrostatic and leak testing required by ASME Section XI are a function of the testing teweratures and the component material. Accordingly, the saximum hydrostatic test pressures will be 1.1 times the operating pressure or about 1139 psig for a reactor coolant temperature greater than 100*F.
Figure 3.6-2 prevides limitations for plant heatup and cooldown when the reactor is not critical. The thernal limitations consider marimum heatup l
and cooldown rates of 100'F/hr in any one-hour period.
l Figure 3.6-3 establishes operating limits when the core is critical.
These limits include a margin of 40*F as required by 10 CFR 50 4pendix t.
Fast neutron irradiation affects the fracture toughness of the reactor vessel saterial. In order to prevent non-ductile failug~*~c types of infors;ation are needed:
l a) a relationship between the change in RTm and the accu =ulated fast neutron fluence, and Millstone Unit 1 B3/46-1 Amendment No. F, F W
- w 9705070250 970429 PDR ADOCK 05000245 P
PDR m
June 15,1992 3.7'CONTAhNMENTSYSTEMS 1
BASES containment is nomally slightly pressurized during periods of reactor operation assuring no air ir-leakage through the primary containment. However, at least once a week, the oxygen concentration will be detemined as added assurance.
l 7.
Containment Hich-Ranae Radiation Monitors 1
Thecontainmenthigh-rangeradiationmonitors(CHRRM)ensurethat adequate infomation is available to monitor and assess containment radiation levels during and following an accident. Area Radiation l
Monitor (ARM) #12 at the control rod drive removal hatch will be utilized as the preplanned alternate method of monitoring containment high-range radiation in the event that the CHRRMs are i
not available. Due to inaccuracies induced by cable bias under j
specific conditions, the CHRRMs do not meet the low end accuracy requirements specified in Regulatory Guide 1.97, " Instrumentation i
for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs During and Fo11cwing an Accident". Under conditions of j
high temperature and low radiation, the operator would not rely on the CHRRMs for indication of containment radiation to assess an l
accident. Since these inaccuracies do not have an adverse affect on i
the ability of the operator to assess accidents, the intent of 1
Regulatory Guide 1.97 is met. This capability is consistent with the recommendations of NUREG-0737, " Clarification of TMI Action Plan 3
Requirements," dated November, 1980.
1 8.
Containment Pressure Monitors The containment pressure monitor ensures additional information is available to monitor and assess the containment pressure from
-5 psig to at least 3 times containment design pressure.
This capability is consistent with the recommendations of NUREG-0737,
" Clarification of TMI Action Plan Requirements," dated November 1980.
B.
Standby Gas Treatment Systems The standby gas treatment system is designed to filter and exhaust the reactor building atmosphere,to the stack during secondary containment isolation conditions.
Both standby gas treatment system fans are designed to automatically start upon containment isolation and to maintain the reactor building pressure to the design negative pressure so i
that all leakage should be in-leakage.
Cach of the two fans has j
100 percent capacity.
romfd High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers. T h :h r:::1 th:21d indi::t: : :y:t:: 1::E tight-rr ' ' rr t"- ! ;--cc-t by-r ':te ' r th hr:::1
- d;;rk. ;
- :d : MEPt Of'ici:::y Of :t 10::t 99 p:r ::t r;;n;1 cf 00P v.. M. 6. TM 1;k r;u r.,
- rkr,
- :;1: t::t r;;;11: ;Muld indi::t Millstone Unit ]
B 3/4 7-5 Amendment No. J, M,57 con
3.7. C'ONTAINMENT SYSTEMS January 14,1997 RAtFC l
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'[: k a ge e e ova emc1ency of the HEPA filters and charcoal adsorbers. 4
[uscgr M Only one of the two standby gas treatment trains is needed to clean up the reactor building atmosphere upon containment isolation.
If one train is found to be inoperable, there is no imediate threat to the containment -
system performance, and reactor operation or refueling operation may continue while repairs are being made. During RUN, STARTUP/ HOT STANDBY and HOT SHUTDOWN, OPERABILITY of the standby gas treatment system is required.
