ML20083Q199

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs for Conversion to TS Based on NUREG-1434, Improved BWR6 Ts
ML20083Q199
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 05/18/1995
From:
CENTERIOR ENERGY
To:
Shared Package
ML20083Q182 List:
References
RTR-NUREG-1434 NUDOCS 9505250171
Download: ML20083Q199 (31)
v • d • e


Text

ISOLATION ACTUA) .A INSTRUMENTATION J

5 0 ANN bel L TRIP FUNCTION P E RI 5S(STEkLa) _C040 ION ACTION ' .

mg7h PRIMARY CONTAINMENT ISOLATION Gao , ,

N [ev $ b* D h ![ok*$ i 2) 2 1, 2, 3 and M l' oS *

@G ~~

Q ow,Leve$

feactor V skei Wak!r (Div Level -

ion 3) 4(d) 1, 2, 3 and I (O

28 F k j bS$ o $kbf~ 2 1,2,3 20 R m

osm h$$[o $f*"#*~"9 4(d) 1, 2, 3 28 F

$ ntal C0 3

'iu$ation-HighE!haus$"hl!nE*

lad 2*) 1, 2, 3 and / MS/t

}$ctpggsgelWaterLevel- 3 g(O

$$0$ ifs $$ 2(*) 1, 2, 3 and 2d/k

" 1, 2, 3 and *

, '3' . bS0$loSk 1(*) 28 lf. MAIN STEAM ISOLATION -jy

a. {ea tggsgel Water Level - up , ,

2D

N! YakkN H 2 1, 2 h*eSshheI$0* 2 1 2+ E

$o0 kgh 2/line 1, 2, 3 &D 1, ** **

$e d%. Condenser Vacuum - Low 2 24-() I o R.% Main Steam Line Tunnel 1,2,3 M &D e

0

_ Temperature - High 2

$*Smh$$uhe"*hS90 2 1, 2. 3 &

f$ $!m Ofa "$ ll$gP

  • 2 1,2,3 2t C 3 S '*- *]
g. M. Manual Initiation h O

2 1,2,3 N- 6 / qr

o>m

$07 m

TABLE 3.3.2-1 (C:ntinued) "sQ

! esr r -

ISOLATION ACTUATION INSTRUMENTATION -

g y7 HINIMUM OPERABLE CHANNELS APPLICABLE -

k$

  • TRIP FUNCTION OPERATIONAL E

~

PER TRIP SYSTEM (a) _ CONDITION ACTION t-3.' SECONDARY CONTAINMENT ISOLA M

a. Reac r Vessel Water Level Low, Level 2 \ 2 1, 2, 3 and # 25
b. Drywell P ssure - High 2 1,2,3 25 nual Init ation 2 1.2,3 2
4. ~ '

REACTOR WATER CLEANUP SYSTEM ISOLATION

(

a. A Flow - High 1,2,3 b.

1 rf H 'q A Flow Timer 1 C, ct, e t.

1 1,2,3 I h 3 EquipmentAreaTemperatureD y 9, 3, h High 1 1,2,3 ft MF Equipment Area A \

~d. _

Temperature - High 1 -

1,2,3 y Reactor Vessel Water Level - r3,3,y n 1.g. Low, Level 2 ,

'g ,2 1,2,3 'h{

Ambient Temperature - High 1 1,2,3

,7 o g; Nain Steam Line Tunnel A

}{

g Temperature - High _

1 (ig 2,_3

?. A y k. SLCS Initiation 1 1, 2,@ 1 27 E Manual Initiation y f .h 2 1,2,3 pH 26 A i4 r

w h

s'R N i

TABLt 3.3.2 l ic:et.. a ) ' -

1ym g ISOLATION ACTUATION INSTRUMENTATION wg$ 'g

< o MINIMUM APPLICA8LE 'wA i

0PERABLE CHANNELS OPERATIONAL i 'E i  !

-4 TRIP FUNCTION PER TRIP SYSTEM (a) _ CONDITION ACTION E

REACTOR CORE ISOLATION C0OLING SYSTEM ISOLATION
a. RCIC Steam Line Flow - High 1 1,2,3 f'/427 h RCIC Steam Supply Pressure -

Low 1 1,2,3 ) 27 RCIC Turbine Exhaust -

01aphrage Pressure - High 2 1,2,3 2 7s h~ RCIC Equipment Room Ambient Tamarature - Hfgh 1 1,2,3  : 7 w i. m. n. e qu w ..; L , ' a - - -

'- ~ i m

1 ~

T;-;;.et. ; "' ? _

-+- t, 2, *

~

H Y f. Main Steam Line Tunnel g Ambient Temperature - Mfgh 1 1,2,3 (7 (g. Main Steam Line Tunnel "

L_ A Temperature - High 1 1, 2. 3 2t Q Main Steam LineTimer Temperature Tunnel 1 1,2,3 2 p3, y

{ ,

Q RNR Equipment Room Ambient Temperature - High L t o t-j 1/ Area 1,2,3 2 F

k j. RHR Equipuent RoomT

-- - o a Temperaturet - High 1/ Area 1,2,3 2C y ,

E RCIC Steam Line Flow High 'g-

, Timer 1 1,2,3 2 n

Dryvell Pressure - High Y. 1 1,2,3 N! 27 k, g. Manual Initiation 1 1,2,3 G/// 26 @

l.3(,kWG.[9Cl(. 51etd Ge4l.w-H %3 krms (o.C. (Q.wt $3ske,15o1 d low ))

EMA

-T.*f'? - ? ?. 5-1 Ira-tL.u)

' ' 1 ;*A h ISOLATION ACTUAT!0N INSTRUMENTATION

.gs-3E 4 -RT e MINIMUM APPLICABLE d g OPERABLE CHANNELS OPERATIONAL $

q TRIP FUNCTION PER TRIP SYSTEM (a) _ CONDITION ACTION E

" h RHR SYSTEM ISOLATION

a. RHR Equipment Area Ambient 1 Temperature - High (F3. 5. t..Ilu i c, i t 1/ Area 2, 3
b. RHR Equipment Area A - --

Teeperature - High 1/ Area Y 2, 3 e 1o .

O_u t i /.' RHR/RCIC Steam Line '

I.% t.n -

Flow - High 1 1,2,3 8

- .jt-fe-) c,;3 c,in R h Reactor Vesss1

- Low, Water s

  • Level level 3 2@ 2 2 2g-E Q R.acter

@I

...iqiFHR CusM MF"* * "" '

m 4#

2 1,2,3 h/tf h Drywell Pressure - High 2 1,2,3 284/4 h Manual Initiation 2 1,2,3 t]sIlj p F

a w R 1,3 g -

E ,

t

PY-CEI/imk-1951L b D 3.3.b [

ha 5 Tc.N ' 3.3 b .l - l

. TACK 2 1 2 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

1. PRIMARY CONTAINMENT ISOLATION
a. Reactor Vessel Water Level - Low, Level 2 NA 1l9 b.

c.

d.

Drywell Pressure - High Containment and Radiation - High ell Purge Exhaust Plenum Reactor Vessel Water Level - Low, level 1 NA

< 10(*

NA --

I m) 3 31.

