ML20111C329
| ML20111C329 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/04/1985 |
| From: | Edelman M CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8501090252 | |
| Download: ML20111C329 (162) | |
Text
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P.O. BOX 5000 - CLEVELAND, OHlo 44101 - TELEPHONE (216) 622-9800 - ILLUMINATING BLDG.
- 55 PUBLIC SQUARE h!E Cbb[bl_N[Ib bbEb
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Serving The Best Location in the Nation MURRAY R. EDELMAN VICE PRESIDENT NUCLEAR January 4, 1985 PY-CEI/NRR-0162 L Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Perry Nuclear Power Plant Docket No. 50-440; 50-441 Draft Technical Specifications, Rev. 1
Dear Mr. Youngblood:
As a result of ongoing reviews and design programs, Cleveland Electric has revised the Perry Technical Specifications (draft), originally submitted to you July 31, 1984. Attached are revised pages to replace those previously submitted.
These technical specifications are subject to ongoing reviews and verification programs. We look forward to working with the NRC staff in establishing technical specifications which are suitable for incorporation into an operating license.
I Very truly your,
Murray RI Edelman Vice President Nuclear Group MRE:nje Attachments cc: Jay Silberg, Esq.
John Stefano (2) f 9{)
J. Grobe S. Brown 3,
1 t
8501090252 850104 PDR ADOCK 05000440 PDR L
Perry Technical Specifications Revision 1 (RI) - Summary of Changes Page Section Number Page Section Number 3/4 12-14 Table 4.12-1 3/4 6-31a Table 3.6.4-1 (Con-6-31b tainment and Drywell 6-31c Isolation Valves) 3/4 12-1 3.12.1, action a 6-31d 6-31e 6-31g 3/4 11 Table 4.11-2, notes 6-31h (e) and (f) 6-311 6-31j 6-31k 3/4 11-4 Table 4.11-1, note c 6-31o 3/4 3-37 Item B. I.c 3/4 6-4 j.
3/4 3-33 Item B.1.c 3/4 6-34 3/4.6.5.2 6-35 3/4 6-22 3.6.3.1.b.3, 3.6.3.1.a B 3/4 6-7 B 3/4.6.5.1 3/4 6-23 4.6.3.1.c 3/4 8-12
- b. 2 and c.3, d. 2.a and d.2.b.
3/4 3-87a 4.3.7.10.1 3/4 10-2 3.10.2.d 3/4 8-13 d.2.c 3/4 3-55 Items 5.b 3/4 3-57 3/4 8-23d Table 3.8.4.2-1 3/4 3-58 8-23f 8-23k-p 3/4 3-57 Items 1.a and 1.b 3/4 3-30 Table 3.3.3-1 3/4 3-28 Tables 3.3.3.-1, item C.1.c 3-29 3.3.3.-2, 3-32 4.3.3.1-1 3-33 3-36 3/4 1-3 3.1.3.1 Action 3-37
P:ga 2 Change Summary g
Section Number g
Section Number 3/4 4-4_
3.4.1.4 3/4 11-17 Action Statement, 4.11.2.7.2 1-13 1.20 3/4 5-8 Action b, footnote **,
- 3. 5. 3.a.1,
- 3. 5. 3.a. 2 3/4 11-14 Action a, 4.11.2.4 3/4 '7-29a Table 3.7.7.5-1
-3/4 7-29b 3/4 2-6a 3.2.2 3/4 7-29c 4.2.2 3/4 3-30 Table 3.3.3-1 3/4 3-38 Table 4.3.3.1-1 3/4 2-1 4.2.1 3/4 2-9 4.2.3 3/4 9-8 4.9.6.c i
3/4 6-41 4.6.6.3.b.3, 3/4 6-42 4.6.6.3.d.4, 4.6.6.3.e,
-3/4 6-7a 4.6.1.4.b, 3/4 7-6 4.6.6.3.f 4.7.2.c.3, 4.6.1.4.a.2 3/4 7-7 4.7.2.e.3, 4.7.2.f.
4.6.1.4.c.2 3/4 7-38 4.7.2.g. 4.7.11.1.b.3, 3/4 7-39 4.7.11.1.d.4, 4.7.11.1.e, 4.7.11.1.f, 3 /4 -- 5-6
- 3. 5. 2.e. 2 4.6. 6. 3.d. 2, 4. 7. 2.e.1,
- 4. 7.11.d. 2 3/4 3-82 Action a and b 3/4 1-18 3.1.5 action a.2 3/4 6-33 4.6.5.1.b.3 3/4 6-36 4.6.5.3.b.3
- _ w-Pago 3
~,
n Change Summary 5
Pajgt Section Number Page Section Number 3/4 9-18 3.9.12.c 3/4 3-34 Table 3.3.3-2.(Item D)
~
3/4 9-18
.4.9.12.a 3/4 3-102 3.3.8 action a, 4.3.8.2.a.1.b 3/4 4-la,b,c 3/4.4.1.1 3/4 1-19 4.1.5.d.4, 4.1.5.b. 2 3/4 3-109.
3/4.3.10 3/4 1-20
~ Figure ~3.1.5-1 3/4 110 3/4 3-111-3/4 3-94 Table 4.3.7.12-1, 13/4 9-4 4.9.2.c, footnote * -
. items (1)4 and (2)3.
J/4 8-3 4.8.1.1.2.a.4, 13/4 6-3 4.6.1.2 3/4 8-8 4.8.1.1.2.f 3/4 '5-9 4.5.3.2 B 3/4 6-5'
-3/4.6.3 1-12 1.51 6-12
- 6. 6.1.a 3/4 6-27 4.6.3.4.b.1 3/4' 3-18 Table 3.3.2.-2 (items 3/4 3-4a Action 9 4c and 4d) 3/4 3-8 Table 4.3.1.1-1 items 2b and 2c.
m
P2gs 4 Change Summary g-Section Numiscr h
Section Number 3/4 8-16 3.8.3.1.a.1.b, 3/4 3-3a Table 3.3.1-1 3/4 8-18
- 3. 8. 3.1.a. 2.b,
Item 12 3.8.3.2.a.1.b,
- 3. 8. 3. 2.a. 2.b 2-2 2.1.4 Action 3/4 8-18 3.8.3.2.a.1.e 3/4 8-16 3.8.3.2.a.2.e,
3/4 3-9 Table 4.3.1.1-1 3.8.3.1.a.1.e, footnote (f), item 11
- 3. 8. 3.1.a. 2.e.
3/4 8-4 4.8.1.1.2.e.2 3/4 6-Sa Actions a,.b, and c 3/4 8-16 3.8.3.1.a.1.c, 3/4 6-5b 3/4/ 8-18 3.8.3.1.a.2.c, 3.8.3.2.a.1.c, 6-1 3.8.3.2.a.2.c footnote **
6-1 6.1.1, 6.2.2.f, 6.5.1.1, 6.8.2 1-2 1.2 6-7 6-13a 1-15 footnote #
6-2 Item f 3/4 3-12a Item 1.e 3-17 6-7 6.5.1.2, 6.5.1.5, 3-23 6-8a 6.5.1.6, 6.5.1.6, 6-8b 6.5.1.7 6-8e 3/4 11-7 3.11.1.4 6-11b 6.5.3 3/4 3-91 Item 4 6-11c 3/4 3-92 Action 103 6-13b 3/4 1-2 4.1. 2.b
TABLE 4.12-1 (Continuedj
,mU TABLE NOTATIONS 4
"This list does not mean that only these nuclides are to be considered.
Other peaks that are identifiable, together with those of the above nuclides,shallalsobeanalyzedandreportedintheAnnualRadiologjcal
~
((1" 4.O*E*Nk.
{tpu,rgtjoS3ec),fgionJg..
m, g
tu a fsy np,M k %
eqert s
Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recom-mendations of Regulatory Guide 4.13, except for specification regarding energy dependence.
Correction factors shall be provided for energy ranges not meeting the energy dependence specification.
b.3
[ fined, for hLLb ta c.Mehik.09urposes of these specification.;, as th f.c /du m*,.
c7 YDntration of radioactive material in a sample that will yiel a net con count, above system background, that will be detected with 95% robability with o 5% probability of falsely concluding that a blank o ervation represen a "real" signal.
For a parti lar measurement system, which may include ra ochemical separation:
p 4.66 s b d
LLD =
exp(-Aat)
Y V
2.22 1
E Where:
LLD is the "a priori" ower limit of d ection as defined above, as picocuries per unit ss or volume s is the standard deviati of th background counting rate or of sthe counting rate of a blank am e as appropriate, as counts per
- minute, E is the counting efficiency as ounts per disintegration, V is the sample size in u ts of mas or volume, 2.22 is the number of sintegrations p minute per picocurie, Y is the fractional adiochemical yield, w n applic:ble, A is the radioac ve decay constant for the p ticular radionuclide, and at for envir mental samples is the elapsed time b ween sample collection or end of the sample collection period, d time of qV counting Typica values of E, V, Y, and at should be used in the lculation.
RI dTNI0/CIb2 Lai 2 3/4 12 ^
/
m
F Ob G We W Oa dnn m.4ec psty,; as,.
g O
y a
O t
kO
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM
-LIMITING CONDITION FOR OPERATION
'~
3.12.1 The radiological environmental monitoring program shall-be conducted' as specified in Table 3.12-1.
APPLICABILITY:
At all times.
AC~
a.
With the-radiological environmental monitoring program not being conducted as specified in Table 3.12-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radio-logical Environmental Operating Report required by Specification g ~6.9.1.-1&, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
b.
With the level of radioactivity as the result of plant effluents'in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a
_..m W
Special Report that identifies the cause(s) for exceeding the limit (s)
~
~
and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose
- to A' MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3.
When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) concentration (2)
+ ***> 1.0 reporting level (1) reporting level (2)
When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if-the potential annual dose
- to A-MEMBER OF THE.PUBLIC is equal'to or gre.ater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3 11.2.3.
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported a'nd described in the Annual Radiological Environmental Operating Report, c.
With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, identify loca-tions for obtaining replacement samples and add them to the radiological 2
environmental monitoring program within 30 days.
The specific
-*The methodology and pararreters used to estimate the potential annual dose to a MEMBER.OF THE PUBLIC shall be indicated in this report.
i
TABLE 4.11-2 (Continued)
TABLE NOTATIONS bSampling and analysis shall also be performed following shutdown, st'artup, or a THERMAL POWER change exceedinc 15% of RATED THERMAL PO R within a al sis sk.m 44 6b P 5 M 1 ul casebb '
H (**I*d Ms wherwe*d 4 g MhMee NJm0} Eda eble p.AMyw stb 4.t e#+saf adivity 64 easJ 6y a 4*Me # 3 io y
J; d
.cSamples shall be changed at~least once per,7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing, or after removal from sampler.
Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least i
7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10._
This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
T-ittur gr:5 :=;1= :h:1' 5: tchr :t !=:t :=: pr ' d:y: 'r=
th:
- " f :nt
!= t :t r t: d:t:r'n: triti = 7:l:n n '- th: ;ntil:t'
=h:=t 'r= th: :p=t f=? ;=1==.;h:==r :p=t f=1 i: in th: :p=t fuel ;=1.
The ratio of the _ sample flow rate to the sampl'ed stream flow rate shall be known for the time period covered by each dose or dose rate calcula-i-
tion made in accordance with Specifications 3.11.2.1, 3.11.2.2, and O-3.11.2.3.
e
.f r es,,
The principal gamma emitters for whic he LLD specification applies include'the following radionuclides: Kr-87, Kr-88, Xe-133, h-122, g Xe-135, and Xe-138, in =b1: ;= r:1:== =d M -Ei, F:-59, C:-58,
/L 0, 50, in-55, M -00, !-131, C:-121, 0:-127, C:-l'1 =d C:-li' '-
.U/ i;di= =d prtini t r-1:==. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those cf the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive fluentReleaseFeport ursuant I
to Specification 6.9.1.M. NvWasik ***
L'D Gr +4* *a'W =
k ry*d
,..Q. W;di. % 'ItSt M "sadskaS Mt k rtWt*0'6 O fMit* NAN W
% g
- 4, y Wisi LLb h
vs.kts ska4 =t k ed n da.
- M d*,onkeldias.
i b.;y ;,; J w sedu, f,e. hcl-4k U-D tW" a lies indO N'*0 4
Fs 5'l, f* *88 C* '#r":
"H,1* '3'. 9I 0 I C' 4't '4 rJ.
- A - 5+
s f
CSl44.TMs bes nd me4a N al M
- " b 0
P
$ u aee* %ab' tvd *,sM hkaM"*d 4
lw;pft,)gaf,,
jg
&. # a e. % aknwa h,;+*~d*!"
t& Y r"~+ '
ugj gq k 9,4J g serg vind t 4 un IM f., #4 udik %$
< & 7 4 6,il tw4 k 'e d 5 tk. r y nel M 60 "
eum. um r,
WAO ::NOT0N N" CLEAR LNIT 2 3/4 11-11 Fl t,
O C6c fi e. wk
.s cra mewY as swn r, sym s.t.<-r.
g O
1 I
J l
l O
(
I
\\
O l
TABLE 4.11-l (Cor.tinued)
@fm TAELE NOTA 110NS
?g The principal gamma emitters f or which the LLD specification applies
- c include the following radionuclides:
Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, Cs-134,.Cs-137, Ce-141, and Ce-144.
This list does not mean that only these nuclides are to be considered.
Other gamma peaks that are identifiable, together with those of the above nuclidet, shall also be analyzed and reported in the Semiannual Radioactive Effluent Releasej hW WI6<
Report gAur ugnbto gLSci.fication 6.9.1.#alsr-(M en M,g.te 4.Aw,tm3 d ' W, compos"ite sample-is one %, u u ",'"" " 3,,, w. _
u.
. a.
.m
"' " """*,w '" "
,,, +
A in which the quantity of liquid sampled is procortional.to 'the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.
This may be accomplished through composites of grab samples obtained prior to discharge after the tanks have been recieculated.
.-Eis p ;h
'% /
7ta.av - uwer i
'I :: "TCi? ':UCL:/"
U!CT 2 3/4 11-4 p
~,
MrrMed kra f ~ s s w s'n Rgure5l.I-I.
O 09 -il ue amkr g
0 O
l DD O
O O
O TABLE 4.3.3.1-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS w
CHANNEL OPERATIONAL
.g CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH q
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED B.
DIVISION 2 TRIP SYSTEM 1.
Low Le. iv, Level 1 S
M R
1, 2, 3, 4*, 5" kk' C b -"*)
b.
S M
R 1, 2, 3, 4*, 5*
Drywell Pressure - High S
M R
1,2,3 c.
Reactor Vessel Pressure-Low 1 LPCI Pump (B) Start Time Delay d.
i Relay NA M
Q 1, 2, 3, 4*, 5*
,y e.
LPCI Pump Discharge Flow-Low 5
M R
1, 2, 3, 4*, 5*
(f.
9f visi= :;.";ur ".er. iter M
M (NA) 1, 2, 3, 4', 5*)
54 Manual Initiation NA
$M(b)y NA 1,2,3,4*,5*
J gy M
k
-/-
2.
AUTOMATIC DEPRESSURIZATION SYSTEM i
TRIP SYSTFM "B"#
a.
Low L;.; Lw, Level 1 S
M R
1, 2, 3 5.
" rf.;;ll " r: s n rc-;;i p.
S M
1, 2, 3 n
b.e: ADS Timer NA M
Q 1, 2, 3 c 4:
i Low, Level 3 S
M R(a) 1, 2, 3 A e.
LPCI Pump (B and C) Discharge Pressure-High S
M R(,)
1, 2, 3
'(M(b))(4)-
NA 1, 2, 3 i
e f.
Manual Initiation NA f q.
Nm.J m:b; +
NA M
NA i
2., 3 i
i 20 R\\
. ~ _ _ - _ _
O O
O TABLE 3.3.3-2 (Continued) g.
3.<
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE
.E TRIP FUNCTION TRIP SETPOINT VALUE H
8.
DIVISION 2 TRIP SYSTEM y
1.
RHR B AND C (LPCI MDOE) 16.T ly.3 a.
Reactor Vessel Water Level - Low Lu La,
1-(150) inches
- 1-(152) inches Level 1 1.bs n.BB D " "'" U ) b.
Drywell Pressure - High
< (1.00) psia uno
< (1.00) psig _ s9 o
( r) psig, decreasing l
i (') psig, decreasing { { fl5 i seconds c.
Reactor vessel Pressure-Low' d
d.
LPCI Pump (B) Start Time Delay Relay 5151 seconds i
e.
LPCI Pump Discharge Flow-Low 1.(Noo) gpm 1 $2so).gpa j
--- ( f.
Divisien 2 L e ? w i L..ite-2(
) (valts)
>(
) (vaits- ..2 g. Manual-Initiation NA HA k ,y U L 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"
- n. s iq.3 a.
Reactor Vessel Water Level - Low Lu L=, 1-(150) inches
- g (102) inches i
Level I b. O rf.cIl Pre;e r; ' (1.00) p;i-405 ' {1.01) pr a 16 7 i rria @(F 90,) < f D econds -(F T,} < $ Dconds bc ADS Timer .1.4} inches
- i (10.0PTnches IP l c 4.
Reactor Vessel Water Level-Low, Level 3 l aA LPCI Pump (B and C) Discharge Pressure-High > $125) psig, increasing > QHpsig, increasing l e #. Manual Initiation HA HA {.g. ew rav.t n6 NA l D
4 e i to*v -Ap At ual % w ~a A uaa ua %.a st O l l l l ['
i . CONTAINMENT SYSTEMS ~ 3/4.6.3 DEPRESSURIZATION SYSTEMS SUPPRESSION POOL LIMITING CONDITION FOR OPERATION 3.6.3.1-The suppression q001 shall be OPERA 8LE with the pool water: 6h 3 M 2M 3 and (13",i_, ft, equivalent to a a. Volume between {1-_145) ft l 1evel between X18' 0"4 and X18'6"t, and a l b. Maximum average temperature of X95)*F during 0PERATIONAL CONDITION 1 l or 2, except that the maximum average temperature may be permitted l to increase to: 1. 11051*F during testing which adds heat to,the suppression pool. l 2. 1110i*F with THERMAL POWER less than or equal to 11(% of RATED i THERMAL POWER. 3. (120)"i with the.;;ia etn.. line ieeleti;n velvs :1;nd f:11=S; : : rr. i APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 3. _ ACTION: a. With the suppression pool water level outside the above limits, 3 restore the water-level to within the limits within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours'.and in COLD SHUTDOWN within the following 24 hours. b. In OPERATIONAL CONDITION 1 or 2 with the suppression pool average water temperature greater than X954*F, restore the average tempera-ture to less than or equal to 1951*F within 24 hours or be in at '~ least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, except, as permitted above: 1. With the suppression pool average water temperature greater than fl051*F during testing which adds heat to the suppression l-pool, stop all testing which adds heat to the suppression pool and restore the average temperature to less than 1954*F within i 24 hours or be in at least HOT SHUTDOWN within the next 12 hours I and in COLD SHUTDOWN within the following 24 hours. 2. With the suppression pool average water temperature greater than: a) X95)*F for more than 24 hours and THERMAL POWER greater o than 114% of RATED THERMAL POWER, be in at least HOT i SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. b) X1104*F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode. i A - 3. Yf*h +ha ="aaression pool average water + 7 :7.t.e @ ater ! V than (120)*F. daae=~ rrn, me rus s i prn :e e u====1 to less thz aos psig within 12 hours. 4 PERRY - UNIT 1 3/4 6-22 P1
e e , _. g y, m 4.s..,. c sm_,,,,,... s.c.u.s.3_ v.._s-a w a e. aa..~ \\ s., u - .c2,
3 CONTAINMENT SYSTEMS h LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) i c. With one suppression pool water temperature instrumentation channel in any pair (s) of temperature instrumentation channels in the same sector inoperable, restore the inoperable channel (s) to OPERABLE status within 7 days or verify suppression pool water temperature to be within the limits at least once per 12 hours. d. With both suppression pool water temperature instrumentation channels i in any pair (s) of temperature instrumentation channels in the same sector inoperable, restore at least one inoperable water temperature instrumentation channel in each pair of temperature instrumentation channels in the same sector to OPERA 8LE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.6.3.1 The suppression pool shall be demonstrated OPERA 8LE: a. By verifying the suppression pool water volump to be within the limits at least once per 24 hours. I ~.Q b. At least once per 24 hours in OPERATIONAL CONDITION 1 or 2 by verifying the suppression pool average water temperature to be less than or equal to 195t*F, except: 1. At least once per 5 minutes during testing which adds heat to the suppression pool, by verifying the suppression pool average i water temperature less than or equal to 1105t*F. l 2. At least once per hour when suppression pool average water I temperature is greater than or equal to 195t*F, by verifying: a) Suppression pool average water temperature to be less than orsCoqual to 1110(*F, and pool b) THERMAL POWER to be less than or equal to 13,4% of RATED THERMAL POWER after suppression average water temperature has exceeded 195(*F for more than 24 hours. c..&. At least once per 30 minutes following a scram with suppression pool average water temperature greater than or equal to 1951*F/, by J;rifyir.; ;;.T. pre;;ier, peel seereg; aeter M ;;r;tur; 1;;; t.Dr. e7 ;Q;el te 07vN. o l l PERRY.- UNIT 1 3/4 6-23 R\\
4 4
- V.
. t.\\ '5,T. \\o i \\ % o%.% regwich 4'. AeAc&,;. w h.d.it Mek. t ac e. - aue s'..sta. duseg udde, o g. c..L o m s@ _W. Sa<m aid,Q 0 9tt.ABL L bg g=r g.c e.# w og
- CH ANNE L fuNc% gat on o n e.
.c not s (#sh.,As h e m e'~ svM N h 'TST 3 (o tws<7kt. NM CtrO oh basi ocu we t g 5 Sco ts,eaek ' -(#sgcwA M ww b se ke'.AeQ. biq oeh ce=i N ^., \\wsAt w,&c ~ val J. ose schMa.Q. -b w A, s b.0 5 q' wa -,< ~.. ~,,,w sw u u _n,& 3 c,nc ut 3 % qugi .w-b
- b. t.'<.3 CH69utt
_ Fu o c. t. w a t sse-- <s e d coun Sa n oo<ns - exe..Awg 7 &a.3s, k m m. _, w
- 3..s u
1 f SPECIAL TEST EXCEPTIONS 3/4.10.2 R00' PATTERN CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod pattern control system (RPCS) per Specification 3.1.'4.2 may be suspended by means of the individual rod position bypass switches for the following tests: j a. Shutdown margin demonstrations, Specification 4.1.1. b. Control rod scram, Specification 4.1.3.2. c. Control rod friction measurements, d. Startup Test Program with the THERMAL POWER less than (.T)% f WED-THEAM4t-PGWER.% 9.ecs low p.a.c se teo:A, { APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With the requirements of the above specification not satisfied, verify that the RPCS is OPERA 8LE per Specification 3.1.4.2. [ SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed on control rod groups by the RPCS are bypassed, verify: a. Within 8 hours prior to bypassing any sequence constraint and at i least once per 12 hours while any sequence constraint is bypassed, that movement of the control rods from X75 5 R00 DENSITY to the RPCS low power setpoint is limited to the established control rod sequence for the specified test, and i I b. Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit technical staff. O PERRY - UNIT 1 3/4 10-2 4 i L\\ l'
e 3.m. s. c. ,. ~,a.,, m,a 4 m.g ~_.,. O ce. 3. i. cm.s., s s c._,...a.x s.m.s. i l l l 6 i i {
O O. O t TABLE 3.3.6-1 g E CONTROL R00 8 LOCK INSTRUMENTATION =* MININUM APPLICABLE OPERABLE CHANNELS OPERATIONAL E TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION i '1. RDO PATTERN CONTROL SYSTEM e a. Low Power Setpoint pRut- % % Tower 2 1, 2 60 3 1 b. Ir.tc. n dict: ":I Wiif r: ! 2 1,f 60 L?etter Setpoint i 1 2. APRM i a. Flow Blased Neutron Flux - Upscale 6 1 61 b. Inoperative 6 1, 2, 5 61 c. Downscale 6 1 61 l d. Neutron Flux - Upscale, Startup 6-2, 5 61 .w s* 3. SOURCE RANGE MONITORS i "J. Detector not full in(a) +5 -3r+ 2 a. z 5 -61 u u b* % scale (b) -+ 1 -e + i -61 g } r l Inoperatig()b) -4 1 -e;c [2 41 & i c. 43 g p g., Downscale d. 2. 5 eg 4. INTEREDIATE RANGE MDNITORS a. Detector not full in 6 2, 5 61 b. Upscale 6 2, 5 61 c. Inoperative 6 2, 5 i 61 Id) l d. Downscale 6 2, 5 61 5. SCRAM DISCHARGE VOLUME l a. Water Level-High 12)r. 1, 2, 5* 62 l I b. Scr= Trip M:;. (2) (?, 2,) S* S2 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW l a. Upscale + f. 1 62 j 5_ In:;:rative 2 1 02 -- l c-(C. r:t:r) (0:x :::1:) 2 1 62-M i l V\\ 1 i
O O O TABLE 3.3.6-2 -3 4.< CONTROL ROD BLOCK INSTRilMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1. R00 PATTERN CONTROL SYSTEM H a. Low Power Setpoint -ff9)E of RATED THERMAL POWER,4 '420 + 15 *- OM of RATED THERMAL Rws_-g,p Pos.r 2o + t% o POWER
- u b.