4 Standby gas treatment system OPERABILITY is also required during COLD SHUTDOWN or REFUELING when situations exist where a significant release of fission products can be postulated, such as moving the fuel cask, irradiated fuel or other loads in containment; or when performing CORE ALTERATIONS or operations with a potential for draining the reactor vessel when the vessel contains irradiated fuel. During a REFUELING OUTAGE, when reactor coolant temperature is less than or equal to 212' F and secondary containment integrity is required, two off-site power sources (345 kV or 23 kV) and one emergency power source would provide an adequate and reliable source of power and allow diesel or gas turbine preventative maintenance. Likewise, one source of offsite power (345.kV or 23 kV) and two emergency power sources provide an adequate and reliable source of power.
~
C.
Secondary Containment The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The reactor building provides secondary containment during reactor operation, when the drywell is sealed and in service; the reactor building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling.
Because the secondary containment is an integral part of the complete containment system, secondary containment integrity is required at all times that primary containment is required. Secondary containment integrity is also required when activities having the potential of significant fission products release, such as movement of the fuel cask, irradiated fuel, or other loads in containment are performed. Administrative controls ensure that loads j
moved in containment, which may result in significant release of fission products, are evaluated to determine if secondary containment is required.
1 D.
Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary i
containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.
l Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.
I MILLSTONE UNIT 1 B3/47-6 Amendment No. 77 98 0247
4.7 CONTAINMENT SYSTEMS 00T 311%6 BASES j
A.. Primary Containment t.
i The water in the suppression chamber is used only for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a once per shift check of the temperature, volume, and differential pressure is adequate to assure that adequate heat removal capability is present.
The interior of the suppression chamber is coated with :n ;;;, paint to prevent rusting.
The inspection of the coating during-mefoe refueling outage, ;; pre;i;;;;1, ;r.;; p;r,;;r, assures the coating is intact.
En;;rt:n:: atth th': ty;: ef :::t'n; :t ft:: fe:'.ed ;: : :t'n; :t:ti:n;
-fedfeste that the 'e:pectfer '9 tere:1 f: rde urte.
fat M-~
The interior of the drywell is coated with _.r:d" tfer r::f:t::t :;::3, 5;;1;r :nd ':n; fut :r'ne.uhite.
The drywell interior surface's will be inspected during each refueling outage to determine whether the coating is deteriorating such that it could be stripped off in the event of a immfer LOCA,
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Such an event could lead to the release of debris that may affect operation of the ECCS.
The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss of i
a coolant acci. dent.
The peak drywell pressure would be about 43 psh, which would rapidly reduce to 25 psig within 30 seconds following the pipe break.
Following the pipe break, the suppression chamber pressure rises to 25 psig within 10 seconds, equalizes with drywell pressure and therefore rapidly decays with the drywell pressure decay.
tThedesinpresfureof e dryw 1 and A sorpt pi chamb f is 6 psif. Thef design eak rpte is 0.
per d atagressur of 62 p ig.
ind cated/
abov, the pressure - sponse df the frywell nd supp,r ssion ham r fol'owing y
cident'would b the same afte he contaip6ent prpfsure re/r about J secondf.
Bas d on an cp culat sponsefiscussepabove the rimar j
contai ent' pre erati i test ressurps were cposen.
Alsothedywella/
base on the p mary co tainme press erespposeand e fac tha nd p
supp ession amber f nction as a unft, the p imary onta) ent ill by
+m, dad me unit rader__th the iMividual ompon ts sgparat y.
/
90%
1.2.%J The design basis loss of coolant accident was e a ua t the primary containment maximum allowable accident leak rate of day at 43 psig.
The analys, showed that with this leak rate and standby gas treatment system filter efficiency of 90% for halogens, for particulates, and a::::'n; th: 'i;;i;; pr:d::t r;1;;;; 'r;;ti;;; ;;;ted in T:" 10^44, the maximum 4eee4 whole body ; ::'n; :!:rd drre f: thert 5 ::: trd t'e pcN (3 maximum 4e4e4 thyroid dose i; ;i;;t 125 7;; :t th: :'t: i: nd:ry :;:r en e ;cre e de et'e-tre herrt.