O l

e. Manual Initiation NA
2. MAIN STEAM LINE ISOLATION LAI -

/

a. Reactor Vessel Water Level - Lowsh(evel 1 < 1. @M hk 6 Main Steam Line Radiation - N1ahr' !i < 1.0* 4 /10l*> b
c. Main Steam Line Pressure - Low - 1. 0* Rfl0(a),*,* )
d. Main Steam Line Flow - High e.

(10 ,

Condenser Vacuum - Low (1%

f. Main Steam Line Tunnel Temperature - High NA (g. Main Steam Line TunneT A Temperature - High 7
n. surbine Building Main Steam Line Temperature - High NA
i. Manual Initiation NA v 5 16-
q. n.
3. SECONDARY CONTAINMENT ISOLATION I
a. Reactor Vessel Water Level - Low, level 2
b. Drywell Pressure - High NA A NA L c. Manual Initiation NA j
4. REACTOR WATER CLEANUP SYSTEM ISOLATION
a. A Flow - High NA
b. A Flow Timer NA
c. Equipent Area Temperature - High NA (d. Equipment Area a Temperature - High
e. Reactor Vessel Water Levet - Lvw, Level 2
f. Main Steam Line Tunnel Ambient NA f _ Temperature - High Main Steam Line Tunnel A Temperature - High N
h. SLCS Initiation NA
i. Manual Initiation NA PERRY - UNIT 1 3/4 3-21

ei;cytnn<u-19siu L.L o ~3 3 . q . \

2' a[ ,

7,yg 3. 3_3_3 go., e m_g)

'~'

, ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
a. RCIC Steam Line Flow - High NA
b. RCIC Steam Supply Pressure - Low NA
c. RCIC Turbine Exhaust Diaphragm Pressure - High NA ,
d. RCIC _ Equipment Room Ambient Temperat_ure - Hich_ NA,

_ , _...___ ,-_-w r - m __ _ mm -

"" ~

"55airtim?iUENIA$fint Tem NA (g. max $perature _High team _Line_ Tunnel A Temperature - High hPMifn Steam Line Tunnel'Tiaiipiratiire Tiiiiir

~ CA N

D ##,,

  • NA
1. C: a (I.RHR_ RHREquipment _ Rooe_

Equipment Room Ambient- High A Temperature

_ TemperatureKA NA- High

k. RCIC Steam Line Flow High Timer -

NA , i

1. Drywell Pressuu - High NA
m. Manual Initiation NA
6. RHR SYSTEM ISOLATION a RHR Equipment Area Ambient Temp _erature - Hi NA 7 RHILEquipment_AreaATempetature - High~gh7

~

c. RHR/RCIC Steam Line Flow - High NA
d. Reactor Vessel Water Level - Low, Level 3 NA
e. Reactor Vessel (RHR Cut-in Permissive)

Pressure - High NA g

f. Drywell Pressure - High NA ( 3(!m,f l )
g. Manual Initiation NA va 4

N (a) Isolation system instrumentation response time specified includes the diese7 ,

generator starting and sequencejoadiDG delayu A T T S.;,; w Wpp A (b1Radiationdetectorsareexempt 1. rwsponse time testing.J Response time  :

y snall be seasureu i,-- uwsector output or the input of t w first electronic /

e omponent in the channel, f*generatordelaysassumed. Isolation system instrumentation response time for 2* Isolation system instrumentation response time for associated valves a scent MSIVs. _

fIsoi' ation system instrumentation response time specified for the Trip

__ Function actuating each containment isolation valve shall be added to the isolation time for each valve to obtain ISOLATION SYSTEM RESPONSE (TIME for each valve.

PERRY - UNIT 1 3/4 3-22 Amendment No. 44

'N r M . M M-wwPop)

TABLE [d.2.2.1-13.6 l -l Tyt,g g,3g of,7 j -

g ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS - -

5 S R 3%f./ " " 'l CHANNEL 4 At3

' r I OPERATIONAL i

CHANNEL FUNCTIONAL l,.9 CONDITIONS IN WHICH c

5 TRIP FUNCTION CHECK TEST CALIBRATION { SURVEILLANCE REQUIRED PRIMARY CONTAINMENT ISOLATION '

a. Reactor Vessel Water Level - -

Low, Level 2 SI 1t-Q 4 'R(b)4 1, 2, 3 and #

b. Drywell Pressure - High 5 -i M- Q -2 4 R(b)-7 1, p , 3
c. Containment and Drywell Purge Exhaust Plenum Radiation - '7 High Si tt-Q'l R-T 1, 2, 3 and *
d. Reactor Vessel Water Level - t Low, Level 1 S -- t (C 7 4'R(b).7 1, 2, 3 and #
e. Manual Initiation NA R :- d NA ~ 1, 2, 3 and

S, h7 jWO q

4s R(b)' 1,2,3 1 Inw. Level 1_ _. _

l

[b. Main Steam Line Radiation "- = = - - - - - - ,

[

w

((. p C g.

Hioh _

n Steam Line Pressure -

Main Steam Line Flow - High

_S S-'

._M tt g

~

L 4,

s R

(b)-

,7 1, 2 N 3

1,2,3 d e'. Condenser Vacuum - Low S Mga 1, 2**, 3**

e_ I' Main Steam Line Tunnel Temperature - High S *Q- 7 R -4 1,2,3 f* g' Main Steam Line Tunnel A Temperature - High S'I , 4 Q 'l R"4 1,2,3 g t(. Turbine Building Main Steam (v Line Temperature - High S ~7 R4 1,2,3 ht Manual Initiation NA 3 NA 1,2,3

- 3' # 5 e n ..

f L g 4 T A 3.a.t. H 3 gg

& JAhi AT tla ps Q $5 N J

~;

W e-

PY-CEI/hRR-195~1L Attachment 2 Page 8 DISCUSSION OF CHANGES CTS: 3.3.2 - ISOLATION ACTUATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.9 These changes to this section have been previously proposed in a letter to the NRC, PY-CEI/NRR-1439 L, dated September 28,1992, or have been previously approved by the NRC in Amendment 44 to the license. The letter proposed eliminating the Main Steam Line Radiation Monitors (MSLRM) instrumentation trip signal to the MSIV. However the instrumentation requirement was proposed to remain in the specifications to fulfill a note requirement to the Table. Amendment 44 to the license removed numerous notes to the Isolation instrumentation table, including the note affecting the MSLRMs. The No Significant Hazards Consideration for the proposed changes is still valid. Therefore this change is considered administrative for purposes of this submittal.

A.10 This comment number is not used for this station.

A.11 This comment number is not used for this station.

A.12 This comment number is not used for this station.

A.13 The Area Temperature and Area Differential Temperature Functions are separately identified in the proposed Specifications by area. Since this does not change the required instrumentation but is only a change in presentation of the same requirements, this change is considered administrative.

A.14 The current ACTION for this function requires both isolation and declaration of the affected system as inoperable. Since the instrumentation for isolation of RWCU is to assure OPERABILITY of the SLCS, the "affected system" can be either the SLCS or the RWCU. These choices are reflected in the proposed Required Actions. Therefore, this change is considered administrative.