Ir.te.- :f t.t; S:f "ithdrawal -f79)E of RATED THERMAL POWER -f74)% of RATED THERMAL POWER"
- Lisiter Setpoint lo4-o,-is l o +- oi - l'~
2. APRM a. Flow Blased Neutron Flux - Upscale < 0.66 W + T42 N" < 0.66 W + 145 58 b. Inoperative NA NA c. Downscale 1 15 N of RATED THERMAL POWER 1 13)% of RATED THERMAL POWER d. Neutron Flux - Upscale Startup 1 112 N of RATED THERMAL POWER 1.T14 M of RATED THERMAL POWER S 3. SOURCE RANGE MONITORS -Y a. Detector not full in MA g 5 NA ;.6 5 5 b. Upscale < fe x.'10 1 cps <9x10ycp, c. Inoperative HA NA d. Downscale 1 13k cps 1 ft)i,gcps 4. INTERMEDIATE RANGE MONITORS a. Detector not full in NA NA b. Upscale 1 4108/125K division of full 1 4110/1251 division of full scale scale c. Inoperative NA NA i d. Downscale 1 (5/1251 division of full 1 13/125i division of full scale scale 5. SCRAM DISCHARGE VOLUME wm n.85 i a. Water Level-High h(22.5) inches < f34) inches l h I -._..7 _,7. 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW j a. . upscale <.4108 N of rated flow < 11115 of rated flow b. !.gr;th; h HA c. (C_ ;:reter) (0:x ; cele) '_ (-10)% -flow-deviation 1 (11)E fl~e &"lation-- l "The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). -The-trip-settir.g of this f r.ctiee e st be e.einteiaed in ecceide.ae with ',g eificetien 3.2.2. g l ~ 14 5'e t e-A ' a r<_ %e corr, p %em.gak,* of,turb.~ -.c Orrt r4=,e prerr a r e -b r + cs. po we r levels. 'T^* 5 e .....a +,u... e..,i., n e, A h -4 \\ I
O O O TABLE 4.3.6-1 3 IE CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL i 5 CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH Q TRIP FUNCTION CHECK TEST CALIBRATION ') SURVFILLANCE REQUIRED I j 1. R00 PATTERN CONTROL SYSTEM a. Low Power Setpoint NA S/U S/U(b) g ) g - _ - *4% tower D Q 1, 2 RwL-W b. )
- " a u ' ' "- -- - '
NA j -Li itei Setpoint D(c) g(d)M q j,p 2. APRM a. Flow Blased Neutron Flux - Upscale TNAL S/U(b) g jgy j t* b. Inoperative NA S/U(b),M NA 1, 2, 5 c. Downscale 1NAK S/U ,N 1Qt 1 ] d. Neutron Flux - Upscale, Startup INAK S/U ,M MK 2, 5 i 3. SOURCE RANGE MONITORS a. Detector not full in NA S/U(b) W NA 2, 5 I b. Upscale NA S/UIU),W Q 2, 5 ), c. Inoperative NA S/UI ),W NA 2, 5 i d. Downscale NA S/U ,W Q 2, S 1 4. INTERMEDIATE RANGE MONITORS i S/U((b),gb) y NA 2, 5 a. Detector not full in NA l b. Upscale NA S/U(b)* Q 2, 5 c. Inoperative NA S/U NA 2, 5 d. Downscale NA S/U(b) 5. SCRAM DISCHARGE VOLUME a. Water Level-High NA XM W R 1, 2, 5* 5. Scr;; Tr'; "yp :;; NA M NA (I, 2, O* l 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW a. Upscale NA S/U ,M Q 1 i n b. Ir g cetive NA ';/U"'" g,M NA 1 c. (Em:*er) (Sc=';cale) Q l
O m.3 c.a_a + e, - e.- < m. w, , ~4 o c, o 6s but tewod.Q sta.h, A. 7= t t ) %s min \\ on, a#A -A rade au % c u.', g % repw A. i l l 9 l l l l
0 0 O TABLE 3.3.3 A EMERdENCYCORECOOLINGSYSTEMACTUATIONINSTRUMENTATION MINIMUN OPERABLE APPLICABLE i CHANNELS PER OPERATIONAL Q TRIP FUNCTION TRIP FUNCTION (,) CONDITIONS ACTION A. DIVISION I TRIP SYSTEM 1. RHR-A (LPCI MODE) & LPCS SYSTEM a. Reactor Vessel Water Level - Low La L , Level 1 2 1, 2, 3, 4*, 5* 30 b. Drywell. Pressure - High 2 1,2,3 30 c. LPCS Pump Discharge Flow-Low (Bypass) Tlk 1, 2, 3, 4*, 5* 31 d. Reactor Vessel Pressure-Low (LPCS Permissive) i1K 1,2,3 32 4*, 5* e. Reactor Vessel Pressure-Low (LPCI Permissive) Tlk 1, 2, 3 32 g 4*, s- -aa-f. LPCI Pump A Start Time Delay Relay ilk 1, 2, 3, 4*, 5* 32 T g. LPCI Pump A Discharge Flow-L4 (Bypass) 11) 1,2,3,4*,5* 31 y G.. Civisiei. I "ee."we, L..ite. (2) 1, 2, 3, i*, 5* 3th h+. Manual Initiation 41 %":y:t--) 1, 2, 3, 4*, 5* .f35K ,s 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"# a. Reactor Vessel Water Level - Low 4. iv.i, Level 1 2(b) 1, 2, 3 30 b. Orf.211 Pres:ere "igh 2
- 1, 2, 3 30 6 -e.
ADS Timer 1 ! 1, 2, 3 32 c.d. Reactor Vessel Water Level - Low, Level 3 (Permissive) X14- ' 1, 2, 3 32 i d,e. LPCS Pump Discharge Pressure-High (Permissive) fit ~4 1, 2, 3 32 i e -P. LPCI Pump A Discharge Pressure-High (Permissive) (4) 2-1,2,3 32 j f.g. Manual Initiation 2 (1)/(velee),1, 2, 3 35 g,lr. ma L6ba i L 2, s 1E 10 R\\
O O O TABLE 3.3.3-1 (Continued) E E EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE APPLICABLE E CHANNELS PER OPERATIONAL Q TRIP FUNCTION TRIP FUNCTION,) CONDITIONS g ACTION s* 8. DIVISION 2 TRIP SYSTEM 1. RHR 8 & C (LPCI MODE) a. Reactor Vessel Water Level - Low, -Lew-Low, Level 1 2 1, 2, 3, 48, 5* 30 b. Drywell Pressure - High 2 1,2,3 30 c. Reactor Vessel Pressure-Low (LPCI Permissive) 112/vsive 1, 2, 3 32 4*, 5=- - d. LPCI Pump (B) Start Time Delay Relay ill 1, 2, 3, 4*, 5* 32 LPCI Pump Discharge Flow - g (Bypass) (1)/ pump 1,2,3,4*,5* 31 e. t' ( '. Givision 2 Gu. Fower riunii.or 2 1, 2, 3, 4*, 5* 34) V -g. Manual Initiation 113/(ejetes)-1, 2, 3, 4*, 5* 1354 4 + 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"# a. Reactor Vessel Water Level - Low -Lew-tew, Level 1 2 1,2,3 30 b. Crya;11."ressure High 2 1, 2, 3 30 l b e. ADS Timer ill 1, 2, 3 32 c t. Reactor Vessel Water Level - Low, Level 3 (Permissive) 111 2-- 1, 2, 3 32 d.e. LPCI Pump (B and C) Discharge Pressure - High (Permissive) pump 1,2,3 32 e 4. Manual Initiation 2 (1)/(valve) 1, 2, 3 35 g.g. W+u i L kN t-l t, 2, 3 35 g i 2 l l K1
i ~ 9, t O l t JON h % #8 NwhkMe GNfb, NS h L$ M gw hrgM I Tec h, $ l l l I O
O O O TABLE 3.3.3-2 o EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE E TRIP FUNCTION TRIP SETPOINT VALUE G A. DIVISION 1 TRIP SYSTEM y 1. RHR-A (LPCI M00E) AND LPCS SYSTEM l C. 5 14 3 a. Reactor Vessel Water Level - Low L = Lc/.;, 1-(150) inches
- 1-(152) inches Level 1 l.c,8 1.88 b.
Drywell Pressure - High 5 th89)- psig 5 (1."4) psig c. LPCS Pump Discharge Flow-Low 1 Ti35c%.gpm 1 1:2=).gpa d. Reactor Vessel Pressure-Low.(t ec s pei m m oe 1 ff t NoK psig, decreasing 9 c,soi psig, decreasing . (Lect g.m.u.W) 5 gimo) psig, decreasingd ((590) psig, decreasing e. Reactor Vessel Pressure-Low f. LPCI Pump A Start Time Delay Relay 5 15Y seconds 5 f 6 y seconds w} g. LPCI Pump A Discharge Flow-Low 1119oo) gpm 1 h p ) gpa -(h. Divi;ic ! 0;;.";nr 't..ite,r 1( ) (vcits) >( ) (vcit:)) u4 h +. Manual Initiation NA RA y-n / + / 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" 16.5 Jy.1, a. Reactor Vessel Water Level - Low L= L=, 1-(150) inches
- 1-(!52). inches Level 1 5.
0 3.;;11."rs ;;r; - liigh
- (1.00) p;'-
ius
- (1.00) p; d conds fg 11 1 b -c.
ADS Timer (T 00,) < 4 [ seconds (F aa.) ? inches <f c -d. Reactor Vessel Water Level-Low, Level 3 m.I @ ii. C inches
- E(10.0
- m. l n5 4 -c.
LPCS Pump Discharge Pressure-High 1 (1451 psig, increasing l f440Fpsig, increasing e c #. LPCI Pump A Discharge Pressure-High > 1125k psig, increasing 2 fMe) psig, increasing S.g. Manual Initiation HA NA H5 g tr. hwa L%%W NA NA e M
n-s V TABLE 4.3.3.1-1 qg EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS s CHANNEL OPERATIONAL c-CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH i [ TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED A. DIVISION I TRIP SYSTEM 1. RHR-A (LPCI MODE) AND LPCS SYSTEM a. Reactor Vessel Water Level - Low Low Low, Level 1 S M R 1, 2, 3, 4 *, 5* b. Drywell Pressure - High S M R 1, 2, 3 c. LPCS Pump Discharge Flow-Low S M R 1, 2, 3,'4*, 5* !(Lecs k-o *) d. Meactor Vissel~ Pressure-Low' S M R 1, 2, 3,-4", 5* gpc 1, u r -a.-)- e. NeaCLor Vessel l'ressure-LoF' S M R 1, 2, 3, 4*, 5* f. LPCI Pump A Start Time Delay Relay NA M Q 1, 2, 3, 4*, 5* i g g. LPCI Pump A Flow-Low S M R 1,2,3,4*,5* j (h. Divi: ice. I S :..::r-MonRcr NA M (NA) Ir2r3 7 *r5*)- 4 J, w 4. Manual Initiation NA iM(b3)fRt NA 1, 2, 3, 4*, 5* h g f s y 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"# a. Reactor Vessel Water Level - Q(,) 1, 2, 3 l Low L: L w, Level 1 S M b. Drf.::11 "re:;;;ure : igh 5 N 1, 2, 3 l n i b r. ADS Timer NA M Q 1, 2, 3 c 4. Reactor Vessel Water Level - Low, Level 3 S M R(a) 1, 2, 3 i d p. LPCS Pump Olscharge j ' Pressure-High S M R 1, 2, 3 e J. LPCI Pump A Discharge Pressure-High S M R(3) 1, 2, 3 l f g. Manual Initiation NA iM(b))fR}. NA 1, 2, 3 J Monu 1 LkW;t N_p M NA t,
- 2., 3 g s.
i
m_-. e.s.s - % ~ e.. d ' O 1 l I I l l 'Nww--, "-e
p (M (m.) uf G Table 3.6.4-1 Contninment end Drywell Icoletien 'Valvrs 2 n
- a. CONTAINMENT AUTWATIC IS01ATION VALVES e
Valve-Penetration Valve Maximum Applicable ' Secondary Te st 2 } Number' Number Group (c) Isolation Time Opera tional Containment Pres sure 4 (Seconds) Condition Bypass Path (Psig) (Yes/No) IB21-F016 P423 6 15* 1,2,3 Yes 11.31 IB21-F019 P423 6 15* 1, 2, 3 Yes 11.31 1821-F022A P124 6 5(i) 1, 2, 3 No 11.31' 1B21-F0228 P416 6 5(i) 1, 2, 3 No 11.31 IB21-F022C P122 6 5(i) 1, 2, 3 No 11.31 IB21-F022D P415 6 5(1) 1, 2, 3 No '11.31 IB21-F028A P124 6 5(1) 1, 2, 3 No 11.31 IB21-F028B P416 6 5(i) 1, 2, 3 No 11.31 IB21-F028C P122 5 5(1) 1, 2, 3 No 11.31 (p 1B21-F02 8D P415 6 5(i) 1, 2, 3 No 11.31 ]E IB21-F067A P124 6 22.5* 1, 2, 3 No 11.31 IB21-F067B P416 6 22.5* 1, 2, 3 No 11.31 1821-F067C P122 6 22.5* 1, 2, 3 No 11.31 j' 1B21-F06 7D P415 6 22.5* 1, 2, 3 No 11.31 i ta j ji 1D17-F071A P201 1 3 1, 2, 3 Yes 11.31 ID17-F071B P201 1 -b 3 1,2,3 Yes 11.31 1D17-F079A P201 1 ,24r 3 1, 2, 3 Yes 11.31 1D17-F079B P201 1 3&- 3 1,2,3 Yes 11.31 1D17-F081A P317 1 -E 3 1, 2, 3 Yes 11.31 1D17-F081B P317 1 -b 3 1,2,3 Yes 11.31 1D17-F089A P317 i WE9 3 1, 2, 3 Yes 11.31 ID17-F089B P317 1 WFF 3 1,2,3 Yes 11.31 LE12-F008 P421 4 33 1,2,3 No 11.31 i IE12-F009 P421 4 33 1, 2, 3 No 11.31 1E12-F011A P105 2 60* 1, 2, 3 No (b) 1E12-F011B P407 2 60* 1,2,3 No (b) 1E12-F021 P408 2 90 1, 2, 3 No 11.31 1E12-F023 P123 4 90* 1, 2, 3 No 11.31 ] 1E12-F024A P105 2 90 1, 2, 3 No (b) 1E12-F024B P407 2 90 1, 2, 3 No (b) 1E12-F037A P113 4 180* 1, 2, 3 No 11.31 1E12-F037B P412 4 180* 1,2,3 No 11.31 4 g I&CMIS4/S/ cap / ;gg
L.) V('% CN m Velve Penatretion Valva (c) Maxirum Applicsble See:ndity Tact Number Number Group Isolation Time Operational Containment Pressure ] (Seconds) Condition Bypass Path (Psig) n (Yes/No) 7) f IE12-F042 N P113 4' 27. 1, 2, 3 No 11.31 g IE12-F042 P412 4 27-1,2,3 No 11.31 Z IE12-F064 P105 4 8 1,2,3 .No (b) } 1E12-F064BN P407 4 8 1,2,3-No (b) IE12-F064 N P408 4 8 1,2,3 No (b) 1E21-F011N P105 17 20* 1, 2, 3 No (b) 1E21-F012. P105 2 180* 1, 2, 3 No (b) 1E22-F004N P410 16 27 1,2,3 No 11.31 1E22-F012M P409 16 5 1, 2, 3 No 11.31 1E22-F023 P409 1 180* 1, 2, 3 No 11.31 E 1E32-F001A P124 10 12.5* 1,2,3 No 11.31 1E32-F001E P416 10 12.5* 1, 2, 3 No 11.31 1E32-F001J P122 10 12.5* 1, 2, 3 No 11.31 p IE32-F001N P415 10 12.5* 1, 2, 3 No 11.31 h IE51-F013N P123 9 15 1, 2, 3 No 11.31 N P104 9 5 1, 2, 3 No 11.31 1E51-F019 1E51-F031 P101 9 30 1,2,3 No (b) y 1E51-F063 P422 9 M So 1, 2, 3 No 11.31 1E51-F064 P422 9 10 1, 2, 3 No 11.31 1E51-F076 P422 9 15* 1, 2, 3 No 11.31 LE51-F077 P106, P107, P115, P429 9 7.5 1, 2, 3 No 11.31 1G33-F001 P131 7 15 1, 2, 3 Yes 11.31 1G33-F004 P131 7 15 1, 2, 3 Yes 11.31 1G33-F028 P424 7 15 1, 2, 3 Yes 11.31 1G33-F034 P424 7 15 1, 2, 3 Yes 11.31 1G33-F039 P132 7 15 1, 2, 3 No 11.31 1G33-F040 P132 7 15 1, 2, 3 No 11.31 1G33-F053 P419 7 15 1, 2, 3 Yes 11.31 1G33-F054 P419 7 15 1, 2, 3 Yes 11.31 1G41-F100 P203 1 30 1, 2, 3 Yes 11.31 1G41-F140 P301 1 30 1, 2, 3 Yes 11.31 1G41-F145 P301 1 30 1, 2, 3 Yes 11.31 h I&CMIS4/S/ cap / 91
O Cr Q Velva .Penstration-Valva Maxirum ' Applicable Second::ry T2st Number Number Group (c) Isolation Time Operational Containment Pressure ' p i (Seconds) Condition Bypass Path (Psig) 3 (Yes/No) ' IG50-F272. P420 1 20*. 1, 2, 3 Yes 11.31 f 1G50-F277 P420 1 20* 1, 2, 3 Yes 11.31 'd 1G61-F075 P417-1 15*' 1, 2, 3 Yes '11.31 ~ 1G61-F080 P417 1 15* 1, 2, 3 Yes 11.31 1G61-F165 P418 1 15* 1,2,3 Yes 11.31 IG61-F170 P418 1 15* 1, 2, 3 Yes 11.31 r IM14-F040 .V313 8 4 1, 2, 3, *
- Yes 11.31 IM14-F045 V313 8
4 1, 2, 3, *
- Yes 11.31 i
1M14-F085 V314 8 4 1, 2, 3, *
- Yes 11.31 1M14-F090 V314 8
4 1, 2, 3, *
- Yes 11.31 l
1M14-F190 V313 8 4 1, 2, 3, *
- Yes 11.31 IM14-F200 V314 8
4 1, 2, 3, *
- Yes 11.31 9
T 1M17-F015 P114 5 5 1, 2, 3 Yes 11.31 I IM17-F025 P208 5 5 1, 2, 3 Yes 11.31 IM17-F035 P428 5 5 1, 2, 3 Yes 11.31 IM17-F045 P436 5 5 1, 2, 3 Yes 11.31 j IP11-F060 P108 1 30 1, 2,.3 Yes 11.31 ~ 1P11-F080 Pill 1 30 1, 2, 3 Yes 11.31 IP11-F090 Pill 1 30 1, 2, 3 Yes 11.31 l! 1P22-F010 P309 1 15* 1, 2, 3 Yes 11.31 1P43-F055 P310 2 30 1, 2, 3 Yes 11.31 l IP43-F140 P311 2 30 1, 2, 3 Yes 11.31 l IP43-F215 P311 2 30 1, 2, 3 Yes 11.31 IP50-F060 P404 1 30 1, 2, 3 Yes 11.31 j IP50-F140 P405 1 30 1, 2, 3 Yes 11.31 IP50-F150 P405 1 30 1, 2, 3 Yes 11.31 IP51-F150 P308 1 15 1, 2, 3 Yes 11.31 IPS2-F160 P305 1 .Fr 3 1,2,3 No 11.31 1P52-F170 P312 1 .Fr 3 1,2,3 No 11.31. l IP52-F200 P306 2 30* 1, 2, 3 Yes 11.31 F I&CMIS4/S/ cap / yg 1
.= - O O O Valva Penstration Valva Maximum Applicable Secondgry Text Number. ' Number . Croup (c) Isolation Time. Operational' Con tainment - Pres sure 4 (Seconds) ' Condition Bypass Path. (Psig) ro (Yes/No) Q, .e {- IP53-F010 P305 1-M3 1, 2, 3 No - 11.31 g-1P53-F015 P305 1 M3 1,2,3 No. 11.31 2. IP53-F020 P312 1 25 3 1,2,3 No: 11.31 } '1P53-F025 P312 1
- 53 1,2,3 No 11.31 1P53-F070 P305 1-25 3 1,2,3 No 11.31 IP53-F075 P312 1
Er3 1,2,3 No 11.31 IP54-F340 P210 1 20 1, 2, 3 Yes 11.31 IP86-F002 P117 1 30* 1, 2, 3 Yes 11.31 d 1 i T o-N I&CMIS4/S/ cap / O-O 10) b. CONTAllMENT MANUAL ISOLATION VALVES 9 - [ A Valve Penetration Valve Maxistan.. Applicable Secondary ' Test g Group ) j Isolation Time Operational Containment Pressu re Number Number. (Seconds) Condition Bypass ' Path (Psig) (Yes/No) Z j IB21-F017(d) P423 NA 1, 2, 3 Yes 11.31 1821-F065A(",f P121-100* 1, 2, 3 (f) -(b) I ~ 1821-F065B P414 100* 1,2,3 (f)' (b) IV-3N(4 valves)(3) P124..P416, NA 1, 2, 3 No 11.31 P122, P415 1C11-F083(*)) P204. 12.5* 1, 2, 3 Yes 11.31 1C11-F128 (d P204' NA-1,2,3 Yes 11.31 1C41-F518 P315 NA 1, 2, 3 Yes 11.31 1C41-F519(d) P315 NA 1, 2, 3 Yes 11.31 ID23-F010A(*)) -P434 rES-3 1, 2, 3 No (a) 1D23-F010B(* P320 ,45-3 1, 2, 3 No (a) e ID23-F020A(*) P434 -riE5 3 1,2,3 No (a) 3 1D23-F020Bf",f P320
- 49 3 1,2,3 No (a)
(b P433 rat 3 1, 2, 3 No (a) ID23-F030A(*) 1D23-F030B P319 '.45 3 1,2,3 No (a) 1D23-F040A(") P433 .45 3 1,2,3 No (a) 1D23-F040f**)) P319 .M 3 1, 2, 3 No (a) 1D23-F050 P425 -r45 3 1,2,3 No (a) 1E12-F004A(*) P102 120*. 1, 2, 3 No (b) 1E12-F004B(*) P402 120* 1, 2, 3 No (b) 1E12-F027A P113 60* 1, 2, 3 No 11.31 1E12-F027B P412 60* 1, 2, 3 No 11.31 1E12-F028A P113 60 1, 2, 3 No 11.31 1E12-F0288 P412 60 1, 2, 3 No 11.31 1E12-F042Cfd P411l '27 1, 2, 3 No 11.31 1E12-F05 P411 NA-1, 2, 3 No 11.31 1E12-F061 d) P123 NA. 1, 2, 3 Yes 11.31 1E12-F073A(**)) I P118 15* 1,2,3 No. 11.31 1E12-F073B P431 15* 1, 2, 3 No 11.31 I 1E12-F105 *) 'P403 120* 1,2,3 No - '(b) ~~ I&CMIS4/S/ cap / / ,m b, kJ U Velva-Pen 3tration Maxitum Applictblo Sectndtry T3ct Number Number Isolation Time Opera tional Containment Pressure (Seconds) Condition Bypass Path (Psig) - to (Yes/No) P 1E21-F001((*)) P103 120* 1,2,3 No (b) 1E21-F005
- P112 27 1,2,3 No 11.31 i
C 1E21-F013(d) P112 NA 1,2,3 Yes 11.31 {. 1E21-F510 P105 NA 1, 2, 3 (f) (b) P112 NA 1,2,3 No 11.31 a 1E21-F517(d) 1E21-F519 P112 NA 1,2,3 No 11.31 1E21-F523((d) ~ P105 NA 1, 2, 3 (f) (b) d) 1E21-F526(d) P103 NA 1, 2, 3 Yes (b) 1E21-F527 P103 NA 1, 2, 3 Yes (b) 1E22-F015(*) P401 24 1, 2, 3 No (b) 1E22-F021(d) P410 NA 1, 2, 3 Yes 11.31 1E22-F510((d) d) P409 NA 1,2,3 Yes 11.31 P409 NA 1, 2, 3 Yes 11.31 IE22-F511(d) g 1E22-F513 P410 NA 1, 2, 3 Yes 11.31 ~ 1E22-F517(d) P410 NA 1, 2, 3 No 11.31 1E22-F519 P410 NA 1, 2, 3 No 11.31
==fd = m 1, 2, 3 m m d I 1E22-F528 P401 NA 1, 2, 3 Yes (b) 9 IE22-F529 P401 NA 1,2,3 Yes (b) LE32-F025A((d) d) P124 NA 1, 2, 3 No 11.31 P416 NA 1,2,3 No 11.31 1E32-F025E(d) 1E32-F025J P122 NA 1,2,3 No 11.31 1E32-F025N (d) P415 NA 1, 2, 3 No 11.31 1E51-F013(*) P123 15 1, 2, 3 No 11.31 1E51-F019(*) P104 5 1, 2, 3 No 11.31 1E51-F034 P123 NA 1, 2, 3 Yes 11.31 1E51-F068 P106, P107, P115, P429 30 1, 2, 3 No 11.31 1E51-F072 P422 NA 1, 2, 3 Yes 11.31 1E51-F082(d) P106, P107, P115, P429 NA 1,2,3 Yes 11.31 1E51-F083 P106, P107, P115, P429 - NA 1, 2, 3 Yes 11.31 1E51-F503 P106, P107, P115, P429 NA 1, 2, 3 Yes 11.31 iE31 364
- 2
- 2423 A
2 I" 11 31 i' 2,' 3 1E51-F507(d) P104 NA 1, 3 Yes 11.31 1E51-F508 P104 NA 1,2,3 Yes 11.31 1E51-F542 P106, P107, P115, P429 NA-1,2,3 Yes 11.31 1E51-F543(d) P106, P107, P115, P429 NA 1, 2, 3 Yes 11.31 N I&CMIS4/S/ cap / g)