The rece!te-t dere t'at "er'd ^er>'
'a*
th: dur:ti;n ;f th: ;;;id:nt it th: 10 ; ;r'!*'-
-t er: ' :: ch:1; i:d, ;;d 155 r;; ;;i ;n t:t:1 thyr:fd d:::.
'hu:, t'^re
--d:::: r:;:rted cr: th:
tr' r-that ureld be ex;::ted '- th: unh:1; htC WWM R imW:a no q uM Millstone Unit 1 B 3/4 7-7
l 1
scnury 44, am l'
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4.7 CONTAIMENT SYSTEMS BASES k
ev' of desigit baj d's: ifMlan cc ent.
es dos a so i
s 'd on e ass s
. no holdi
.n the : c ayc t nt resul (
in hrec rol o
i hon p c
ts froy e primary ontai t
thro g f tp nd st ek the nvi s.
Ther's the sp led i
prima 1,nme leak ra ahd fil ffici ncy r serv and i
provi n
twe n ex e doYf-s doses a 0 CF uide ines.
l The ton duct s u e te def id in TID Q was used i te j
n o 1-lity e eered s f y fet es neluding' shielding nd iter /1ing.5
/
j allowab1 r test lea,
- eis1, ay at a ssure of e
4 i),
is vaj e for th e
cond n was ed fr the taum,
ab1)ke i
a a i eht, eak r of ounde, 5%/ day, correcte r the
/-
j ffe so ainee t ironnen acci t an test on ti s.
j In the cc e
- case, e contal t at s are int in Mul be compose fsi ma t air le of rygen, who as4mder.
t condit n the s at ph would ir or nitr t
nt cond ions.
Co ering e differe in fxt c
os on as t
eratures,,1 e r
te cor cti act gg)(ed s
.8 s*
/ etermined f 0m the 7
de co ainment tn l
Al ough the d se sicu t ns
~
~'gNt at t,e accident leak rate uld g
}
a lowed inc se s bout ay, h
guld a thyro a
dose lu given n CFRN00 d be exce AestablJshi th limit c 1.2%/ day ro ides adeq te m '
no sif f to as r i
heal a
pa'fet of ene pu c.f It is her insi y
th llowa' ' leak r ud v
'e si ic romJhe
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t.
tainme gn ue to take antage the dett n leaV-tinhte dt _ /
L > nab'
" of th structure aver / ts serv &e lifetisie./ Additional' margin to = nta + the. containment ist the *r t.: t' n dittin is achieved by establishing the allowable operational leak rate. The operational limit t
is derived by multiplying the allowable test leak rate by 0.75, thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.
The periodic retest schedule for performing Type A tests is consistent with the requirements of 10CFR50, Appendix J, paragraph III.D.
If two consecutive periodic Type A tests fail to meet acceptance criteria, 10CFR50, Appendix J, paragraph III.A.6(b) requires an increase in Type A test frequency.
However, if Type B and C leakage rates constitute the identified contributor to the failure of the two Type A tests, an exemption to the increased Type A test requirements of 10CFR50, Appendix J, paragraph III.A.6(b) may be requested by following the guidance of Information Notice No. 85-71.
C.
43 20583 eakag'e ara e ti on nt Ve i
f' Leak Rate train
/
hillstone Unit 3 8 3/4 7-8 Amendment No. 60 ears
4.7 CONTAINMENT SYSTEMS November 10,,1993 BASES The combined leakage rate for all penetrations and valves subject to Type B and C tests is limited 'to 60% of the allowable test leak rate, in accordance with 10CFR50 Appendix J.
1 l
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Millstone Unit I B 3/4 7-Ba Amendment No. 68 cii
1 NOV 0i m3 4.7 CONTAINMENT SYSTEMS l
RASES
?
The penetration and air purge piping leakage test frequency, along j
the containment leak test, which is performed at a test pressure of at i
least 43 psig (p Whenever a doubil-), is adequate to allow detection of leakage trends.