RELOCATED SPECIFICATIONS R.1 Qhis comment number is not used for this statio sua e PERRY - UNIT 1 13 12/13/93

. , . 1 PY-CEI/f3RR-1951L Attachment 2 Page 9 DISCUSSION OF CHANGES CTS: 3.3.2 - ISOLATION ACTUATION INSTRUMENTATION (Insert lla) 1 R.1 The differential temperature instruments proposed to be relocated are i not assumed to function to mitigate any accident described in Chapters 6  !

or 15 of the USAR. These differential temperature instruments are provided only to detect and initiate isolation of a 25-gpm-equivalent steam leak. However, these instruments constitute only one method of detennining steam leakage in their respective areas. In addition to the temperature monitoring, excess reactor coolant leakage can be detected by low reactor water level, high process line flow, high differential flow, and various other plant specific methods. Several of the BWR-6s have performed studies and analyses which support the relocation of the differential temperature instruments from the Technical Specifications.

The differential temperature instruments are neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).

The differential temperature isolation instruments are neither used for, nor capable of, monitoring a process variable that is an initial condition of a DBA or transient analyses.

The differential temperature isolation instruments are not used as part of the primary success path in the mitigation of a DBA or transient. No pressure-temperature analyses, radiation dose calculations, or equipment qualification parameters take credit for operation of these differential temperature instruments. In addition, adequate redundancy is available to perform their functions by other methods.

Although the overall isolation instrumentation Function satisfies Criterion 3 of the NRC's Final Policy Statement on Technical Specification Improvement, these differential temperature instruments are not assumed to function to mitigate any DBA or transient.

PERRY - UNIT 1 11a 5/10/95

j

'." CRER System Instrumentation  ;

3.3.7.1 1 PY-CEI/NRR-1951L Table 3.3.7.1-1 (page 1 of 1) Attachment 2 Control Room Emergency Recirculation System Instrumentation Page 10 APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SFECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEP ACTION A.1 REQUIREMENTS VALUE

1. Reactor vessel Water 1,2,3, 2 B SR 3.3.7.1.1 E 14.3 inches Level - Low Low Low, (a) SR 3.3.7.1.2 '

Level 1 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5

2. DryweL L Pressure - High 1,2,3 2 B SR 3.3.7.1.1 5 1.88 psig SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5
3. Control Room 1,2,3, 1 C SR 3.3.7.1.1 s 800 cpe Ventilation Radiation (b) SR 3.3.7.1.2 Monitor SR 3.3.7.1.4 SR 3.3.7.1.5 3

(a) During operations with a potential for draining the reactor vessel.

(b) During CORE ALTERATIONS, operationswithapotentialfordrainingthereactorvessel,[movementof irradiated fuel assemblies in the primary containment or fuel handling building, r 1 13.3.7.1 l dz

( c. ..$ z-J d

4 PERRY - UNIT 1 3.3-73 AmendmentNo.fh j E i l

^ "

3/4.4.1 PECIRCULATION SYSTEM RECIRCULATION LOOPS klRUMS%M M.J2JHuuW -.

?

3.4.1.1 ' The,_ reactor coolantsys tem recircula tion-loop (sLshalLbe..in ...

- opg_ra r tica' vi th the to tal core flow grea ter-than-or-equal-to -45Fof-rated-core gp' n

5) 149*r-or- THERMAL POVER l-esrthan or equal to the r limi thspeci-fied-in-Pigure pt

-3. 4.1.1-1 -and -e i t he r : , e -i G.

f

. . --_ q

% dx L- '

.-)' L s-p".

o - / b](\,),

x M m.Q %

, ~

Twos recirculation

~ ,, t . m.lopps.\ s E <perat-ing,-or'

~ . , 1 bk .w A-singl3 recirculation loop operating with-the 'vIledng)

"7~Iimits and cu..didwacS&p Mm.hb AM.'

N h6 3 'l ) The MINIMUM CRITICAL POVER RATIO (MCPR) Saf -

I adjusted for single recirculation loo p p ion per g

\ , specification 2.1.2, and b) The +13x mum Average Planar Li ear Beat Generation Rate j L (MAPLHGR) mits adjusted f single recirculation loop

operation pe'rsthe CORE OF TING LIMITS REPORT  %

in accordance Ulth Spec ication 3.2.1, and j

b) v - .

p j- c) The Average Pover Range Monitor (APRM) Scram and Rod Block

\ l.b 9 -

Trip Setpoint andfilov'ahle Value equations adjusted to 7

those valiggs ap Iicable fo'rgingle recircu on loop operation .p Specification 2.2.1 and 3.3.6 M L4 o '7--v 2 '2- A-volume tric -r irculation loop drive-f a less-than.or- ual .

N .e .2 ~' 2 to 48,500 g r-and M w-Q<J NM h ~ t c, . M 2 :' l '^ :.\.h ,

1 3. The re Manual '-' -

D Po*__ ion rculation flow control.,_

.Coritrol)-mode,_and_f system in the 12bp \ . '

(

._ THERMAL POVER less-than-or-equal to.2500 Megavatts-th 5.id APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 .

ACTION:

a.

L(.o 1imits initial Upon arisi/setp6fn'GTSi>Ecifieatiorii) entry into single loop operation, 2.1.2, 2.2.1T,M@ adjustments to the @

'3. % I t4 cTE tQ. 3. 6 h hours, or declare the astectated A all'be equipment TiiplimWid'vithin[

inoperable (or declare the associat_e.d_1_ mits to b'* i ,__ _.J _

s h-(be "not satisfied"l/

specificagons --- and take the ACTIONS required by the applica~bg h See Special Test Exception 3.1 bk

    • To function lly implement these protective functio during entry into single loop ration, APRM gain adjustments may be e in lieu of  %

adjusting the A M Scram and Rod Block Flov Biased Setp ints for an interim period of hour pb KRT PERRT - UNIT 1 3/4 4-1

h 1."id M.Ahik$A_1;iikkMAW ddYJddfS'lh '

M '

h it i -

70 _ _ ._.-, .,

. - .__- t. _.. ___

...&.t 4

m, . ,,,

4, g' = .-

-,p4. a a,41 j4.r, .,

p. . - ._

4,

.n . - ,_._..  ;  ; 4 ., _ i

.Jy . g.gpg .

a ,4 N$

74 h

y. -sadY. s,

' ' i' P.Mi@

~ ~

! us k go

.. _t , m a... -

(. 4 _w ,,c ..d.u_ ..p44

+- >-

1 -

g n.

1 i .

_.3_- .A.

--+ .

r ,,,s -:A .77++t w__ j.4 . ' e WQgi 1 < _

. .. _;-.-. -q {a t- .

u o _-_- 5 ---

"#  ! f

~~

442'4 I - k '

4 50 ., 1

^

1 3

i 1 . - p rg r 3 u m.

a i i.

1, .:

1..

c s

s. 14 3

uM

_a4_ .