Q D_ f~% V O V Valva Pen 3tratica Maxi:zum -Applicable Seccndtry Tact Number Number Isolation Time. Operational Containment Pressure j (Seconds) Condition Bypass Path (Psig) p, (Yes/No) 1E51-F562((d) '4 P101 NA 1, 2, 3 - Yen (b) h d) IE51-F563 P101 NA 1, 2, 3 Yes (b) 1E61-F504((d) P120 NA 1, 2, 3 No .11.31 d) l 1E61-F505 (d) P109 NA 1, 2, 3 No 11.31 l ~~ 1E61-F514(d) P119 NA 1, 2, 3 No 11.31 LE61-F517 P319 NA 1, 2, 3 Yes 11.31 1E61-F520((d) P319 NA 1,2,3 Yes 11.31 1F61-F523 d) P317 NA 1, 2, 3 Yes 11.31 j 1E61-F525(d) P317 NA 1,2,3 Yes 11.31 j 1E61-F549 P317 NA 1, 2, 3 Yes 11.31 1E61-F550 P317 NA 1, 2, 3 Yes 11.31 1 1E61-F551 P319 NA 1, 2, 3 Yes 11.31 IE61-F552 P319 NA 1,2,3 Yes 11.31 g 1G41-F528(d) P203 NA 1,2,3 Yes 11.31 1G43-F050A(*)) ~ P102 v46-4 1, 2, 3 No (a) 1G43-F050f**) i Y P402 545-3 1,2,3 No (a) 7 1G43-F060 P401 ,4&- 3 1,2,3 No (a) 2 1G43-F504A(d) P102 NA 1,2,3 Yes (b) 1G43-F504 (d) P402 NA 1,2,3 Yes (b) d) 1G43-F505 (d)- P401 NA 1, 2, 3 Yes (b) P102 NA 1, 2, 3 Yes (b) 1G43-F506A(d) IG43-F506gd) P402 NA 1, 2, 3 Yes (b) 1G43-F507 P401 NA 1,2,3 Yes (b) 1G43-F510(d) P102 NA 1,2,3 Yes (b) 1G43-F511(d) P102 NA 1, 2, 3 Yes (b) i 1G43-F512fd P402 NA 1,2,3 Yes (b) 1G43-F513 P402 NA 1, 2, 3 Yes (b) { 1G43-F514 ((d) d) P401 NA 1, 2, 3 Yes (b) 1G43-F515 P401 NA 1, 2, 3 Yes (b) 1G61-F633((d) d) P417 NA 1, 2, 3 Yes 11.31 1G61-F634 P418 NA 1,2,3 Yes 11.31 N IM14-F602 V314 NA 1, 2, 3 No 11.31 1M14-F603(d) V313 NA 1,2,3 No 11.31 I I&CMIS4/S/ cap / R1
Velva Pen 2tretion Maxirum Applictble. Second2ry Test Number Mumber Isolation Time . Operational Containment Pressure M (Seconds) Condition Bypass Path (Psig) ,g (Yes/No) f 1M17-F055((*)) P434 r34 1, 2,-3 No .(a). 1M17-F065
- P320 346 3 1,2,3 No (a)
C T 1M51-F090(*) P302 30* 1, 2, 3 Yes 11.31 IM51-F110(*()) P302 30* 1, 2, 3 Yes 11.31 1M51-F210A
- P425 s45-3 1,2,3 No (a)
IM51-F210B(*)) P318 ~,35-3 1,2,3 No (a) IM51-F220A(I") P425 E3 1, 2, 3 No (a) IM51-F220B(*) P318 24 3 1, 2, 3 Nc (a) IM51-F230A P425 res 3 1,2,3 No (a) IM51-F230B *)) C P318 .36-3 1, 2, 3 No (a) I* P425 v45-3 1, 2, 3 No (a) IM51-F240A(*) IM51-F240B P318 ,45 3 1,2,3 No (a) IM51-F250A(*)) P425 e3 1,2,3 No (a) v iM51-F250B(* P318 .-i65-3 1,2,3 No (a) 2 IN27-F555A(d) P121 NA 1, 2, 3 No -O+ 1 \\,3 g e Id) 4 1N27-F555B P414 NA 1, 2, 3 No -( M W *>l IN27-F751 P106, P107, P115, P429 NA 1, 2, 3 Yes 11.31 IN27-F793(d) P106, P107, P115, P429 NA 1, 2, 3 Yes 11.31 1P11-F539(d) P108 NA 1,2,3 Yes 11.31 1P22-F576(d) P309 NA 1, 2, 3 Yes 11.31 1P43-F693(d) P310 NA 1,2,3 Yes 11.31 1P50-F570(d) P404 NA 1, 2, 3 Yes 11.31 1P50-F571(d) P405 NA 1,2,3 Yes 11.31 iP51-F521(d) P308 NA 1, 2, 3 Yes 11.31 IP52-F556(d) P306 NA 1, 2, 3 Ye's 11.31 IPS2-F644(d) P305 NA 1, 2, 3 No 11.31 IP52-F645(d) P312 NA 1, 2, 3 No 11.31 1P53-F030(*) P305 r2+ 3 1,2,3 No 11.31 1P53-F035(C*) P305 M-3 1,2,3 No 11.31 iP53-F040 *) P312 M3 1, 2, 3 No 11.31 1&CMIS4/S/ cap / R1
d[' /b v Volva . Penstratien , Maximum Appliccble SecondEry Test Number Number Isolation Time Operational Containment Pres sure (SeconJs) Condition Bypass Path (Psig) 4 (Yes/No) m) 1P53-F045(*)) P312 vt'J-3 1,2,3 No 11.31 IP53-F536 ((d) d P305 NA 1, 2, 3 -No-h s 11.31 4 IP53-F541 P312 NA 1,2,3 -No-Me 5 11.31 .. { IP53-F558(d) d) P305 NA 1, 2, 3 No 11.31-IP53-F559(d) P305 NA 1, 2, 3 No 11.31 IP53-F562((d) 4 P312 NA 1, 2, 3 No 11.31 IP53-F563 P312 NA 1, 2, 3 .No 11.31 1P53-F567 ((d) d) P305 NA 1, 2, 3 .No 11.31 1P53-F569 P312 NA 1,2,3 No 11.31 IP53-F570((d) d) P305 NA 1, 2, 3 -No-Y L 11.31 1P53-F571 P312 NA 1,,2, 3 -Ne-y e s 11.31 IP54-F726 P406 NA 1, 2, 3 Yes 11.31 1P54-F727 P406 NA 1, 2, 3 Yes 11.31 5d) P210 NA 1, 2, 3 Yes 11.31 1P54-F1097 w2 1P57-F015A(*) P304 15* 1, 2, 3 No 11.31 1P57-F015B P116 15* 1, 2, 3 No 11.31 e 1P57-F523A P304 NA 1, 2, 3 No 11.31 y IP57-F523B(d) P116 NA 1, 2, 3 No 11.31 IP86-F526(d) P117 NA 1, 2, 3 Yes 11.31
- ~~
IP87-F037((*) P401 ,Fr-3 1,2,3 No (a) IP87-F065 *I P318 .46-3 1,2,3 No (a) IP87-F071((*) P318 -,Fr 3 1,2,3 No (a) IP87-F074 *) P318 36-3 1,2,3 No (a) IP87-F077 P318 ,3& 3 1,2,3 No (a) IP8F-F167 P401 NA 1, 2, 3 Yes (b) "I IP87-F04 9(*)) P413 ,Fr 3 1,?,3 Yes 11.31 iP87-F055(* P413 25-3 1,2,3 Yes 11.31 1P87-F046 ((*) P413 .34 3 1,2,3 Yes 11.31 IP87-F052 *) P413 .34 3 1, 2, 3 Yes 11.31 7J ~~ I&CMIS4/S/ cap / M
O O O c. (YrHER CONTAINMENE ISOLATION VALVES 'd [ Valve Pene tra tion Applicable Seconda ry . Test 3 Number Number Operational Containment Pressu re f. Condition Bypass Path (Psig) (Yes/No) .S 3 IB21-F032A((8) .Jie- \\\\[K) 44+ )ta g P121 1, 2, 3 IB21-F032B 8) P414 1,2,3 .Ne-y,5 -(b.)-gg3g jd IC11-F122(8) P204 1,2,3 Yes 11.31 IC41-F520(3) P315 1, 2, 3 Yes 11.31 IE12-F005 P106, P107, P115, P429 1,2,3 No (h) l 1E12-F025A P106, P107, P115, P429 1,2,3 No (h) i 1E12-F025B P106, P107. P115, P429 1, 2, 3 No (h) 1E12-F025C P106, P107, P115, P429 1,2,3 No (h) 1E12-F041C(E) P411 1, 2, 3 No 11.31 j 9 IE12-F055A P106, P107, P115, P429 1,2,3 No (h) l2 1E12-F055gg) P106, P107, P115, P429 1,2,3 No (h) LE12-F550 (8) P421 1, 2, 3 No 11.31 P118 IE12-F558A(8) 1,2,3 No 11.31 , c' 'h IE12-F558B P431 1, 2, 3 No 11.31 I I 1E21-F006(8) P112 1,2,3 No 11.31 1E21-F018 P106, P107, P115, P429 1,2,3 No (h) j 1E22-F005(8) P410 1, 2, 3 No 11.31 1E22-F035 P409 1,2,3 No (h) 1E51-F066(8) P123 1,2,3 No 11.31 l IG41-F522(8) P203 1, 2, 3 Yes 11.31 IM17-F010(8) P114 1,2,3 Yes 11.31 IM17-F020((8) 8) P208 1, 2, 3 Yes 11.31 P428 IM17-F030(8) 1,2,3 Yes 11.31 IM17-F040 P436 1,2,3 Yes 11.31 IN27-F559A(E N P121 1,2,3 -Ne le s 11.31 IN27-F559B 8) P414 1, 2, 3 -43e 3,3 td 11.31 i ? I&CMIS4/S/ cap / RI
k' yn.5,,% eqc.A.- NOTES: a. Isolation valve for instrument line which pentrates the containment, conforms to tly requirements y re of Regulatory Guide 1.11. The i,ISI) program will provide assurance of the operability and integrity Q of the isolation provisions. Type "C" testing will not be performed on the instrument line iso-lation valves. The instrument lines will be within the boundaries of the Type "A" test, open to f i the media (containment atmosphere or suppression pool water) to which they will be exposed under C postulated accident conditions. o g e b. lydrostatic leak test at? 1.10Pa. d nuec iAeA mA c. See Specifiction 3.3.2, Table 3.3.2-1, for isolation signal (s) tha i. vec ica-each valve group. d. Test connection valve. c. Remote manually controlled valve, f. Bypass leakage through this valve is accounted for in the leakage measured during system leak testing per NUREG 0737, item III.D.1.1. W2 g. Check valve, r-h. Relief valve /,act locally icatabic. Teatmd lu mus s mc; dismiivu J u m ius T, ym A imot. yJ 1. See section 3/4.4.7, "Miin Steam Line Isolation Valves". O j. IV-3N is the label for each of the four fill valves in instrument lines associated with the four MSIV penetrations. One IV-3N valve is associated with each MSIV penetration. s
- Standard closure time, based upon nominal pipe diameter, is approximately 12 inches / min for gate valves and approximately 4 inches / min for globe valves. Times listed will be revised using the results from pre-operational ISI stroke time tests.
- When handling irradiated fuel in the primary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
f *E Nf-8 "O [ %ree wqboos %% ha o t *- 3 er ou (b d bc' Ad:: Saah 5 4<- ( %< h - W re.c petJkcdb~ar w h t= % lb. Accd t 9 M d " ). %m skdes c,rc was cQ . Loc o ,a d% rEdh.-tig,M ]pC -t o hs, O( oM p)
- k. % Se.Lc6u chi vaim aw %pr Ak9e
( 41 4 S, c g r u i h l) bw P** *^ I D one I6CMIS4/S/ cap /-15 D^* M "h DD' C M hh b5 SM % i'*O" a*A %%AE hpu bAege - 6 ( 4 % c h k_ vedut s. gg
CONTAINMENT SYSTEMS SURVEILLANCE' REQUIREMENTS (Continued) 2. Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test. 3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P,, g psigt.er P ' I ) Eii5' EE i=; = 'n y t t d. Type B and C tests shall be conducted with gas at P,-(15.0) psig*, atintervalsnogreaterthan24monthsexceptfort$stsinvolving: 1. Air locks, 2. Main steam line isolation valves, (':. ?enetretiene seing centinvece leeksge-monitoring syste;s,)
- 3. %
Valves pressurized with fluid from a seal system,k 4 -fr. frECCS and RCICP containment isolation valves in hydrostatically tested lines whicn penetrate the primary containment,-and-5'.f8. Purge supply and exhaust isolation valves with resilient c"*a y,a} s,e,aJs $3 g 3,,3 _,, te 6. e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3. f. Main steam line isolation valves shall be leak tested at least once per 18 months. -(g. 'Typ= B periodic. tests are not required fo enstrathns~c56Einuously monitored by tne Fri netration Pressurization System, nea=%d um sys en is OPERA 8LE per SpecificTtton%444-- G.. Tg: " +act< for oenetrations employing a continuous Jeakage mon % system shall be conductaa n Q;--(1 T'ps4gr-at._ inter _vAls no greater + M e.,e= pur 4 years.) 96 Leakage from isolation valves that are sealed with fluid from a.;eal system may be excluded, subject to the provisions of Appendix Ji og men so Section III.C.3,.when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P, v2., % t l: (IC.5)psig,andthesealsystemcapacityisadequatetomainfain system pressure for at least 30 days.4 h.-f: XECCS and RCICK containment isolation valves in hydrostatically tested lines which penetrate the primary containment shall be' leak tested q at least once per de months. 24 .e.(*: . Purge supply and exhaust isolation valves with resilient material seals snall be tested and demonstrated OPERABLE per Surveillance Requirements 4.6.1.8.3. and 4.6.1.8.4.K j 7 l K-h The' provisions of Specification 4.0.2 are not applicable to 24 month Q, or 40 210 month surveillance intervals. "Unless a hydrostatic test-is required per Table 3.6.4-1. - j. Coded _ e.dM p.vMns e wguawqA l PERRY - UNIT-1 3/4 6-4 h* o^u fer 24 %%. I Rl
O O 3 4 m % t - e e. m;_ c. % f
CONTAINMENT SYSTEMS
- O
. CONTAINMENT HUMIDITY CONTROL-(Cpti;nel for Irce St;nding St;ci Centsin.cnt)- LIMITING CONDITION FOR OPERATION 3.6.5.2 Containment temperature and relative humidity shall be maintained above the curve shown in Figure 3.6.5.2-1. APPLICABILITY: Whenevery-PRIMARY CONTAINMENT INTEGRITY is required for Specification 3.6.1.1. ACTION: 1. With containment relative humidity not above the curve shown in Figure 3.6.5.2-1, close the containment purge inlet and outlet isolation valves within I hours. 2. With the containment temperature and relative humidity not above the curve shown in Figure 3.6.5.2-1: a.' InOPERATIONALCONDITION1,2or3,restorethetemperature/and-relative humidity to a condition above the curve within (B k _ () hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours. 9 b. At all other times, either: 1. Maintain an unobstructed opening (s) in the containment that equals or exceeds the flow area provided by two open vacuum breakers,' or - 2. Deactivate the containment spray header isolation valves by ctos6. oA tema one en a eack coe= 6..+ 5 pray syt 3 w d5 suggs[hea6er t a de e.cpiq % power tw.A.c ogermA or. SURVEILLANCE REQUIREMENTS f 4.6.5.2 Containment temperature and relative humidity shall be verified to be above the curve shown in Figure 3.6.5.2-1: a. At least once every 24 hours. b. . By -verifying the temperature end 'r-f dity instrumentation 0PERABLE by performance of a: 1. CHANNEL CHECK at-least once per 24 hours, l 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and 3. CHANNEL CALIBRATION at least once per 18 months. PERRY - UNIT 1 3/4 6-34 81
l l L t. -.. ~._. = 1 e 100 ........t... g ,4 .....i..............,...,........... .... i.........,,,,, gg j >^ A i y . i....... q z_ i .........................i........,, i..,................ g. .' ' ' '. ' '.',.,e..' ',.' ' '# 3 i....... g ll #
- ,#,i W
I * * ) .i f l= +->di+ ---i ..l , e. .b t s l< q j,, J $4 4 -~- o .J 4 Oc e e e
- e
._..e e .f. f. ...................... f.... ..,................... s. 3.....,,., y, .....f. .................. f.... ..... i.......................... f........,..... .................. f...,....,,...,..,,,,,, .......................... f.......................,, 3 ...... p......................... ........................ v...., ......f.....,. UNACCEP Taut.E.'.'.'.. ..f. OrfilATION . e.............. to ..........f. r............. i 0 so 7: so to too 11a 12e l Initial Temperature (*F) FIGURE 3. 6.'s. 2-1 CONTAINMENT TEMPERATURE VS RELATIVE HUMIDITY 1 l l GE-STS (B'a'R/6) 3/4 6-35 Q 'l .J
q CONTAINMENT SYSTEMS 0 BASES 3/4.6.5 VACUUM RELIEF 3/4.6.5.1 CONTAINMENT VACUUM RELIEF -(-FE~ ST*J:0!NO ST CL CONT * :: NT) (Optisi,el) Vacuum breakers are provided on the containment to prevent an excessive i vacuum from developing inside containment during an inadvertent or_ improper x operation of the containment spray. Four vacuum brhkers and their associated isolation valves are provided. Any two vacuum breakers provide (440)% vacuum relief. 100 The containment vacuum relief system is designed to prevent an excessive -vacuum from being created inside the containment following in advertent initiation of the containment spray system. By maintainina temperature /-end _ s relative humidity above the curve shown on Figure 3.6.G.2-1, the maximum containment vacuum created by actuation of both containment. spray loops will be limited to approximately 0.7 psig. 3/4.6.5.2 ORYWELL POST-LOCA VACUUM BREAKERS N i..i G.1-e; ; t.14. vee.. cer t., vi r,et.r) d* Mere,b.L(te5 hare "M dipdk od = 5 5 Drywell post-LOCA vacuum breakers are provided on the drywell to prevent drywell flooding due to : ' :tr: : t br _ :nph;r: in:id: ::nt i,.n at and to equalize pressure between the drywell and containment.befer; th; ;; ts;tibie
- M c--tM :y:t= ;;,.r;;;;r; are ;terted.
Two drywell vacuum breakers and their associated isolation valves are provided. Any one vacuum breaker can provide full vacuum relief capability. W out of drywell hydrogen control system) Drywell p CA' vacuum breakers are provided on th ell to prevent drywell flooding an instrument break re inside containment, and to act in conjunction the c le gas control system to provide mixing concentration ding 4% (by volume) of the drywell j following a LOCA. Il' Two d vacuum breakers and their associated isolatio ves are pro Any one vacuum breaker can provide full vacuum relief capa ty I PERRY - UNIT I.' B 3/4 6-7 gg
v l S. % \\ E2.t -soo s ad i s.w t-Son bes3 c.We$ err d d L ea<.L suf f ) t A L \\e.a d, So o q eres oi, o-m Wune o( \\"2. 5 M o \\is for oi \\ead % kowc s. ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 1110(-volts, or battery overcharge . with battery termina1' voltage above 1150K-volts, by verifying that: 1. The parameters in Table 4.8.2.1-1 meet the Category B limits, 2. There is no visible corrosion at either terminals or connectors, or the eennee44en resistance of there 't--- is lees tt.:n GZ x Z *) ;t.;.;, 2nd euw ca.u,-t..c u .a derQ coor<da I is lesr A u egud to So x to-' o b a's. 3. The average electrolyte temperature of 'e repr;;;atativ; a r.t;r)- -efconnectedcellsisabove60*FR. Io c. - At least once per 18 months by verifying that: 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, 2. The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material, f .5o s 3. The resistance of eachlcelly-to-cells and terminal connection is less than or equal to 1140-x 10 s4 ohns, -and-l iR91-soos,-so o r,-soa s,.4 --soo9 4. The, battery chargerswill supply at least (400) amperes at a i e ck - minimum of fl253: volts for at least 14k hoursh awa 1 d. At_least once per 18 months, during shutdown, by verifying that either: 1. The battery capacity is adequate to supply and mairitain in OPERA 8LE status all of the actual emergency loads for the design duty cycle when the battery is subjected to a battery i service test, or i 2. The battery capacity is adequate to supply a dummy load of the L following profile while maintaining the battery terminal voltage greater than or equal to flo7k volts. 1It.w1.- s os t a) Battery 44A), greater than or equal to X82s)t amperes; 1Rgt-soo3 ' batteryM, greater than or equal to X5MX amperes; and f cattery tie), greater than or equal to 118) ampares during a tt-soas the initial 60 seconds of the test. \\nw'L-soot b) Battery 44A), greater than or equal to (n.1) amperes; ggs So 3 / oattery 149}, greater than or equal to itt9) amperes; and i gn Se,,g/ Dattery tiG),. greater than or equal to (tr.) amperes during the remainder of the first* hour of the test.