4
{
equipmenthatches,gasketedpenetration(primarycontainmenthead, and the suppression chamber access hatch) is broken j
and remade that the se,als are performing properly.the space between the The test pressure of at least 43 l psig is consistent with the accident i
analyse t -anddhe ~ maxinum preoperational leak rate test pressure.71t is e, ec".ed th 1
.o tnt-i mat's r vgive, penetr ons an seals wo 4hemaj\\$he
<ty k
j cfor vildin Howev, it is ssible at 1b(kag ihto et part d be in
~
ft \\facil y coul
' occur.
uch 1 hgp pat thatN N affee i
,si ficantl. the'40 quen'tes acc!de s-art-to-inimir 10CFR50, Appendix J requirements. personnel air lock door se Monitoring the nitrogen makeup requirements of the inerting system leaks in a very short time.provides a method of observing leak rate tr This equi from service for test and maintenance,pment must be periodically removed but this out-of-service time will be kept to a practical minimum.
Surveillance of the su of operability checks,ppression chamber-drywell vacuum b valves.
disc to open and close and to. functionally test th system.
the fact that these valves are normally closed and tests or accident conditions.
The refueling outage surveillance tests are performed to check that the the calibration of the position indication system. valve w Measuring the force during an accident. required to lift the valve assures that the valve will fu refueling outage assures that deterioration of the valve int This test interval is based on equipment quality a experience.
Millstone Unit 1 eies 5 3/4 7-9 Amendment No. JJ, 67
_ _ _ =. _ -
NOV O I Tfff 4.7 CONTAINMENT SYSTEMS
)\\
BASES i
B.
Standby Cas Treatment System and l
C.
Secondary Containment Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 7 inches of water, at the system design flow rate, will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.
Heater capability and pressure drop should be determined at least once per operating cycle to show system performance capability.
@ $5KT' C [ f4 gjt/
54 g/rI46
) l Millstone Unit 1 B 3/4 7-91 Amendment No. pp, 67 eies o
m 0
,s
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January 10, 190s p,..,,.s....no-<
i 4.7 CONTAIMEN SYSTEMS g,;fe,, rj s pec/hr./ /.,,af c
)
ggggg (festoles( 9ttei/UsWow,
( AN5%/ ASNIG. N5to - 1980]
(1N1 PLACE L6A@
j The frequency of tests and sample analysis are necessary to sboMhat the HEPA filters and charcoal adsorbers can perform as evaluated. & Tests of i
the charcoal adsorbers wit hal enated hydrocarbon refrigerant shall he perfermed in accordance with Indine removalI%5TMD360*3-i efficiency tests shall follow = ca-r "he charcoal adsorber 1989 n
efficiency test procedures should allow for the removal of one adsorber i
tray, emptying of one bed from the tray, mixing the adsorbent thoroughly A
if M
and obtaining at least two samples.t Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.
If fQ7.*.
test results a6 all adsorbent in the s{ stem shall be replaced with an adsorbent qualified - --- -- '- " - ' ^' *^a"'"--"
l jDTC edieddeM1m The replacement tray for"tieNdsorber tray removed for theM j
test should meet the same adsorbent ouality.
-h-Mh" W1 i
riN DLCATG #&.McMArt. GFFIC.Itf N Cf LE% h4 95%) {To AH51.MSMu Nso9 - 1%ci)
! :lN+tME trMTests of the HlIPA filters with D0P aerosol shall be pertorinea in 1
accordancepe n tistsNteementetty Any HEPA filters found defective shall be replaced with filters qualified d
i estesgatetseydeWe=4ette Although the SGTS desi n flow rate is 1100 i hlN SCFM, the DOP test at reduced flow rate is actual $y more sensitive MSE/496 because diffusion is the primary mechanism of small particle collection.
i i glo-1966 The lower limit for test flow rate (500 SCFM) is based on test instrument sensitivity.
All elements of the heater should be demonstrated to be functional and 8
operable during the test of heater opacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
D.