44, tg.

e ri i 4F" g i i

- , 1 i

I

. i t ' .

r j i

4.. +. ,4. 4-1 1 i i m-

,, j a ' ' '

j-G.J tf+T.gQ

t. -

.- 2 1 pm g r .

---,.4 a444

,s W, j 1 > _

if g, _-

y ._--

j , . __ ..- _.a. . 4 4.g. s p: y> - 'I~~ 4

-~7iItal 3 -___.-__--

ir #

, '- '! I II O i 4m il  ::..

P 7

a i

, 1l  !

,, . 2.

'j'h'li IerM* '""' "

L I

N f4k \

$ .1

: ^-

z 30 I h I I '

i i1

  • 1 1

_& -++

k p.

Z l

, -? . -

g ,

> t- 1 i i C.tti!:

0

. 11 ' - Tfit -

44:

u .4. u++. .

5 2o . 4-k + k 4 .-.

7 > 1 .-pu,

_s - _ - i!4,. i i_& a .

titir*tti ttt'- It+ -

4;a g .'t.

10 -

4,1ta .

._ a m.a

,. 4

_4

.I yp..

./ _

i j4 ...4

_a_9...w 44 4 4 4 .

-+- .. i 4+

4

-+

_ - 4-._- __- +- - & . t J.. .- ._

L

,,, q. 4f +.

,f ,44 m>m

_,,, _ 4 _ _,._-

i.._.--._ . - _._-., . _ __

.J . 9 .- ' ' '

0

,1 ., ,

50 60 70 $g?

30 40 oao om CORE FLOW (% RATED) e h64 THERMAL POWER VERSUS CORE FLOW A@

u-

> v:

FIGURE 3.4.1.1-1 ~

fseA f p U.l-t

PY -C E1/ NMM- 19 51 L Attachment 2  ;

Page 13 DISCUSSION OF CHANGES -

CTS: 3.4.1.1 - RECIRCULATION LOOPS ADMINISTRATIVE (continued)

A.6 The specified changes were proposed in a letter to the NRC, PY-CEI/NRR-1353 L, dated June 28, 1991. Due to the extensive changes requested in that letter, Attachment 3A of the submittal letter is inserted into this proposal and then .

revised accordingly. Footnote ** to that change request is not being carried forward into this change request. The No Significant Hazards Consideration proposed in the letter are still valid. Therefore the change is considered administrative for purposes of this submittal.

A.7 This comment number is not used for this station.

A.8 The existing action to "immediately initiate action to reduco. . .within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />" is proposed to be revised to " initiate I

action to restore ...Immediately". The existing requirement would appear to provide a two hours in which the power / flow ratio could exceed the limits, even if capable of being returned to within limits. Also, if the parameters are incapable of being restored to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the existing action would appear to result in the requirement for an LER. The intent of the action is believed to be more appropriately presented in proposed Required -

Action D.1. This interpretation of the intent is supported by the BWR Standard Technical Specification, NUREG-1434. As an enhanced presentation of the existing intent, the proposed change is deemed to be administrative.

A.9 The footnote explaining that the numbers given are preliminary and could be changed by a Technical Specification change is proposed to be deleted. This footnote is unnecessary since a Technical specification change can always be submitted if new data determines it is warranted. The footnote does not contain any technical data necessary for the plant operation.

A.10 This comment number is not used for this station.

A.11 A new SR is added to periodically check that the LCO requirements are being met. Since the LCO must always be met, and the operator maintains awareness of the reactor power and recirculation flow parameters, this change is considered

__ administrative in nature.

hus RELOCATED SPECIFICATIONS None in this section.

PERRY - UNIT 1 2 12/13/93 I

  • PY -CE1/ t4HK- 19 d iL Attachment 2 Page 14 4 DISCUSSION OF CilANGES CTS: 3.4.1.1 - RECIRCULATION LOOPS (Insert 2a)

A.12 The current LC0 description of the allowed region of operation, i.e., "...

with the total core flow greater than or equal to 45% of rated core flow, or THERMAL POWER less than or equal to the limit specified in Figure 3.4.1.1-1..." is now graphically represented on ITS Figure 3.4.1-1 as Region II. Since these regions are equivalent, the proposed change is deemed to be administrative.

PERRY - UNIT 1 2a 5/10/95

PY -CE1/ fSH-1951L Attachment 2

, Page 15

' REACTOR COOLANT SYSTEM l C. O 3.h . C

) OPERATIONAL LEAKAGE

(,(

LIMITING CONDITION FOR OPERATION l 3.4.3.2 Reactor coolant systes leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE. .I A/s6E 7 p ,9
b. 5 gpa UNIDENTIFIED LEAKAGE. N *
  • i

{

c. 25. gpa IDENTIFIED LEAKAGE averaged.over any 24-hour period. Ly

$R d. 0.5 gpa leakage per nceinal inch of valve size up to a maximum of gy,4, f @5 non ed infrom Tableany reactor coolant 3 4.3T-Ives system rar.co pf%m v. pressure isolation valvQ - L APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 3. 4- . [p effhorI ACTION:

Fest a.wg soc, N pcA y,g a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within twnc 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

LCo b. With any reactor coolant system leakage greater than the limits in b 3M and/or c, above, reduce the leakage rate to within the limits within

@ 00 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and M'O C in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

g c. With any reactor coolant system pressure isolation valve leakage greater y *q than the above limit, isolate the high pressure portion of the affected W OA system fros_. the low pressure portion within 4 ho_urs by use of at le*

one ott.ce(closed manual or deactivated automaticKor check" valve Q,(\ ,

NOO or be in at least HOI SHUTOOWN within the next if hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

T.N iEG:t L t- Q 'l.N .% h C.t ot4 9 Q os E l E psce m.s.s e,ecc t g

9SEtt bCo 3.N. C bWc $ '

Which have been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.

l PERRY - bNIT 1 3/4 4-10 t i

'

3.4.6 RCS Pressure Isolation Valve (PIV) Leakage LCO 3.4.6 The leakage from each RCS PIV shall be within limit.

APPLICABILITY: MODES 1 and 2, .

MODE 3, except valves in the residual heat removal (RHR) i 1 , shutdown cooling flow >ath E- :: -:__: . __ __ - mw yr - . - - c: : =: Dwhen in the shutdown cooling mode of operation.

3 c/ dv/.ag % -

^ ^ '#

ACTIONS )

NOTES-------------*---------------------

1. Separate Condition entry is allowed for each. flow path.
2. Enter appitcable Conditions and Required Actions for systems made inoperable by PIVs.

CONDITION REQUIRED ACTION '

COWLETION TINE A. eakage from one or (;-- -------NOTE-----------

more RCS PIVs not Ive used to satisfy OM '~s. within limit, '

Each Re Wquire ired etion A.1 and ion A.2 shall k O ne or m ,c have been v fled to meet p*W f *g S SR 3.4.6.1 an e in the k reactor coolant essure '

% ,N % boundary [or the h /

,i pressure portion of .