- PERRY - UNIT 1 3/4 8-12 gg
ELECTRICAL POWER SYSTEMS O SURVEILLANCE REQUIREMENTS (Continued) lR.91-soo s c) Battery -(4A-), greater than or equal to (32sk amperes; 1Rn-soodr battery YHF), greater than or equal to (stQ amperes; and tEti-soos/ Dattery tHP), greater than or equal to (ist amperes during the remainder of the 48} hour test. 2. e. At least once per 60 months during shutdown by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. At.this once per 60 month interval, this performance discharge test may be performed in lieu of the battery service test. f. At least once per 18 months during shutdown performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its. average on previous performance tests, or is below 90% of the manufacturer's rating. O 1 l I l PERRY-UNh1 3/4 8-13 g\\
y TA BLE
- 3. 6. 4. 2 - l (Con /-)
y 480 Y LcAb OVERC UKRENT PROTEC TICA/
- '~
N msgv + 5eco m b y
- 4 lElz-F042 A E F B-o 1 Comp. X IR24-Scal sec 19. F3
- Fo426 EF b- 07 Comp. X IR34-Soa6 sec 18 F8 -FS37A EF B-07 ccmp.33 IRaf -sca l sec -19 F4- ~ FS 37 6 EF b- 07 conp. 3J IR24 - SORG Sec 18 I9 l E 51 - F0% EF- )-c - 07 Comp. 6 1R.24-so23 sec 14 FS I F 11 - E 012 A F - l-E - li 8.24 (H5I-PB%B Pal O iFIS-E003 F-l- C- 08 "l3 IH S I -P844B P3 - E 005 F-l - F-fi
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- ...... _3_ - - _. _ _ _. - - _ _ -.. _ - - -. - - - -
p I L 9 REACTIVITY CONTROL SYSTEMS 3/4.1.3 CONTROL RODS CONTROL' ROD OPERA 8ILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods shall be OPERABLE. APPLICA81LITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: a. With one control rod inoperable due to being immovable, as a result of excessive friction or mechanical interference, or known to be untrippable: 1. Within one hour: a) Verify that the inoperable control rod, if withdrawn, is separated from all other inoperable control rods by at least twocontrolcellsinalldirections.NM ht m # b) Disarm the associated directional control valves ** either: 1) Electrically (by bypt; = th: """ :=?y::r :: d), or 2) Hydraulically by closing the drive water and exhaust water i isolation valves. ~ 2. @ Comply with Surveillance Requirement 4.1.1.c. ) Oth; raise, t; ir, et h u t = T 0""T"""". d thir, the ru t 12 h u n. 2. 2: _ Restore the inoperable control rod to OPERA 8LE status within 48 hours. 1 o%<lh ;'u'%' hie' ' "k"Ts'Krk" J '11'%"'GX WQ s. b. With one or more control rods trippable but inoperable for causes other than addressed in ACTION a; above: - 1. If the inoperable control rod (s) is withdrawn, within one hour: a) Verify that the inoperable withdrawn control rod (s) is separated from all other inoperable withdrawn control rods by at least two control cells in all directions, and b) Demonstrate the insertion capability of the inoperable withdrawn control rod (s) by inserting the control rod (s) at least one notch by drive water pressure within the normal operating range". Otherwise, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves ** either: a) Electrically -(t, typeesir.ii er, the ZZ e. ei zer ..G, or i b) Hydraulically by closing the drive water and exhaust water isolation valves. "The inoperable control rod may then be withdrawn to a position no further withdrawn than its position when found to be inoperable.
- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERA 8LE status.
1 E PERRY - UNIT 1 3/4 1-3 2 El
a ~ - m .2 O O 2..u a ~ m <s...~ t smo a-, "++ "'^" y,,.3.., d
1 RA010 ACTIVE EFFLUENTS p,, O MAIN CONDENSER s LIMITING CONDITION FOR OPERATION @nlaak tale, ef-ds swief na M44h8 freak n,Wp p 6sa,p sF,W a,1 s.In,rcier,wd re.ise 3.11.2.7 The gr::: r:dit::ti it;
- t: 't:t: :nd.':r ;:- :} Of th: 7:5!: ;::::
measured at the main condenser air ejector shall be limited to less than or equal tq43G millicuries /second af ter 30 minutes decay. @ S$8 g APPLICABILITY: "t ;1' tin:.1).,eig.adMeser de ep4er epw.tik ACTION: @ O cesJame dr tild*! ~ g< \\ t n e. With the ;r::: r:dioacti ity rate of the specified noble gases at the
- t'-::
- t;;r j:t ::nd:n::e discharge exceeding 34G illicuries/second, restore the gross radioactivity rate to within its limit ithin 72 hours or be in at least HOT CT'"
'c within the next 12 hours, 358 (D silvtco.ad g SURVEILLANCE REQUIREMENTS 8) The - d' : + '"'ty rate of noble gasesI a,wss,v srs,sas se ont W ri- : r sies.se a) w-ss,w c,u *n 4.11.2.7.1 .ireWe A.,3 s, sit.n -:j::t:- shall be continuously monitored in accordance with Specifica-k tion 3. 3. 7.yl.h I te lea 4.11.2.7.2 The ;-4u
- d:s e:::tivity rate (beta and/or gamma) of the specified A()
noble gases from the main condenser air ejector shall be determined to be within the limits of Specification 3.11.2.7 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the discharge (prior to dilution and/or discharge) of the main condenser air ejector: a. At least once per 31 days. Within 4 hours following an increase, as indicated by the condenser b. air ejector noble gas activity monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level, in the nominal steady-state fission gas release from the primary coolant. a l '#,0M;N0 ION N00t.EA" - UN:? ? 3/4 11-17 PittN-wwwi g
de.(ete - (gp) 7upp 3 4,q.,e ar c,4;mf (ps4x i. ). MA m emMwrd dest) dt.< ~f th w p ' 4 a SM sy gas tn44 systen brys fr*~ % e><4sime 4 da[usti O WA!!di 4 Fast hf_f sysfx. ,,, ut 4,,sd 6y N' fE' f udt t mf vuf l l l t i i O l l l i l t O
e EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 SUPPRESSION P00L l LIMITING CONDITION FOR OPERATION 3.5.3 The suppression pool:eheM4e-4PEAA8hEt
- a. su b eie...wi :
a.i. In OPERATIONAL CONDITION 1, 2 and 3 with a contained water volume of at least (135,50B1") fta, equivalent to a level of (10.")f W6". l
- 106,
+.2.InOPERATIONALCONDITION4and5Nc."ithacontainedN,tervolumeofat 104'So 8 eleastT33, 000) fta, equivalent to a level of -(10'n), except that the suppression pool level may be less than the limit or may be drained provided that: +.W No operations are performed that have a potential for draining the reactor vessel, i it.tw The reactor mode switch is locked in the Shutdown or Refuel
- position,
+.tc) The condensate storage tank contains at least (150,000),available gallon,s of water, equivalent to a level of (3o)%, and 4(4) The HPCS system is OPERABLE per Specification 3.5.2 with an O-OPERA 8LE fiow Path caPabie of taxia suctioa from the :oadeasate storage tank and transferring the water through the spray sparger to the reactor vessel.
- b. L e septe k -w be chu%s b og..b.J d.b*^4*
j. APPLICA6ILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4,and-5(a
- ACTION:
a. In OPERATIONAL CONDITION 1, 2 or 3 with the suppression pool water level less than the above limit, restore the water level to within the limit within I hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, y,we-b. In OPERATIONAL CONDITION 4 or 5& with the suppression pool water level less than the above limit or drained and the above required conditions not satisfied, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel and lock the reactor mode i switch in the Shutdown position. Establish E '"........ CONTAINMENT INTEGRITY within 8 hours. y.. a,3.r me.y /d ar**" TM n;;r;ni= ;x? i: 7.;t rgkM " M OM""."'.! ;rnitd ".h.; the reactor vessel head is removed, the cavity is flooded ( r 5 i q f? x i d 1 . tt: 7 _.s
- rn
- i r. ;xi), the ;;
- r xt:f r r.; ft:1 ;x! gates ape" removed,t-the get l
- ful ;xi gets ereh-(when the cavity is flooded), and the water levelg 0 is maintained within the limits of Specification 3.9.8 and 3.9.9. g,,, m., a... x g..w PERRY - UNIT 1 3/4 5-8
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) TABLE 3.3.3-1 (Continued) 3 .g EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRt#ENTATION MINIMUM OPERA 8LE APPLICA8LE j CHANNELS PER OPERATIONAL y TRIP FUNCTION TRIP FUNCTION,) CONDITIONS ACTION g +* C. DIVISION 3 TRIP SYSTEM 1. HPCS SYSTEM a. Reactor Vessel Water Level - flow,4ew, level 2k 4 1, 2, 3, 4*, 5* 36 b. Drywell Pressure - High**- 4 1, 2, 3 36 i c. Reactor Vessel Water Level-High, Level 18k M (c) I 1, 2, 3, P, 5* M% d. Condensate Storage Tank Level-Low 2(d) 1, 2, 3, 4*, 5* 37 d) e. Suppression Pool Water Level-High 2 1,2,3,4*,5* 37 f. Pump Discharge Pressure-High (Bypass) XIK 1, 2, 3, 4*, 5* 31 g. HPCS System Flow Ra e-p (.";mh;!x) (,B gm,) Q ],2,3,4*,5* { 3 s. ._.,__.._...m., s., !? 1. Manual Initiation * * $1)/;y;t;;) 1, 2, 3, 4*, 5* T351 4- + l r- ,jg MINIMUM APPLICA8LE ) TOTAL NO. CHANNELS OPERABLE OPERATIONAL l OF CHANNELS TO TRIP CHANNELS CONDITIONS ACTION D. LOSS OF POWER z 2 2 l 1. 4.16 kw Emergency Bus Undervoltage 4/ bus -1/ bus -1/ bus 1, 2, 3, 4**, 5** 38 (Loss of Voltage) 2 l 2. 4.16 kw Emergency Bus Undervoltage -3/ bus 2/ bus 2/ bus 1, 2, 3, 4"*, 5** 39 l (Degraded Voltage) (a) A channel may be placed in an inoperable status for up to 2 hours during periods of required surveillance without placing tha trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter. j (b) Also actuates the associated division diest.1 generator.(=d ik r;;r;;;icr. ;;;! nhg ;y;te:). (c) Provides signal to close HPCS pump d h 9 r,"_ valve only. ,6fW. n j (d) Provides signal to HPCS pump suction valves only. When the system is required to be CPERA8LE per Specification 3.5.2 or 3.5.3. Required when ESF equipment is required to be OPERA 8LE. 3 Not required to be OPERABLE when reactor steam done pressure is less than or equal to 1100K psig. -M-Alafw enlyg ,_ Dr well ~ ~ g.s W sNnW t.ac ti.. i or j es A Pressure - High and Manual Initi tion are not required to be OPERABLE with indicated reactor .. W, vessel water level on the wide range instrument greater than Level 8 setpoint coincident with s s.- ... - u. e s.,. <J u. s.,
O O O TABLE 4.3.3.1-1 (Continued) 3 EMAGENCY CORE COOLING SYSTEM ACTUATION INSTRtMENTATION SURVEILLANCE REQUIREENTS 4 CHANNEL - OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH d TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED -4 C. DIVISION 3 TRIP SYSTEM 1. HPCS SYSTEM a. Reactor Vessel h ter Level - I R *) 1, 2, 3, 4*, 58 $ tow tour, Level 2K S M IR ") 1, 2, 3 b. Drywell Pressure-Highi*
- 5 M
c. Reactor Vessel Water Level-High, TR1,) 1,2,3,48, 5* g Level E.k. 151 M d. Condensate Storage Tank Level - g Low S M R(,) 1, 2, 3, 4*, 5* e. Suppression Pool Water Level - High S M R(,) 1, 2, 3, 4*, 5* w h f. Pump Discharge Pressure-High iSK. M (RK 1, 2, 3, 4*, 5* g. HPCS System Flow Rate-Low TSt M (R) 1, 2, 3, 4*, 5* (i.
- .. 2 M " r r 5. iter "A
(NA) 17 2,-3r4*,5*)- b +. Manual Initiation **- NA iM(b))(R)- NA 1, 2, 3, 4*, 5* 0. LOSS OF POWER 1. 4.16 kw feergency Bus Under-NA NA R 1, 2, 3, 4**, 588 voltage (Loss of Voltage) 2. 4.16 kw Emergency Bus Under-5 M R 1, 2, 3, 4**, 5** voltage (Degraded Voltage) Not required to be OPERABLE when reactor steam done pressure is less than or equal to fl00K psig. When the system is required to be OPERA 8LE per Specification 3.5.2. Required when ESF equipment is required to be OPERA 8LE. (a) Calibrate trip unit at least once per 31 days. f(b) Manual initiation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least once oer 31 days _as_a part of circuitry required to be tested far_ automatic system actuation.1 M_ The i 3,.%. Soc - I tion of Drywell Pressure - High and Manual Initiation are not required to N'" h h h I be OPERA 8LE with indicated reactor vessel water level on the wide range a4bD. instrtment g eater than Level 8 s g coincident with the reactor pres-gg g ,a n
REFUELING OPERATIONS O -3/4.9.6 REFUELING PLATFORM LIMITING CONDITION FOR OPERATION 3.9.6 The refueling platform shall be OPERA 8LE and used for handling fuel assemblies or control rods within the reactor pressure vessel. APPLICA8ILITY: During handling of fuel assemblies or control rods within the reactor pressure vessel. ACTION: With the requirements for refueling platform OPERABILITY not satisfied, suspend use of any inoperable refueling platform equipment from operations involving the handling of control rods and fuel assemblies within the reactor pressure vessel after placing the load in a safe condition. SURVEILLANCE REQUIREMENTS 4.9.6 Each refueling platform crane or hoist used for handling of control rods Q or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERA 8LE within 7 days prior to the start of such operations with that crane or hoist by: a. Demonstrating operation of the overload cutoff on the main hoist when the load exceeds (1200 2 50) pounds. b. Demonstrating operation of the overload cutoff on the frame mounted and monorail hoists when the load exceeds (500,* 50t pounds. por c. Demonstrating operation of the uptravel mechanical stop on the frame mounted,and monorail hoists when uptravel brings the r; ;f ( :ti n) = ;;;- ij u ; :::t below the (==1 ?_;I ;n, _,; seel) a:W 95 4
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05 %. y.gpe % % c..sg d. Demonstrating operation of the downtravel mechanical cutoff on the toA et fa main hoist when grapple hook down travel reaches (4k inches below fuel assembly handle. y c y " *".rh e. Demonstrating operation of the slack cable cutoff on the main hoist cc 3 1% when the load is less than (50 2 10) pounds. 1 & Sie s ~5 f. . Demonstrating operation of the loaded interlock on the main hoist when the load exceeds 1445 2 50) pounds. g. Demonstrating operation of the redundant l'oaded interlock on the O main heist when the load exceeds $550 2 50X pounds. PERRY - UNIT 1 3/4 9-8 l D
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~ l V CONTAINNENT SYSTEMS MSIV LEAKAGE CONTROL SYSTEM (0pti;;;;)- LIMITING CONDITION FOR OPERATION 3.6.1.4~ Two independent MSIV leakage control system (LCS) subsystems shall be OPERA 8LE. APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 3. i ACTION: i With one MSIV leakage control system subsystem inoperable, restore the incperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. . SURVEILLANCE REQUIREMENTS 4.6.1.4 Each NSIV leakage control system subsystem shall be demonstrated 0PERA8LE: a. At least once per 31 days by verifying: 1., 8 lower OPERA 8ILITY by starting the blower (s) from the control room and operating the blower (s) for at least 15 minutes. -O cHiater0$ ERA 8ItITY by demonstrating eiectrica, cont 4nuity of 2. the heating element circuitry /, % m M t h % * ab d W M dr**5 8 per During.savo={.o~pereieach C0 D SHUTDOWN (,e %..if not performed within the pre b. 92 days, t,y cjcikg-tect, r te, ..uel-end-automatic motor operated-i niv; thr;;;;h ;t 1;;;t ;;;; ;;.;..,.1;t; ej;1e-of-fv14-trevel) fin accordance with Specification 4.0.5). c. At least once pgr 18 months by: 1. Performance of a functional test which includes simulated actuation of the subsystem throughout its operating sequence, and verifying that each automatic valve actuates to.its correct position,,the blower (s) start (s). o.6 2. Verifying that the blower (s) develop (s) at least the below required vacuum at the rated capacity,;;d the he t:r (' r;:r;- 4eee' -"- '- ' ( )*f within ( ) ;;;in;t;;) (&;;; (3.7) 0; -(0.5)
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[ EMERGENCY CORE COOLING SYSTEMS '3/4 5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE: a. The low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel. b. Low pressure coolant injection (LPCI) subsystem "A" of the RHR system with a flow path capable of taking suction from the suppres-sion pool and transferring the water to the reactor vessel. 4 c. Low pressure coolant injection (LPCI) subsystem "B" of the RHR i system with a flow path capable of taking suction from the suppres-sion pool and transferring the water to the reactor vessel. d. Low pressure coolant injection (LPCI) subsystem "C" of the RHR system with a flow path capable of taking suction from the suppres-E sion_ pool and transferring the water to the reactor vessel. e. The high pressure core spray (HPCS) system with a flow path capable of taking suction from one of the following water sources and transferring the water through the spray sparger'to the reactor . O i: i. From the suppression pool, or W aty "A h 4 @, a.1 2. When the suppression pool level is less than the"li;;;it er is d7;ine;, from the condensate storage tank containing at least 1150,0001 available gallons of water, equivalent to a level of (So)L APPLICA8ILITY: OPERATIONAL CONDITION 4 and 5*. ~ ACTION: a. With one of the above required subsystems / systems inoperable, restore at least two subsystems / systems to OPERA 8LE status within 4 hours or suspend all operations that have a potential for draining the reactor vessel. . b. With both of the above required subsystems / systems inoperable, suspeno CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel. Restore at least one subsystem / system to OPERA 8LE status within 4 hours or establish C^"^."Y CONTAINMENT INTEGRITY within the next 8 hours. N"* " ' i "TheECCSisnotrequiredtobeOPERA8LEprovidedthatt ee actor vessel head, is removed, the cavity is flooded, the ;;g ;;ateirant
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removed, th C::t '=1 ;::1 ;:te: Or: 7: :c;d, and water level is maintained f within the limits of Specification 3.9.8 and 3.9.9. a w %,, c.am. A j ...s, PERRY.- UNIT 1 3/4 5-6 8\\
V INSTRUMENTATION i O huo inutwt ocot O CHLORINE (?"".". = O DETECTION SYSTEM (^ptf:::1) LIMITING CONDITION FOR OPERATION ad ewg\\ect od d e 3.3.7.8 Two independent chlorine W
- a h) detection system subsystems shall be OPERABLE with their falareMtrip) setpoints adjusted to actuate at a:
o ** o.q a. Chlorine concentration of less than or equal to f5} ppa (, andt. t
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Ammon4e concentration of less than or equal to ( ) ppm):. A_PPLICA8ILITY: All OPERATIONAL CONDITIONSF.44 ACTION: ewye.s ov.i A e a. With one chlorine fand/or one emmenfet detection system subsystem inoperable, restore the inoperable detection subsystem to OPERABLE status within 7 days, or within the next 6 hours, initiate and 2:t en, control room ; ;ci,;acy filtr:q m'agl., ut-maintain operar. ion -i et 1: t ^a system,ewboys4em in the ( be ht ha) mode of o '*' Y** x.["'c bration. yns /2.< ) b. With both chlorine Tand/or NYd dete?ction system subsystems inoperable, within one hour initiate and maintain operation of-et %s. Od heet :n; control room r :rg ery *ffitr:ti^ system,-a.,ej;t r in the (Schth9 mode of. operation. ^^' * *^ (m15 /2t ) ftec ca&.. cm rye t3 c. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS es %e o Nh e-s 4.3.7.8 Each of the above required chlorine fand ammonist detection system subsystems shall be demonstrated OPERABLE by performance of: l a. CHANNEL CHECK at least once per 12 hours, b. CHANNEL FUNCTIONAL TEST at least once per 31 days, and c. CHANNEL CALIBRATION at least once per 18 months. l O + maa e. % -u ' ' s y %& %% o' e em c.u.-..A. PERRY - UNIT 1 3/4 3-82 R\\
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m CONTAINMENT SYSTEMS SUREVILLANCE REQUIREMENTS 4.6.5.1 Each containment vacuum breaker shall be: a. Verified closed at least once per 24 hours. b. Demonstrated OPERA 8LE: 1. At least once per 31 days by: a) Cycling the va:uum breaker and isolation valve (+) through at least one complete cycle of full travel. b) Verifying M)ithe) position indicator 4e3 OPERABLE by observing expected valve movement during the cycling test. 2. At least once per 18 months by: a) Verifying the force required to open the vacuum breaker, from the closed position, to be less than or equal to -(4rp psid, and o.u f) b) Verifying-(5:th)ithek position indicator (+)-0PERABLE by v performance of a CHANNEL CALIBRATION. 3. By verifying the OPERA 8ILITY of the vacuum breaker isolation valve differential ressure actuation instrumentation with the opening setpoint (6-5) psid by performance of a: >,00 a) CHANNEL CHECK at least once per 24 hours, b) CHANNEL FUNCTIONAL TEST at least once per 31 days, and c) CHANNEL CALIBRATION at least once per 18 months. O PERRY - UNIT 1 3/4 6-33 g
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CONTAINNENT SYSTENS O ORYWELL POST-LOCA VACUUN BREAKERS LIMITING CONDITION FOR OPERATION 3.6.5.3 All drywell post-LOCA vacuum breakers shall be OPERA 8LE and closed. l APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION: ' With one drywell post-LOCA vacuum breaker inoperable for opening but a. known to be closed, restore the inoperable vacuum breaker to OPERA 8LE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. b. With one drywell post-LOCA vacuum breaker open, restore the open vacuum breaker to the closed position within I hour or be in at least HOT SHUT-DOWN within the next 12 hours and in COLD SHUTDOWN within the following i 24 hours. With the position indicator of an OPERA 8LE drywell post-LOCA vacuum c. breaker inoperable, verify the vacuum breaker to be closed at least once per 24 hours by (er. ;1tu #r."e ;;;;th:d). Otherwise, declare the vacuum '^'t" * * ' breaker inoperable. SURVEILLANCE REQUIRENENTS O 4.6.5.3 Each drywell post-LOCA vacuum breaker shall be: a. Verified closed at least once per 7 days. b. Demonstrated OPERA 8LE: 1. At least once per 31 days by 1 a) Cycling the vacuum breaker and isolation valve (s) through at least one complete cycle of full travel, b) Verifying 6th) ithel position indicators OPERA 8LE by observing expected valve movement during the cycling test. 2. At least once per 18 months by: a) Verifying the pressure differential required to open the vacuum breaker, from the closed position, to be less than or equal to g psid, and b) Verifying (4eth).itheX position indicators OPERA 8LE by performance of a CHANNEL CALIBRATION. 3. By verifying the OPERA 8ILITY of the vacuum breaker isolation valve differential pressure actuation instrumentation with the opening setpoint 1 (q.",)N; $NTDD;r; p:i; by performance of a: OE.^ZL--04EGM-et-1T -e)
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CHANNEL FUNCTIONAL TEST at least once per 31 days, and b A) CHANNEL CALIBRATION at least once per 18 months. PERRY - UNIT 1 3/4 6-36 m
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REACTOR COOLANT SYSTEM O IDLE RECIRCULATION LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal to 1100FF, and: I a. When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to X50k*F, or b. When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is less than or equal to X50FF and the operating loop flow rate is less than or equal to 1501% of rated loop flow. APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4,P vsA re..i. e s*: s.% ACTION: T D* i With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop. i 7 SURVEILLANCE REQUIREMENTS l l 4.4.1.4 -The temperature differentials and flow rate shall be determined to be within the. limits within 15 minutes prior to startup of an idle recirculation loop. O PERRY - UNIT 1 3/4 4-4 R\\
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A 'i 'f ~ i Di ?!NITioC (Contine1) SOLIDIFICATION OW -&-+k SOLIDIFICATION shall be the conversion of radioactive wastes from licuid syster.s to a homogeneous (uniformly distributed), monolithic. immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing). OFFSITE DOSE CALCULATION MANUAL (00CM) h 119*-??- The OFFSITE 00SE CALCULATION MANUAL shall contain the methodology anc caru,eters used.in the calculation of offsite doses due tc radioactive gaseous anc liquic ef fluents anc in :ne calculation of of gaseous anc liquid effluent ecr.i toring alart/ trip setecints. GASECUS RACWASTE TREATMENT SYSTEM f6 (.tc %g, b) Jr G/.SE005 RA0VASTE TREATMENT SYSTEM is eay system designed and installed to recuce racioactive gasecus ef fluents by collecting primary coolant system .offgases from the primary system and providing for ce ay or holdup for the l purpese of recucing tne total radioactivity prior te release to the environment. 70tR!JTION EXHAUST TREATMENT SYSTEM ,_s I 1 '.2' -A VENTILA: CXHAUST TREATMENT SYSTEM is any system desi r . instaliec 1 o. te reduc 2 gaseous radioi... aa or radioactive material i .iculate form in es through charcoal accerters k* ef flucnts by passing ventilatto.. vont exha - e ar.c/or.HEFA filters for the purpose . ving iodines er particulates from tne gaseous exhaust stream - # .o the relea - the environment (such a system is not cons - _. to have any effect en nocle 3 affluents). Engineered i iafety Fe'*.. ;ESF) Etm:sene-i: :leanu: systems are not con, wrec :: te "~'~.un. ION EX.wAUST 'REATvEN S'fSTEM c cconents. PURGE - PURGINS l 36 'D-l' "rce a confinemen*. te a:.. tun tencerature, pressure, humidity, concentration tra-:ontro'.iec ;-: cess Of cisenarging air or gas a R3:.',3 'i PUF.GE ar d or other operating cone:::en, in sucn a manne-that reolacement air or gas is ecuirec :: _ri fy tne ::-t ' nement. @3jG - 1.51 . ;-; VENTING is the centrollec process of discnarg:ng tir or gas from a confinement to maintain tamoe ture, pressure, humicity, concentration or other operating concitien, in sucn a manner-that replacement air or gas is net-provicec or rectirec dur ng VENTING. Vent, usec in system names, does not i imply a VENTING process. m \\v WWIn dId I ~'I3 TT em _qu n y TRi
' RADIOACTIVE EFFLUENTS 'I ) (81:FGM) -s GASE0US RADWASTE TREATMENTaSYSTEM LIMITING CONDITION FOR OPERATION (eFF6%) The GASE0US RADWASTE TREATHENT 5YSTEM shall be in operation in 3.11.2.4 3 e,ither the normal or charcoal bypass mode. The charcoal bypass mode shall not be.used unless the offgas p,ost-treatment radiation monitor is OPERABLE as specified in Table 3.3.7.y -1. 0 0 -APPLICABILITY: Whenever the main condenser ste - jet' air ejector (c'!!cust 4^" system is in operation. ACTION: (orn M ) a. With the GASEGUS RADWASTE. TREATMENT 5YSTEM not used in the normal 3 mode for more than 7 days, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following i nf ormati on: 1. Identification of the inoperable equipment or subsystems and the reason for inoperability, 2. Action (s) taken to restore the inoperable equipment to OPERABLE (v) ' status, and 3. Summary description of action (s) taken to prevent a recurrence. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REOUIREMENTS (orF3M ) 4.11.2.4 The GASEOUS RACWASTE TREATMENT 5YSTEM shall be verified to De -in g operation in either the normal or charcoal bypass mode at least once per 7 days wnenever the main conce,ser ;.c r jet air ejector 'c ac a t-;-.) system is in operation. _,3_ 'l u ' *y p:CTON MUCLEra - UMP 2 3/4 11-it g) - ft AM -bN tT 1
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POWER DISTRIBUTION LIMITS 3/4.2.$ MINIMUM CRITICAL POWER RATIO '0,tional 7 00% Ortien A) LIMITING CONDITION FOR OPERATION w 3.2.4 The MINIMUM CRI CAL POWER RATIO (MCPR) shall be equal to or greater than both MCPR and MC limits at indicated core flow and THERMAL POWER as l f z P 2_ showninygures3.2+1ang33,g2gg9.yiggg,,393phe5.,g9.L..eap_,,,,,, j ..............y,, ..,,, s y.. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to (25)% of RATED THERMAL POWER. ACTION: -(er e end-of-cycle recirculation pump trip system in le per Specificatio o eration may conti e provisions of Specification 3.0.4 are no rovided that, within one hour, MCPR is determine qual to or grea h MCPR and MCPRp n Figures 3.2.3-1 and 3.2.3-2 by the EOC-RPT inoperab e n . 4br. main turbine bypass system inoperable per Specification 3.-7.9,' operation may d the provisions of ification 3.0.4 are not applicable provided that, w1 nx9 ~r PR is determined to be equal O to or greateW-both 14CP.Rj and MCPR, as s ures 3.2.3-1 and o .Lar3-2 b'y the main turbine Dypass inoperable curve.) With MCPR less than the applicable MCPR limit shown in Figures 3.2.6-1 -er P and 3.2.462, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours or reduce THERMAL POWER to less than (255 of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS l 1 4.2.0 MCPR shall be determined to be equal to or greater than the MCPR limit' determined from Figures 3.2.3-1 and 3.2.362: 2. 2 a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is operating c. with a LIMITING CONTROL ROD PATTERN for MCPR. h -d. % 9twiscac q s p [ p M.o q ce ed apha PERRY - UNIT 1 3/42-6[ n
POWER DISTRIBUTION LIMITS 3/4.2 POWER DISTRIBUTION LINITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION j 3.2.1. All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1,3.2.1-2,and3.2.1-3f, r . APPLICABILITY: OPERATIONAL CONDITI3N 1, when THERMAL POWER is greater than or equal to 125 5 of RATED THERMAL POWER. . ACTION: 5,c:; E 3.1 i With an APLHGR exceeding the limits of ff;=4 A.0^
- 3. 2.1 - 1, 0. 2.1 -2, ;;r 0. 2.1 -- 0, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 1 25 % of RATED THERMAL POWER within the next 4 hours.