Primary containment Isalation Valves Those large pipes comprising a portion of the reactor coolant system, whose failure could result in uncovering the reactor core, are supplied with automatic isolation valves (except those lines needed for emergency core cooling system operation or containment cooling). The closure times specified in the Technical Requirements Manual (TRM)l rupture of any are adequate to prevent loss of more coolant from the circumferentia these lines outside the containment than from a steam line rupture.
Therefore, this isolation valve closure time is sufficient to prevent uncovering the core.
MILLSTONE L5(IT 1 B 3/4 7-10 Amendment No. U, 78
==
. _ -.. _. _ - - - - - ~ _ -.. _ -
l s
i t
1 TECH SPEC BASES INSERTS B 3/4 7-5 Insert
'A' HEPA filters are purchased at a 99.97% removal efficiency for particulate larger than 0 '.icron.
The HEPA in-place DOP leak test ensures filter bypass leakage is less than 1% (removal A 99%).
The leak test is required for the annual surveillance, when a HEPA filter is replaced, or after structural maintenance that could affect the HEPA or seal integrity.
This test is not required when a charcoal sample is l
taken, or charcoal trays are replaced.
The charcoal adsorber in-place halogenated hydrocarbon leak test ensures filter bypass leakage is less than 1% (removal 2 99%),
l and must be performed for the annual surveillance, when charcoal l
trays are installed, or after structural maintenance that could affect the charcoal cell or seal integrity.
This test is not applicable to HEPA filter replacement.
The laboratory analysis of used (exposed or weathered) charcoal should indicate a methyl iodide removal efficiency of at least 95% to ensure sufficient adsorption margin until the next efficiency test is performed.
Installing new charcoal with a methyl iodide removal efficiency 2 99%, in lieu of sampling the used charcoal, satisfies the surveillance requirement to perform a sample analysis, since it provides the same degree of assurance i
that charcoal efficiencies are above the efficiency limit.
B 3/4 7-9a 4 #/t/76 g',,)M Insert
'B' i
However, an administrative lim +
es been established at s5 l
inches of water to ens equate filter loading margin for 920 hours0.0106 days <br />0.256 hours <br />0.00152 weeks <br />3.5006e-4 months <br /> (200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> year plus 30 day post LOCA) of train operatio e the last performed surveillance.
B 3/4 7-10 Insert
'C' The charcoal filter air flow distribution test is required only if inlet piping, filter plenum, or outlet piping geometry is altered, or if maintenance is performed that alters flow distribution.
The flow distribution test is not applicable to the HEPA filters.
l l
1
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Docket No. 50-245 B16393 1
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4 i
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1
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4 Millstone Nuclear Power Station Unit No.1 3
]
Retvoed Bases Paaes 4
j April 1997
'3.6 ' PRIMARY SYSTEM BOUNDARY BASES l
A.
Thermal Limitations l
The reactor vessel has been analyzed for thermal conditions encountered i
during heatup and cooldown operations conducted within the specified differential temperatures and rate limits. Heatup and cooldown operations throughout plant life at uniform rates of 100*F/hr were considered in the temperature range of 100*F to 546*F and were shown to be within the requirements for stress intensity and fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code (1965 Edition). The allowable number of reactor vessel closure bolt pre-loading cycles is 80. The allowable number of heatup and cooldown cycles during the service lifetime is 260.
B.
Pressurization Ta=nerature B.I.a Inservice Hydrostatic and Leak Tests Operating limits for the reactor vessel pressure and temperature during normal heatup and cooldown, and during inservice hydrostatic and leak testing were established using 10 CFR 50 Appendix G, January 1992, and Appendix G, of the 1992 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code.
For the purpose of this analysis the reference temperature, RTuor, of the reactor vessel material is based on the impact test data taken in accordance with the requirements of the Code to which the reactor vessel was designed and manufactured. For the reactor vessel beltline region, a RTuor of 137.2*F was calculated for 32.0 EFPY based on surveillance capsule 300', results.
For the remainder of the reactor vessel, a RTuor of +40'F was used as this is the maximum NDT temperature (NDTT) permitted by the reactor vessel purchase specification.
Figure 3.6-1 establishes the minimum temperature for hydrostatic and leak testing required by the ASME Boiler and Pressure Vessel Code,Section XI.