( system].

l (Continued)

    • Q  %  % * * * * ., l
  • W t. VR) ko y fdi>C'g Lg,'a j h d.L

) M s%U %- % wk U S

) % m A S L 3.'t. G .t s n .)

w. . w c t. a . ,

1 u d 3 9<a>e c k 4 .c; e>.,W k.iw Uu A ' g . , N J _ S w s,U C^ .

' ' ' ' 5;[.~ ( 33/2",/02 Un/0 % ' '

~ ~ 3.61I - J J l l

1

P i -U E 1/ tJ MM- I n 11.

. Attachment 2

, Page 17 RCS PIV Leakage B 3.4.6 1

i BASES (continued)

APPLICptlJY In H0 DES 1, 2, and 3, this LCO applies because the P!V M leakage potential is greatest when the RCS is pressurized.

L% In H00E 3, valves in the RHR flowpath are not required to

(}ecde meet the requirements of this LCO when the RHRgmode of operation.

q .

or .{,o., In MODES 4 and 5, leakage limits are not provided because r the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. Accordingly, the potential for the consequences of reactor coolant letkage is far lower during these MODES.

4 ACTIONS The ACTIONS are modified by two Notes. Note 1ha$been provided to modify the ACTIONS related to RCS PIV flow

' paths. Section 1.3, Completion Times, specifin once _a yMM -

3 Condition has been entered, subsequent + + r, subsystems, components or variables expressed in the Condition, CV discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for the Condition of RCS PIV leakage limits exceeded provide appropriate compensatory measures for separate affected RCS PIV flow paths. As such, a Note has been provided that allows separate Condition entry for each affected RCS PIV flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system OPERABILITY, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perforta its safety function. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems. .

A.1(D If leakage from one or more RCS PIVs is not within limi the flow path must be isolated by at least one closed -

O manual, deactivatedt-%tomatic, or check valve within __

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> . @u t . ,.w Q.7,_n.2 m _ __.- m . v d'"= $  %] 3

{ ._ X W j n 3 N 4_d^iM N ryy (continued)

-BWR/6. STS B 3.4-28 Rev.- G. G3/2e/.G2

@ "' RCS P!V Leakage Page 18 B 3.4.6 BASES . l n

CM ACTIONS A .16 7 (continued) '

h v .y, \ iA d w s 1.: p' :.'i;d ( :...g e g 9 ; :';; ;h;',o; .o mj fer 1

N{S

\

l(rp2 anu ..wu n um um gyeo y gp cypcp2 -

, _ _on

c. . . ., m . . . . . . .m r-.

e v= :- ~ww wa Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows time for these actions and restricts the time of operation with leaking valves.

RegtNrgd Action A.2 specifies that the double isolation barrienof two valves be restored.by closing another valve qualifiedyor isolation or restori' one leaking P!VK The g 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Compit tion Time after excee the time requit d to complete the Requ ed Action, the l' the limit considers probability of a cond valve failing du g this time period, and the lo grobability of a press boundary rupture of the low pregsure ECCS piping when verpressurized to reactor pressure (Re 8) --

8.1 and 8.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The~ Completion Times are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS Perfonnance of leakage testing on och RCS PIV is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpa per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable (continued)

W/G Sik 8 3.4-29 R:n 0, G5/29/92

_ __m ________m - _ _ . - - _ - . _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ - - - _ _ _ _ _ _

'. PY-CE1/NHH-1951L Attachment 2 Page 19 .

INSERT B28A

! A check valve may be used for this purpose if leakage past the

!- check valve did not exceed the allowable leakage limit at the last refueling outage, or after the last time the valve was j , known to have opened, whichever is more recent.

Af - un cia.,yd frm awk Jul~ 5k l

l l

I .

l e

INSERT PERRY - UNIT 1 B 3.4-28 12/13/93

e .

  • PY-CE1/NRK-1951L Attachment 2 Page 20 DISCUSSION OF CIIANGES TO NUREG-1434 TS 3.4.6 - RCS PRESSURE ISOLATION VALVE LEAKAGE CljANGE/ IMPROVEMENT TO NUREG STS (continued)

C.5 These changes are proposed to revise specific terminology to that which is generically preferred for application to the BWR/6 plants. The BWR LCOs do not use the term " train",

however, " division" is used in several places.

C.6 This change is to provide clarity and prevent including a list containing excessive procedural type details in the Required Action of the Technical Specifications.

Noteyeleted from Require (Action and a' ropriate sti ulations FC.7 from NDTE added to Bases.

(

T s is consis nt with this ,ind of statementNuade other places Technical _ S eqifications. j C.8 Correct mode of IUIR system operations inserted into Bases \

Applicability statement. )

C.9 The reference is deleted from the text since the limit being /

discussed could not be identified in the reference. ,'

r ',/

(' e ,,smd ed uh)M

{p l y OnJ Id*

~

l o.w i

  1. 4 1

1 PERRY - UNIT 1 9 12/13/93

4 a * % &e 1/ 6 4t\f5 - 1 J J 1 L.

M t achtremt 2

.N _. . . . , _ N St.p tc4( Muh-

~ cuN uu nntn s $Y$TEMS LCo ~3 b. l . I l_ C o ~S ,6 l. 3

' PRIMARY CONTAltmENT LEAKAGE #6

- I ( I,j LIMITING CONDITION FOR OPERATION L l' b a 2- Primary containment leakage rates shall be limited to:

]C gg

a. f An overall integrated leakage rate of less than or equal to . 'La* l g g(q *g I 0.20 percent by weight of the primary containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> atP,.(ll.31ps

\.

b A combined leakage rate.cf less than or equal t[0.60 Le or all penetrations and all valvesjexcept for< main steam iWe iso atT3]D falves and valves which are hydrostaticallDeak tested.I subject'to 43.(

lype B and (test n pressufizedTo P,,Ql.31 psigKg 3 A,9

'~3.(,,l,

, Less than or equal toh5 tcf per hour for anyye_ main ste,am linb--

( through the isolation valves when tested at P,. Ql.31 psigg

d. A combined leakage rate of less than or equal to 0.0504 L, for all 3kj penetrations that are secondary containment bypass leakage paths F 'O when pressurized to the required test pressure. U.13

.;,"'q'g' 3 . A combined leakage rate of less than or equal to 1 gpm times the g total number of containment isolation valves in hydrostatically tested I tL J1 nes ch penetrate the primary containment, when tested at 1.10 Pa ' l (12.44psig. g APPLICABILITY: OPERATIONAL CONDITIONS 1. 2* and 3.

ACTION: . 3'hd With: U 3 . 4 .1. 3 A (~Tt W % E Ll

a. The measured overall integrated primary containment leckage rate exceeding
b. The measured combined leakage rate for all penetrations and all valves except for main steam line isolation valves and valves which are hydrosta i ally leak tested, sub ect to Type B and C tests exceeding .60 L, y @ug l.----___.