udAqu t by sJr y a.dk e -% flow N e+b A M e ta t-c. ot at -& pw McAsi Huc a u _ f] &chr- (Mh9 hc4 ) A Hyu 3.2.i-y %, tau ncp q v.r.. u.i-s, i SURVEILLANCE REQUIREMENTS ~~ 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits .. -. i;; urn 3. 2.1 - 1, 3. 2.1 - 2, =d 3. 2.1 - 2: o( 5 (n i b sL..- 3,2.. l. p a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is l operating with a LIMITING CONTROL R00 PATTERN for APLHGR. i, d P=+oe,t ej. Sq,,. M -4.0.y d u, i are wot a s. l' { Lo t. . PERRY - UNIT 1 3/4 2-1 i. ,.,Nl
-POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 413.4).kw/ft. APPLICABILITY: '_0PERATIONAL CONDITION 1, when THERMAL POWER is greater than or i equal to (253% of RATED THERMAL POWER. ACTION: With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours or reduce THERMAL POWER to less than.f25 W of RATED THERMAL POWER within the next 4 hours. -SURVEILLANCE REQUIREMENTS 3 4.2.4 LHGR's shall be determined to be equal to or less than the limit: a. At least once per 24 hours,- I b. Within 12 hours'after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating on a LIMITING' CONTROL ROD PATTERN for LHGR. d. 'T b greddon e4 Sect:QaA-4.o.4 au w o N o w J:. v.c. i 10 PERRY - UNIT 1 3/4 2-9 Ri 4 ,e e ,,,v- - -,., -,. -, - -,,,,,... - - - -. -,, + - -. - -. -, - -. -, - - - - - - - -,,,.. - -.,. - -
k CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b. . At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone -communicating with the subsystem by: 1. Verifying that the subsystem satisfies the in place penetration and bypass _ leakage testing t::; ting acceptance criteria of less 0.05' than W and uses the test procedure guidance in Regulatory Positions C.5.a. C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1979.,, and the system flow rate is-(20C0) cfm
- 10L 2 coo 2.
Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 5; and 2cos - .t 3. Verifying a subsystem flow rate of (2000) cfm i 10% during syster g operation when tested in accordance with ANSI M510 M75. \\%o. I LO After every 720 hours of charcoal adsorber operation by verifying c. within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than {2*)E o.rts % d. At least once per 18 months by: 1. Performing a system functional test which includes simulated automatic actuation of the, system throughout its emergency operatingsequencefortheb ' a) CLOCAf,end 4) T::1 u; ling n idnt. l g ~ 2. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than inches Water Gauge while operating the filter train at a flow rate of fE3003 cfm i 10L zooo 3. Verifying that the filter train starts and isolation dampers open on each of the following test signals: a. Manual initiation from the control room, and b. Simulated automatic imitation signal. O Y5:!!0E9M.e. ! i..M.. !.!.!.T sE !F..., e..t.i. f r. -*"'-**" 6' **""'""- y....... 4-5. Verifying that the heaters dissipate f974}4 (L O) N when tested in accordance with ANSI N510-4474.1 % PERRY.- UNIT 1 3/4 6-41 g
s. ? CONTAll# TENT SYSTEMS ~ O-SURVEILLANCE REQUIREMENTS (Continued) After each complete or partial replacement of a HEPA filter bank by e. verifying that the HEPA filter bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than (2)E 0.05% in accordance with ANSI N510-1975 while operating the system at a flow rate of (2000) cfm i 10%. 1*l80 zooo f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than o.oS (13% in accordance with ANSI N510-19Mr for a halogenated hydrocarbon refrigeranttestgaswhileoperating[the system at a flow rate of -(2000) cfm i 10%. g row O l l N .05% value applicable when a HEPA filter or charcoal adsorber efficiency s assumed, or 1% when a HEPA filter or charcoalysorber efficiency o of 95% or is assumed in the NRC staff's safet m aluation. (Use the ( value assumed fo charcoal adsorber effi y if the value for the !~ HEPA filter is differen the char adsorber efficiency in the . NRC staff's safety evaluation. (** 0.175% value applic when a boiler adso f 99% is assumed, or. l 1%.value appl e when a charcoal adsorber effic of i 95% is d, or 10% value applicable when a charcoal a er e ency of 90% is assumed in the NRC staff's safety evaluati L l. l-L 1 !~ PERRY - UNIT 1 3/4 6-42 R\\ J
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p PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter _or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-cating with the subsystem by: 1. Verifying that the subsystem satisfies the in place penetrationf d D and bypass leakage testing acceptance criteria of less than-(^f% and uses the test-procedure guidance in Regulatory Positions C.5.a C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is (20^^)-cfm 1 10%. Soooo 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than p ; and uocoo 3. Verifying a subsystem flow rate of (20^^) cfm + 10% during sub-system operation when tested in accordance witE ANSI N510-1W9: Mo. d. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Posi-ton C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Po'sition C.6.a of -Regulatory Guide 1.52, Revision 2,_ March 1978, for a methyl iodide penetration of less than p %. e. At least once per 18 months by: Verifying that the pressure drop across theba.s l L 1. combined HEPA filters l-and charcoal adsorber banks is less than M inches Water Gauce m m while operating the subsystem at a flow rate of (2000T cfm 1 10%. l: ene rech M 4.s 2. . Verifying that on each of the below (i::L.. -) mode actuation l 6,g4. "- ' "gnals, the subsystem automatically switches to' the test si " mode of operation and the isolation d ::: close w,Ty1 within Co) seconds: da t r i a) (Chlorine detectioni, b) (fr :.ie deu - $ -), 1 % % c ox% aM..a, c) Lem A and l, d) h rhb.#. L . PERRY _- UNIT 1 3/4 7-6 g L .-.2..
w e.z..., - m..+ .. - m_ _,_ e _ _m. O O
t PLANT-SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) Verifying that on each of the below (pressurization),sode a on test. signals, the subsystem automaticaMr switches to the (pre ation) mode of operation a (control room is maintained at tive pressure inch W.G.) relative to the outside atmosp duri 4 system operation at a flow rate less than or equal cfs: a) (Smoke det n) -b) Ai ake radiation monitors, and ~ goo wva *S%-1o% 3-4r . Verifying that the heaters dissipate 0.5) _ (0. % ) L when tested in accordance with ANSI N510-19Mr. tagge, f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetr ion and bypass leakage testing acceptance criteria of less than in U'" accordance with ANSI N510-19 5 while operating the system at a flow rate of (2000) cfm 1 10%. I W Boooo
- .O g.
After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass leakage testing acceptance criteria i
- o. 5 of less than W % in accordance with ANSI N510-M95 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow l
rate of '0000) cfm 1 105. gg go 30000 I l l. / l value applicable when a HEPA filter or charcoal adsorcer efficiehcy of ( 995 is , or 1% when a HEPA filter or charcoal adsorberffficiency of 955 or less is in the NRC staff's safety e flon. (Use the value f assumed for the charcos rber efficienc value for the HEPA filter is different from the e rber efficiency in the NRC staff's safety evaluation). z0 o 17s= iue aaiic ai a charcoai ds-icieacy of esx is ssum d. or 1% value a when a charcoal adsorber effic f 95% is assumed, i or 10E v pplicable won a charcoal adsorber efficiency o assumed p C staff's safety evaluation. PERRY - UNIT 1 3/4 7-7 Rt
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e CONTAINMENT SYSTEMS pd SURVEILLANCE REQUIREMENTS (Continued) e b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by: 1. Verifying that the subsystem satisfies the in place penetration and bypass leakage testing testing acceptance criteria of less I' tha M and uses the test procedure guidance in Regulatory Positions C.S.a. C.5.c and C.5.d of_ Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is (2300) cfm i 10%. 1500o 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Geide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than (2a-)%; and I5on i 3. Verifying a subsystem flow rate of (2000) cfm i 10% during syster operation when tested in accordance with ANSI N510-4975. %o. O-After every 720 hours of charcoal adsorber operation by verifying c. within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than y. d. At least once per 18 months by: 1. Performing a system functional test which includes simulated automatic actuation of the operating sequence for ther, system throughout its emergency a) LOCA, :nd h} Fuel handling accident. q,3 l Verifyingthatthepressuredropacrossthe/)combinedHEPAfilters 2. ( and charcoal adsorber banks is less than (6 inches Water Gauge while operating the filter train at a flow rate of (2000) cfm i 10%. 15o00 3. Verifying that the filter train starts end iseletisa desper; -ope on each ;f th: f;ilewing test signels. e Manual initiation from the control room 7. end-5. Si=}eted nteaatic imitetien eiv.l. e i. Y:rifying th;t th filter ;;;1ing bype:: dampers can be-manually-
- e =e th: f= un bc = =14y-stagt g.,
'l S. Verifying that the heaters dissipate 497-3-) t (1.0) he when tested in acc.ordance with ANSI N510-1975. IatBo. l l PERRY - UNIT 1 3/4 6 M 1-3s ai t
\\ 4 CONTAINMENT SYSTEMS O SURVEILLANCE REQUIREMENTS (Continued) .e. . After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than W l in accordance with ANSI N510-19PS while operating the system at a flow rate of 12300) cfm i 10%. liBo 150co f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than I- '*)% in accordance with ANSI N510-MPS for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of (2000) cfm
- 10%.
. i s ooo RBo r l f / 1 j. ,s .x ( 05% value applicable when a HEPA filter or charcoal adsorbeF efficiency l of assumed, or 1% when a HEPA filter or char oal'adsorber efficiency I of 95% or s assumed in the NRC staff's y evaluation.. (Use the t i value assumed for harcoal adsorber ciency if the value for the i HEPA filter is different the oal adsorber efficiency in the NRC staff's safety evaluatio (** 0.175% value appl e when a boiler adsorber 1 99% is assumed, or 1% value ap le when a charcoal adsorber effic f 95% is d, or 10% value applicable when a charcoal a r e ency of 90% is assumed in the NRC staff's safety evaluatio j A O l \\ PERRY - UNIT 1 3/4 4-*t f- "I 'M gg
LEACTIVITYCONTROLSYSTENS -'b - 3/4.1. 5 STAN08Y LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5. The standby liquid control system" g e ns W s shall be OPERA 8LE. APPLICA8ILITY: OPERATIONALCONDITIONS1,2and5*. ACTIO_N: a.- In OPERATIONAL CONDITION 1 or 2: %d;sgt m 1. With one,.. ;;.,.7 ;..;.y. inoperable, restore _3 4, g .n.. the inoperable ;7 :n'/r ;;ph;iv;,el'v+ to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours. w.w sub mee m 2. With the standby liquid control system";t.% r ie; inoperable. Q \\J. restore th:' g;t r to OPER/.8LE status within 8 hours or be in one suspbe at-least H0T SHUTDOWN within the next 12 hours. b. In OPERATIONAL CONDITION 5*: 1.- With one ; 7 :28:Y2$e$h f : :h; inoperable restore _s,og the inoperable ; 7 r.d/r e-pieeive.el e to OPERA 8LE status within 30 days or insert all insertable control rods within the next. hour. was subw% 2. With -the standby liquid control system +4 hem $ee inoperable, insert all insertable control rods within one hour. SURVEILLANCE REQUIREMENTS seso+ws 4.1.5 The standby ' liquid control system shall be demonstrated OPERABLE: a. .At least once per 24 hours by verifying that; a eu,y A.,. i
- 2
'1. The temperature of _the sodium pentaborate solution'is ef hh the li it; ef-Tigr; 2.1.5--1. gue' h or e@ Ao 'o *F 4 2. The available volume of sodium pentaborate solution is geoaten j th:2 Or :;ri t: ( ) ;;1h::. eh %= Q*ds y F 4't L'5-8 3 3. The heat tracing circuit is OPERA 8LE by determining the. temperature of the (pump suction pipingt to be greater than or i: equal to 170)'F. r 'ifith any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. LO j PERRY - UNIT 1 3/4 1-18 O
A O 'I ~ 55 % .h t'i g a S tev(c,%; y g, O i 0 1 ( i p 1
l REFUELING OPERATIONS O'. 3/4.9.12 INCLINED FUEL TRANSFER SYSTEM LIMITING CONDITION FOR OPERATION '3.9.12 The inclined fuel transfer system (IFTS) may be in operation provided that: 9%5 The access doord'o. 4 %.c f all rooms through which the transfer system a. penetrates are closed and locked. All access h interlocks *a ga sehr b. are OPERABLE. g, ave V i c. The Versa blocking valve located in the fuel building IFTS hydraulic l power unit is OPERA 8LE. hete m.,e r g., ,3 o l -AWIFTS pri=rj : d :::: d j carriage position : d liquid 1:p e.wi<_ d.
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- e. ~ e ' h keylock switch'which provide [ IFTS access control-transfer system -e. lockout g 0PERA8LE. sI-l fl::warw y Ping light / outside of h.accessdoor/*isee OPERABLE. f. APPLICA8ILITY: When the IFTS containment blank flange is removed. < O ACTION: With the requirements of the above specification not satis ed suspend IFTS operation with the IFTS at either terminal point. The provisions of Specifi-cation 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS b i 4.9.12 Within.4 hours prior to the operation of IFTS and at least once per - 12 hours thereafter,- verify that: i i %w.% x.v is is,, o p..$ - ~ e. All GC;&is desi listsilGCks eis CEEEAOLE. f 5. The '!:r:: bic king -valve--in - t'. F 1 kMding-IFTSr hydr:ulic-power-viii i, is OPERA %E. Bt k *H ** *. a t ed. is L
- a. z'.
4 H IFTS pri n rj nd ::: =d:rj carriage position r.: 1;=1 'adi:: tees- -ere-OPERA 8LEr et each carr;a3e posh,., w d \\cos t L \\ ipa o#= j leon h s.r sJ be O P E R e 6 t,E. he Pr/ e k : witch dich ;=vid:: IRS :::::: :: t=1-tr=:fer :ystem-l. 1 l l- -4ect:;t i; O?EAASEE-l %e a 44 is /' b d.
- H fMi. light /outsideof'accessdoorIseeOPERA8LE.
] PERRY - UNIT 1 3/4 9-18 O
l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation. APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2". ACTION: red y oMacM pqe a. one reactor coolant system recirculation loo n operation, immed a tiate measures to place th n at least HOT I SHUTDOWN within 2 hours b. With no reactor system recircu o s in operation, immediat itiate measures tc place the unit n t STARTUP hours and in HOT SHUTDOWN within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by: a. Verifying that the control valve fails "as is" on loss of hydraulic pressure Tat the hydraulic control uniti, and b. Verifying that the average rate of control valve movement is: 1. Less than or equal to 111 5 of stroke per second opening, and 2. Less than or equal to T11% of stroke per second closing. "See Special Test Exception 3.10.4. ~ PERRY - UNIT 1-3/4 4-1 A R\\
('h O 3.w.\\, \\ A c T se a - a. With one reactor coolant system recirculation loop not in operation, imediately initiate an orderly reduction of THERMAL POWER to less than or equal to \\ima as specified in Figure 3.9. \\. \\-l, and be in at least HOT SHUTDOWN within the next 12 hours. b. With no reactor coolant system recirediation loops in operation, imediately initiate an orderly reduction of THERMAL POWER to less than or equal to %e limit-as specified in Figure 3.9. n.1 - l , and initiate measures to place the unit in at least' STARTUP within 6 hours and in HOT SHUTDOWN within the next 6 hours. O l' I l l l I i l ~ ~. -, u-a
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Fi. REFUELING OPERATIONS o SURVEILLANCE REQUIREMENTS (Continued) b. Performance of a CHANNEL FUNCTIONAL TEST: 1. Within 24 hours prior to the start of CORE ALTERATIONS, and " b least once per 7 days. 2. 0.7 4 c. Verifying that the channel count rate is at least + cps: 1. Prior to control rod withdrawal, 2. Prior to and at least once per 12 hours during CORE ALTERATIONS, and 3. .At least once per 24 hours. / d. Verifying, within 8 hours prior to and at least once per 12 hours 7 during, that the RPS circuitry " shorting links" have been removed for that the rod pattern control system is OPERABLEK during: 1. The time any control rod is wi' hdrawn," or t 2. Shutdown margin demonstrations. 4 .q [exceptthat: p h1. During spiral unloading, the required count rate may be persitted to be less than-S cpsi
- n
() 2. - -Priortoandduringspiralloading,ungilsufficientfuelhas ') been loaded to maintain at least-4-cps, the required count rate ( may be achieved by: OG b a) Use of portable external source, or b) Loading up to 2 fuel assembTies" in ::!h :;nteini.-- l t i %{ - 4 : rt:d :::tr:1 ::d: :- : d in-64M. Q. 4 og Jo A ( ( -%e..w SAM c._Os c.Md3 ins rkA
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m l ~' ' -- r l-k 9c odi6e A siyd % nJiu. eatto B 2. o % rd u, m S e 95. L - " Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2. - = - These fuel assemblies eay be loaded with the SRM count-rate less than 3 cps. ( PERRY - UNIT 1 3/4 9-4 l l t t
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s. _ _ _. CONTAINMENT SYSTEMS ~ LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued) restore: The overall integrated leakage ratefet-to less than or equal to a. 0.75 L, er 0.75 L, :: 1;; le, and t b. The combined leakage rate for all penetrations and all valves listed in Table 3.6.4-1f, except fort imain steam line isolation valves 9 fand)-(valves which are hydrostatically leak tested per Table 3.6.4-1,$. subject to Type B and C tests to less than or equal to 0.60 L*, and M The leakage rate to less than (11.,, (4C.0) scf per hour for 4any c. one) ( 11 ': r) main steam linc((s) th=t;;5 th:) isolation valve (-st, and d. _The combined leakage rate for all penetrations shown in Table 3.6.4-1 as secondary containment bypass leakage paths to less than or equal to((00).L,,and e. The combined leakage rate for all XECCS and RCICK containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to X1 gpa times the total number of such valvest (3 g,;), prior to increasing reactor coolant system temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using.the methods and provisions of Z I %?5.0 - (1070); hNsr_/ANs Ec. 8 - Re t. l' a. Three Type A Overall Integrated Containment Leakage Rate tests shall l be conducted at 40 + 10 month intervals during shutdown at P, H.M l~ -(-15} psig, ;r ;t " 7 ( } ;:4, during each 10 year servic$ period. .Thethirdtestofkachsetshallbeconductedduringtheshutdown for the 10 year plant inservice inspection. b.- If any periodic Type A test fails to meet 0.75 L er 0.75 L, ::
- pplic;ti;, the test schedule for' subsequent Typ$ A tests s fall be reviewed and approved by the Commission.