Test pressures for inservice hydrostatic and leak testing required by ASME Section XI are a function of the testing temperatures and the component material. Accordingly, the maximum hydrostatic test pressures will be 1.1 times the operating pressure or about 1139 psig for a reactor coolant temperature greater than 100*F.
Figure 3.6-2 provides limitations for plant heatup and cooldown when the reactor is agi critical. The thermal limitations consider maximum heatup and cooldown rates of 100*F/hr in any one-hour period.
Figure 3.64 establishes operating limits when the core is critical.
These limits include a margin of 40*F as required by 10 CFR 50 Appendix G.
Fast neutron irradiation affects the fracture toughness of the reactor vessel material.
In order to prevent non-ductile failure, two types of information are needed:
a) a relationship between the change in RTuor and the accumulated fast neutron fluence, and Millstone Unit 1 B 3/4 6-1 Amendment No. 7, 77. 77, oase
3.7 CONTAHBIENT SYSTEMS I
BASES containment is normally slightly pressurized during periods of
]
reactor operation assuring no air in-leakage through the primary l
containment. However, at least once a week, the oxygen concentration will be determined as added assurance.
}
7.
Containment Hiah-Ranae Radiation Monitors i
The containment high-range radiation monitors (CHRRM) ensure that adequate information is available to monitor and assess containment radiation levels during and following an accident. Area Radiation l
Monitor (ARM) #12 at the control rod drive removal hatch will be utilized as the preplanned alternate method of monitoring containment high-range radiation in the event that the CHRRMs are a
i not available. Due to inaccuracies' induced by cable bias under j
specific conditions, the CHRRMs do not meet the low end accuracy i
requirements specified in Regulatory Guide 1.97, " Instrumentation i
for Light Water Cooled Nuclear Power Plants to Assess Plant and i
Environs During and Following an Accident". Unq conditions of l
high temperature and low radiation, the operat would not rely on i
the CHRRMs for indication of containment rad mion to assess an accident.
Since these inaccuracies do not have an adverse affect on l
the ability of the operator to assess accidents, the intent of l
Regulatory Guide 1.97 is met. This capability is consistent with 1
the recommendations of NUREG-0737, " Clarification of TMI Action Plan Requirements," dated November, 1980.
8.
Containment Pressure Monitors The containment pressure monitor ensures additional information is available to monitor and assess the containment pressure from l
-5 psig to at least 3 times containment design pressure. This J
capability is consistent with the recommendations of NUREG-0737, j
" Clarification of TMI Action Plan Requirements," dated November 1980.
t B.
Standby Gas Treatment Systems The standby gas treatment system is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions.
Both standby gas treatment system fans are designed to automatically start upon automatic containment isolation and l to maintain the reactor building pressure to the design negative pressure so that all leakage should be in-leakage.
Each of the two fans has 100 percent capacity.
High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of NILLSTONE UNIT 1 B 3/4 7-5 Amendment No. J. #, F,
02M
i 3.7 CONTAIMENT SYSTEMS BASES particulates to the environment and to prevent clogging of the iodine adsorbers. Operation of the fans significantly higher than design flow (21210 cfe) could change the removal efficiency of the HEPA filters and charcoal adsorbers. HEPA filters are purchased at a 99.97% removal efficiency for particulate larger than 0.3 micron.
The HEPA in-place DOP leak test ensures filter bypass leakage is less than 1% (removal 2 99%). The leak test is required for the annual surveillance, when a HEPA filter is replaced, or after structural maintenance that could affect the HEPA or seal integrity. This test is not required when a charcoal sample is taken, or charcoal trays are i
replaced.
The charcoal adsorber in-place halogenated hydrocarbon leak test ensures filter bypass leakage is less than 1% (removal 2 99%), and must be performed for the annual surveillance, when charcoal trays are installed, or after structural maintenance that could affect the charcoal cell or seal integrity.
This test is not applicable to HEPA filter replacement.
l The laboratory analysis of used (exposed or weathered) charcoal should indicate a methyl iodide removal efficiency of at least 95% to ensure sufficient adsorption margin until the next efficiency test is performed.