$ .2,( )

l c. The measured leakage rate exceeding (25 scf per hour for any one. main d'Q steam line through the isolation valv'es or

g3

d. The combined leakage rate for all penetrations that are secondary containment bypass leakage paths exceeding 0.0504 L, or
e. The measured combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary ,

containment exceeding 1 gpm times the total number of such valves:

i Special Test Exception 3.1 D <-- i PERRY - UNIT 1 3/4 6-3 Amendment No. 44 1

e . . P r -C h r/ nx A- 4 3 3'4 L

,

  • Attachment 2 Page 22

. DISCUSSION OF CHANGES p CTS: 3.6.1.2 - PRIMARY CONTAINMENT LFAKAGE 3 ( 1, j CIb j ADMINISTRATIVE (continued)

A s\u , 13 n ,'T y u] be ihs 34s.{'.

A.3 I These changes have already been proposed tox the NRC in a

' letterq PY-CEI/NRR-1576 L dat'ed March 1, 19q3. The .75 proposed to be deleted from LCox 3 .6.1.2 a. i merely a different presentation of what has been proposed in qhe letter to the NRC, N In that request.It wassproposed, among other changes, to insert wording into this re'qqirement to clarify that the requireiiiegt was only for as-leYt leakage rhtes. The Therefore as found lealcago rates up to La are%c'lceptable.

proposed change made fnqhis present submitta presents a act same clearer approach, but st;ill contains the requirements, that a leak ra'tg of La is acceptable ,

found, but has to be made less than orMqual to .75 La as -left. The No Significant Hazards Considerat' ion written as b~ art of that j

~

nubmittal is still valid. Thereforc this change is considered "

/

administrative.

A.4 The deletion of these numerical values were already proposed in a letter to the NRC, PY-CEI/NRR-1510 L dated June 24, 1992.

The No Significant Hazards Consideration written as part of that submittal is still valid. Therefore this change is considered administrative.

A.5 The format of the proposed Technical Specifications does not include providing " cross references." The existing reference to "See Special Test Exception 3.10.1" serves no functional j purpose, and therefore its removal is purely an administrative i difference in presentation. l A.6 This comment number is not used for this station. -

1 A.7 This comment number is not used for this station.

A.8 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

PERRY - UNIT 1 5 12/13/93

w -a1/rike- m 1L Attactm.ent 2 REACTOR COOLANT SYSTEM page 23 50 3.M \\.S IDLE RECIRCULATION LOOP STARTUP 3 S q. 1 %.1\.y LIMITING CONDITION FOR OPERATION S R. ~3 M \\ . 3 3.4.1.4 An idle recirculation loop shall not be started unless the ~ s.ht temperature differential between the reac tor pressure vessel steam space coplant and the bottoe head drain line co_o,JAnt/Is less than or equal to f(00*F",and: _ . . _ _ . _ . .- --

a. When both loops have been iditD unless the temperature differential f G.3.M . 'd .y eween the reactor coolant w< thin the idle loop to be started up '

and the coolant in the reactor pressure vesseys less than or equa h

Qo 50 F, or ,_  !

b. N When only one loop has been idle, unless the temperature differenti5 N1 between the reactor coolant within the idle and operatino recirculation) kloops is less than or equal to 50*FRnd the operating soap flow rat 3 l Lis less than or equal to 50% of rated loop flow. --

3 /2. 3 .9.il,3 N. A e

[LSE3.M.u.9 NA<

\.--> APP LICAB I LITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

Lu ACTION:

With temperature differencesMiow ratWexceeding the above limits, suspend startup of any id!* recirculation loop.

SURVEILLANCE REQUIREMENTS S R. 3.% I 1. 3

~

S E S.N.M.N

4. 4.1. 4 The temperature differentialsCand_ flow rat $ shall be determined to be within loop.

the limits within 15 minutes prior to startup of an idle recirculation S L 3.M .\\.*L % <

  • Below 25 psig, this temperature differential is not applicable.

l PERRY - UNIT 1. l 3/4 4-6 Amendment No.42 I

a .. ,

i

  • PYJ CE17Tmbyyg Attachtnent 2 Page 24 ,

. 1 DISCUSSIOff OF CHANGES l CTS: 3.4.1.4 - IDLE RECIRCULATION LOOP STARTUP ADMINIfzTBliT.LVE A.1 Thermal stresses on vessel components during recircula, tion loop startups are dependent on the temperature difference between  !

the idle loop coolant and the RPV coolant. Proposed ,

SR 3.4.11.4 ensures the temperature difference between any loop to be started and the RPV coolant is acceptable. A requirement to monitor the temperature difference between an idle loop and an operating loop is unnecessary and can be deleted as it is redundant to the loop-to-coolant requirement of SR 3.4.11.4.

However, the loop-to-coolant temperature check may use the operating loop temperature as representative of " coolant temperature."

BELOCATED SPECIFICATIONS lione in this section.

TJ&lH11 CAL CHANGE - MORE RESTRICTIVE None in this section.

TECHNICAL CEANGE - LF4S_EESTRICTIVE

" Generic" _ hc 4 wd d bb Mc34) 4 LA.1Who details re ting to RPV prec e and temperatur- mits have been r ocated to the -ssure and Tempera e Limits Report ( ). These li s for system oper on are also descri in the USAR. anges to the PTLR be controlled by e provisions the proposed Re ting Requirements Q ion in Chapt a of the Technical S ecifications.

12.2 The flow rate of an operating loop during idle loop startup is not a required initial condition of any transient analysis.

The speed requirement is an operational limit to reduce the probability of scram during idle loop startup and is therefore covered by plant specific procedures.

" Specific" None in this section.

PERRY - UNIT 1 8 32/13/93

._, ___ _ ___-._ _ m - --_ __._. . _ . _ . - - _ . _

  • ~; , yi-dE1/ tmM u b 11 Attachment 2 Page 25 - L AS r*

ti DISCUSSION'OF Cl!ANGES i CTS: 3.8.2.1 - D.C. SOURCES - OPERATING ADMINISTRATIVE I (continued)

A.8 The present wording changed to. clarify that inter-cell .

connections'also include inter-rack'and inter-tier connections where required. Since this does not represent any change to present requirements, but is merely a clarification, the change is considered administrative.  ;

RELOCATED. SPECIFICATIONS None in this section.

l . TECHNICAL CHANGE - MORE RESTRICTIVE

.1 In the event a battery experiences a discharge (not a momentary d7 transient as in the. starting of a large load) which brings the battery terminal voltage below 110 volts, confirmation of continued battery OPERABILITY should be made sooner than 7 days.

The proposed change to*(3" hours allows. sufficient time to plan for an' unscheduled surveillance and. complete the performance of 6 the surveillance of cell parameters without undo haste. Since

.43 this confirmation of OPERABILITY is revised to make that determination sooner, the change is conservative.

7 M.2 The allowance to correct the Category B limit for temperature is being proposed for deletion based on the BWR Standard Technical Specification, NUREG-1434, presentation and IEEE-450 recommendations.

M.3 Limitations are proposed to be imposed on this allowance: the utilization of charging current limited to 7 days, and a  :

requirement to measure actual specific gravities at the end of this period. These restrictions will assure excessive reliance l on charging current is not made.