If two consecutive Type A tests fail to meet 0.75 L er 0.70 L, es opp i:: :, a Type A test t shallbeperformedatlealtevery18monthsuntiltwoconsecutive o Type A tests meet 0.75 L er 0.75 L as appli:d!:, at which time theabovetestschedule$ayberesuke,d. c. The accuracy of each Tjpe A test shall be verified by a supplemental test which: 1. ' Confirms the accuracy of the test by verifying that.the difference .n-. between the supplemental data and the Type A test data is within -U 0.25.L,. + 6,~ed % NteJJ<. -r of io cr e so. PERRY - UNIT 1 3/4 6-3 L
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CONTAINMENT SYSTEMS BASES 3/4.6.3 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the drywell and containment pressure will-not exceed the design pressure of 1301 psig andJ151 psig, respectively, during primary system blowdown from full operating pressure. The suppression pool water volume must absorb the associated decay and tois structural sensible heat released during a reactor blowdown from (10007 psig. Using conservative parameter inputs, the maximum calculated containment pressure during and following a design basis accident is below the containment design pressure of 115s psig. Similarly the drywell pressure remains below I 3\\8,5%uthe design pressure of 1301 psig. The maximum and minimum water volumes for j the suppression pool are 62",351) cubic feet and (135,11SI cubic feet, US A L respectively. These values include the water volume of the containment pool, horizontal vents, and weir annulus. Testing in the Mark III Pressure Suppression Test Facility and analysis have assured that the suppression pool temperature will not rise above 185'F for the full range of break sizes. Should it be necessary to make the suppression pool inoperable, this shall 3 .only be done as specified in Specification 3.5.3. I Experimental data indicates that effective steam condensation without excessive load on the containment pool walls will occur with a quencher device
- T'
'and pool temperature below 200'F during relief valve operation. Specifications have been placed on the envelope of reactor operating conditions to assure 'the bulk pool temperature does not rise above 185'F in compliance with the containment structural design criteria. In addition to the limits on temperature of the suppression pool water, i operating procedures define the action to be taken in the event a safety relief valve inadvertently opens or sticks open. As a minimum this action shall-include: (1) use of all available means to close the valve, (2) initiate l suppression pool water cooling, (3) initiate reactor shutdown, and (4) if [ other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open' safety relief valve to assure mixing and uniformity of energy insertion to the pool. The (containmenti ( d) (dr,,=li7 spray system consists of two 100% gP capacity h, each with three spray rings located at different elevations about the inside circumference of the icontainmentt (2nd) (d w il). RHR A War pump supplies oneM and RHR pump B supplies the other. RHR pump C cannot supply the spray system. Dispersion of the flow of water is effected by 350 l' bor nozzles in each tTain, enhancing the condensaton of water vapor in the lcon- .tainmentt ( W ) (d fr!!F volume and preventing overpressurization. Heat l rejection.is through the RHR heat exchangers. The turbulence caused by the ' spray system aids in mixing the containment air volume to maintain a homogeneous I mixture-for H control. 2 l ~The suppression pool cooling function is a mode of the RHR system and l: functions as part of the containment heat removal system. The purpose of the l system is to ensure containment integrity following a LOCA by preventing exces-sive containment pressures and temperatures. The suppression pool cooling mode (' is designed to limit the long term bulk temperature of the pool to 185'F [ PERRY - UNIT 1 8 3/4 6-5 ( 3\\
CONTAINMENT SYSTEMS O -SUPPRESSION POOL MAKEUP SYSTEM LIMITING CONDITION FOR OPERATION 3.6.3.4 The suppression pool makeup system shall be OPERABLE. APPLICABILITY: -OPERATIONAL CONDITIONS-1, 2 and 3. ACTION: a.' With one suppression pool makeup line inoperable, restore the inoperable makeup line to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. 2b. With the upper containment pool water level less than the limit, restore the water level to within the limit within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. ~ c. With upper containment pool water temperature greater than the limit, restore the upper containment pool water temperature to within the limit within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours.and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENT 5 4.6.3.4 - The suppression pool makeup system shall be demonstrated OPERABLE: 1 a. .At least once per 24 hours by verifying the upper containment pool water: l: 1. Level to be greater than or equal to feiWgnd 2. Temperature to be less than or equal to (426)*F. l. b. -At least once per 31 days by verifying that: The dN2Ds$ ora 0e/f: SED)$,bfYpool gate is-removedp A % 1. Govuer Each valve, manual, p epower operated or automatic, in the flow p.. w m ow u., 2. path that is not locked, sealed, or otherwise secure in position, is in its correct position. c. At least once per 18 months by performing a system functional test -which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual makeup of water to the suppression pool may be excluded from this test. l-O PERRY - UNIT 1 3/4 6-27 %\\ . e -
O O C; TABLE 3.3.2-2 (Continued) o ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE I TRIP FUNCTION TRIP SETPOINT VALUE e z j . Q 3. SECONDARY CONTAIMENT ISOLATION a. Reactor Vessel h ter Level - 12 9. 8 \\17. C. Low 4ew; Level 2 -> - (i a}-inches * '1 '53)-inches t I b. Drywell Pressure - High $ (1.73) psig 5 (,t.eg psig j .c.. i. u___.u..._.,_ u__. n.. Ext:::t ":d!:ti r. "!;h High 5 (4. 5) -"/hr**- 1 '5.5) "/;..'* 1 ,g ra_ _ i. u. A 1 4. ,..a.m.m.. n_-1 c..___ Cat.e et ::edietier, iiigt. High- .1 (35) ;;;::/hr** - i '35) ^/;.. c.-e. Manual Initiation NA NA i 3 / + y i
- l
~ T 4. REACTOR WATER CLEANUP SYSTEM ISOLATION gg 1'!. I i a. A Flow - High 5168K gpm 1.'(JJ-) gpa w n p b. A Flow Timer iMT ' )1' ) seconds iM ( )1' ) seconds c. Equi p.t "ree T- ; retare Otigh 1' )"T 1( )*T Se_e_ seO d. .m...-... .y. 2s r.,- 2s, ,i 9% e. Reactor Vessel Water Level - 129 8 12r6 j Low 4ew, Level 2 '51) inches * '53) inches 1 j f. Main Steam Line Tunnel (12.9 113.9 Ambient Temperature - High 5 f--)*F $ +--)"F i~ g. Main Steam Line Tunnel 61.6 5 61 4 j A Temp. - High 5 -(---)* F 5 4--)*F j h. SLCS Initiation NA NA l 1. Manual Initiation NA NA U / .+ f i N 3 1 a
.( . _ \\, k; ET ACTot = Whit t C Ls Awu p sMSitM ~TsotAtiow
- r. s.
w.v. C. IN'ae d-P ten. "Teger.be- %)k
- 1. bwcw ax
- s. o. ~, Pu u, woms, g m,q -
f 138,9 W oe NesA hom 2 RVo C u ' Ditti+J boms, M MiW Valu e. 3 \\33 9 . Ro., ; Dt wiw ' Recc.w. T.*R. Row y]. A_/ g; g wA htex 't> T ' *^ t - O s p -
- c ^
g, . ft w c w WK ~R=.* d 7'1.f.5 i 18,y 2; Rva cu, 94 M. E**'*5, NS W E d 30,6 5 1314 9est.h.- 3' Rweu tto,w hou s, tf Mi+3 Valv e R***, t1M 4 Rec ewe r %E R e.u. d 71.f.5 6734 ./ 6 ?t A tV - ud t v i 3/4 3-IB b U"" 9*5* ) \\
LO O 1 l TA8LE 3.3.3-2 (Continued) a 4' EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION'SETPOINTS ALLOWABLE } TRIP FUNCTION TRIP SETPOINT VALUE C. DIVISION 3 TRIP SYSTEM y 4 Y T,,127. 6 O j 1. HPCS SYSTEM \\29.8 7 i
- a. ' Reactor Vessel Water Level - XLow 4-ow, 1-(51-)- i nches*
1-f6& Finches Level 2K t.ss i.es l b. Drywell Pressure - High $ ((1.0^) psio 2.s9.5 5 (1. %) psia _.22i.7 c. Reactor Vessel Water Level - High, Level 16K < - 4 Winches * < '52.0 Tinches-4'* 5 d. Condensate Storage Tank Level - Low i (X+3) i=h:. 5 (X) f chr-** 30'oo S j 3 {v-3) inchee#)B' 35" 3 ((Y) i A; ;; e. Suppression Pool Water Level - High i s ' 6.s l f. Pump Discharge Pressure - High 1 fins) psig 1 tzo) psig g. HPCS System Flow Rate - Low > 1725) gpa > T(ook gpa j t* (h. HPCS-Su; ";.er ".ersiter 5-( }-(volts) 5-( )-(velts)) w -1. Manual Initiation HA HA i Y -fr e 4 w D. LOSS OF POWER i { 1. 4.16 kw Emergency Bus Undervoltage a. 4.16 kv Basis - Soto 1 151 l (Loss of Voltage) 'f0))- god '2%0)1(151) volts (2%^)i(~;15) volts t b. 120 v Seeis - -(04)i(4.5) velis (34)i(")'.'1t; l b -e: < K sec. time 3.o to.ss delay 3.0 t o.o ig < fle-)-sec. time delay 2. 4.16 kv Emergency Bus Undervoltage. Edd --. 4.16 kv Basis - a 3s32. t 4o (Degraded Voltage) ~ i --- (3727)i(^,) volts (: 727)1'21) volts -k 120 ; ", ;is - g.ns--v.sfagv.caj 4 M S.O ! o.25 mio, (1^5.5)i(0.50) ;;It'; b ***-
- C.
5,0 2 0.5 n*w. sinuin n r-s-s-.., time delay -(10)i(1) ;ec. time delay 1 "See Bases Figure B 3/4 3-1.' ~ M N *~ ia, n5 t 1.5 5e c Loe A % C-l (**X i ; v;!;; th;t :n;;res ;d;;;;t; %"0" ;;d precludes air entry due te-verleningD)- da.D ) i ( fY i'; (&}-inch::; ;b;ve-normal eter le.el.) i mmrt__ _ ___ i. 2_ ..i i si. .mit is__ 2_s_ vo_ j g erws was us u s.ww.as w i nes wwsey vesseys suseya vs s e sa sess tev esvua russays s u s umy a w w wss e i. s uus wwwey. u nsw I YOlIIGOC Ch^^^^ C M thO Z Z$-^^^ thOt ^^i)) lot ie Se)t ii e tiip. '_Gwei ^v'G's Ieys es.,d$1 e..A ^w^sll iese t j [ in decreased ti!p ti 5.)~ s
1 INSTRUMENTATION C 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM (0pti;nal) l I ~~ LIMITING CONDITION FOR OPERATION ~ 3.3.8 At least one turbine overspeed protection system shall be OPERA 8,Er-5 Combe
- A APPLICA8ILITY:
OPERATIONAL CONDITIONS 1 and 2. ' jl' c ACTION-
- * *
- A ' *k y of Stop r a.
With one turbine control valve}"one turbine thr;ttis stop valve ce l
- t;-b b
- 7
- h
- :t : top v h; per high pressure turbine steam leed bac 1 ino)erable,and/or with one turbine 5t:n:ptr' valve per low pressure _
,km' tur)1ne steam +eed-inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours or close at least one valve in the b ~ affected steamNeed or isolate the turbine from the steam supply within the next 6 hours. b. With the above required turbine overspeed protection system otherwise inoperable, within 6 hours isolate the turbine from the steam supply. SURVEILLANCE REQUIREMENTS nb 4.3.8.1 The provisions of Specification 4.0.4 are not. applicable. 4.3.8.2 The above required turbine overspeed protection system shall be demonstrated OPERABLE: a. At least once per 7 days by: 1. Cycling each of the following valves through at least one complete cycle from the running position: a) For the overspeed protection control system; 1) Four high pressure turbine control valves, and Six C o mb i c e.C 'i n$ e r a e P T 2) -Fete low pressure turbine intercepter valves b) For the electrical overspeed trip system and the mechanical overspeed trip system; 1) Four high pressure turbine thr;tt h-stop valves, SU \\ow comb *=a Wter m b Ne 2) T;;r high pressure turbine-rch::t stop valves, 3) Four high pressure turbine control valves, and Q Sh c a +.k D '.+ w c. e t V 4) -Fote low pressure turbine interceptee valves. PERRY - UNIT 1 3/4 3 90lo t R\\
@ %s. Lc o, 6 R, ced Tdle. nv=ftry 4-r< ~ b e s'6f*^i a'in 7 4 W-575 nu%he<g, Q p Hou %Awsv offys Hydrym Morb r (t% 4 - No tz a{e) does not monitor re6 area hr dose asessani. radidan u d 4Ls is sot O O l l l 9 l
7 REACTIVITY CONTROL SYSTEMS i : SURVEILLANCE REQUIREMENTS (Continued) b. At least once per 31 days by; sh peM.<A4 1. Verifying the continuity of the explosive charge.
- *% * ^
551F 2. Determiningthattheavailable[weightofsodiumpentaborateisi greater than or equal to (",00, lbs and the' concentration ef-5:r;; in ahthr. is within the limits of Figure 3.1.5-1 by chemical analysis.* l 3. 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. a-c. Demonstrating that, when tested Xpursuant to Specification 4.0.5K (:t h =t e..e. p., % dey;), the minimus' flow requirement of 141.2) gpmiat a pressure of greater than or equal to $1220k psig is met. e= r e* ~ e d. At least once per 18 months during shutdown by; d s sh ees 1. Initiating e A
- ef the stan liquid control system }eepe, bv including h explosive valve and verifying that a flow path e
from the pumps to the reactor pressure vessel is available by pumping domineralized water into the reactor vessel. The l .( replacement chargesfor the explosive valver shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch success-1 fully fired. "ett, injntie-hep; et,un % t.est.d in % re..th:, Demonstrating that'thjNor., 12. pump relief valve setpoint is less than'or equal to (10.., psig and verifying that the relief valve does not actuate during recirculation to the test tank.5 3.
- Demonstrating that all heat traced piping between the storage tank and the reactor vessel is unblocked by (pumping from the storage tank to the test tanks and then draining and flushing the piping with domineralized water.
t w exp,a,a o e nd * *g i5 g 4. Demonstrating hatthestoragetankheater[4rlOPERABLE by verifying temperature rise cf the sodium pentaborate 4 i solution in the storage tank by et heet "F wittiin l
- i=t= after the heaterfere energize is "This test shall also be performed anytime water or boron is added to the o*F solution or when the solution temperature drops below the !!
- it ;f Tigere -[.,.1.5-1.
- This test shall also be performed whenever both heat tracing circuits have been
-found to be inoperable and may be performed by any series of sequential, over- . lapping or total flow path steps such that the' entire flow path is included, r O. i i PERRY - UNIT 1 3/4 1-19 Rt
e l l l t i Sa-s 4 - T< usa s, sg,. O l i I i
/\\ ,-~ d U - Is f REGION OF APPROVED - 14 ( VOLUfE - CONCENTRATION k \\\\\\ g g0VERFLOW S G.9 t%RG i ta VOLUME 13 LOW LEVEL ALARM k \\\\\\ ( 3 12.9 "g5 MINIMUM REQUIRED vg CONCENTRATION LINE =.u h 5 12 0 8;;
- u em v a.
- 11 4603 4843 5046 10 4400 4500 4600 4700 4800 4900 5000 5100 5200 V - NET TANK VOLUME (GALLONS) 73 e f-
- 3. i. s - i
%A. %h b-~h 5 W-C "
- t'- / V '""'
R ' 1 * ~ #'
Am-.u .a u _m
- a.
~<- 2 .4 O f Q\\ b $m h I l 1 t i l i t I
r" iLj TABLE 4.3.7.J1-1 (Continued) TABLE NOTATIONS g g..J 4 4et.44 4. % b k U 4 % c.i The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic (1) isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured' levels above the alarm / trip setpoint. 2. High voltage abnormally low. 3. Instrument indi:ates a downscale failure. 4 Instrument controls not set in operate modef o<.cgi lc Wp *Hy goM. (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm setpoint. 2. Instrument indicates a downscale failure. 3. Instrument controls not set in operate modet enee+ b Wp v*H9e PM. The initial CHANNEL CALIBRATION shall be performed using one or more (3) reference standards certified by the National Bureau of Standards (NBS) or q' using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit qj calibrating the system. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. -(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. GHANNEL CHECK shall be made at least once per 24 hours when continuous, periodic, or batch releases are made. N: CP ^""E L cL9CT!C'^ '. TE st ;h:!' 0';; d:ren;tr:t: that cut tic ::c'atic-h(5) 2^^r ciati ^ cccu- '"; c ' t" 4' ^f this p 2 t *" ey 2 ^ d c ^^t ee ' ec ^ 212 *= a fc?! J g :^^ditf^^: er!!!!: 'kn $1 ?"-* 4 '"*^^4a' 1. ? art rat '^direte' ~~esurad t=" ate 9^"a 11 i mz_ 2. ut;s, gig 3;g 399-3 7 In;t m :.; 2 mcicatn ; du. ;;0! ' i! J;- q v iT 4RM -u N d I .. r. : ;.. ~0 4 LU ^ ";T 0-3/4 3-Ser R\\ qq
U i ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be: a. Determined OPERA 8LE at least once per 7 days by verifying correct breaker alignments and indicated power availability, and b. Demonstrated OPERA 8LE at least once per 18 months during shutdown by transferring / rxMy ;r.d n" rtid% unit power supply from the normal circuit to the alternate circuit. 4.8.1.1.2 Each of the above required diesel generators shall be demonstrated OPERABLE: 3 a. In accordance with the frequency spectfied in Table 4.8.1.1.2-1 on a STAGGERED TEST BASIS by: 1. Verifying the fuel level in the day rd n;is x xted fuel tanke. 2. Verifying the fuel level in the fuel storage tank. 3. Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day r d r ;f= n = t d f x! tanke. {:- 4. Verifying the diesel starts from ambient condition and 882. re.{., inst s. TcTerates to at least -(^^O,, in less than or equal to ~ o^a. @ \\ eem seconds.* The generator voltage and frequency shall be_ io be im.s.. o k 14160K i 1420) volts and 160) * (1.21 Hz within (4&Tseconds uu,,..;.. q, after the start signal. The diesel generator shall be started for this test by using one of the following signals: lo a) Manual. b) Simulated loss of offsite power by itself. c) Simulated loss of offsite power in conjunction with an ESF actuation test signal. d) An ESF actuation test signal by itself. 5. Verifying the diesel generator is synchronized, loaded to f looo areater than or equal-to (nr.tir..; reti@ kw for diesel .:.r.oo g,g,,g aeneratorsW and 44) and inr.tf== r;tir.?) kw {or diesel generator 44,63 in less than or equal to.4609 seconds, and " t 1 - 5 ** ' - operates with this load for at least 60 minutes. 6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses. 17. Verifying the pressure in all diesel generator air start receivers to be greater than or equal to -(450) psig.K .b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour by checking for and removing accumulated water from the day.and h r;f n n=td f=? tanke. . w~n PERRY - UNIT 1 3/4 8-3
-- = 1 ELECTRICAL POWER SYSTEMS
- O
~ SURVEILLANCE REQUIREMENTS (Con'tinued f. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting all three diesel p generators simultaneously, during shutdown, and verifying that all three diesel generators accelerate to at least M rps; in less than or equal to y seconds. 4,% a 3,a,,..,a., i s n-s. a dm keA g yqi, gr a y. J.r i g. At least once per 10 years by: , a g.%. 3.., g, 1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a 1 sodium hypochlorite)or *W'A solution, and 2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection NO of the ASME Code in accordance with ASME Code Section 11 Article IWD-5000. 4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid. shall 2 be reported to the Commission pursuant to Specification 6.9.M Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number Q-of failures in the last 100 valid tests, on a per nuclear unit basis, is greater 'than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Ravision 1, August 1977. O PERRY - UNIT 1 3/4 8-s gs 1
EMERGENCY CORE COOLING SYSTEMS ( SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression pool shall be determined OPERA 8LE by verifying the water level to be greater than or equal tp4; eppli;;ti:: (10'","I et leeet
- ;;r Oi h
- u n 16'6" A
-(--) at least once per '(12K hours.
- 5. #
4.5.3.2 With the suppression poqV1evel less than the above limit or drained inOPERATIONALCONDITION4or58,atleastonceper12housf i a.t l g r. Verify the required conditions of Specification 3.5.3.& to be satisfied e.-er b. ";rify f:: tact: ::nditi r.:
- te b; ::tisf4ed:-
un 40% -h sq.s r.. go.\\ +.t t e gw e. R Ao b* ormet. L o pe s oi. e.l. conAl+.a s + d go.s A o.a' per \\1 k wc M
- h j
so. A 6.At. corbNoas + -t
% e soil
v PERRY - UNIT 1 3/4 5-9 El r QOCESS CONTROL PROGRAM (PCP) ,3(') (.35 +90- The PROCESS CONTROL PROGRAl-1 shall contain the ::r-!' ;, 2^:D'i' f:=;!:tha cate....Ratia b n:s 00;.:0:r =T 0" c f mi:::tiv: ette: ' e- 'i;;id ;yst;;; ;; :::. ;:.k current prins(a., SupD att, M t5 Sad e,n j dehmarnaFou to te ude f, cass,s -rbt e ctss,n .d c'4 ef Seb3 radiefhe n/a4tc.s fated e^ dea,estr ed ce vf a dd*I of c5ej solid pHM dill le m,9 sked so su<1 a **y a s t= 5 blded u,of /i as6de tooqft w w0k (oCFQ20 to CPRloI to efK'1I, al kO*'O ** Side reqvlaflon5 burial {eenJ re oire 5, y{ p%f r5siccina.ds f $ d*I oSN c( Y a ionc$VL wssic, governi^ f t SITL 'boJND g y N oddMN d
- 1. 9 3 T he sgu Bod W 1 560 k Ost Ec 1
(s'u d, nor.4hrwssa. c o n + <,lled % 6 I:c e n n e, isnek omed or i vNRe sTnecTen etare ( s Ls0 be t n3 are*-a.k or tejo^d ik
- l. 5 0 l}n JNKESTAtLTEb MEA Snt MWY mes b wk'k is et c dc.lled b 6 l'ceva be &
y for%Lof~protstdtei of,' slid?deds N tAysore 4= r ala.Es
- 64 s
g rs4ioat.tQs 4 ;<Ls, ee u4 atsa. u'EN In ~%. snit?'B00t40kc,y vsed fu t'e d esto%l $ va,eten -(,*- indvstvl0, eeeueiJ, isstitoh4J ar al{or-rstrerNnJ f aefases. VENTILATION EXHAUST TREATMENT SYSTEM l *Il 44-A VENTILATI0li EXHAUST TREATMENT SYSTEH is any system designed and installed to reduc 2 gaseous radiof orfine or radioactive material in particulate form in ef fluents by passing ventilation or vent exhaust gr.ses through charcoal adsortiers and/or HEPA filters for the purpose of removing lodines er particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any ef fect on noble gas ef fluents)j [c. 4(e. Wbery butOlap radkw5fc 6,e Ut'n, o,nd cwAcwWeA ou,e s eMM*- 6 3 Sys% my,and prt#% y asom m f u d al/afwell wrje 59%. Safety Feature (EST) atmospheric cleanup systems are not consicered to be VENT!!.ATION EXHAUST TREATHENT SYSTEM ccmponents. h) b D ** Q N \\*T \\ l-l'L R\\ e \\o. Fo be% (g 6.h.kd bec==w -N- %sW s+o p w.6 : h vt goiMa s4E.br s*>Acket. ioA r \\r 4-Af CM 7e1 <Ps 311 b km d *M'I - $2., 24 phac u O O ADMINISTRATIVE CONTROLS n EVENT V 6.6 REPORTABLE seGUARENCE ACTION Euw-rs 6.6.1 The following actions shall be taken for REPORTABLE SectfRRENCES: mn..A w. % rquitw,,is og s,h soa2.o( iocre.SS a. The Commission shall be notified /and a report submitted pursuant to the requirements of Spe;ifi;;tica C.0, ;nd red..a so.n w so c ra. t,a s o 8, s%c l'o 8 C-I Each REPORTABLE GGEORRENGE 7: wiring 'i-5 0:_-i::f:n shall be reviewed by the M :,r :ti'ic:tien te th: b. and the results of this review shall be submitted to the (CN ZO) and the 4Vice President - N5 Nuclear 0;;r:ti:::). Gr.-e. 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated: a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour. The XVice President - Nuclear 0;Ge..g:r;tiens) and the (CN MC) shall be notified within 24 hours. wsac %RC b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the -(BAG}. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon unit components, systems or structures, and (3) corrective action taken to prevent recurrence. c. The Safety Limit Violation Report shall be submitted to the Commission, cMsCNMG), and the (Vice President - Nuclear Op;ratiens) within a 14 days of the violation. 6"" d. Critical operation of the unit shall not be resumed until authorized by the Commission. O PERRY - UNIT 1 6-12 R\\ Cq TABLE 3.3.1-1 (Continued) REACTOR PROTECTION SYSTEM INSTROMENTATION ACTION ACTION 1 Be in at least HOT SHUTDOWN within 12 hours. ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the SHUTDOWN position within one hour. ACTION 3 - Suspend all operations involving CORE ALTERATIONS
- and insert all insertable control rods within one hour.