Installing new charcoal with a methyl iodide removal efficiency 2 99%, in lieu of sampling the used charcoal, satisfies the surveillance requirement to perform a sample analysis, since it provides the same degree of assurance that charcoal efficiencies are above the efficiency i
l limit.
Only one of the two standby gas treatment trains is needed to clean up the reactor building atmosphere upon containment isolation.
If one train is found to be inoperable, there is no immediate threat to the containment system performance, and reactor operation or refueling operation may continue while repairs are being made. During RUN, STARTUP/ HOT STANDBY and HOT SHUTDOWN, OPERABILITY of the standby gas treatment system is required.
Standby gas treatment system OPERABILITY is also required during COLD SHUTDOWN or REFUELING when situations exist where a significant release of fission products can be postulated, such as moving the fuel cask, irradiated fuel or other loads in containment; or when performing CORE ALTERATIONS or operations with a potential for draining the reactor vessel when the vessel contains irradiated fuel.
During a REFUELING OUTAGE, when reactor coolant temperature is less than or equal to 212' F and secondary containment integrity is required, two off-site power sources (345 kV or 23 kV) and one emergency power source would provide an adequate and reliable source of power and allow diesel or gas turbine preventative maintenance.
Likewise, one source of offsite power (345 kV or 23 kV) and two emergency power sources provide an adequate and reliable source of power.
MILLSTONE UNIT 1 B 3/4 7-6 Amendment No. pf, pp.
0299
i 3.7 CONTAINNENT SYSTENS 1
BASES C.
Secondary Containment The secondary cont'ainment is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The reactor building provides secondary containment riuring reactor operation, when the drywell is sealed and in service; the reactor 4
building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary containment is an integral part of the complete containment system, secondary containment integrity is required at all times that primary containment is required.
Secondary containment integrity is also required l
when activities having the potential of significant fission products release, such as movement of the fuel cask, irradiated fuel, or other loads i
in containment are performed. Administrative controls ensure that loads moved in containment, which may result in significant release of fission 4
j products, are evaluated to determine if secondary containment is required.
D.
Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment. Closure of one of the valves in each line would be i
sufficient to maintain the integrity of the pressure suppression system.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.
MILLSTONE UNIT 1 B 3/4 7-7 Amendment No.
0299
l 4.7 CONTAllglENT SYSTEMS BASES i
f A.
Primary Containment The water in the suppression charber is used only for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a once per shift check of the temperature, volume, and differential pressure is adequate to assure that adequate heat removal capability is present.
The interior of the suppression chamber is coated with paint to prevent rusting. The inspection of the coating during refueling outage assures the coating is intact.
The interior of the drywell is coated with paint. The drywell interior l
surfaces will be inspected during each refueling outage to determine whether the coating is deteriorating such that it could be stripped off in the event of a LOCA. Such an event could lead to the release of debris l
that may affect operation of the ECCS.
i The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss of a coolant accident. The peak drywell pressure would be about 43 psig which would rapidly reduce to 25 psig within 30 seconds following the pipe break.
Following the pipe break, the suppression chamber pressure rises to 25 psig within 10 seconds, equalizes with drywell pressure and therefore rapidly decays with the drywell pressure decay.
l The design basis loss of coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.2% day at 43 psig.
The analysis showed that with this leak rate and a standby gas treatment system filter efficiency of 90% for halogens, 90% for particulates, the maximum whole body and maximum thyroid dose are within 10CFR100 limits.
Additional margin is achieved by establishing the allowable operational leak rate. The operational limit is derived by multiplying the allowable test leak rate by 0.75, thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.
The periodic retest schedule for performing Type A tests is consistent with the requirements of 10CFR50, Appendix J, paragraph III.D.
If two consecutive periodic Type A tests fail to meet acceptance criteria, 10CFR50, Appendix J, paragraph III.A.6(b) requires an increase in Ty)e A test frequency. However, if Type B and C leakage rates constitute t1e identified contributor to the failure of the two Type A tests, an exemption to the fm used Type A test requirements of 10CFR50, Appendix J, paracjah III.A.6(b) may be requested by following the guidance of Information Notice No. 85-71.