M.4 Proposed Required Action A.1 requires a more immediate check that pilot cell electrolvtA~1evel and float voltage are within limits. Required Action A.2 proposes a periodic re-verification that all cell parameters are within limits. These restrictions provide added assurance of adequate battery capabilities for the period allowed to completely restore the cell parameters.

M.5 The frequency for Surveillance 4.8.2.1.f has been chang'ed from 18 months to 12 months. Since the change will require the test on a more frequent bases, the change is considered more

, restrictive.

1 PERRY - UNIT 1 19 12/13/93 i

. 4 PY-CEI/NRR-1951L Attachment 3 Page 1 of 6 Summary of Perry Nuclear Power Plant Totally and Partially Relocated Technical Specifications Existing Title Relocation Relocation TS Document Control 1.2 Operational Conditions ORM/ Bases 650.59 4.0.5 Inservice inspection and Testing Programs ORM 650.55a 650.59 3/4.1.2 Reactivity Anomalies ORM/ Bases 650.59 3/4.1.3.1 Control Rod Operability ORM/ Bases 650.59 3/4.1.3.2 Control Rod Maximum Scram Insertion Times ORM/ Bases $50.59 3/4.1.3. 3 control Rod Scram Accumulators ORM/ Bases 650.59 3/4.1.3.4 control Rod Drive Coupling ORM/ Bases 650.59 3/4.1.3.5 Control Rod Position Indication ORM/ Bases 650.59 3/4.1.3.6 Control Rod Drive Housing $upport ORM/ Bases 650.59 3/4.1.4.1 Control Rod Withdrawal ORM/ Bases 550.59 3/4.1.4.2 Rod Pattern Control System ORM/ Bases 650.59 3/4.1.5 Standby Liquid control System ORM/ Bases 650.59 3/4.2.1 Average Planar t.inear Heat Generation Rate USAR/ Bases 650.59 3/4.2.2 Mininum Critical Power Ratio USAR/ Bases 650.59 3/4.2.3 Linear Heat Generation Rate USAR/ Bases $50.59 3/4.3.1 Reactor Protection System Instrumentation ORM/ Bases 650.59 3/4.3.2 1 solation Actuation Instrumentation ORM/ Bases 650.59 3/4.3.3 Emergency Core Cooling System Actuation ORM/ Bases 650.59 Instrumentation J4.3.4.1 ATWS-RPT System Instrumentation ORM/ Bases 650.59 J/4.3.4.2 EOC-RPT System Instrumentation ORM/ Bases 650.59 3/4.3.5 RCIC System Ir.strumentation CRM/ Bases 650.59 3/4.3.6 Control Rod Block Instrumentation ORM/ Bases 650.59 3/4.3.7.1 Radiation Monitoring Instrumentation ORM/ Bases 650.59 3/4.3.7.2 Seismic Monitoring Instrumentation ORM/USAR 650.59 3/4.3.7.3 Meteorological Monitoring Instrumentation ORM/USAR 650.59 3/4.3.7.4 Remote Shutdoun Instrumentation ORM/ Bases $50.59 3/4.3.7.5 Accident Monitoring Instrumentation ORM/ Bases 650.59 3/4.3.7.6 Source Range Monitors ORM/ Bases 650.59 3/4.3.7.7 traversing In-Core Probe System ORM/USAR 650.59 l

1

> a  :

, .

  • 1
  • 1 i

PY-CEI/NRR-1951L Attachment 3 Page 2 of 6 l Existing Title Relocation Relocation TS Document Control 3/4.3.7.8 Loose-Part Detection System ORM/USAR 550.59 3/4.3.7.9 Radioactive Liquid Effluent ORM/USAR 650.59 Monitoring Instrumentation 3/4.3.7.10 Radioactive Gaseous Effluent Monitoring ORM/U3AR 650.59 Instrumentaion 3/4.3.8 Turbine overspeed Protection System ORM/USAR 650.59 3/4.3.9 Plant Systems Actuation Instrumentation ORM/USAR/ Bases 650.59 3/4.4.1.1 Recirculation Loops ORM/ Bases 650.59 3/4.4.1.4 Idle Recirculation Loop Startup ORM/ Bases 650.59 3/4.4.2.1 Safety / Relief Valves ORM/ Bases 650.59 3/4.4.2.2 S/RVs Low-Low set Function ORM/ Bases 650.59 3/4.4.3.2 Operational Leakage ORM/ Bases 650.59 3/4.4.4 themistry ORM/ Bases 650.59 3/4.4.7 Main Steam Line Isolation Valves ORM/ Bases 650.59 3/4.4.8 Structural Integrity USAR 650.59 3/4.4.9.1 RHR Hot Shutdown ORM/ Bases 650.59 3/4.4.9.2 RHR Cold Shutdown ORM/ Bases 650.59 3/4.5.1 ECCS-Operating ORM/ Bases 650.59 3/4.5.2 ECCS - Shutdown ORM/ Bases 650.59 3/4.5.3 Suppression Pool ORM/ Bases 650.59 3/4.6.1.1.1 Primary Containment Integrity - Operating ORM/ Bases 650.59 3/4.6.1.1.2 Primary Containment Integrity - Shutdown USAR/ Bases 650.59 3/4.6.1.3 Containment Air Locks ORM/ Bases 650.59 3/4.6.1.4 Main steam Isolation valve Leakage Control CRM/ Bases 650.59 System 3/4.6.1.7 Primary Containment Average USAR/ Bases 650.59 Air Temperature 3/4.6.1.9 Feedwater Leakage Control System USAR/ Bases 650.59 3/4.6.2.2 Drywell Bypass Leakage ORM/ Bases 650.59 3/4.6.2.3 Drywell Air Lock ORM/ Bases 650.59 3/4.6.2.4 Drywell Structural integrity ORM/ Bases 650.59 3/4.6.2.6 Drywell Average Air Temperatare ORM/ Bases 650.59 3/4.6.3.1 Suppression Pool ORM/ Bases 650.59 3/4.6.3.2 Containment spray ORM/ Bases 650.59

. i a e

PY-CEI/NRR-1951L Attachment 3 Page 3 of 6 Existing Title Relocation Relocation TS Document Control 3/4.6.3.3 Suppression Pool Cooling ORM/ Bases 650.59 3/4.6.3.4 Supp ession Pool Makeup ORM/ Bases 650.59 3/4.6.4 Containment Isolation valves OPM/ Bases 650.59 3/4.6.5.1 Containment vacuum Breakers USAR/ Bases 650.59 3/4.6.5.2 Containment Humidity Control USAR/ Bases 650.59 3/4.6.5.3 orywell Vacuum Breakers USAR/ Bases 650.59 3/4.6.6.1 secondary Containment Integrity ORM/ Bases 650.59 3/4.6.6.2 Annulus Exhaust Gas Treatment System USAR/ Bases $50.59 3/4.6.7.1 Containment Hydrogen Recombiner Systems ORM/ Bases 650.59 3/4.6.7.2 Combustible Gas Mixing System ORM/ Bases 650.59 3/4.6.7.3 Containment and Drywell Hydrogen Ignition ORM/ Bases 650.59 System 3/4.7.1.1 Emergency service Water System ORM/ Bases 650.59 (Loops A, B, C) 3/4.7.1.2 Emergency closed Cooling Water fystem ORM/ Bases 650.59 3/4.7.2 Control Room Emergency Rectreulation System ORM/ Bases 650.59 3/4.7.3 Reactor Core Isolation Cooling System USAR/ Bases 650.59 3/4.7.4 snubbers ORM 650.59 3/4.7.5 sealed Source contaminati:,'i ORM 650.59 _