ACTION 4 Be in at least STARTUP within 6 hours. ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours or in at least HOT SHUTDOWN within 12 hours. ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to less than the automatic bypass setpoint within 2 hours. -p ACTION 7 Verify all insertable control rods to be inserted within one V hour. ACTION 8 Lock the reactor mode switch in the SHUTDOWN position within one hour. roh ACTION 9 Suspend all operations nvolving CORE ALTERATIONS *, and insert all insertable control and lock the reactor mode switch in the Shutdown position within one hour. "Except movement of IRM, SRM or special moveable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2. l PERRY - UNIT 1 3/4 3-4a gg i O O OL TABLE 4.3.1.1-1 Aj REACTOR PROTECTION SYSTEM INSTRUE NTATION SURVEILLANCE REQUIREE NTS CHANNEL OPERATIONAL E CHA8AIEL FUNCTIONAL CHANNEL ColWITIONS IN WHICH Q FUNCTI000AL UNIT CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED w 1. Intermediate Range Monitors: a. lieutron Flux High S/U.S,(b) S/U(c),W R 2 5 W R 3,4,5 b. Inoperative NA W NA 2,3,4,5 2. Average Power Range Monitor:(#) IC) a. Neutron Flux - High, S/U S,(b) S/U ,W SA 2 Setdown S W SA 3, 5 b. Flow Biased Simulisted w 5,D (h)) i ge) y y(d)(e),34,g(i) j -f .2 Thermal Power - High c. Neutron Flux - High 5 -6NN,W W(d), SA 1 d. Inoperative NA W llA 1,2,3,5 3. Reactor Vessel Steam Dome IU) -) Pressure - High 5 M R 1, 2 4. Reactor Vessel Water Level - I9) Low, Level 3 S M R 1, 2 5. Reactor Vessel Water Level - I9) 1 High, Level 8 5 M R 6. Main Steam Line Isolation Valve - Closure NA M R 1 i 7. Main Steam Line Radiation - High 5 M R 1,2(j) XRt(g) 1,2(J) 8. Drywell Pressure - High TSt M TABLE 3.3.1-1 (Continued) Ag REACTOR PADTECTION SYSTEM INSTRISENTATION E APPLICABLE MINIIRM
- 1 OPERATIONAL OPERABLE CHANNELS FtBICTIONAL INIIT CONDITIONS PER TRIP SYSTEM (a)
ACTION y 9. Scram Discharge Volume Water. Level - High 1I93 2 1 a. n..i.; m e / t.r. e " *. S ~~ 2 3
- 10. Turbine Stop Valve - Closure 1(h) 4 6
- 11. Turbine Control Valve Fast Closure, IhI Valve Trip Systes 011. Pressure - Low I
2 6 t'
- 12. Reactor Mode Switch Shutdown Position 1, 2 2
1
- tYl 3, 4 2
7 5 2 3
- 13. Manual Scram 1, 2 2
1 3, 4 2 8 { 5 2 9 - s. nm - s ,,m z 5 41 3 a-G O u.n. t., .a.1 '" 9 aa s +-t l mum a x,, I i 1 l l I l l l l l f l
.c,
-n--..- .__-.----,-,_,,.n._ . _, - - -_. -,,.. _ _,,., - - -,. - -, - - - - - ~,,
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued) REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel. APPLICA8ILITY: OPERATIONAL CONDITIONS 3, 4 and 5 ACTION: With the reactor vessel water level at or below the top of the active irradiatedfuel,manuallyinitiatetheECCStorestorethewaterleve1A*Ne Depressurizing the reactor vessel,'if r;gir;d. Comply with the requirements of Specification 6.7.1. L Eces p h. l O I O PERRY - UNIT 1 2-2 gg
u O O O TABLE 4.3.1.1-1 (Continued) 3 REACTOR PNOTECTION SYSTEM INSTRUDENTATION SURVEILLANCE REQUIREBENTS 'E CHAf00EL OPERATIONAL CHAISITL FUNCTIONAL CHANNEL COISITIONS FOR )Af!CH k FIBICTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED ) 9. Scram Discharge Volume Water I9) 1,2,5(k) a Level d".fahfw., *M 4-N A M R n. N
- 10. 7urbine Stop Valve - Closure NA M
1 R
- 11. Turbine Control Valve Fast Closure Valve Trip System 011 M
Pressure - Low NA M R 1 1
- 12. Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,5
- 13. Manual Scram NA M
NA 1,2,3,4,5 5:' (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. (b) The IM and SM channels shall be determined to overlap for at least 11/24 decades during each startup after entering OPERATIONAL C00GITION 2 and the IRM and APM channels shall be determined to overlap for at least $1/21 decades during each controlled shutdown, if not performed within the previous 7 days. (c) Within 24 hours prior to startup, if not performed within the previous 7 days. (d) This calibration shall consist of the adjustment of the APAN channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THEIMAL POWER. Adjust the APM channel if the absolute difference is greater tEan 2% of RATED THERMAL POWER..t; AT channel-gairedjestment-mode-in templiance-with f+;;fficeth; 3.2.2 2:11 r. t h !=?W h tu=in!q th 2; lete-4ifforence.- (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal. mp (f) The LPRNs shall be calibrated at least once per 1000 effect.;; Nil ;r:7 5: r; (ET."") using the TIP system. (g) Calibrate trip unit at least once per 31 days. ( T /C rete, m ly) 44 g. Th) Verify measured sawe-flow to be greater than or equal to established som flow at the existing. flow control valve position. G ro.t. (i) This calibration shall consist of lverifyingt ( 9.j= ts ;t, = m k J vi) the 661 second simulated thermal power time constant. (j) This function is not required to be OPERA 8LE when the reactor pressure vessel head is removed per Specification 3.10.1. (k) Withanycontrolrodwit(drawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. 3 b. N x. s N~ s m q R g,5 ~ M
i CONTAINMENT SYSTEMS 4 *" '. " ) ' * - + e H,c a a.* g 4 stt J M . og.,. b ei
- % % h*W O
, c- ~ < ~ x<< - - - CONTAINMENT AIR LOCKS ad o AJ.c A. pn p g,. % read.c u.u d. LIMITING CON 0! TION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERA 8LE with: { i a. Both doors closed except when the air lock is being used for normal 4 transit entry and exit through the containment, then at least one air lock door shall be closed, and 2.5 sc{ h b. An overall air lock leakage rate of less than or equal to 0.0', l., at P,,Q.gpsig. APPLICA81LITY: OPERATIONAL CONDITIONS 1, 2*;end-3, acA %. ' ACTION: a. With one containment air lock door inoperable: -1. Maintain at least the OPERA 8LE air lock door closed and either restore i the inoperable air lock door to OPERA 8LE status within 24 hours or lock the OPERA 8LE air lock door closed. O ~~ 2. Operation may then continue until performance of the next required i overall air lock leakage test provided that the OPERA 8LE air lock door is verified to be locked closed at least once per 31 days, i in oeimi.e4 Ac cowtim..J \\,1... e b
- 3..
Otherwise,"be in at least H0T SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. 5.4. The provisions of Specification 3.0.4 are not applicable, goonaam.wAt cowo m.a L,1., c 3; b. With the containment air lock inoperable,* except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERA 8LE status within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- h ana containment air lock door inflatable seal syntam_ air-flasF pressure instrumenun.iv., Cxx' iaaa--%,mitiFe the inoperable channel to 0Pf818'M.t.i.v. w1 thin 7 days or verify ai, firesk-pressures _
-te v. 3,30 psig at least once per 12 hours. 5** ud gap "See Special Test Exception 3.10.1. 4 ww u s. Al.i' D fu.k a bug WWeQ. s ^ % Q' '*' n L a d" "'E s O --s , c -.t a 'm < - ~,, a r< " e-+ "'" y na+ r..n + a. PERRY - UNIT 1 3/4 6-5(L R\\
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4 6.0 ADMINISTRATIVE CONT,ROLS 4 6.1 RESPONSIBILITY %me e, %rt, WA o t* <b*^' D * &^*~ > The (Ple..; %,erint;;i..t) shall be responsible for overall unit 6.1.1 T operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Supervisor for,during his absence from the control room, a designated individual)'shall be responsible for the control room command function. A management directive to this effect, signed by the (high;;t Isal of cuir-.atej r: : p t) shall be reissued to all station personnel on an annual basis. / w ti,s.a,.4 - 6.2 ORGANIZATION g,gl ' C o (eo e. m g, emn cor g orde. 6.2.1 The effd te organization for unit management and technical support shall l be as shown en Ff;;r; 5.2. ' '. a cqw a.4 % Fsac. UNIT STAFF A cw.r 6 -4 6 E5 u 6.2.2 The unit organization shall be as shown :n Pf;;r: 5.2.2 ' and: a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1; '. O b. At ieast oes iicensed Operator shaii be in the coetroi room waea fusi is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Operator shall be in the control room; c. A Health Physics Technician
- shall be on site when fuel is in the reactor; d.
ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation; e. A site fire brigade of at least five members shall be maintained on i site at all times". The fire brigade shall not include the Shift Supervisor, the Shift Technical Advisor, nor the two other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and ~ "The Health Physics Technician and fire brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours, in order to accommodate unexpected absence, provided immediate action is taken to fill 'the required positions. O I PERRY - UNIT 1 6-1 g
ADMINISTRATIVE CONTROLS O 10 NIT STAFF (continued) f. Administrative procedures shall be developed.and implemented to limit the_ working hours of unit staff who perform safety-related functions Xe.g., licensed Senior Operators, licensed Operators, health physi ciste, auxiliary operators, and key maintenance personnelk. inMdans [ J The amount of overtime worked by unit staff members performing l safety-related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).t' I y g l [ dequate shift coverage shall be maintained without routine he ! -v ' us of overtime. _The objective shall be to have operating per onnel work normal 8-hour day,'40-hour week while the. unit is op ating. Howeve in the event that unforeseen problems require su tantial amounts overtime to be used, or-during extended peri s of shut-down for_re eling, major maintenance, or major unit. difications, F on a tempora basis the'following guidelines shall e followed: - 1. .An-individua 'should not be permitte'd to w k more than 15 hours straight, exc1 ing shift turnover time. p' 2. An individual shoul not be permitte to work more than 16 hours V in any 24-hour period, ner more th 24 hours in any 48-hour period,' nor more than 7 hours i any seven day period, all' excluding shift turnover e. 3. A break of at least eight our should be allowed between work - periods, including shif turnove time. ~ + - 4. Except-during exten d shutdown peri s, the use of overtime should be conside d on an individual b is and not for the entire staff on shift. Any deviation fr the above guidelines shall be a horized by the (Plant Superi ndent) or his deputy, or higher leve of management, in accordan~ with established procedures and with doc entation of' i- - the basis r granting the deviation. Controls shall be
- cluded in the pro ures such that individual-overtime shall be revie d monthly-
- by th Plant Superintendent) or.his designee to assure-that ces-siv curs have not been assigned. Routine deviation from the ove g delines is not authorized.] W s 2 1 f e-y PERRY:. UNIT 1-6-2 gg gwg ww -weegy,ry t-'yr
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l ADMINISTRATIVE CONTROLS 6.4 -TRAINING it g 4 Ge6.,J 59rk c Aretrainingandreplacementtraininghogramfortheunitstaffshall 6.4.1 be maintained under the direction of the (pe;iti:n titic), shall meet or exceed,s,5 1e requirements and recommendations of Section -P-} of (en ANSI Standard eccept. Ml; to N the %",0 :t:ff) and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 N NRC letter to all licensees, and shall include familiarization with relevant M industry operational experience. 6.5 REVIEW AND AUDIT ~ d by which independent review and audit of unit operations i accomplishe ke one of several forms. The licensee er assign j this function to an o tional unit separate ependent from the group having responsibility for unit on utilize a standing committee composed of individuals from wi the licensee's organization. Irrespective of th od used, the licensee shall sp the details of each functiona nt provided for the independent review and au i ss as ated in the following example specifications. 6.5.1 (U"IT "EVIC'd C"^U" (U^C)) 9am oggewous ggvi u c3 % m sg (foc c ) NCTION 7O Rc g,,,p,s, p,,t1 p g,,g n,p,g, g 6.5.1.1 The (URG) shall function to advise the ("lant Seperintendeat) on all pd _ matters related to nuclear safety. jie.N;i. / Mea,< s Maaayr,4: Pbd 3 COMPOSITION as
- e. h ea Tec kaM De ed-Jt.
Po(c. 6.5.1.2 The i&RG) shall be composed of the: G. c y,y,, g,3,,, syr g,a.A y w,,,% % a,a 3, 4,A 7 ("leat ceperiatendeat) MacaSer,9nr3 Pia A OP" A* 0* P M'# Chairman: Sr; rvie:r) G,e..ca 5+'es e, oea'M a$ s,"*.. Member: (Oper:ti::: Teckad S'*,** Member: (T::h-i:21 Supervi;;r) Ge..,.L 59As.% E Supe rvi ser) G= a='A M"1"'@R"#,* *"*
- 5"'
- e. c Member:
(".:f-' ::::: Mer.ber: ("1: t Instrrr:nt :nd C ntral Casin;er) Redar %*r Member: ("leat cieer En;;i ::r} Cweer A Sgeres% C w ae r, L E= C Member: 9bk ;(Health Physicisty 9t d'.b ^
- s..A..
ALTERNATES ' 6. 5.1. 3 All alternate members shall be appointed in wribngh the (URGh Poncc - Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in 4URG) activities at any one time. Fo Rc. MEETING FREQUENCY Po#C 6.5.1.4 The (UR4 shall meet at least once per calendar month and as convened by the (URG) Chairman or his designated alternate. Potc. QbORUM Poe.r Po4c 6.5.1.5 The quorum of the f0RG) necessary for the performance of the fGAG) responsibility and authority provisions of these Technical Specifications e shall consist of the Chairman or his designated alternate and four members including alternates, w & _ ch.,', w, 5,co< os rneub rr pr % % CGr= is pres,q. PERRY - UNIT 1 6-7 R.\\ .. -. - -. ~. -.
ADMINISTRATIVE CONTROLS RESPONSIBILITIFS g g 6.5.l.6 The (URG) shall be responsible for: / Review of (1) all proposed procedures required by SpecificationA.8 and changes thereto, (2) all proposed programs required by Specifica-ion 6.8 and changes thereto, and (3) any other proposed prMedures or ch es thereto as determined by the (Plant Superintende ) to affect nuc1 safety; b. Review o all proposed tests and experiments that fect nuclear safety; c. Review of a roposed changes to Appendix A Te nical Specifications;. __ d. Review of all p osed changes or modificati s to unit systems or equipment that aff t nuclear safety; e. Investigation of all v lations o.f the echnical Specifications, including the preparatio nd forvar ng of reports covering evaluation and recommendations to prev t re trence, to the (Vice President - Nuclear Operations) and to the ompany Nuclear Review and Audit Group); f. Review of events requiring hou ritten notif.ication to the -._ Commission; 4 g. Review of unit operati s to detect pote ial hazards to nuclear safety; h. Performance of spe ~ 1 reviews, investigati , or analyses and reports thereon as reque ed by the (Plant Superinten t) or the (Company ,,}g . Nuclear Review nd Audit Group); V i. Review of t Security Plan and implementing procedu. s and submittal of reco ded changes to the (Company Nuclear Review'a Audit Group) and j. Rev w of the Emergency Plan and implementing procedures and mittal the recommended changes to the (Company Nuclear Review and A t roup). 6.5.1.7 The (URG) shall: See cdd.acd he Recommend-in writing to the (Plant Superintendent) appr royal of items considered under Spscificationg 6.a. through
- d. pri their implementation.
I b. Render determinat in writing with ard to whether or not each item considered under Sp icati .5.1.6.a. through e. constitutes an unreviewed safety questio Providewrittennojifcationwithin24ho ~to the (Vice President - c. Nuclear Operat afs) and the (Company Nuclear Re ' and Audit Group) of disagreement etween the (URG) and the (Plant Super ent); however, the 1(nt. Superintendent) shall have responsibility for r tion such disagreements pursuant to Specification 6.1.1. ( PERRY - UNIT 1 6-8a. N .,,r... an-n,,,. ,---...n.,n-----e
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-ADMINISTRATIVECONTRblS [) RESPONSIBILITIES . PoRC 6.5.1.6 The -eE-snall oe responsible for: a. Review of all Administrative Procedures; /;mw M.s S V b. -Review of.ne safety evaluations for (1) procedures (2) change to /dstrut'.., F procedures', equipment, systems or facilities, and (3) tests or exper-iments comoieted uncer the provision of 10 CFR 50.59 to verify that such ac.icas did not constitute an unreviewed safety question; '/ sh sif acAeEet /M sirecA de c. Revi?w of proposed procedures
- and changes to procedures", equipment, system or facilities which may involve an unreviewed safety questiqn as defined in 10 CFR 50.59 or involves a change in Technical Specifications; s
d. . Review of proposed test"or experiments which may involve an unreviewed safety question as defined in 10 CFR 50.59 or requires a change in Technical Specifications; e. Review of proposed changes to Technical Specifications or the Operating License; f. Investigation of all violations of the Technical Specifications including the forwarding of reports covering evaluation and recom- , - (] mendations to prevent recurrence to the Vice President-Nuclear and s V to the Mfinfh Ac. leu Safety Review Co v.tte e desop sqc4c et y _g. Review of. reports of operating abnormaliti,es, deviations from expected
- performance of plant equipment and of unanticipated deficiencies in the ~ design or operation of structures, systems or components that affect nuclear safety; h.
Review of all REPORTABLE EVENTS; i. Review of the plant Security Plan and shall submit recommended changes to the NGABi-Acteu 5 % h #.,~ Cw.'Aec; aa %,\\e *,e% metad =es j. Review of the Radiological Emergency Response Plan #and shall submit k Jseia C.ddhe; recommended changes to the ft&Rfh-Wc.Lew fye.t3 k.- Review of-changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL, and Radwaste Treatment Systems; 1. Review of any accidental, unplanned or uncontrolled radioactive release _ including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent 8 recurrence and the forwarding of these reports to the Manager 7 Oc 1 my Plant,and to the NSAB'. Wc.tew $$ty Review C.. ;ttee, -n-1 Teg be#ts ',) PERRY C O L' W - UNIT 1 6-S b ( a,s ,) Rg a ~
7-ADMINISTRATIVE CONTROLS RESPONSIBILITIES.(Continued) m. Review of Unit: operations to detect potential hazards to nuclear safety; -n. Investig'ations or analysis of special subjects'as requested by tne Chairman of' the NSra; :nd Waeu 5. jet) hview C. J Jkes, 9:. i c; c f L'r i t h rb i n: 0.-cr:pred " ret:: tier 7:liabi'::. ' reg : re.!;ien: ther:tc. PORC 6.5.1.7 The eRe.Shal1: ['fd "gS s Pe rt3 a. . Recommend -in writing to the Manager! C_!!;w Plant approval or disapproval of items considered under Specifications 6.5.1.6a. ! t h ro u g h e., i,, J., k., g., -aad-e-above; l oed b.~ Render. determinations in writing with regard to whether or not each item conside' red under Speci fications 6. 5.1.6b. through e., and -. above, constitutes an unreviewed. safety question; and c. Provide written notification.within 24 hours to the Vice President- - Co mu tb e F erog Nuclearland the Nuclear Safety Review 30ird-of disagreement between the'4RE and the Manager, C 1?:acy Pldnt; however, the Manager, co nce i p1 d O I I 6 "".y P'oirt shall have responsibility for resolution of such [ pj,,16ee-disagreements pursuant to Specification 6.1.1 above. w g,, A 2 op. b rA~* +k ~ fg g g (ne.a y .C;..: ~ .~
- 3. PTOCE53 CDNTROL PROGRM1 implemen8aSton.
~ ^ ~ h.: OFF5ITE DOSE CALCULATION MANUAL implementation. Quality Assurance Prograin for effluent and environmental monitoring, ,g, ~ using the guidance in Regulatory Guide 4.15, Deceeder 1977. /= & heen m ) 6.8 PROCEDURES *AND PROGRAMS g /insscu: % 6.8.1 Written procedures'shall be established, implemented, and maintained covering the activities referenced balow: 1 a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. b. h: :;;110:51: prn; durn r;;;ir:d t; i pi;n nt th: 7:quir;nnt: cf . ~..... b e. Refueling operations. c d. Surveillance and test activities of safety-related equipment. 6 -e. Security Plan implementation. e -f. Emergency Plan implementation. 5 S Fire Protection Program implementation. wy espem M %g&c< &s, 6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed by the (HRf& and shall be approved by the XPlant S.p.. -:nte. dent 3 prior to implementation and reviewed periodically as set forth in administrative pro-cedures. fo u ~/ orary changes to peoseduces of Specification 6.8.1 may be made pro-y- / O Ta $at at ar riai" i * $= a t4Tt r d b. The change is approved by t ers of the unit management staff, at least one of whom s a Sen erator license on the unit , affected; and c. ange is documented, reviewed,by the (U=), an roved by the v.... ,..rir.te..de..t) within 14 days of implementation. ~ 3 6.8.+ The following programs shall be established, implemented, and maintained: l a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the XMPC-I, 45, RHR, RCIC, hydregen recerLiner, es, pr;n:: : 7'q, :: t:1..nr.t ad :t:ndt3g;strestr:nthsystems. The program shall include the following: trw.tes 1. Preventive maintenance and periodic visual inspection requirements, and per 6.c 2. Integrated leak test requirements for each system at-refuel Mg cycl:-intervals.er lese. O PERRY ~- UNIT 1 6-13 a Rt g- --.--g
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ADMINISTRATIVE CONTROLS -s()
- 6. 5.- 3 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.3.1, Activities which affect nuclear safety shall be conducted as follows:
A * $, d.,,, /E;it e am a. Procedures" required by Specification 6.8 and other ;rocedures%nicr. affect. plant nuclear safety, and changes thereto, shall be prepared. reviewed and approved. Each such procedure or procedure' change v shall be reviewed by a qualified individual / group other than the , f, yg,,, F individual / group which prepared the procedure or procecure change, \\ r which prepared the proceduref or procedure' change.
- m du
- cther-)
but-who may be from the same organization as the individual / group 5 eta u t b +t 9 har id+h tr;ti.; "r ::dur:', shall be approved by the appropriate _ o gp.g[. At ~ Cnema.L Sqee-ter/ U:p;rtment " :d as designated in writing by the' Manager, E !!r eyy M,),m,3-Met +. The Manager 9 Cali =y'o!: t, shall approve Administrative
- "M*"'*, i Procedures /-S;curity Pl:n impica. eating pro::dur: and Radiological p-
"S ' # ' " Emergency Response Plan implementing prc::duY:;. uTemporary changes \\ g g,;, t_o procedures;which do not change the intent of the approved proce-dures7 shall be approved for implementation by two members of the plant staff, at least one of whom holds a Senior Operator license, -m p.mg and documented. The temporary changes shall be approved by the 9.q N. A i original approval authority within 14 days of implementation. For 7e h,t 'o changes to procedures 1which may involve a change in intent of the tefA **d, V - approved proceduresi, the person authorized above to approve the M WUwe procedure shall approve the change prior to implementation; 6ecdy tw A,e s m hoc 5j gWe t e'u W... A cq b. Proposed changes or modifications to plant' nuclear safety-related e u d.iie s, structures, systems and components shall be reviewed as designated by the Manager, eg!:=y Picnt. Each such modification shall be f*\\'. reviewed by a qualified individual / group other than the individual / b '*"'*1 group which designed the modification, but who may be from the same 3 D't*
- d-organization as the individual / group which designed the modifica-tions.