The combined leakage rate for all penetrations and valves subject to Type B and C tests is limited to 60% of the allowable test leak rate in accordance with 10CFR50 Appendix J.
MILLSTONE UNIT 1 B 3/4 7-8 Amendment No. pp, 0299
f 4.7 CONTAllelENT SYSTEMS BASES l
t i
The penetration and air purge piping leakage test frequency, along with the containment leak tests, which is aerformed at a test pressure of at i
least 43 psig (P Whenever a doubl,)-gasketed penetration ()rimary containment head,, is adequat ow detection of leakage trends.
1 e
i i
j equipment hatches, and the suppression ciamber access hatch) determine is broken and remade, the space between the gaskets is pressurized to that the seals are performing properly. The test pressure of at least 43 psig is consistent with the accident analyses and the maximum 4
preoperational leak rate test pressure.
l Personnel air lock door seal testing is performed in accordance with 10CFR50, Appendix J requirements.
Monitoring the nitrogen makeup requirements of the inerting system provides 4
l a method of observing leak rate trends and would detect gross l
1eaks in a very short time. This equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be l
kept to a practical minimum.
1 Surveillance of the suppression chamber-drywell vacuum breakers consists of operability checks, calibration of instrumentation and inspection of the valves.
The monthly operability tests are performed to check the capability of the j
disc to open and close and to functionally test the position indication j
system. This test frequency is justified based on previous experience and j
the fact that these valves are normally closed and are only open during
)
tests or accident conditions.
The refueling outage surveillance tests are performed to check that the valve will perform properly during the accident condition and to verify e
i the calibration of the position indication system. Measuring the force required to lift the valve assures that the valve will function properly i
during an accident.
Inspection of a select number of valves during each I
refueling outage assures that deterioration of the valve internals or I
misalignment of the disc does not impair the proper operation of the valve.
l This test interval is based on equipment quality and previous equipment experience.
B.
Standby Gas Treatment System and
}
C.
4 Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 7 inches of water, at the system design flow rate, will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability, pressure drop should be i-determined at least once per operating cycle to show system performance j
capability.
)
MILLSTONE UNIT 1 8 3/4 7-9 Amendment No. pp, O299 d
,w
4.7 CONTAINMENT SYSTEMS j
BASES l
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated.
In-place leak tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI /ASME N510-1980.
Iodine removal efficiency tests shall follow ASTM D3803-1989, as well as the conditions and requirements specified in the technical specification.
The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples or removal of a test canister.
Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.
If test results indicate removal efficiency less than 95%, all adsorbent in the system shall be replaced with an adsorbent qualified to ANSI /ASME N509-1980. The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. The charcoal filter air flow distribution test is required only if inlet piping, filter plenum, or outlet piping geometry is altered, or if maintenance is performed that alters flow distribution. The flow distribution test is not applicable to the HEPA filters.
In-place leak tests of the HEPA filters with DOP aerosol shall be performed in accordance with ANSI /ASME N510-1980. Any HEPA filters found i
defective shall be replaced with filters qualified pursuant to ANSI /ASME N509-1980. Although the SGTS design flow rate is 1100 SCFM, the D0P test at reduced flow rate is actually more sensitive because diffusion is the primary mechanism of small particle collection. The lower limit for test i
flow rate (500 SCFM) is based on test instrument sensitivity.
4 All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
D.
Primary Containment Isolation Valves Those large pipes comprising a portion of the reactor coolant system, whose failure could result in uncovering the reactor core, are supplied with automatic isolation valves (except those lines needed for emergency core cooling system operation or containment coolin ).
The closure times specified in the Technical Requirements Manual (TRM are adequate to i
prevent loss of more coolant from the circumferenti 1 rupture of any of i
these lines outside the containment than from a steam line rupture.
Therefore, this isolation valve closure time is sufficient to prevent uncovering the core.
4 1
4 MILLSTONE UNIT 1 B 3/4 7-10 Amendment No. 77, 77, 02M