3/4.7.6 Main Turbine Bypass System ORM/ Bases 650.59 3/4.7.7.1 Fuel Handling Building Ventilation System ORM/ Bases 650.59 3/4.7.7.2 Fuel Handling Building (FHB) Integrity ORM/ Bases 650.59 3/4.8.1.1 AC Sources - Operating ORM/ Bases 650.59 3/4.8.1.2 AC Sources - Shutdown ORM/ Bases 650.59 3/4.8.2.1 DC Sources - Operating ORM/ Bases 650.59 3/4.8.2.2 DC Sources - Shutdown ORM/ Bases 650.59 3/4.8.3.1 onsite Power Distribution systems - operating ORM/ Bases 650.59 3/4.8.3.2 Onsite Power Distribution Systems - Shutdown ORM/ Bases 650.59 3/4.8.4.1 Containment Penetration Conductor Overcurrent ORM/ Bases 650.59 Protective Devices l 1

3/4.9.1 Reactor Mode Switch ORM/ Bases 650.59 3/4.9.2 Instrumentation ORM/ Bases 650.59 l 3/4.9.4 Decay Time ORM 650.59 l

l

, ..'e i 1

l PY-CEI/NRR-1951L Attachment 3 Page 4 of 6 Existing Title Relocation Relocation TS Document Control 3/4.9.5 Communications ORM 650.59 3/4.9.6 Refueling Platform ORM 650.59 3/4.9.7 Crane Travel ORM 650.59 3/4.9.8 Water Level - Reactor Vessel ORM/ Bases 650.59 3/4.9.9 Water Level-Spent Fuel Storage and Upper USAR/ Bases 650.59 Containment Fuel Pools 3/4.9.10.1 Single control Rod Removal ORM/ Bases 650.59 3/4.9.10.2 Multiple Control Rod Removal ORM/ Bases 650.59 3/4.9.11.1 Residual Heat Removal and Coolant Circulation ORM/ Bases 650.59

- High Water Lev i 3/4.9.11.2 Residual Heat Removal and Coolant Circulation ORM/ Bases 650.59

- Low Water Level 3/4.9.12 Inclined Fuel Transfer System ORM 650.59 3/4.11.1.1 Liquid Effluents USAR 650.59 3/4.11.1.2 Dose USAR 650.59 3/4.11.1.3 Liquid Radweste Treatment System USAR 650.59 3/4.11.1.4 Liquid Holdup Tanks USAR 650.59 3/4.11.2.1 Dose Rate USAR 650.59 3/4.11.2.2 Dose - Noble Gases USAR 650.59 3/4.11.2.3 Dose - I-131, 1-133, Tritium and USAR 650.59 Radionuclides in Particulate Form 3/4.11.2.4 Gaseous Radwaste (Offgas) Treatnent USAR 650.59 3/4.11.2.5 Ventilation Exhaust Treatment Systems ORM/ Bases 650.59 3/4.11.2.6 Explosive Gas Mixture ORM/ Bases 650.59 3/4.11.2.7 Main Condenser ORM/ Bases 650.59 3/4.11.3 Solid Radwaste Treatment ORM/ Bases 650.59 3/4.11.4 Total Dose ORM/ Bases 650.59 3/4.12.1 Monitoring Program ORM/ Bases 650.59 3/4.12.2 Land Use Census ORM/ Bases _

8350.59 3/4.12.3 Intertaboratory Comparisen Program ORM/ Bases 650.59 5.1.1 Exclusion Area, unrestricted Area for Liquid Or<M/USAR 650.59 Effluents, and Site Boundary for Gaseous Effluents 5.2.1 Containment Configuration ORM/USAR 650.59 5.2.2 Design Temperature and Pressure ORM/USAR 550.59 i

l

4 . 6 .

.. 1 s PY-CEI/NRR-1951L Attachment 3

, Page 5 of 6 Existing Title Relocation Relocation TS Document Control 5.2.3 secondary Containment ORM/USAR 650.59 5.3.2 Control Rod Assemblies ORM/USAR 650.59 5.4.1 Reactor Coolant System Design Pressure and ORM/USAR 650.59 s Temperature 5.4.2 Reactor Coolant system volume ORM/USAR 650.59 5.5.1 Meteorological Tower Location ORM/USAR 650.59 5.6.1 Fuel storage Criticality ORM/USAR 650.59 5.7.1 Component Cyclic or Transient Limit ORM/USAR 650.59  !

6.1.2 Line of Authority Directive ORM 650.59 6.2.2.a, Minimum Shift Crew Composition and $RO ORM 650.54(m) '

Table Requirements 650.59 '

6.2.2-1, 6.2.2.d 6.2.3 Independent Safety Engineering Group ORM 650.59 l 6.2.4.1 shift Technical Advisor ORM $50.59 i 6.4 Training ORM 650.59 650.55 i

6.5 Review and Audit ORM 650.54(a) 6.6 Reportable Event Action ORM 650.59  ;

650.73 B

6.7 safety Limit violation ORM 650.72 650.73 650.59

6. 8.1. c Security Plan Implementation ORM 650.54(p) 6.8.1.d Emergency Plan Implementation ORM 650.54(q) 6.8.1.e Process Control Program Implementation ORM/00CM 650.59 6.3.1.g Radiological Environmental Monitoring Program ORM/00CM 650.59 Implementation 6.8.1.h fire Protection Pro; ram Implementation CRM 650.48 650.59 LC 2.C(6) i 6.8.1.1 Regulatory Guide 4.15 ORM 650.59 6.8.2 Review and Approval Process ORM 650.54(a) 6.8.3 Temporary Changes ORM 650.59 6.8.4.b In-Plant Radiation Monitoring ORM 650.59 620 6.9.1.1,2,3 Startup Rtports ORM 650.59 l

l

_ . _ _ _ _ . . _ _. . . ~ . _ . _ . _. _ - - ~ . - _

o..a PY-CEI/NRR-1951L Attachment 3 Page 6 of 6 Existing Title Relocation Relocation TS Document Control 6.9.2 Special Reports ORM/ Bases 650.4 650.59 650.7?.

6.10 Record Retention ORM 650.54(a) 620 671 6.11 Radiation Protection Program ORM 650,59 S20 6.12 High Radiation Area ORM 650.59 620 6.13 Process Control Program PCP 650.59 620 661 671 6.14 offsite Dose calculation Manual 00CM 650.59 ORM - Operational Requirements Manual USAR - Updated Safety Analysis Report ODCM - Off-site Dose Calculation Manual LC 2.C(6) - License Condition 2.C(6)

PCP - Process Control Program

Retrieved from "https://kanterella.com/w/index.php?title=ML20083Q199&oldid=2745092"