Proposed modifications to plant nuclear safety-related structures, systems and components _shall be approved prior to implementation by the Manager;5Callowey Plent, peq gig Dep%,Jq. c. Proposed tests and experiments which affect plant nuclear safety and are not addressed in the Final Safety Analysis Report or Technical Specifications shall be prepared, reviewed, and approved. Each such test or experiment shall be reviewed by a qualified individual / group other than the individual / group which prepared the proposed test -beforeimplementationbythe[jandexperimentsshallbeapproved or experiment. Proposed test Manager,, Colle ;y P!:nt; Peit N d h p A w d s. 3 s s
- q t/
PE RN C A JJ,2 - UNIT 1 6-11 b (w ha I R)
ADMINISTRATIVE CONTROLS ) an6 meet er exce* 4 ** M9'@' (' Y' " '" ^ * ' ~~' ACTIVITIES (Continued). 'of SecAiu 4.9 o{ ANrr tJes. '#9 21. d. Individual's resconsible'for' reviews performed in accordance with Scecifications 6.5.3.la., e.5.2.'t., and 6.5.3.lc., snall be
- .O f' p he fi x;ly designated by the 2.W@A
~0eN c' *": pian. : n: ;; cat Serc[ Manager, CII e-ey '" n t. Each web review 'snall include a deter-y mina. ion of wnetner or not additional, cross-disciplinary, review is gcI we4 p Plant 3, p,.t, necessary. If deemed necessary, such review shall be performed by Wacwapa qualified personnel of tne aporcpriate discipline; wdk Sye @L,, -e. Each review shall include a determination of whether or not an [f,f,f,'"e,', unreviewed safety question is involved. Pursuant to Section 50.59,..A 65.u.c. 10 CFR, NRC approval of items involving unreviewed safety questions shall be obtained prior to the Manager, Cal!y y o!:nt, approval g for implementation; and s Per: 9\\ad.'DegA ce #5 3 f. The Plant Security Plan and Radiological Emergency Response Plan, mtwt%ug and implementing pr :: dure, shall be reviewed at least once per r 12 months. Recommended changes to the implementing ; r:cedure: Ms % cA'**- shall be approved by the Manager, Cella-;j 0 :nt. Recommended _ _g changes to the Plans shall-be reviewed pursuant to the requirements pj of Specifications 6.5.1.6 and 6.5.2.,6-and approved by the Manager, 7,,,, g C a 'yi a-a, II:nt. NRCapprovalshall/beobtainedasappropriate. g 9ecr 9hJ: 'Tec6s'ut h p.A eed.. 7 ,m y \\v] n h g.p Pit 69 07.LLARAY -' UNIT 1 6-11 c. W 99) R\\
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F_ REACTIVITY CONTROL SYSTEMS
- 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION
-3.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY.and the predicted ROD DENSITY shall not exceed 1% delta k/k. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With the reactivity equivalence difference exceeding 1% delta k/k: a. . Within 12 hours perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected. b. Otherwise, be-in at least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS' '4.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall be verified to be less than or i-equal to 1% delta k/k: a. -During the first startup following' CORE ALTERATIONS, and b.- At least once per 31 ;ffecthe fell pw;r ip during POWER OPERATION. r L \\oco Swt/r 1-PERRY - UNIT 1 3/4 1-2 M i.
(- ' ELECTRICAL POWER SYSTEMS fl 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS V DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following power distribution system divisions shall be energized: with ti br ;i;rs sp;n (beth) betu;;n redund;nt bu;;; within the unit (;nd ht =:. =lt: :t t, : ::. : ::: tic,C: t g ;- t-A- o r, 5 F i. A - o s, E F-t - E - o s, a. A.C. power distribution: E F-t- A - W, EF-t - B -o 1, E F-t 6-oB EP-t-s -oK ass EF-I-8-toft F-t-D $o ? 1. Division 11P, consisting of: E B-t- M *1 Ek-t- M 3) 4160 volt A.C. bus (f unn ).dd a M % v.aA.c.buerc.ty) 480 volt A.C. MCCs ( ). EF-t-B - o' ' 8 5 6 *
- A d -e) 120 volt A.C. distribution panels n X480 volt MCCs" Er-t-e.
a a. Er-t-A-o1 g ) (;a;79;;;g 773; in;;7;;7 y cc ::ted te O.C. E :ica (1)* :nd 400 valt bu:(:as) - -) ;- _F e) no vow A.C. bus EV-l-A persua S.e 'mwer*ee-W 4-son c. &.s 5 Tyh".*.cD b"* I. 0 C b*5 EO - t - B-o". ? Division'62t, consisting of: As-t-t-on E F -t-e-o e, e c. l-c-o% sc-s o* } ES-i-c - it,, t e-i o-o i, g r i.e. os' 4 a) 4160 volt A.C. bus TE.att ). ) f f-I-b *% a.4 f f r-g.io/p.g.o.,3 f ei -61 aat f.x-t-e t xc4) 480 volt A.C. MCCs (-
- nd t
d-e-) 120 volt A.C. distribution paneisiin 1480 volt MCCs U-t-S;a' oa EST-c-o' and ) (energized frc: inverter # ~. (j ei no 9.w 6.c. ps tv-t-s e....yu n %...o.c m q. sos c.~. m a,, -eennected to 0.0. T5Tsica (2)* sad 400 voit bus (sas) ). 3.
- Division 131, consisting of:
RC b** u -t-8-= S M a) 4160 volt A.C. bus (EHt3gEF-t-E-i uA t c-t- t. s 480 volt A.C. MCCs ( ik-i-ci b) c) 120 volt A.C. distribution panels in '$480 volt MCCs *g,-i-t. j t (2n:rgized fre invertar #
- onnected t; 0.0.-
Ohi ica (3)* :nd 400 velt b;;(:::) ). b. D.C. power distribution: 1. Division 111, consisting of 125 volt D.C. distribution panel ( ). EO-t- A-or a.n. tAtc. - E.0- t - A - og. o 2. Division 12K, consisting of 125 volt D.C. distribution panels ( ). ED S -o b a4 EO l-6.o s, acA MCC. ED - t - B - 09, 3. Division $31, consisting of 125 volt D.C. distribution panel ( ). Kn - sori. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. %"One inverter may be disconnected from its D.C. source for up to'24 hours for the purpose of performing an equalizing charge on the associated battery _ g bank provided (1) its bu:::/"COs/p 15 are 0"C"J"LC and energized, and u (2) the bases /"CC;/penele essecie"Id with the other batterj banks-are-OPERABLE and energized.t h oh r Wet Asr q w o 'n H %ou.aacc E, - t-3.. f,,.,,,,, y g,, j t ac 2-e..,. PERRY - UNIT 1 3/4 8-16 VJ
9, as a.43 + m e. a o -.3 s..s _. 4 . s s., 4 m ssm. O 1 l i l i I l I
v ELECTRICAL POWER SYSTEMS h: DISTRIBUTION - SHUTDOWN-LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following power distribution system divisions shall .be energized: a. For A.C. power distribution, Division fly or Division 12k, and when ~ the HPCS system is required to be OPERABLE, Division 13k, with: g EG-l. A-o ;, E F-i-A-08, E F-i A - 09, '1. Division lit consisting of: 1 E H - A-n., E F B - o7, C F- / - S - o s
- y a) 4160 volt A.C. bus (Edit )
gg ""
- A C -
c.-&) 480 volt'A.C. MCCs ( u- -ni )g '"" E F-t 6 d -e) 120 volt A.C. distribution panels in (480 volt MCCs E F-8 8, * ' r . a+a E F-t-se " ~ 04 EFp-A-ol qqd -) (;;;rgi Od 77; inVcrt0r f
- ==:t;d t; 0.C.UEEision (1) and ?00 velt bes(1e5)
). r e) no 9.a g,c, w, tv-i. g...,763 go. ;o., %, g gy. so n,,,,,,,,3 g, M% *" 6.c. 2. Division X2). consisting of: V c *' t ' - t - A - o r.. . bgte, p p.g.c ,- Ef-t-c.o y g g-g. g.o g t r-t - c. oi, ad Ec i-o. a) .4160 volt A.C. bus 1f u n. X. J t r-t- c -it', t r-i -o-oi, s c-i-o - os, c-b) 480 volt A.C. MCCs ( =d )P B +*t al EF-B-t -'*N F-*-5-'@ is-i-s o.A E w.-i.es 1~ - d-e) 120 volt A.C. distribution panetsT in 1480 volt MCCs EF-'- 0,-o1 55 $ -o1 end 3-(eneriiized fre;i inverter # ccmerted te D.C.E;f = (3) =d 400 voit bs:(;c:) }. e3 vio y W A.c. ws tv-t- e eu,@ A z, ,~,,4,,. ig. iq. wig c u, A,1 .() 3. Division $3) consisting of: w t.c. dos e e-t-e,-os, a) ' 4160 volt A.C. bus TE dis J.7 -l-1-1 EF od EF-l-f.1. 480 volt A.C..MCCs ( ). s t i-c-i b) c) 120 volta.C.'distributionpanele9n(480voltMCC,sp--g-, k (=:rgi::d fr;;; in:rt:r f connected t D. C,- Divizi= (3) =d '00 V;1t bs(se;) ). -b. For D.C. power distribution, Division il):.or Division.12K, and when the HPCS system is required to be OPERA 8LE,. Division X34, with: 1.- Division ill. consisting of 125 volt D.C. distribution panel ( ). Eo-t-6-o6 ca. tAcc. to-t-A-o3 2. Division 42K consisting of 125 volt D.C. distribution l-panels ( ). to-i _ s.os na to-i-6-os, oA We so-t-B-ov. 3. Division 13k consisting of 125 volt D.C. distribution panel (. ). 1 CA 2.- sovr. APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *. [' "When handling irradiated fuel in the ;G = d ri ;Gt i R Gt. l' eMi.g c-eh~ e h or p).~ + + %o 3-W ke r- ' EF - t-B-w/ E F-t-o - vo sk.u W e.,[q,p -(c., A.e N '5 ' - ( I or 2. - pow. r. PERRY -UNIT 1 3/4 8-18 Rt p
=. =. ELECTRICAL POWER SYSTEMS ( SURVEILLANCE REQUIREMENTS (Continued) c. - a+ 1aast once per (31 [if ground water table is ma"=1 t er highTf-than the bottom or tn= r O) (^^) ye by removing accumulated r:te, i.um sne fuel storage tank (s). Sef pd -d. leastonceder92 days.andfromnewfueloilpriortoadditionto pqe th torage tanks, by obtaining a sample in accordance with ASTM-0-1975, and by verifying that-the sample meets the foll ng minimum quirements and is tested within the specified time imits: 1. As soon s sample is taken from new fuel or prior addition to the sto e tank, as applicable, verify in ac dance with the tests sp fied in ASTM-0975-77 that the le has: a) A water and diment content of less an or equal to 0.05 volume pe nt. 4 b) A kinematic viscosi 9 40*C greater than or equal to 1.9 centistokes, but s n or equal to 4.1 centistokes. c) A specific gravity as-e ied by the manufacturer 9 60/60*F of greater than or al to ' but less than or equal to or an AP ravity 9 of greater than or_ equal to degr but less than or ual to degrees. 2. Within one wee after obtaining the sample, erify an impurity 'O-- level of le than 2 mg of insolubles per 100 when tested in accordan with ASTM-D2274-70. 3. With two weeks after obtaining the sample, verify t the o r properties specified in Table 1 of ASTM-0975-77 a egulatory Guide 1.137, Position 2.a, are met when tested accordance with ASTM-0975-77. -e. At'least once per 18 months, during shutdown, by: I ~ 1. ' Subjecting the diesel to an inspection in accordance with smwer..e GM 6 e ' prepared in conjunction with its manufacturer's recommendations for this class of standby service. i 2. Verifying the diesel generator capability to reject a load of -m. greater than or equal to (bigt &;;1: z gery la=N kw for " ' ' * ' ^ diesel generator 14A), greater than or equal to (largat ingT *,* " 'S x;rger.cy leeO kw for7Tesel generator t4B-), and greater than or equal to (1:rgu t ; ingle ;. rgency i n O kw for diesel _l 22mo lin-Soon - generator 'f4C4 while maintaining ivoltage at X41601 i X420) volts _/and =frquxy :t (50) 1 (1.2) ll;) 4enthe speed gwa of Inr W5 G y, s ted mM difference between nominal speed and the overspeed trip eva + +ou setpoint or 15% above nominal, whichever is lessh o 3. Verifying the diesel generator capability _ to reject a load'of ou-Loai A i looo -- ...ing). kw for die.sel generators f4At and f1() and e............ y.g,g wooo ?;;rtin== reting) 'kw for diesel generator -tifd without - ng,y tripping. The-generator voltage shall not exceed X4784f volts h m-soot *Mduring and following the load rejection. ~ tRW Soot 6,3 d . 59W M& t t 2.t - So o t E -PERRY . UNIT 1 3/4 8-4 l R\\ c
l 1 b^ M 'I l hew. 6 - L (, r..c -b % UL s =~ 3 b d cy a >b 4 w c.#s s%k wh bsst S, ~5/4. 8 L t ( c % TN 0-1) O l O' I
r DEFINITIONS V CORE ALTERATION
- 1. 7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vesse jl. Suspension
-of CORE ALTERATIONS shall not preclude completion of the movementfof a component to.a safe conservative position. C %A %g q gm CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY e c).D /.[c 1.8TheCOREMAXIMUMFRACTIONOFLIMITINGPOWERDENSITY(CMFLPD)Y11b'ethD highest value of the FLPD which exists in the core. CRITICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the MGEXLK correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly m, operating power. 1,lo --p DOSE EQUIVALENT I-131 .Wh
- l 1.1h DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries A
per gram, which alone would produce the same thyroid dose as the quantity and , g,M -isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. A The thyroid dose conversion factors used for this calculation shall be those U listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." s DRYWELL INTEGRITY t 1.11 DRYWELL INTEGRITY shall exist when: All drywell penetrations required to be closed during accident a. conditions are either: 1. Capable of being closed by an OPERABLE drf.cl1 automatic isola-tion system, or 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as pro-vided in Table 3.6.4-1 of Specification 3.6.4. 74e 6 b. -AH drywell equipment hatchee-are closed and sealed. ~ m conciam iAs rQ~.k A 6 e. The drywell air ock is -0FERA",LE p....at te Specification 3.6.2.3. e 4. The drywell leakage rates are within the limits of Specification 3.6.2.2. A
- c. A.
6A % up'*M i 4-e. The suppression pool is GTERA::LE pun un t te Specification 3.6.3.1. O i+. The seaiins mechaaism associated with each dryweii Peaetr t4ea; e.g., welds, bellows or 0-rings, is OPERA 8LE. "c. % ays S.A w im akd a4 w a.A, -PERRY - UNIT 1 1-2
t TABLE 1.2
- ~
OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE 1. POWER OPERATION Run Any temperature
- 2..STARTUP Startup/ Hot Standby Any temperature 3.
HOT' SHUTDOWN Shutdown #'*** > 200*F 4. COLD SHUTDOWN Shutdown #'"'*** 1 200'F 5. REFUELING
- Shutdown or Refuel **'#
$ 140*F 10 - cua un,-.au
- The reactor mode switch may be placed in the'Run g Startup/ Hot Standby,or L i uJL
_ position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technica11y' qualified member of the unit technical staff. NThe reac : i mode switch may be placed in the Refuel position while a single control. c drive is being removed from the reactor pressure vessel per SpecificatNo 3.9.10.1.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
- See Special Test Exceptions 3.10.1 and.3.10.3.
- The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled provided that the one-rod-out interlock is OPERA 8LE.
- See Special Test Exceptions 3.10.1 and.3.10.3.
PERRY - UNIT l' M g I-iG
7 -_ _.. l. [ . TABLE 3.3.2-1 m m E ISOLATION ACTUATION-INSTRIMENTATION ' VALVE GROUPS MINIMUM APPLICA8LE E OPERATED BY OPERABLE CHANNELS OPERATIONAL. Q TRIP FUNCTION SIGNAL PER TRIP SYSTEM (a) CONDITION-ACTION H 1. PRIMARY CONTAllMENT ISOLATION a. Reactor Vessel Water Level-(b)(c) ,u,_t Low 4ew, Level 2 -f3)s-is-el,5,7,8 2-1, 2, 3 : d P 20 sue _s two b. - Drywell Pressure -- Hioh -(2, 01'"'"' f 2.5,8,9 2 1,2,3 20 2 mettExhaustPlebmab c.4..~.a. aw g) <u s s t it M' "'"' & 2 1, 2, 3 and
- 21f-
-(c. um Radiation - High 1, 2, 3.and g Ds.w ta.,s.,,a,9 d. U nual Initiation (2,3,5)-- $2)/(;;r:;) - 1221 . e. bcs,- %.1 LAW Leod _ 2 2. 1, 2, 3 2o w Loa,Leoet i .1 2. MAIN STEAM LINE ISOLATION 'i' a. Reactor _ Vessel Water Level-g Low L a t w, Level 1 (1, 5) C-2 1,2,3 20-b. Main Steam Line 6 Radiation - High (1, 7)Id) 1/fliner 1,2,3 23 c. Main Steam Line l Pressure - Low -(+) 6 _ 1/flinet 1 24 d. Main Steam Line W Flow - High -(+) 6 2/flinel 1,2,3 23 l e. Condenser Vacuum - Low -(4} G 2 1, 2,** 3** 23 3 f. Main Steam Line Tunnel Temperature - High -(4)- 6 2 1,2,3 23 i g. Main Steam Line Tunnel j A Temperature - High -(+) 6 2 1,2,3 23 l h. Manual Initiation -t1, 5, 7) 6-42b"gg)- 1,2,3 122k l l 1. ww., t.,at, pg. 6 2. 1, 2., 3 23 S%n t.' +.e. W.cf a -re.,e,, w
1
- 0) Nf LCO tw nbar fu rwk e.onsISied NA TM1b-ST5 numh*'*).
4.4 per f 4 4 R T a.ife 11 5 - 4 1 t-D's (sustfluities) g I & V d *f C 0 O $$ (& S 9 O
n. TABLE 3.3.2-2 -g ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE 1. PRIMARY CONTAINMENT ISOLATION w a. Reactor Vessel Water Level - 129.8 12 L C Low 4:ew, Level 2 1--(413-inches
- 1-f53}-inches I.(oB l.8B b.
Drywell Pressure - High 5 (1.73) psig 1 (-1-93) psig C d, -, A. o a O r p. a E,r3e .fc. -Man 6 Exhaust Plenum Radiation - High $ i2t mR/hr** 5 4 4 ) mR/hr**K d. Manual-Initiation NA NA R e. Re.c%e \\lemt W iu LeM - 2 16.5 kack,5 114.3 bck.5 Lo w, Le.2, t i T 2. MAIN STEAM LINE ISOLATION _gu a. Reactor Vessel Water Level - 16.5 19 3 Low-Lew-Lew, Level 1 - 1-(100) inches
- l-(152) inches-3 3.6 b.
Main Steam Line Radiation - High 5-(Gr O x full power background 5 -(+9) x full power background '7 RL 7 3L 7_ c. Main Steam Line Pressure - Low 1-(-859) psig 1 (638) psig 183 191 d. Main Steam Line Flow - High $ 4004 psid 5 f3 M) psid 8.5 7.c e. Condenser Vacuum - Low 1 (9-6} inches Hg. (:t::!:t: 1 f6-7-)- inches Hg. (:t::ine pr :::r:).tvacuunt-pr ::..;) Tvacuusk f. Main Steam Line Tunnel ti2.9 11 3.9 Temperature - High i {---}'F 1 t--)*F g. Main Steam Line Tunnel
- c. t.c 5 42 4 a Temp. - High 14--)* F
$ f---)*F h. Manual Initiation - NA NA 70 1. i a r b.,,. ~6 % ___A g u,,.g. g- $ its.9,' n.- t:. a,n rne..s,
m, @cyc sR nua,f $ emi,, unesd du wP-su n,b.'g. cp TNPP -ar n*me -nd no, der O @ Cbne.l feedGn dest - a 0) c% b PG)7erGr. col d gm (Ac. a wra etar as Ja;ma ssky prawsn Add
- sug,acJ 6y deg ein ca z O
0
e O O O I' TABLE 4.3.2.1-1 Ag ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL j CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH y TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 1. PRIMARY CONTAINMENT ISOLATION a. Reactor Vessel Water Level - Low tw, Level 2 S M R 1,2,3':d" b. Drywell Pressure - High S M R 1,2,3 fc. EE Nst P e 7 ' Radiation - High S M R 1, 2, 3 and *K d. Manual Initiation NA XM(a)MA} NA 1, 2, 3 and k '* R e. Eeoe%c d'uel Wde r Leo.1 - 5 M R 1, 2, 3 Low,Leoe.1l w 2. MAIN STEAM LINE ISOLATION a. Reactor Vessel Water Level - Low-4:ow 4ew, Level 1 S M R 1,2,3 b. Main Steam Line Radiation - High S M R 1,2,3 c. Main Steam Line Pressure - Low S M R 1 d. Main Steam Line Flow - High S M R 1,2,3 e. Condenser Vacuum - Low S M" R 1, 2"*, 3** f. Main Steam Line Tunnel l, Temperature - High S M R 1, 2, 3 g. Main Steam Line Tunnel a Temp. - High 5 M R 1, 2, 3 iM ") W I h. Manual Initiation NA NA 1,2,3 bh% Mail 5 M R \\,2, 3 i. hwe Tic m Li., e H sq k y r,_c., m,<
4 .m... -,.m A_ O co u ~ c.s.t.s.s u a u sa. a O l 0
EI a I i 1-RADI0 ACTIVE' EFFLUENTS .q LIQUID HOLDUP TANKS
- LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in any outside
~ ' temporary: tanks shall be limited to the limits calculated in the 00CM such that a complete release of the tank contents would not result in a concentra-tion at the nearest offsite potable water supply that would exceed the limits specified in 10 CFR Part 20 Appendix B Table II. APPLICABILITY: At all times. ACTION: a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours reduce the tank contents to within the limit, and describe the events leading to this condition in the'next Semiannual Radioactive Effluent Release Report. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ) v. SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed. tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. ..(jh.
- Tanks inclucta in'this Specification are tnose outcoor tanks tnat are not surrounded oy liners, dikes, or walls capable of nolding the tant contents and (nat do not nave tank overflows and surrounding area drains ConncCtcC to the licuia racwaste t.-eatment system.
g \\ y fft93-Swsr i "*!"!"CTC" NUC LCA mGT 2 3/4 11-7 gg L'
3 ~ y _) ~ x-g 82-TABLE 3.3./.Jd l ij +, RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION =: c a: - yj { M1NIMUM ' o S CHANNELS INSTRUMENT I: e OPERABLE ACTION -4 d P 1. GROSS RADIDACTIVITY MONITORS PROVIDING ALARM AND !2 AUTOMATIC TERMINATION OF RELEASE ] 'pis.4 ye LJiden 4,4kr-EsWDiulary Liquid Radwaste Eff?uc..t '.ir.c g a. 1 100
- t1 5.
Tu rt ir.: Suildini; Sump 1/S u...,, 101 - a 2. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE (D g.cmer ut-y Weiss Wder Lost A RJtd.'.n nut e a. 5:rv. : '..' tcr Sy:tc; C f f l ucc.t Li r.: 1 101 Y (9 z r ur pr;ic: .s.,vic, wahr L-p B R.448 4 w M b' )(. "eter Syste-Ef"uent L! c 1/L::; 101 3 3. FLOW RATE HEASUREMENT DEVICES ~ S,,gg jagggp Ef f),7fygir,i a. 1 102
- g. M.ap ft-M Ifud~ 4--
I WL ,.)5hk/[5[.f'bhDNU.'^^ 1 102 De. at i e.9~<y J1*~ ~J~ ri.. s-..r,, ~' Iot e l e I N e '. y i j
- e m
'" '~
d> )L i e TABLE 3.3.7.g-1(Continued) J ACTION STATEMENTS With the number of channels OPERABLE less thar) required by th Minimum Channels OPERABLE requirements, effluent release ACTION 100 - 'this pathway may continue for up to 30 days provided that prio to initiating a release: ~ At least two independent samples of the batch are analyze in accordance with Specifications a. and i b rs of the facility .At least two technically qualif ed mem estaff indepe b. the discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway. With the number of channels OPERABLE less than required by Minimum Channels OPERABLE requirement, effluent release ACTION 101 - t this pathway may continue for up to 30 days provided that, a least once per 12 hours, grab samples are f i <j~L ~ of detection of at least 10-/ microcurie /mt. u- ' e -( With the number of channels OPERABLE less than required b Minimum Channels OPERABLE requirement, effluent releas ACTION 102 - this pathway may continue for up to 30 day 4 Pump performance curves generated in place may be releases. used to estimate flow. O h ,.,do;,.Gigi, i,0aEisa Unii 2 - 3/4 3-g 9' EN - u N rr i .}}